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TRANSCRIPT
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SAFETY EVALUATION OF THE
CRYSTAL RIVER
UNIT 3
FLORIDA POWER CORPORATION
DOCKET NO. 50-302
U.S. ATOMIC ENERGY COMMISSION DIRECTORATE OF LICENSING
WASHINGTON, D. C.
Issue Date: JULY 5, 1974
-- ------.,
SAFETY EVALUATION
BY THE
DIRECTORATE OF LICENSING
U. S. ATOMIC ENERGY COMMISSION
IN THE MATTER OF
FLORIDA POWER CORPORATION
CRYSTAL RIVER UNIT 3
DOCKET NO. 50-302
July 5, 1974
1.0
2.0
3.0
TABLE OF CONTENTS
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT ...••...... 1. 1 Introduction .•••....................•.....•.•..... 1.2 General Plant Descriptidn .....• · .....•...••........ 1.3 Comparison with Similar Facility Designs •.•......• 1.4 Identification of Agents and Contractors .........• 1. 5 Summary of Principal Review Matters ......•.•......
1.5.1 Site .... ~ ................................ . 1 • 5 . 2 Criteria • . . . . . . . . . . .... ·. . . . • . . . . . . . . . . . . • . 1. 5. 3 Design Bas is Ace iden ts .•.................. 1. 5. 4 Radioactive Releases ....................•. 1.5.5 Organization •...........................•. 1. 5. 6 Financial Qualifications .......•.....•....
1.6 Facility Modifications as a Result of Regulatory Staff Review .... : ... ~ ••....•.•.....•.......•....
SITE CHARACTERISTICS .....•...••.•....•.... · .•.•...•.••... 2 .1 Geography and Demography .••.....••......•......... 2.2 Nearby Industrial, Transportation and Military
2.3
2.4
2.5
Facilities ..................................... . Meteorology •..•....•...........••......•.....•.... 2.3.l Regional. Climatology ....................•. 2. 3. 2 Local· Meteorology ....•...................• 2.3.3. Onsite Meteorological Measurements
7f2.3.4 '-k--2.3.5
2.3.6
Pro gram . ............................... . Short-term (Accident) Diffusion Estimates. Long-term (Routine) Diffusion Estimates ... Conclusions ...................•.•.....•...
Hydrology ..................••...........•.....•... 2.4.1 Hydrologic Description ..........•.......•. 2~4.2 Floods ..................................•. 2.4.3 Water Supply .....•.•.....••....•....••.•.. 2. 4. 4 Ground Water ..........................•... "2 .·4. 5 Con cl us ions ...•...•.......• · .....•.•....... Geology, Seismology, and Foundation Engineering ...
DESIGN CRITERIA - STRUCTURES, COMPONENTS, EQUIPMENT, AN'D SYSTEMS • . . • . . . . . • • . . • . . . . . . . . ." . . . . . . • . . . • • • . . • • . . .
3.1 Conformance with AEC General Design Criteria (GDC} .•••.•....•••.•••••.••••••••...•••••••••••.
3.2 Classification of Structures, Components and
1-1 1-1 1-2 1-4 1-5 1-5 1-6 1-6 1-6 1-7 1-7 1-7
1-8
2-1 2-1
2-6 2-7 2-7 2-7
2-8 2-9 2-10 2-11 2-11 2-11 2-13 2-15 2-16 2-16 2-16
3-1
3-1
Systems ... ........... ., ..... · ...... ,, .............. · 3-1
ii
3.3 Wind and Tornado Design Criteria.................. 3-2 3.4 Water Level (Flood) Design Criteria............... 3-3 3.5 Missile Protection Ctiteria....................... 3-3 3.6 Protection Against Dynamic Effects Associated
With the Postulated Rupture of Piping........... 3-5 3.6.1 Criteria for Protection Against Dynamic
Effects Associated With a Loss-of-Coolant Accident (LOCA).......................... 3-5
3.6.2 Postulated Breaks Outside Containment...... 3-6 3. 7 Seismic Design. . . . . . . . . • . . . . . • . . . . . . . . . . . . . . . . . . . . 3- 7
3. 7 .1 Seismic Input. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 7 3. 7. 2 Seismic Analysis.. . . . . . . . . . . . . . . . . . . . . . . . . . 3- 7 3.7.3 Seismic Instrumentation Program............ 3-8
3.8 Design of Category I (Seismic) Structures......... 3-9 3. 8 .1 Foundations.. . . . . . . . . • . . . . . . . . . . . . . . . . . . . . . 3-9 3.8.2 Seismic Category I Structures.............. 3-9 3.8.3 Containment................................ 3-11
3.9 Mechanical Systems and Components................. 3-14 3.9.1 Dynamic System Analysis and Testing.······~ 3-14 3.9.2 Structural Integrity of Pressure ~etaining
Components............................... 3-16 3.9.3 Components Not Covered by ASME Code........ 3-18
3.10 Seismic Qualification of Seismic Category I Instrumentation and Electrical Equipment......... 3-19
4 . 0 REACTOR ................................................ . 4 .1 Summary Description ............................... . 4. 2 · Mechanical Design ....................... • ......... .
4.2.1 Fuel .•...................................... 4.2.2 Reactor Vessel Internals ................... ~
4. 3 Nuclear Design •...•................................ 4.3.1 Nuclear Analysis ........................... . 4.3.2 Power Distribution ......................... . 4.3.3 Moderator Temperature Coefficient .......... . 4.3.4 Control Requirements ....................... . 4.3.5 Stability ...............................•...
4.4 Thermal Hydraulic Design ...............•...........
5 .0 REACTOR COOLANT SYSTEM ........ .- .............•........... 5 .1 Sunnnary Description ...........................•.... 5.2 Integrity of Reactor Coolant Pressure
Boundary (RCPB) ........•.........••......•....... 5.2.l Fractu~e Toughness ......................... .
4-1 4-1 4-1 4-1 4-3 4-4 4-4 4-5 4-6 4-7 4-8 4-10
5-1 5-1
5-1 5-1
5.3 5.4
5.5 5.6 5.7 5.8
iii
5.2.2 General Material Considerations ..•.•..•..... 5. 2. 3 Water Chemistry Control. •.....••. ~ .••.•..••• 5.2.4 Control of Stainless Steel Welding ..•.•••.•• Reactor Vessel Integrity ..........•....•.••.•.•.•.. Reactor Coolant Pressure Boundary (RCPB) Leakage
Detection System ....•....•...•••.•.........••.... Inservice Inspection Program •......•........•••.•.. Pump Flywheel •••.•.....•.•.•••......••.••..•••....• Loose Parts Mani tor .•.•..•......•.•.•.....•...••••. Pump Over speed •.......•......•..•...•.•.•......•.•.
6.0 ENGINEERED SAFETY FEATURES ...••.••..•..•...••...•.•....• 6 .1 Gen er al .•.•..•••.•..•............••...••.......••.. 6. 2 Containment Systems ..•••...........•..•.....••.••..
6.2.1 Containment Functional Design .•.•••...•.•••. 6.2.2 Containment Heat Removal Systems •.......•••. 6.2.3 Containment Isolation Systems .•...........•. 6. 2. 4 Combustible Gas Control Systems ...•....•.... 6.2.5 Containment Leakage Testing Program .•.....•. 6.2.6 General Material Consideration (Compatibility
with Coolant) ........•••....•.•....•..••.. 6.3 Emergency Core Cooling System (ECCS) •....•.....•••.
6. 3.1 Design Bases .•.•...•.•.•....••....•......••• 6.3.2 System Design ..•...........•..•...••.....••. 6. 3. 3 Performance Evaluation .............••••.•••. 6.3.4 Tests and Inspections ••...•.•••.•.••...•..•• 6. 3. 5 Conclusions ••....•.•.•...•..•.....•...•.. • ..
7.0 INSTRUMENTATION AND CONTROLS ...••...•....•..•..•....••.. 7. 1 General ••••.••..••.........•.•.......•.. · ..........• 7.2 Reactor Protection System (RPS) ..•.....•....•.••••. 7.3 Engineered Safety Feature (ESF) Systems •.•.•.•.•.••
7.3.l Core Flooding Tank Isolation Valves .•.•.•••. 7.3.2 Steam Line Break Isolation (SLBI) .....•••...
7.4 Systems Required for Safe Shutdown •..........•••... 7.5 Safety Related Display Instrumentation ••..•...••.•. 7.6 RHR Overpressure Protection Interlocks ••..••..•...•
~. 7 Control Room Ventilation .•.....•.....•.•.•...•..••. 7. 8 Environmental Qualifications .•..•..•.•........••••. 7.9 Separation and Identification of Safety Related
5-4 5-5 5-6 5-6
5-8 5-8 5-9 5-10 5-11
6-1 6-1 6-1 6-1 6-6 6-8 6-9 6-10
6-11 6-12 6-12 6-13 6'-15 6-17 6-18
7-1 7-1 7-1 7-1 7-2 7-2 7-3 7-4 7-5-7-5 7-6
Equipment. . • . . . . . . . . • . . . . . • . • • • • • . . . . • • • . • • • . • • • • 7-6 7.9.1 Reactor Protection System (RPS) Cable
Separation. . . . . . . • • . . . • . • • . . • . . • • . • • . . . . • . 7-6
7.9.2 7 .9. 3 7.9.4
iv
Switchgear Rooms Flooding ••.••.•••..•••••••• Battery Rooms Separation •••...•••••••.••••.• 230 KV Switchyard Breakers Control Power
7-7 7-8
S~par~tioti~............................... 7-8 7.10 Control Systems ••.•...• ··-··················........ 7-9 7.11 Anticipated Transients Without-Scram (ATWS)........ 7-10
8. 0 ELECTRIC POWER. . ••••••. '· ••.•....•••• ·• ~· ..••••• _ •. .' •• ~ • • . . • • • 8-1 8.1 General ..... ;'....................................... 8-1 8. 2 Off site A-C Pol¥ er System .•••••••••••••.•.••••.• ~ . . • 8-1 8.3 Onsite A-C Power Systems ••.. _ •.•••• _~................ 8-3 8.4 D_-C _Power System .• _ .••••••...••. _.................... 8-7
9 .0 AUXILIARY SYSTEMS....................................... 9-1 9 .1 Fue_l Storage and Handling.......................... 9-3
9.2
9.3
9.4
9.1.1 New Fu,el Stor_age •••••••••••• _.··········"••··· 9-3 9 .1. 2 Spent Fuel St.or_age .•. • ••.••••••.. ·'· .• •. • . . • • • • • • • • 9-3 9 .1. 3 Spent Fuel Pool Cooling and C_leanup
9 .1.4 Water 9.2.1
. S.ystems •.•••••••••• ~ ••••••• -· ••••••.•• · .••••• Fuel Handling -System ....•... -......••.••••••••
Systems .•.• .- ...•••••••••. ~ .• -.• •.••.•.•.•.••••••. Nuclear Services Cooling Water
System (NS CWS ) •..••...•..•...•••...•.•..•• 9.2.1.1 Nuclear Service Seawater
9-6 9-7 9-8
9-8
Subsys tern. . . . . . . . . . . . . . . . . . . . . . . . . 9-9 9.2.1.2 Nuclear Service Closed Cycle
Cooling Water Subsystem.......... 9-9
9.2.2 Decay Heat Services Cooling System ..•.••.•.• 9.2.2.1 Decay Heat Service Seawater
Cooling System (DHSSCS) ••••.••••• 9.2.2.2 Decay Heat Closed Cycle Cooling
Water Subsystem (DHCCCWS) .•..•••.
9.2.3 9.2.4
9 . 2 . 2 . 3 Con cl us ions ........................ . Ultimate Heat Sink (uHS) .•.••.•• ~ •••.•.••••• Condensate Storage Facility ••••••. ~ ••••••••.
Process Auxilia~ies .... ........................... . 9.3.1 Chemical Addition and Makeup Systems •..••••• 9.3.2 Storage Of Compressed Gases ••••.••.•• ···~··· Air-Conditioning, Heating~ Cooling and Ventilation
Systems •••••..••••••• .•.•.. · •. ., .•.••••••••••••.•••
9-10
9-11
9-11 9-11 9-12 9-13 9-13 9-13 9-15
9.4.1 9.4.2
&J9.4.3
Control Complex Building •• ~ ~ ••••...•••.•.••• Fuel Handling Area •....••••...•••••••.•••••• Engineered Safety Feature and Other
9-16 9-16 9-17
Essential Equipment. • • . . • • . • • . . • • • • . • • • . • • 9-18 9.5 Other Auxiliary Systems............................ 9-19
v
9 .5.1 Fire Protection System........................... 9-19 9.5.2. Diesel Generator Fuel Oil Storage, Transt:e+
and Auxiliary Systems.......................... 9-21
10 .O STEAM AND. POWER CONVERSION SYSTEM •••••.••• • ••••.••.•••.•. 10.1 Summary Description .. -.:· ............................. . 10. 2 Turbine Genera tor •....••••.••.•••.•••...•.•••••••• 10 .·3 'Main Steam '!$upply ·Sy.stem ••••...••....••.•••..•••.• 10. 4 Steam and Power Conversion Subsystems ••••• ~· •••••••
10. 4. 1· General . ... • ........ -............. .::. ....... . 10.4.2 Turbine Bypass, Steam ••••••.•••..•...•••..• 10.4.3 Circulation Water System •••.••..•••••.•••• 10. 4. 4 Auxiliary Feedwater Sys tern (AFS ) .••••••••••
.11. 0 RADIOACTIVE WAS TE MANAGEMENT •.•••..••••.••••..••••••••.• 11.1 Sunnn.ary Description . .............................. . 11. 2 Liquid Wastes . ................ -.... ~ ............... . 11. 3 Gaseous Wastes ........ · ............................ . 11. 4 Solid Wastes ....... ~ ........................... · .. . 11.5 Design ........................................... . 11.6 Process and Area Radiation Monitoring Systems ••••• 11.7 .Radiation Protection Management •••..•••..•••.••••• 11. 8 Con cl us ions . ........................... ~ ......... •
12. 0 RADIATION PROTECTION •••••••••.••• ~ •••. ; •.••••.•••.•••••. 12 .1 Shielding . .......... o ••••••••.•••••••••••••••••••••
12.2 Ventilation .. o .................................... .
12.3 Health Physics Program ••.••..•......••.••..••.••..
10-1 10-1 10-1 10-2 10-2 10-2 10-3 10-4 10-4
11-1 11-1 11-3 11-7 11-10 ll-12 11-12 11-13 11-15
12-1 12-1 12-2 12-3
13.0 CONDUCT OF OPERA'J;IONS................................... 13-1 13.1 Plant Organizations, Staff Qualifications and
13.2 13.3 13.4 13.5
T~aining . .................. · .................... . Safety Review and· Audit •••...••••••••.•••..••••••• Plant Procedures and Records ••.•••.••••.••.•.••••. Emergency Planning . .............................. 8 •
Industrial Security .............................. .
13-1 13-3 13-4. 13-5 13-6
14.0 INITIAL TEST AND OPERATION ••••• ·•..•••••••••..••..••••.•. 14-1 14.1 Test Startup Program.............................. 14-1
¥:-is.a ACCIDENT ANALYSIS....................................... 15-1 . . 15.1 General ....... e••••••••••••••••••••••••••••••••••• 15-1 15.2 Design Basis Accident Assumptions................. 15-3
~5.2.1 Loss-of-Coolant Accident (LOCA) •••• ~...... 15-3 15.2.2 Fuel Handling Accident.................... 15-5 15.2.3 Gas Decay Tank Rupture.................... 15-5
~15.2.4 Control Rod Ejection Accident............. 15-5 .J 15.2.5 Hydrogen Purge Dose....................... 15-7 'I
r
vi
16 .O TECHNICAL SPECIFICATIONS................................ 16-1
17 .0 QUALITY .ASSURANCE....................................... 17-1 17.1 General·........................................... 17-1
· 17. 2 Organization. • • • • • • • • • • • • • • • • . • • • • • • • • • • • • • • • • • • • • 17-1 17. 3 QA Pro gr am . ...................... ~ ............... o 17-4 17. 4 Conclusions. . • . • • • • • • . . • • • • • • • • • • • • • • • • • • • • • • • . • • • 17-7
18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS) • • • • • • • • • • • • • • . • • • • • • • • • • • • • • • • • • • • • • 18-1
19 .0 COMMON DEFENSE AND SECURITY............................. 19-1
20. 0 FINANCIAL QUALIFICATIONS. • • • • . • • • • • • • • • • • • • . • • • • • • • • • • • • 20-1
21.0 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS......... 21-1 21.1 Preoperational Storage of Nuclear Fuel............ 21-1 21.2 Operating License ••• ·•••••••••••••••••••••••••••••• 21-2 21.3 Conclusions....................................... 21-3
22. 0 CONCLUSIONS. • • • • . • • • • • • • • • • • . • • • • • • • • • • • • • • • • • • • • • • • • • • • 22-1
APPENDIX A
APPENDIX B
APPENDIX C
APPENDIX D
vii
APPEND IC I ES
REGULATORY REVIEW OF FLORIDA POWER CORPORATION, CRYSTAL RIVER, UNIT 3
REPORT OF U.S. ARMY CORPS OF ENGINEERS
FINANCIAL .ANALYSIS
BIBLIOGRAPHY
viii
LIST OF TABLES
PAGE
Table 4.1-1 Comparisbn of Thermal and Hydraulic Designs of CR-3 and Oconee~l •....•.•....•• 4-11
Table 6-1 Subcompartment Design and Calculated Pressures • • • . • . • • . • . . • • • . • . . . • . . . . • . . . . • • . 6-5
Table 15.1 Potential Offsite Doses Due to Design Basis Accidents ...••...•.•.•.•.••••..•..•• 15-2
ix
LIST OF FIGURES
PAGE
Figure 2.1 Crystal River Site Plan .••..••...••••• ~ •••.••• 2-2
Figure 2.2 Crystal River Unit 3 Exclusion Area ••••.•••••• 2-3
Figure 2.3 Crystal River Unit 3 Cumulative Population Distribution . . ~ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4
Figure 17 .1 Florida Power Corporation Organization Chart •• 17-9
a-c
ACI
ACRS
AEC
AISC
ANSI
ASME
ASTM
Btu/hr-ft 2
cfs
CFR
Ci/sec
DBA
d-c
DHCCCWS
DHSSCS
DNB
DNBR
ECCS
ERM
ESF
ESSA
x
. ABBREVIATIONS
alternating current
American Concrete Institute
Advisory Committee on Reactor Safeguards
United States Atomic Energy Commission
American Institute of Steel Construction
American National Standard Institute
American Society of Mechanical Engineers
American Society for Testing and Material
British thermal units per hour per square foot
cubic feet per second
Code of Federal Regulations
Curies per second
design basis accident
direct current
decay heat closed cycle cooling water system
decay heat service seawater cooling system
departure from nucleate boiling
departure from nucleate boiling ratio
emergency co~e cooling system
emergency recirculation mode
engineered safety features
environmental science services administration
OF
ft/sec
g
GDC
gpm
HPI
HPIS
hr.
I-131
IEEE
in
IPS
kV
kVA
kW
kW/ft
LOCA
LPI
LPIS
xi
ABBREVIATIONS Cont'd
degrees Fahrenheit
Fire protection system
square feet
cubic feet
feet per second
gravitational acceleration, 32.2 feet per second per second
AEC General Design Criteria for Nuclear Power Plant Construction Permits
gallons per minute
high pressure injection
high p~essure injection system
hour
Iodine 131
Institute of Electrical and Electronics Engineers
inch
Interim Policy Statement
kilovolt
kilovolt amperes
kilowatt
kilowatts per foot
loss-of-coolant accident
Low pressure injection
Low pressure injection system
LPZ
m
2 m
MM
mph
m/sec
MSL
MWe
MWt
mrem
NOAA
NPSH
NSCWS
OBE
PMF
PMP
ppm
FSAR
psi
psig
psia
PWR
QA
xii
ABBREVIATIONS Cont'd
low population zone
meter
square meters
modified Mercalli
miles per hour
meters per second
mean sea level
megawatts electrical
megawatts thermal
one thousandth of a Roentgen equivalent man
National Oceanic and Atmospheric Administration
net positive suction head
nuclear services cooling water system
operating basis earthquake
probable maximum flood
probable maximum precipitation
parts per million
Final Safety Analysis Report
pounds ·per square inch
pounds per square inch gauge
pounds per square inch absolute
pressurized water reactor
quality assurance
rad
RCPB
rem
RHRS
sec/m3
SIA
SIAS
Sr-89
Sr-90
SSE
U-235
UHS
uo2
USGS
xiii
ABBREVIATIONS Cont'd
radiation absorbed dose
reactor coolant pressure boundary
Roentgen equivalent man
residual heat removal system
seconds per cubic meter
safety injection actuation
safety injection actuation signal
strontium 89
strontium 90
safe shutdown earthquake
uranium 235
ultimate heat sink
uranium dioxide
United States Geological Survey
1-1
1.0 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
1.1 Introduction
The Florida Power Corporation (hereinafter referred to as the
applicant) by application dated August 10, 1967, and as subsequently
amended, requested a license to construct and operate a pressurized
water reactor, identified as the Crystal River Unit 3 Nuclear Generating
Plant (hereinafter referred to as CR-3 or the facility) at a site
on the Gulf of Mexico in Citrus County, Florida. The Atomic Energy
Connnission's Regulatory staff (Regulatory staff or staff) reported the
results of its review prior to construction in a Safety Evaluation
Report (SER) dated June 6, 1968. Following a public hearing before an
Atomic Safety and Licensing Board in Crystal River, Florida on July 16, 1968,
the Connnission issued Provisional Construction Permit CPPR-51 on
September 25, 1968.
On February 8, 1971, the applicant filed, as Amendment No. 11,
the Final Safety Analysis Report (FSAR) required by 10 CFR 50.34(b)
as a prerequisite to obtaining an operating license for the facility.
The operating license application is for a core power level of 2452
megawatts thermal (MWt), the same thermal power.considered by the
Regulatory staff in the construction permit review. Our evaluation of
the design characteristics, the engineered safety features, the con
tainment, and the accident analyses has been based on operation at the
'2544 MWt core power level _as described in the applicant's Final Safety
1-2
Analysis Report (Amendment No. 11) and subsequent Amendments 12
through 39 inclusive, all of which are available for review at the
Atomic Energy Connnission's Public Document Room at 1717 H Street,
N. W., Washington, D. C., and at the Crystal River Public Library,
Crystal River, Florida. Before operation of the reactor at any power
above 2452 MWt is authorized, the Regulatory staff will perform a
safety evaluation to assure that the facility can be operated safely
at the higher power. In the course of our safety review of the
material submitted, we held a number of meetings with representatives
of the applicant, the nuclear steam supply system (NSSS) manufacturer,
the Babcock & Wilcox Company (B&W): and the applicant's architect
engineer, Gilbert Associates, Inc. (GAI). These meetings were to
discuss the plant design, construction, proposed operation and per
formance under postulated accident conditions. A chronology of our
review is attached as Appendix A to this evaluation.
1.2 General Plant Description
The CR-3 uses an Oconee class B&W NSSS. Water is heated by the
reactor and flows through two steam generators, where heat is, transferred
to the secondary (steam) system.
The steam produced in the steam generators is used to drive
the turbine generator which converts the heat energy to electrical
energy. After passing through the turbine, the steam is condensed
and the condensate returned to the steam generators to repeat the
1-3
cycle. The condensers are cooled by seawater drawn from, and recir
culated to, the Gulf of Mexico by extensions to the existing intake
and discharge canal.
The reactor qq9lant system is housed inside the reactor (or containment)
building, a cylindrical, prestressed concrete structure with a shallow
dome roof and a tigt reinforced concrete bas~ ~Lab. The inside
surface of the reactor building is sealed with a welded steel liner.
The reactor building provides a barrier to the escape of radioactive
products that might be released from the reactor coolant system in
the event of an accident. In addition, the reactor building is
equipped with a spray system designed to reduce rapidly both the
pressure and the fission product concentration within the containment
after a postulated accident.
Auxiliary systems, including the chemical and volume control
system, the waste handling system, auxiliary coolant systems, spent
fuel storage facility, and components of the engineered safety features
are located in an auxiliary building, adjacent to and abutting the
reactor building.
Rapid react~v~ty control of the reactor will be achieved by
control rods (n.~utrqn absorbers) that will be moved vertically
within the core by individual control rod drives. Boric acid dis
solved in the c.ool~mt will be used as a neutron absorber to provide
steady state req~tivity control. CR-3 is like Oconee in these respects.
1-4
An Oconee type reactor protection system will automatically
initiate approyriate corrective actions whenever plant conditions
monitored by the system reach preestablished safety limits.
Appropriate actuation instrumentation is provided to initiate
closure of isolation valves and operation of the engineered safety
features should these actions be .required. These engineered safety
features include the containment systems with their supporting heat
removal systems, isolation systems, a filtered purge system for com
bustible gas control, an emergency core cooling system (ECCS) that
will prevent the reactor core from overheating for a broad spectrum
of postulated loss of coolant accidents, an emergency feedwater sys
tem, and an emergency electrical supply system.
1.3 Comparison with Similar Facility Designs
Many features of the design of this facility are similar to
those we have evaluated and approved previously for other nuclear
power plants now under construction or in operation especially the
Oconee Nuclear Station - Unit 1 (Oconee-1) which is the lead plant
for this type of B&W NSSS. To the extent feasible and appropriate,
we have made use of these previous evaluations in conducting our
review of CR-3. Where this has been done, the appropriate sections
of this report identify the other facilities involved. Our Safety
Evaluation Reports for these. other facilities also have been pub-
lished and are available for public inspection at the Atomic Energy
Commission's Public Document Room at 1717 H Street, N. W., Washington, D.C.
1-5
1.4 Identification of Agents and Contractors
The appl~cant has arranged for the purchase of equipment and con
sulting, engineering, and construction services for the design and
construction of CR-3. · As sole owner, the applicant is responsible for
the design, construction and operation of CR-3.
GAI has been retained for architectural, engineering, and procure
ment services. GAI is also providing assistance in employee training,
acceptance testing, quality control, and initial start-up of the
plant.
B&W manufactured and delivered the complete nuclear steam supply
system and supplied the initial reactor core (fuel). In addition, B&W
is supplying technical consultation for erection, fuel loading,
testing, and initial start-up of the complete nuclear steam supply
system. B&W is also participating in the training of the initial
plant operating staff.
J. A. Jones Construction Company is responsible for construction of
the reactor building and the balance of plant (non-NSSS).
Dames and Moore, Inc., is the applicant's principal meteorology
consultant for the CR-3 site.
1.5 Sunnnary of Principal Review Matters
Our evaluation included a technical review of the information
submitted by the applicant particularly with regard to the following
principal matters:
1.5.1
1.5.2
1.5 .3
1-6
Site
We evaluated the population density and land, use characteristics
of the site environs and the physical characteristics of the site,
including seismology, meteorology, geology, and hydrology. Our
purpose was to determine that these characteristics have been ade
quately established and appropriately considered in the final design
of the plant, and to further determine that the site characteristics
in conjunction with the design features of the facility are consistent
with the Commission's siting criteria provided in 10 CFR Part 100.
Criteria
We evaluated the design, fabrication, construction and testing
and performance characteristics of the plant structures, systems,
and components important to safety to determine that they are in
accord with the Commission's General Design Criteria, Quality Assur
ance Criteria, Regulatory Guides and appropriate industrial codes
and standards, and to determine that any depart~res from these cri
teria, codes, and standards have been identified and justified.
Design Basis Accidents
We evaluated the expected response of the facility to various
anticipated operating transients and to a broad spectrum of accidents
and selected a few highly unlikely postulated accidents (design basis
accidents) the potential consequences of which would exceed those of
all other accidents considered. We then performed conservative
1.5.4
1.5.5
1.5.6
1-7
analyses of these design basis accidents to determine that the cal
culated potential offsite doses that might result from their postulated
occurrence would be within the guidelines of 10 CFR Part 100.
Radioactive Releases
We also evaluated the design of the systems provided for control
of the radioactive effluents from normal plant operation to determine
that these systems are capable of controlling the release of such
radioactive wastes from the facility within the limits of the
Connnission's regulations (10 CFR Part 20). We further evaluated
·these systems to determine that the equipment provided can be operated
in such a manner as to reduce radioactive releases to levels that are
as low as practicable.
Organization
We evaluated the applicant's engineering and construction organi
zations, plans for the conduct of plant operation, including the
proposed organization, staffing and training program, the plans for
industrial security, and the scope of planning for emergency actions
to be taken irJ. the unlikely event of an accident that might affect
the general public, to .determine that the applicant is technically
qualified, staffed and organized to safely operate the plant.
Financial.Qualifications
We evaluated the financial position of Florida Power Corporation
to determine that FPC has adequate financial resources to operate the
1-8
CR-3 plant in accordance with the activities that would be permitted
and required by an operating license.
1.6 Facility Modifications as a Result of Regulatory Staff Review
During the review of CR-3 several meetings were held with repre
sentatives of the applicants, its contractors, and its consultants to
discuss the facility and the tecnnical material submitted. A chrono
logical listing of the meetings and other significant events is given
in Appendix A to this report. During the course of the review the
applicants proposed or we requested a number of technical and admin
istrative changes. These are described in various amendments to the
original application. We have listed below the more significant
modifications that have resulted from our review. Included are re
ferences to the sections of this report where each matter is discussed
more fully.
(1) Modification of the design basis water level (Section 2.4.2).
(2) Modifications to main feedwater and steam piping, auxiliary and
intermediate building venting areas, cable trays to meet the
high energy piping rupture design criteria (Section 3.6.2).
(3) Installation of a loose parts monitoring system (5.7).
(4) Modification to residual heat removal (RHR) system piping to
assure abundant cooling water to the core in 'the event of a break
in either of the core flooding tank piping (Section 6.3.2).
~9
(5) Modifications to Core Flooding Tank Isolation Valves (Section 7.3.1).
(6) Upgrading of auxiliary feedwater system(AFS) electric components to
satisfy the requirements of IEEE-279 and 308 (Section 7.4).
(7) Modifications to 230 KV Switchyard Breakers (Section 7.9.4).
(8) Modifications to Onsite Power Systems (Section 8.3).
(9) Modification of auxiliary feedwater system (AFS) to make the con
sequences of a single active failure coincident with rupture of a
high energy pipe acceptable (Sections 10.4.4 and 7.4).
2-1
2.0 SITE CHARACTERISTICS
2.1 Geography and Demography
The CR-3 site is situated on a 4,738-acre tract of land located
in Citrus County, Florida, on the eastern shore of the Gulf of
Mexico. The nuclear facility CR-3, along with two fossil fired units
CR-1 and CR-2 (shown in Figure 2.1), is located approximately 7-1/2
miles northwest of Crystal River, Florida and 70 miles north of Tampa,
Florida.
The site region is characterized by mangrove swamps and marshlands
along the coastal areas to gently rolling hills about 16 miles inland.
The minimum exclusion distance specified for CR-3 is 4,400 feet
(1340 meters). The applicant has selected a low population zone
(LPZ) of 5 miles for this site. There are no residents at present
within a 3-1/2 mile radius of the reactor. Figure 2.2 shows the exclusion
area.
Figure 2.3 shows the present and projected year 2020 cumulative
population surrounding the site. The 1970 resident population within
50 miles was 174,218. The applicant projects that this population will
increase to 382,221 people by the year 2020. This corresponds to a
119% increase in 50 years, and is in substantial agreement with the
population projections of the Bureau of Economic Analysis for the area
surrounding the CR-3 facility.
,.--, r--·-·--,.J !
f EXIST UNITS ,. 2 I UNIT"3 l I
SUBSTATION 1 SUBSTATION I
1 1 I . . .·.~.; ....
., h~ "-'I. • ~t, •
• ..,; & :-. ~ :-.; ""··:.~ •• • ·~ ••
FIGURE 2.1 CRYSTAL RIVER SITE PLAN
! -·-~
GRADE ELEV. 96'
200 400
SCALE IN FEET
N I
N
2-3
. '. ~
2. 0 I< -1 :::> c... 0 10.: c.. :!J
> 1-< ..J ::> ::::!: ::J u
1a3 9 8 7 6
5
4
3
2
CRYSTAL RIVER
YEAR 1970
10 20 30 40 50. (Ml LES)
DISTANCE.FROM PLt,NT, MlLES
FIGURE 2-3 . CUi·liULATIVE POPULATION DISTRl3UTION
2-5
The population within the 5-mile LPZ was about 500 in 1970. The
nearest population center (as defined in 10 CFR Part 100) with a
population exceeding 25,000 is Gainesville, Florida, which is located
55 miles NNE of the siteo
At the present time, the land bordering the site is sparsely popu
lated and of a rural nature. The Gulf of Me~ico is used for transporta
tion, boating, and extensive commercial and sport fishing. The major
waterways in a five-mile radius of the site are the Cross Florida Barge
Canal (only a western section has been constructed) and the site
entrance channel which has a maximum 14-foot draft and extends 14 miles
out into the Gulf of Mexico. The public facilities within a 10-mile
radius of the site include the Crystal River Indian Mounds Museum Park,
which is located approximately 6 miles southeast of the site (annual
attendance in excess of 47,000), and four schools.
The major agricultural land use in the vicinity of the site
consists of approximately 60% woodlands and 20% range and pasture lands.
Little of the available land is used for crop production. Recreational
land and water use in the area of the site consists of fishing, boating,
and small game hunting.
We have concluded that lanq and water uses have been adequately
considered and are not critical with respect to interaction with the
operation of this facility. On the basis of the applicant's specified
population center distance, minimum exclusion area distaµce, and low
2-6
population zone distance, and potential radiological dose consequences
of design basis accidents (discussed in Section 15.0 of this report),
we have concluded that the exclusion area radius, low population zone,
and population center distances meet the guidelines of 10 CFR Part 100,
and that the CR-3 site is acceptable.
2.2 Nearby Industrial, Transportation and Military Facilities
There is no significant manufacturing or storage of hazardous
materials within a 10 mile radius of the site. There is a dolomite
quarry located approximately 4 miles from the site which uses approxi
mately 1000 pounds of TNT equivalent per blasting event, but does not
store explosives. There are oil storage facilities at Yankeetovm and at
Inglis which are located 4.3 miles and 5 miles, respectively, from the
site. The two oil fired plants on the site are supplied with oil by
approximately 3 barges per week via a 14-mile long channel and intake
canal. The Seaboard Coast Line Railroad Company tracks are located
approximately 3-1/2 miles east of the site. The closest road, US 19,
passes approximately 3 miles from the facility. There are no airports
within a 5-mile radius of the plant site. Approximately 8 miles southeast
of the site there is a small sod covered airfield which is used by small
aircraft. There are no natural gas pipelines passing near the site.
There are no missile bases near the site.
In view of the large exclusion radius and the low industrial activity
within 5 miles of the site, we conclude that offsite hazardous materials
will not affect the safe operation of CR-3.
2-7
2.3 Meteorology
2.3.l Regional Climatology
The climate along the central Gulf Coast of Flprida is character-
ized by warm, wet summers contrasted by relatively cooler, drier
winters. The Gulf of Mexico has a moderating influence on temperatures
and increases the humidity in the site area. Local circulation is
modified by the Gulf with the establishment of the diurnal land-sea
breeze regime. The area is generally well south of primary storm
tracks, although lows formed over the Gulf during the fall occasionally
.track northeastward through the area. Tropical storms and hurricanes
are not uncommon in the area, and make landfall in the vicinity of
the site about once every 8 and 12 years, respectively. In the
period 1936-1965 there were about 46 cases of atmospheric stagnation
episodes lasting 4 days or more, and 5 cases lasting 7 days or more at
Tampa. Atmospheric dispersion conditions are expected to be better
than the average for all sites in the United States.
2.3.2 Local Meteorology
The site is about 70 miles north of Tampa, in an area where the
topography varies by only 20 feet within 5 miles of the site. Tempera-
tures may be expected to reach 90°F or higher 75 days per year, while
temperatures of 32°F or below may be expected on about 4 days annually.
Monthly mean temperatures at Tampa range from 61°F in January to 82°F
in August. Fifty to sixty inches of precipitation can be expected each
'
L
2.3.3
2-8
year, with about 50% occurring during the months of June through August,
mainly as a result of thunderstorms. These 3 months account for 56 of
the 87 thunderstorm days expected annually. Snow is negligible. In the
period 1871-1971, 52 tropical storms and hurricanes passed within 50
miles of the site. Tornadoes are primarily associate.d with tropical
storms and hurricanes. In the period 1955-1967, 33 tornadoes were
reported within the one degree latitude-longitude square c0ntaining the
site, giving a mean annual frequency of 2.5 and a computed recurrence
interval of 580 years at the site. Also for the period 1955~1967, there
were 26 reports of hail 3/4 inch or greater, and 31 reports of windstorms
with speeds of 50 knots or greater in the one degree latitude-longitude
square containing the site. The "fastest mile" of wind reported at
Tampa was 84 mph in September 1935. The predominant wind flow in the
area is from the northeast through east.
Onsite Meteorological Measureme~ts Program
The current onsite meteorological measurements program consists
of measurements from a 150-ft tower, lor.ated about 1800 feet west of
the nearest building, CR-2, which is the highest building onsite
at about 190 feet. Current instrumentation on the tower does not
fully meet the recommendations of Reguiatory Guide 1.23, "Onsite
Meteorological Programs," dated February 17, 1972. It consists of
Bendix Model 120 aerovanes (nominal starting speed of 2 mph) measuring
wind speed and direction at the 35-ft and 150-ft levels. Although data
2.3.4
2-9
have been recorded at both levels since 1970~ recovery did not meet
the recommended 90% until 1972. The applicant has submitted one year
of onsite data (January 1, 1972..;.;December 31, 1972) with data recovery of
about 97% at both levels. These data were submitted as joint· frequency
distributions of wind speed and direction by atmospheric stability, with
· stability classifications derived from the standard deviation of
horizontal wind fluctuation (sigma-theta) instead of from a vertical
temperature gradient. The expected accident and annual average dis
persion conditions for the site, have been evaluated using the wind and
stability data from the 150-ft level for 1972, with the wind speed
reduced to 33-ft by use of the power law for wind profileS'. It is the
staff's opinion that sigma-theta measurements from 150~ft are more
conservative than the 35-ft measurements. Prior to plant licensing the
applicant is committed (by letter dated October 19, 1973) to establish a
new meteonological measurements program, fully commensurate with the
recommendations of Regulatory Guide 1.23, and consisting of a wind speed
and direction sensor at the two levels, ambient temperature at two
levels, and differential temperature between two levels~ In addition,
this'i~formation will be displayed in the CR-3 control room. Evaluation
will be made prior to issuance of an operating license.
Short-term (Accident) Diffusion Estimates
In the evaluation of short-term (0-2 hours at the exclusion
distance and 0-8.hours at the LPZ distance) accidental releases from
'·
2.3.5
2-10
the buildings and vents, a ground-level release with a building wake
2 factor, cA, of 925 m was assumed by the staff. The, relative concen-
tration (X/Q) for the 0-2 hour time period for onshore flow conditions
which is exceeded 5% of the time was calculated by the staff (using
the model described in Regulatory Guide 1.4, "Assumptions Used for
Evaluating the Potential Radiological Consequences of a Loss-Of-Coolant
Accident for Pressurized Water Reactors," dated December 1, 1970) to be -4 3 2.2 x 10 sec/m at the exclusion distance of 1340 m from the reactor.
This relative concentration is equivalent to dispersion conditions pro-
duced.by Pasquill F stability with a wind speed of 1.3 meters/second.
The relative concentrations for design basis accidents for onshore
flow conditions at the outer boundary of the low population zone (8047
m) were estimated by the staff to be:
1.0 x 10-5 sec/m3 for the 0-8 hour period,
6.8 x 10-6 sec/m3 for the 8-24 hour period,
-6 3 2.75 x 10 sec/m for the 1-4 day period, and
-6 3 7.5 x 10 sec/m for the 4-30 day period.
The relative concentration estima.tes of the applicant are within
a factor of two of the values calculated by the staff. These differ-
ences can be attributed to use of data from the 35-ft level and use of
all directions in the analyses.
Long-Term (Routine) Diffusion Estimates
The staff calculated highest overland offsite annual average relative
concentration value of L 5 x 10-6 sec/m3 for vent releases occurring at the
2.3.6
2.4
2.4.1
2-11·
site boundary (1450 m) east-northeast of the reactor structures.
This value is about a factor of two more conservative than calculated
bJ the applicant. This qifference can be attributed to the use of
only onshore winds from the 35-ft level for the arinual average cal
.culations performed by.· the applicant.
Conclusions
The staff concludes that the data presented in the FSAR for the
period January 1, 1972 to December 31, 1972, are the best available for
the site at this time. The instrumentation used for the wind speed and
direction measurements, and the use of the standard deviation of
horizontal wind fluctuations for classifying stability conditions, are
not fully connnensurate with the recommendations of Regulatory Guide 1.23.
The applicant will establish an onsite meteorological measurements program
that is fully in accordance with Regulatory Guide 1.23, and that displays
appropriate parameters in the control room. The staff will evaluate
this program before an operating license is issued for this facility.
The applicant has been requested to provide at least one representative
year of onsite data (with data recovery of at least 90%) from the new
program to allow the staff to further verify the atmospheric dispersion
conditions.
Hydrology
Hydrologic Description
The site is in northwestern Florida on the Gulf of Mexico, approximately
2-12
70 miles north of Tampa, 40 miles southwest of Ocala, and about 3.8
miles south of the mouths of the Withlacoochee River and the Cross
Florida Barge Canal. Cooling water intake and discharge channels have
been dredged about eight miles into .. the Gulf with dredgings disposed
in spoil bank islands along the channels. The southerly intake channel is
also used for shipment of fuel to the two fossil units located inn:nediately
gulfward of CR-3. The slope of the continental shelf offshore is very
mild with natural depths of 10 feet or less 8-10 miles out from the
site.
Topography in the site vicinity is very flat and storm surges in
the Gulf can cause extensive low land flooding. An island has been
constructed by the applicant for the reactor, control complex,
emergency diesel, and turbine buildings; fire service :pump house;
main transformers, and storage facilities for diesel fuel oil and
water. All safety-related. facilities a.ssociated with the maintenance
of shutdown are on the".island. The switchyard, necessary for normal
operatfon, is at elevation 10 feet U.S. Coast and Geodetic Survey
(u.s·.c.& G.S .• ) mean low water (MLW). Zero (O.O) feet, U.S.C.&G.S.
mean low water (MLW) datum is equivalent to a Crystal River Datum,
used by the applicant, of 88.0 feet. Existing grade around the island
is elevation 10.0 feet MLW. Grade level on the island is elevation
30.5 feet MLW. The crown of the service road around the island is 0.5
foot higher than the island grade. Both public and private water
supplies in the site area are derived from ground water.
2.4.2
2-13
Floods
The flood potential from stream and river flooding, site drainage,
and hurricane~induced surges was investigated by the applicant and the
staff.
Local stream and river flooding is not considered a threat to the
facility because of the relative locations, slopes, and elevations of
streams with respect to the facility's raised island.
Site drainage for the fa~ility island has been designed for
a maximum rainfall rate of 10 inches per hour; somewhat less than
could occur during a local probable maximum storm. However, drainage
in the vicinity of safety-related structures has been sloped away
from the structures to preclude pondage, even during a local prob
able maximum storm. Roof drains have been designed to discharge
directly into the site storm drainage system without roof pondage for
rainfall intensities up to 6 inches per hour. For ponding that could
occur for more severe rainfall intensities of probable maximum
severity, the roof structures will support the ponded water.
During the PSAR review stage, analyses of hurricane flood potential
for the site were performed based on preliminary Environmental Science
Services Administration, ESSA, (now NOAA) Hydrometeorological Branch
estimates of probable maximum hurricane (PMH) parameters. Our con
struction permit Safety Evaluation Report, dated June 6, 1968, noted
2-14
"The applicant has stated, and :we will require, that the plant protection
will conform to the applicable portions of revised ESSA criteria." The
estimated maximum hurricane-induced surge level at that time was 21.4
feet MLW with wave heights up to 9.0 feet higher. Based upon detailed
studies of surge levels which could result from presently accepted NOAA
PMH parameter estimates, the applicant proposed a maximum still water
level of 29.4 feet MLW with wave runup to elevation 35.5 feet MLW.
We and our consultant, the U.S. Army Coastal Engineering Research
Center (CERC), each reviewed the applicant's analysis in detail and
independently estimated that the maximum surge-induced still water level
could reach elevation 33.4 feet MLW with correspondingly more severe wave
action. Our consultant's report is contained in Appendix B. The
staff required the applicant to provide protection against the higher
still water and wave level estimates made by the staff and our con
sultant. As part of that protection, water-proof doors are provided
at vulnerable accesses to safety-related structures, a low wall is
provided around a portion of the north and west sides of the turbine
building, a stepped reinforced concrete cap is provided around the
south, and a portion of the west slopes of the nuclear island, diesel
fuel oil facility exposures are protected, and finally a Tech_nical
Specification requiring plant shutdown and emergency procedures in
anticipation of severe hurricanes will be employed. The applicant esti
mated the maximum wave runup level on exposed portions of safety-related
structures for the higher surge level would be elevation 39.0 feet MLW,
2-15
and has provided assurances that the static and dynamic effects of
such water levels will not adversely affect safety-related structures.
We concur iri the applicant's waverunup estimate, and conclude that
safety-related structu~es~ systems, and components are adequately pro-
tected from severe flooding, providing that the reactor is shut down
in anticipation of severe hurricanes as required by the Technical
Specifications.
2.4.3 Water Supply
Safety-related water supply is to be taken from the intake channel
via pumps located inside buildings on the nuclear island. The pumps are
thus protected from flooding in the same manner as other safety-related
facilities. Minimum pump water level require~ents are 17.1 feet below
MLW in the nuclear 'is.land building sump, and, at a conservative slope
of one foot per mile, 9 feet below MLW at the entrance to the intake
channel eight miles out in the Gulf. The applicant estimated that the
probable minimum low water level at the entrance of the channel would
be 4.7 feet below MLW. This estimate was based upon the water level that
would result from a PMH oriented to produce maximum sustained off shore
winds blowing away from the facility and channel. We conclude that
sufficient water level margin exists such that safety-related water
supply will be available, even under the most adverse hurricane conditions.
This conclusion is based upon the assumption that the intake channel will
be periodically surveyed and dredged as necessary to preclude blockage at
low water levels.
I --
2.4.4
2.4.5
2-1~
Ground Water
Ground water in the site area occurs in small shallow pockets of
relatively small areal extent, and in the large, extensive, and
deeper Floridan artesian limestone a,quifer. In s_ome areas the
Floridan outcrops in the form of springs. The permeability of soils
and rock at the site are very high and are typical of the region;
however, the ground water mapping in the site vicinity indicates ground
water moves toward the Gulf to the west and southwest, and away from
potential users. The water table conditions at the site have been
estimated to vary with tide levels in the Gulf and are generally at or
near elevation 2 feet MLW.
Conclusions
The staff has_ concluded that adequate flood protection from severe
hurricanes has been provided, that heavy local or regional rainfall
should not adv~rsely affect the plant, that sufficient water will be
available for safety-related purposes, and that any accidental
releases of radioactive liquids should not reach any water supply
users.
2.5 · Geology, Seismology, and Foundation Engineering
We and our consultants reviewed the geology and seismology of
this site with respect to faulting, foundation conditions, and intensity
of earthquakes at the construction permit stage of our review. No
new information has been obtained since our construction permit review
2-17
to change our previous conclusions on the acceptability of the site
relative to 1:-ts geology,. seismology, and foundation conditions.
A horizontal ground acceleration of 0.05g was used for the
Operational Basis Earthquake (OBE) and an acceleration of O.lOg for
the Safe Shutdown Earthquake (SSE). These acceleration values remain
adequate for seismic design of the plant structures and components.
All seismic Category I structures are founded on structural.back~
fill. The backfill was em.placed over the Inglis member of the Moody
Branch formation which is a cream colored to occasionally tan, porous,
grandular, biogenic limestone, and dolomite. Bedrock is approximately
20 feet beneath the original ground surface and is of Tertiary Age
(lO to 65 million years ago).
Because the exploratory investigation revealed the presence of
both open and filled solution cavities in the limestone bedrock, the
applicant undertook a program of consolidation grouting. The grouting
extended into the dolomite and bottomed at an average elevation of
10 feet MLW. In a few areas chemical grouting was required.
The seismic Category I backfill below the ground water level consisted
of an uncompacted blanket of groutable coarse aggregate (Brookville lime
rock) • An impervious Visquene membrane was placed on top o.f the aggre
gate, and a load bearing fill of 1500 psi concrete was placed thereon
to the bottom of the foundation mat. The coarse aggregate was pressure
grouted during the first stage of consolidation grouting. Elsewhere
2-18
for above ground water placement structural fill concrete was used.
The staff reviewed and approved the foundation conditions before the
construction permit was issued. Although there were a few changes
from the proposed foundation preparation, those were reviewed by the
staff and found acceptable. No other new facts have been uncovered
during the construction which would affect the~previous acceptance.
We conclude that the foundations are structurally adequate to carry
the applied loads.
3-1
3.0 DESIGN CRITERIA - STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS '
3.1 Conformance with AEC General Design Criteria.(GDC)
This facility was designed and constructed to meet the AEC's GDC,
as originally propose4 in July 1967. The Connnission published the re-
vised GDC in 1971 just before the FSAR was filed. We conducted our
technical review against the present version of the GDC and we conclude
that the plant design acceptably conforms to the current criteria.
3.2 Classification of Structures, Components and Systems
The applicant has classified the plant structures, components
and systems into three principal categories. Seismic Category I
includes those structures, components and systems whose failure might
cause or increase the severity of a loss-of-coolant accident, or result
in an uncontrolled release of radioactivity and those structures,
components and systems vital to safe shutdown and isolation of the
reactor. Seismic Category II includes those structures, components
and systems that are important to reactor operation, but not essential
to safe shutdown and isolation of the reactor? and whose failure could
not result in the release of substantial amounts of radioactivity.
Seismic Category III includes the balance of structures, systems and
components. Seismic Category I items have been designed to withstand the
Safe Shutdown Earthquake (SSE) without loss of function.
3-2
We find these classifications to be acceptable and have concluded
that the applicant placed the safety related structures, systems and
components in their appropriate category, seismic Category I.
3.3 Wind and Tornado Design Criteria
The design wind velocity for the seismic Category I structures is
110 mph at 30 feet above ground based on a recurrence interval of 100
years. The design tornado for such structures is a 300 mph maximum
tangential velocity and a 60 mph translational velocity wind accompanied
by a simultaneous atmospheric pressure drop of 3 psi in 3 seconds. All
seismic Category I components and equipment are protected by being
housed in wind and tornado resistant structures, or are provided with
tornado missile shields.
ASCE Paper No. 3269 was utilized to determine the loads resulting
from these wind and tornado effects. The load factor associated with
wind load is 1.25 against the required ultimate capacity for concrete.
For tornado loads concrete structures have been designed for a load
factor of 1.0. Steel structures have been designed in accordance with
the .American Institute of Steel Construction (AISC) specifications (1963
Edition).
We find that the wind and tornado design criteria, as discussed
above, are conservative and provide reasonable assurance that, in the
event of winds or tornadoes of design level intensity, the structural
integrity and safety function of seismic Category I structures will
3-3
not be impaired. Conformance with these criteria is an acceptable basis
for satisfying the requirements of GDC No. 2.
The seismic Category I structures are arranged or protected such
that wind or tornado damage to structures will not affect the structural
integrity of seismic Category I structures, systems, or components. The
criteria in the design arrangement and the means employed for protection
of seismic Category I structures and other structures comply with the
provisions of GDC Nos. 2 and 4 as related to structures and are acceptable
3.4 Water Level (Flood) Design Criteria
The facility's major structures and buildings are located or
constructed so as to be undisturbed by the maximum water level result
ing from the Probable Maximum Hurricane (PMH). The design hydraulic
force on these structures included both the static and dynamic effects
from the PMH.
The use of these design loading criteria provides reasonable
assurance that, in the event of flooding, the seismic Category I
structures will maintain the required structural integrity and safety
functions. Conformance with these criteria satisfies the requirements
of GDC Nos. 2 and 4 as related to the flood and hurricane design basis
for structures.
3.5 Missile Protection Criteria
The design basis tornado generated missiles include a spectrum of
possible items that could be dislodged during tornadic winds and become
missiles.
~\ 3-4
The selected missiles are a wood plank, a wooden utility pole, a
schedule 40 pipe, and an automobile. In determining missile effects on
structures the applicant has used the NDRC (National Defense Research
Council) formula with modifications suggested in U. S. Army Technical
Manual TM 5-1300 for penetration resistance. Potential interior missiles
are generally controlled by reinforced concrete and steel barriers and
missile shields which are provided with a 25% margin in energy
absorption capacity.
The criteria used in the design of seismic Category I structures
to account for the loadings due to specified missile impacts postulated I
to occur at the site provide conservative design forces such that
missile impacts will not penetrate structures, shields, or barriers
beyond acceptable limits as governed by the strength and resistance
offered by such structures, shields and barriers.
We find that the missile protection design provides reasonable
assurance that, in the event of the generation of the postulated
missiles, resulting loads and effects will neither impair the structural
integrity of seismic Category I structures, nor result in loss of
required functions of safety related systems and components protected
by such structures. Conformance with these design loading cr~teria
is an acceptable basis for satisfying the GDC Nos. 2 and 4.
3-5
3.6 Protection Against Dynamic Effects Associated.with the Postulated
Rupture of Piping
30601 Criteria for Protection Against Dynamic Effects Associated with a
Loss-of-Coolant Accident (LOCA)
The design criteria used for determining the LOCA break locations
and break orientations for the reactor coolant pressure boundary (RCPB)
are acceptable to the staff. Both longitudinal and circumferential pipe
breaks were assumed to occur at any location. Generally, redundancy and
physical separation are employed to assure that .. a single incident will
·not prevent safety-related systems and components from performing their
required safety functions. Where ,physical s.eparation. cannot be
. achieved, shields and restraints are employed to prevent loss _of
required function •.
The design of piping restraints as applied to the RCPB provides
adequate protection of the containment structure, the unaffected reactor
coolant system.components, and systems important to safety that are
either interconnected with the reactor coolant system or are in close
proximity to the RCPB in which postulated pipe failures are assumed to
occur as a design basis LOCA.
The design provisions are such that the combined loadings imposed
by an SSE and a concurrent single break of the largest pipe at one of
the design basis break locations will be within the design capability
of the piping or.its restraints such that multiple failure of piping is
3.6.2
3-6
precluded and the emergency core cooling systems can perform their
required function.
Postulated Breaks Outside Containment
The criteria employed by the applicant to analyse the effects of
high energy pipe breaks outside containment are cons.istent with the
staff position transmitted to the applicant by letter dated December 22,
1972, and are acceptable. The applicant's analysis of high energy
piping breaks outside containment are presented in the report, "Effects
of High Energy Piping System Breaks Outside Reactor Building," October
1973, revised November 1973.
The protection provided against the dynamic effects of postulated
pipe breaks and discharging fluids in piping systems containing high
energy fluids and located outside the containment is adequate to
prevent damage to structures, systems and components to the extent
considered necessary to assure the maintenance of their structural
integrity. Such protection, includes enlargement of the auxiliary
and intermediate building vent areas, rerouting of pipe, installation
of restraints, barriers and jet shields for protection of piping,
electrical penetrations, cable, trays, pumps and valves. There is reason
able assurance that the safety shutdown of the reactor can be accomplished
and maintained, as needed.
The criteria used for the identification, design and analysis of
piping systems where postulated breaks may occur are consistent with
the staff position, meet the applicable requirements of GDC Nos. 1, 2, 4,
14, 15, 31 and 32 and are acceptable.
3.7
3.7.1
3.7.2
Seismic Design
Seismic Input
3-7
The seismic design response spectra indicate amplification
factors of 2.7 at a period of 0.8 seconds, of 2.0 at a period of
0.17 seconds and of greater than 1 for periods ranging from 0.03 to
0.17 seconds with 2% damping. The structure and equipment damping are
in accordance with the damping .factors which have been accepted for
all recently licensed plants including Three Mile Island Unit 1. The
modified time history used for component equipment design is adjusted
in amplitude and frequency to envelope the response spectra specified
for the site.
We conclude that the seismic input criteria used by the applicant
provides an acceptable basis for seismic design.
Seismic Analysis
Modal response spectrum and time history methods for multi-degree
of-freedom systems form the bases for analyses of all major seismic
Category I structures, systems and components. Governing response
parameters are combined by the square root of the sum of squares to
obtain maxima when the modal response spectrum method is used. The
absolute sum of responses has been used for in-phase closely spaced
frequencies.
Two components of seismic motion are considered: one horizontal
and one vertical. The total response is obtained by the absolute sum
of the responses to the two components.
Floor svectra inputs used for design and test verification of
structures, systems and components were generated from the time history
method. Dynamic analysis of vertical seismic systems has been
3.7.3
3-8
employed for all structures, systems and components where structural
amplifications in the vertical direction are significant. System and
subsystem analyses have been performed on an elastic basis. Effects on
floor response spectra of expected variations of structural properties
and damping have been accounted for by widening the response spectra
peak9 by ±10%.
We conclude that the dynamic methods and procedures for seismic
systems used by the applicant provide an acceptable basis for seismic
design.
Seismic Instrumentation Program
The type, number, location and utilization of strong motion
accelerographs to record seismic events and to provide data on the
frequency, amplitude and phase relationship of the seismic response
of the containment structure correspond to the recorrnnendations of
Regulatory Guide 1.12, "Ins trumen ta tion for Earthquake" dated March 10,
1971.
Supporting instrumentation installed on seismic Category I
structures, systems and components will provide data for verification
of the seismic responses determined analytically for such seismic
Category I items.
We conclude that the applicant's Seismic Instrumentation Program
is acceptable.
3.8
3.8.1
3.8.2
3-9
Design of Category I (Seismic) Structures _
_ Foundations
(Refer to Section 2.5 of this report.)
Seismic Category I 'Structures
,The important seismi~ Category I structures other than containment
(discussed in Section 3.8.3 of this report) are the auxiliary building,
the control complex, the diesel generator b:uildir,i-g, the intermediate
buildin~ and the intake structure._. These structures were de_signed to
the same criteria that were utilized for the containment s_tructure
except that a strict application of the ACI 318-63 ultimate strength
design with the Code specified load_factors was used and a portion of
the steel superstructure.of_ the Auxiliary Building was· not designed
against tornado missiles. However, the spent fuel pool is protected by
a tornado missile shield.
-In response to re_quest_s by the Regulatory staff, iacluding staff
positions transmitted to the applicant_ by lett_;er of December 22, 1972,
the high energy pipe breaks hypothesized outside containment and
the related interactions with structures have been addresse_d by the
applicant in a report entitled, "Effects of High Eaergy Piping System
Breaks Outside the Reactor Building," dated October 1, 1973 a:id amended
November 6, 1973. The applicant has reviewed its design to accommodate
these high energy pipe breaks. The revisions made are consistent with
the above mentionP-d Regulatory staff positions. On the basis of our
3-10
review of the revised design, we have concluded that the effects associ
ated with the high energy breaks outside containment now can be adequately
resisted by the structures.
The seismic Category I structur.es were built from a composite of
structural steel and reinforced concrete members. In general, the
structures were designed as continuous systems with slabs, walls,
beams and columns being integrated into the design. The design methods
for reinforced concrete followed the .ultimate strength design provisions
of ACI-318. For structural steel design the AISC Specification was
utilized.
The loading combinations used for the design of these structures
included normal dead and live loads, wind and tornado loads, and
earthquake loads.
The analyses were based on elastic analysis procedures with
the design executed using the ultimate strength design provision of
ACI-318 for concrete and the working stress design provisions of the
AISC Code for structural steel.
Construction practice for the seismic Category I structures was
accomplished.in accordance with ACI-301 appropriately modified to
account for the specialized nature of the construct~on.
It is concluded that the criteria used in the analysis and design
of seismic Category I structures, to account for the loadings and
conditions that are anticipated to be experienced by the structures
3.8.3
3-11
during the service life time, are in compliance with acceptable codes,
standards, antl specifications.
The use of these dP-sign criteria defining the applicable codes,
standards and specifications; the load and loading combinations; the
design and analysis procedures; the structural acceptance criteria;
the materials, quality control and special construction techniques;
and the testing_ and inservice surveillance requirements, provide
reasonable assurance that, in the event of winds, tornados, earthquakes
and various postulated accidents, these seismic Category I strµctures
will withstand the specified conditions without impairment of their
structural integrity and required safety functions. Conformance with
these criteria satisfy the requirements of GDC Nos. 2 and 4.
Containment
The containment is a soil supported prestressed concrete structure
in the form of a right vertical cylinder with a shallow dome and a
conventionally reinforced concrete flat slab base. The inside surface
of the containment is steel lined in order to form a leak tight membrane.
The containment design was based on the concepts of ACI 318-63
using the working stress design procedures for the loading combinations
representing the construction conditions and the normal operating
conditions. Under the various accident conditions including earth
quakes, wind and tornado the des_ign criteria were based on the ultimate
strength design procedures using load factors. The design criteria used
3-12
included the load combinations, stress allowables and analytical pro-
cedures that are consistent with those used on other similar prestressed
concrete containments previously licensed such as Three Mile Island
Unit Nq. 1, Palisades, Point Beach and Turkey Point.
The loads considered in the containment design include appropriate
combinations of dead and live loads, thermal loads, loss-of-coolant ' .
accident induced loads and severe environmental loads such as earthquake
loads, and wind and tornado loads. A test pressure load of 1.15 times
the design accident pressure is also included.
The static analysis for the containment shell was based on
classical thin shell theory. -The allowable stress and strain limits
were those defined in ACI 318-63 and as provided for in the FSAR. For
the loading combinations cited previously, reinforcing bar yield was
the mC?st significant limit. For specific critical areas such as the
equipment hatch area there were additional detailed studies completed
by the applicant. In general, finite element techniques were used in
those situations.
Interior Structure
The interior structures of the containment have been designed for
the same general conditions considered for the containment shell with,
of course, differences in magnitude. The primary shield wall was
designed for a differential pressure of 170 psi; the secondary shield
wall was designed for a differential pressure of 15 psi with capability
3-13'
to 17.5 psi taking the reinforcing steel to yield. The secondary
shield wall was designed in accordance.with. ACI 318~71 which was ~· I,
consistent with using the latest available codes at the time of its
design. (Also see Section 6.2.1 of this report.)
Yhe construction was carried out_ using ACI 301-6~, Specifications
for Structural Concrete Buildings, with ·the modifications enumerated in
the FSAR. Applicable sections of the ASME Boiler and Pressure Vessel
Code, Section III and Section IX were used in conjunct.ion with the
constr.uction and desi~ of the steel liner and penetrations.
Containment Testing
The testing of the containment will be as prescribed in a report
entitled,. "Preliminary Report on Structural Integrity Testing of
Reactor Containment Structure," by Gilbert Asso_ciates, Incorporated,
·dated January 12, 1970. Strain measurement instrumentation consists
of 70 instrumented and embedded steel reinforcing bars and rosettes
on the liner plate at· six general lo.cations with addii::ional rosettes
around three typical penetrations. Displacements will be measured
u.sing jig transits, precisions levels, invar tapes and linear variable
displacement transdueers. Four visual monitoring locations for
cracking are defined to·talling 1230 square feet ofo surface area which
will be closely monitored for cracking. We find this to be an
acceptable containment structural test program.
The tendon surveillance program proposed by the applicant follows
the provisions of Regulatory Guide 1.35, "Inservice Surveillance of
3.9
3.9.l
3-14
Ungrouted Tendons in Prestressed Concrete Structures," dated
February 5, 1973. The Technical Specifications reflect this program
also_. Consequently we find the tendon surveillance program to be ac
ceptable.
The use of these design criteria defining the .applicable codes,
standards and specifications; the materials, quality control and
special construction techniques; and the testing, provide reasonable
assurance that, in the event of winds, tornadoes, earthq.uakes and
various postulated accidents occurring within the containment, the
seismic Category I containment and its internal structures will with
stand the specified conditions without impairment of structural integrity
or required safety functions. Conformance with these criteria constitutes
an acceptabl~ basis for satisfying the requirements of AEC GDC Nos. 2, 4,
16 and 50. We ·conclude that the design of containment and its internal
structures is acceptable.
Mechanical Systems and Components
Dynamic System Analysis an~ Testing
The applicant has designated Oconee 1 as the prototype plant
from which preoperational vibration test results are applicable in
evaluating the design adequacy of the CR~3 reactor internal structures.
This designation is acceptable to the Regulatory staff. Thus, only
confirmatory tests in accordance with Regulatory Guide i.20, "Vibration
Measurements on Reactor Internals" dated December 29, 1971, will be
. ,·.·.
3-15
conducted. This program of confirmatory preoperational vibration testing
of reactor internals is acceptable to the Regulatory staff.
The CR-3 reactor internals are designed to withstand the
dynamic effects of a LOCA due to pipe rupture near the reactor vessel
nozzle during an SSE. The design analyses are described in B&W Topical
Reports (1) BAW-10008-1-Rev. 1, "Reactor Internals Stress and
Deflection due to Loss-of-Coolant Accident-and Maximum Hypothetical
Earthquakes," (2) BAW-10035, "Fuel Assembly Stress and Deflection
Analysis for Loss-of-Coolant Accident and Seismic Excitation." We
evaluated these analyses during our review of the Oconee 1 applica-
tion. We find these topical reports are also applicable to CR-3.
A series of preoperational functional tests will be performed on
piping systems both inside and outside the RCPB, in accordance with
Paragraph I-701.5.4 of ANSI B31.7 Nuclear Power Piping Code. This code
requires that piping be arranged and supported to minimize vibration
and that the designer make appropriate observations under startup or
initial operating conditions to assure that vibration is within accept-
able levels. These tests are to verify that the piping and piping
restraints have been designed to withstand dynamic effects due to
valve closures, pmnp trips, and·operating modes associated with the
design operational transients. The applicant has submitted an acceptance
criterion for these tests. During the turbine main steam stop valve
3.9.2
3-16
closure tests .an,d relief valve opening tests,. ins.trumentation will be
installed to provide test data for comparison with .these .acceptance
criteria· to insure that displacements are within ·allowable limits. If
the acceptance criteria should be exceeded, the applicant has agreed
to take. the necessary steps to bring displacements within a_cceptable
limits. We. find this program to be acceptable and will require that
all such tests and analyses have been satisfactorily completed prior to
issuance of an operating license.
The applicant has stated that he has or will conduct either tests
or analysis.for each item of seismic Category I mechanical equipment
to assure functional capability of that equipment during a seisraic
event. We find this criterion to be acc.eptable and' will review the re
sults prior to issuance of an operating license.
Structural. Integrity of Pressure Retaining Components
Pressure retaining components in fluid systems designat~d seismic
Category I which .are within the boundaries of AECSystem Quality
Group Classifications A, B or C are desigrred to the requirements
of the codes and standards specified in 10 CFR 50.55a or Regulatory
Guide 1.26, "Quality Group Classification and Standards," dated March 23,
1972~ as appropriate. All components are designed to sustain
normal operating loads, anticipated operational occurrences and the
operational basis earthquake _(1/2 SSE) with~n the stress limits of the
code specified. In addition, Quality Group A components are designed
for a limiting primary stress of two-thirds of ultimate strength for
:.J
3-17
the combination of design loads plus SSE and pipe rupture loading.
Quality Group B and C components are designed to sustain the SSE within
stress limits comparable to those associated with the emergency
operating condition of current component codes.
The specified design basis combination of loading as applied to
the design of the safety-related ASME Code Class 1, 2 and 3 pressure-
retaining components in systems classified as seismic Category I
provide reasonable assurance that in the event (a) of postulated.
seismic occurrences, or (b) an upset, emergency, or faulted plant
operation, the resulting combined stresses imposed on the system
components will not exceed the allowable design stress and strain
limits for the materials of construction. Limiting the stresses under
such loading combinations provides an acceptably conservative basis ..
for the design of the system components to withstand the most
adverse combination of loading events without gross loss of structural
integrity. The design load combinations and associated stress and
deformation limits specified for ASME Code Class 1, 2 and 3 components
constitute an acceptable basis for design in satisfying GDC Nos. 1, 2
and 4.
The criteria used for the design and mounting of the safety and
relief valves of .ASME Class 1 and 2 systems provide adequate assurance
that, under discharging conditions, the resulting stresses will not
exceed the allowable design stress and strain limits for the materials
3.9.3
3-18
of construction. Limiting the stresses under the loading combinations
associated with the actuation of these pressure relief devices pro
vides a conservative basis for the design of the system components to
withstand these loads without loss of structural integrity and impair
ment of the overpressure protection function.
The criteria used for the design and installation of overpressure
relief devices in ASME Class 1 and 2 systems meet the applicable
requirements of GDC Nos. 1, 2, 4, 14 and 15. On the basis of our review,
we conclude that the structural integrity pf these pressure retaining
components is acceptable.
Components not Covered by ASME Code
The design and tests performed for the fuel and control rod
assemblies and control rod drives are comparable to those of prior
designs which were found acceptable for Arkansas Nuclear One Unit 1 and
are acceptable for this facility. We find there is reasonable assurance
that the fuel and control rod assemblies and control rod drives will
withstand the imposed loads associated with normal reactor operation,
anticipated operational transients postulated accidents, and seismic
events without gross loss of their structural integrity or impairment
of function. We conclude that the design of the fuel, control rod
assemblies, and control rod drive meet the requ~rements of GDC Nos. 2
and 14.
3-19
3.10 Seismic Qualification of Seismic Category I Instrumentation and
Electrical Equipment
The reactor protection system, engineering safety feature circuits
and the emergency power system were designed to meet seismic Category
I requirements. A seismic qualification program was conducted and
confirmed that all seismic Category I instrumentation and electrical
equipment and supporting structures will function properly during an
SSE and during post-accident operation. The operability of the
instrumentation and electrical equipment was assured by testing. The
design adequacy of their supports was assured by either analysis or
testing.
We find the seismic design aspects of instrumentation and electrical
equipment to be acceptable.
4-1
4.0 REACTOR
4.1 Summary Description
4.2
4.2.1
The design of the B&W reactor for CR-3 is similar to the design
of other pressurized water reactors that we have recently approved
for operation, and is nearly identical to Duke Power Company's
Oconee-1 reactor. The core consists of 177 fuel assemblies having
208 fuel rods each; the design heat output of the core is 2452 MWt,
which is less than the design output of 2568 MWt for the Oconee core.
A unique feature of the B&W design is internal vent valves which
minimize steam binding in the event of loss-of-coolant accident
(LOCA). Full and part length control rods, dissolved boron, and
burnable poison rod assemblies (BPRA) are used for reactivity control.
Mechanical Design
Fuel
The reactor fuel elements, designed and fabricated by B&W,
will employ Zircaloy-clad fuel rods containing uranium dioxide
pellets. All fuel rods are pre-pressurized with helium gas and are
similar to those approved for use in Oconee-1 except for the density
of the fuel pellets. The Oconee-1 fuel prior to operation was 93.5%
of theoretical density (TD), whereas the fuel for the first cycle of
CR-3 is 92.5%T TD.
Fuel elements designed and fabricated by another manufacturer
and used in other power plants have experienced physical changes
r
4-2
(due to fuel densification) that could affect core operating condi-
tions. The conditions of operation for these facilities have been
restricted where necessary to maintain acceptable safety margins.
The staff requires that densification of uranium dioxide fuel
pellets be assumed to occur during irradiation in power reactors. The
initial density of the fuel pellets and the size, shape, and distribu-
tion of pores within the fuel pellet influence the densif ication
phenomenon. The effects of densification on the fuel rod will
increase the stored energy, increase the linear thermal output, increase
"the probability for local power spikes, and decrease the thermal
conductance.
The primary effects of densification on the fuel rod mechanical
design are manifested in calculations of time-to-collapse of the
cladding and fuel-cladding gap conductance. Time-to-collapse calcula-
tions predict the time required for unsupported cladding to become
dimensionally unstable and to flatten into an axial gap caused by
fuel pellet densification. Gap conductance calculations predict the
decrease in thermal conductance due to opening of the fuel-clad
radial gap.
Babcock & Wilcox T@pical Report BAW-10054 entitled, "Fuel Densi-
fication Report, May, 1973," is applicable to all B&W reactors
beginning with Oconee Unit 1 and includes CR-3. The staff's review
and acceptance with modifications of the B&W fuel densification model
4-3
was presented in its report "Technical Report on Densification of
Babcock & Wilcox Reactor Fuels," dated July 6, 1973. This model also
applies to CR-3 and we conclude that it is acceptable.
4.2.2 Reactor Vessel Internals
For normal design loads of mechanical, hydraulic and thermal
origin, including anticipated plant transients and the operational
basis earthquake, the reactor internals have been designed to the stress
limit criteria of Article 4 of the ASME Boiler and Pressure Vessel
Code Section III.
For the loads calculated to result from the LOCA, the SSE and the
combination of these postulated events the reactor internal components
were designed as described in B&W Topical Report BAW-10008, "Reactor
Internals Stress and Deflection Due to a LOCA and Maximum Hypothetical
Earthquake" dated June 1970. These criteria are consistent with
comparable code emergency and faulted operating condition category
limits and the criteria which have been accepted for all recently
licensed plants including CR-3. We find these criteria acceptable.
We find the mechanical design of the reactor internals to be acceptable.
We have also reviewed the selection of materials for the reactor
vessel internals. All materials are compatible with the reactor
coolant, and have performed satisfactorily in similar applications
including the Oconee reactors. Undue susceptibility to intergranular
stress corrosion cracking has been prevented by avoiding the use of
4.3
4.3.1
4-4
sensitized stainless steel according to the methods recommended in
Regulatory Guide 1. 44, "Control of the Use of Sensitized Stainless
Steel," dated May 1973.
The use of materials proven to be satisfactory by actual service
experience and avoidance of sensitization by the methods recommended
in Regulatory Guide 1.44, provides reasonable assurance that the
reactor vessel internals will not be susceptible to failure by
corrosion or stress corrosion cracking.
The applicant has described. the measures that were taken to
ensure that deleterious hot cracking of austenitic steel welds was
prevented. All weld filler metal was of selected composition, and
welding processes were controlled. to produce welds with adequate
delta ferrite, in conformance with the recommendation in Regulatory
Guide 1.31, "Control of Stainless Steel Welding," dated June 1973.
Following these recommendations provides reasonable assurance that no
deleterious hot cracking will be present that could contribute to loss
of integrity or loss of functional capability.
Nuclear Design
Nuclear Analysis
Our review of the nuclear design of the CR-3 reactor was based on
the information provided by the applicant in the FSAR and revisions
thereto, discussions with the applicant and B&W, and the results of
independent calculations performed for us by the Brookhaven National
Laboratory.
4.3.2
4-5
The applicant has described the computer programs and calcula
tional techniques used by B&W to predict the nuclear characteristics
of the reactor design, and has provided examples to demonstrate the
ability of these methods to predict the results of critical experiments
using uo2
and Puo2-uo
2 fuel.
The applicant has also performed analyses, using a two-dimensional
PDQ computer program in conjunction with fuel cycle calculations
obtained with the use of the HARMONY computer program, to provide
estimates of core fuel burnups and first and second cycle and equili
brium core enrichments.
We have concluded that the information presented adequately demon
strates the ability of these analyses to predict reactivity and the
physics characteristics of the reactors.
Power Distribution
Detailed three-dimensional power distribution measurements have
been performed at the B&W Critical Experiments Laboratory. The results
of the applicant's calculations using PDQ07, a three-dimensional computer
program, agree quite well with the measured power distribution. PDQ07
as used by B&W incorporates a thermal feedback in obtaining radial and
axial power distributions for operations involving (1) changes in
control rod positions, (2) various xenon stability and control
conditions, and (3) various reactivity coefficients.
The axial distribution of power was calculated for two conditions
of reactor operation. The first condition is an inlet peak resulting
4.3.3
4-6
from partial insertion of a Control Rod Assembly (CRA) group. This
condition results in the maximum local heat flux and maximum linear
heat rate. The second power shape is a synnnetrical cosine which is
indicative of the power distribution with xenon override rods (part
length rods) withdrawn. Both of these flux shapes have been evaluated
for thermal departure from nucleate boiling (DNB) limitations by the '
applicant. The limiting condition was found to be the cosine power
distribution (peak to average power ratio, P/-:- = 1.5) although the ·p
inlet peak shape has the larger maximum value (P/-:p = 1.7). However,
the position of the cosine peak farther up the channel results in a
less favorable flux to enthalpy relationship and, therefore, the
cosine axial shape has been used by the applicant to determine individual
channel DNB limits.
We have concluded that the analytical methods used to calculate
power distribution are adequate and that core thermal limits are
conservatively based on the most restrictive power peaking factors.
Moderator Temperature Coefficient
The moderator temperature coefficient is slightly positive at
the beginning of the initial fuel cycle due to the use of soluble boron
for reactivity control. Since the moderator temperature coefficient
at temperatures less than 525°F will be less negative (or more positive than
at operating temperatures, the applicant has stated that startup and
4-7
operation of the reactor when the reactor coolant temperature is less
than 525°F will be prohibited except where necessary for low power
physics tests, when special operating precautions will be taken.
The accident analyses, except for the calculation of clad temperature
for the LOCA, uses a maximum positive moderator temperature coefficient
of 0.5 x 10-4 ~K/K°F. The LOCA analysis was performed with a zero
moderator coefficient. The Technical Specifications will, therefore,
prohibit operation above 95% of power unless the moderator temperature
coefficient is zero or less .
. 3.4 Control Requirements
To allow for the typical changes of reactivity due to reactor
heatup, operating conditions, fuel burnup and fission product buildup,
a significant amount of controllable excess reactivity is designed into
the core. The applicant has provided substantial information relating
to core reactivity balances for first and equilibrium cycles for
beginning-of-life (BOL) and end-of-life (EOL) and has shown that neutron
adsorption means have been provided to control excess reactivity at all
times. This is done through control of the concentration of soluble
boron in the reactor coolant and movement of control rods. Fuel burnup
and fission product buildup are partially controlled by fixed B4c
4.3.S
4-8
burnable poison rod assemblies (BPRA) fot the longer first fuel cycle.
These assemblies are used rather than increased concentrations of
soluble boron to prevent the BOL moderator temperature coefficient from
becoming more positive. The applicant has conservatively shown that
the core can be maintained in a subcritical condition by at least 1%
~k/k with operating boron concentrations even with the highest worth
CRA withdrawn. In addition, under conditions where a cooldown to reactor
building ambient temperature is required, concentrated soluble boron
can be added to the reactor coo~ant to produce a shutdown margin of
at least 1% ~k/k with all the control rod assemblies withdrawn from the
core.
On the basis of our review, we have concluded that the applicant's
assessment of reactivity control requirements over the core lifetime
is suitably conservative, and that adequate negative worth has been
provided by the control rods, the soluble boron system, and the
burnable poison rod assemblies to assure shutdown capability for all
conditions.
Stability
The basic instrumentation for monitoring the nuclear power (neutron
flux) level and distribution in the CR-3 core is the same in principle
as for all PWR plants recently licensed for operation including Oconee-1.
Primary reliance is placed on four axially split, out-of-core neutron
detectors that are spaced approximately 90° apart around the reactor
L
4-9
pressure vessel. Also, 52 assemblies of self-powered neutron detectors
are available for in-core mapping. Each in-core assembly can measure
local neutron flux at seven elevations in the core. Normally, the
output of these detectors will be read out through the plant computer;
however, a backup readout system is provided. The applicant has pro
vided for availability of these detectors for monthly calibration of
the out-of-core detector tilt factor. Test results showing that
these in-core detectors have a rated lifetime in excess of 5 years and
a precision of + 5% in determining relative power distribution are
presented in B&W Topical Report 10001 "Incore Instrumentation Test
Program" (August 1969).
We have concluded that the out-of-core detectors are adequate for.
detecting power maldistributions originating from axial xenon instability
and misplaced control rods if the power distribution mapping capability
provided by the in-core detectors is utilized to calibrate the out-of
core detectors pe~iodically and to investigate any power distribution
anomalies detected by the out-of-core detectors.
We have reviewed the applicant's analyses of xenon-induced
oscillations which have been reported in three B&W Topical Reports,
BAW-10010 Part 1 "Stability Margin for Xenon Oscillations Model
Analysis" (August 1969), BAW-10010 Part 2 "Stability Margin for Xenon
Oscillations - One Dimensional Digital Analysis" (February 1970), and
BAW-10010 Part 3 "Stabilit;y Margin for Xenon Oscillations - Two and
- -----r"I
4-10
Three Dimensional Analysis" (April 1970). Those analyses indicated
that, while azimuthal and radial xenon oscillations will not be
divergent, axial xenon oscillations could be divergent at the beginning
of the fuel cycle. The analyses further indicated that axial xenon
oscillations (which are slow changes taking place over several hours)
can be controlled by operator control of the position of the eight
part-length (axial power shaping) rods. In addition, the operator of
the prototype plant, Oconee-1, has performed tests during the initial
startup of that plant and confirmed the as-built stability of this
·core design against xenon-induced reactivity fluctuations.
As added assurance that power mald~stributions will not go un
detected should they occur, the Technical Specifications will (1)
require appropriate axial and radial power distribution monitoring
and control measures to be in effect, and (2) limit the BOL positive
moderator coefficient.
On the basis of our review and with the restrictions to be imposed
by the Technical Specifications we conclude that the nuclear design is
acceptable.
4.4 Thermal Hydraulic Design
With exceptions as stated in Table 4.1-1, the thermal hydraulic
design of CR-3 is identical to that of Oconee-1 which was reviewed
previously and found acceptable. However, since the applicant does
not propose to validate operation of the plant in a single loop
4-11
TABLE 4.1-1
COMPARISON OF THERMAL AND HYDRAULIC DEISIGNS OF
CR-3 AND OCONEE-1
Parameter
Thermal Power (MWt)
Nuclear Heat Flux Hot Channel Factor
2 Heat Transfer Surface (ft )
Average Heat Flux
Maximum Heat Flux
2 (Btu/hr/ft )
(Btu/hr/ft2
)
Average Specific Power at 100% Power (kW/ft)
Design Thermal Output, kW/ft
DNBR at Nominal Conditions
Minimum DNBR for Design Transients
CR-3
2,452
3.12
49,734
163, 725
510 ,296
5.51
16 .83
2.21
1.30
Oconee-1
2,568
3.12
49,734
171,470
534,440
5.656
11.63
2.00
1.30
4-12
. :-i
configuration (i.e., two pumps in one loop running while both pumps
in the other loop are idle), the Technical Specifications will prohibit
single loop operation.
On the basis of our review of the thermal-hydraulic charac~eristics
of CR-3 including comparison with the previously approved Oconee-1,
we conclude that the thermal-hydraulic design of CR-3 is,acceptable. "
5-1
5.0 REACTOR COOLANT SYSTEM
5.1 Summary Description ,I
CR-3 uses a B&W 2-loop nuclear steam suppiy system. In all
important aspects, it is the same as the Oconee-1 system. All
principal components of the system, the physical sizes, the materials
0£ construction, and the basic design codes used for CR-3 are the
same as for Oconee-1. Operating conditions of the systems vary slightly
since CR-3 will operate at a lower core power level of 2452 MWt
compared to 2568 MWt for Oconee-1. However, the design and operating
pressures are the same. On the basis of our evaluation of the CR-3
system and the similarity to the previously approved Oconee-1, we
conclude that the overall design of the reactor coolant system of
CR-3 is acceptable.
5.2 Integrity of Reactor Coolant Pressure Boundary (RCPB)
5.2.1 Fracture Toughness
Compliance with Code Requirements
We have reviewed the materials selection, toughness requirements,
and extent of materials testing accomplished by the applicant to
provide assurance that the ferritic materials used for pressure re-
taining components of the RCPB will have adeq~ate toughness under
test, normal operation, and transient conditions. All ferritic
materials, not including piping, were ordered and tested in accordance
with the requirements of .the ASME Boiler and Pressure Vessel Code,
5~
Section III (1965 Edition and with Addenda through Surnmerel967).-
Piping met the requirements as ASAS Standard B31.7, dated February 1968,
including the Errata dated June 1968. Dropweight NDT data were obtained
for the beltline shell plates of the reactor vessel. It is concluded
that the design of the reactor coolant pressure retaining components
have complied with applicable code requirement.
The fracture toughness tests and procedures required by Section
III of the ASME Code, augmented by the additional dropweight testing
for the reactor vessel, provide reasonable assurances that adequate
Eafety margins have been provided against the possibility of nonductile
behavior or rapidly propagating fracture of the pressure-retaining
components of the RCPB.
Operating Limitations
The reactor will be operated in accordance with Appendix G to Section
III of the ASME Boiler and Pressure Vessel Code, Summer 1972 Addenda,
and Appendix G, 10 CFR Part 50 which will minimize the possibility of
rapidly propagating failure. Additional conservatism exists in the
pressure-temperature limits to be used for heatup, cooldown, testing,
and core operation will be provided because these will be determined
assuming that the beltline region of the reactor vessel has already
been irradiated.
The use of Appendix G of the Code as a guide in establishing safe
operating limitations, using results of the fracture toughness tests
5-3
performed in accordance with the Code and. AEC Regulations, will ensure
adequate safety margins during operating, testing, maintenance, and
postulated accident conditions. Compliance with these Code provisions
and AEC regulations, constitute an acceptable basis for satisfying the
requirements of GDC No. 31.
Reactor Vessel Material Surveillance Program
The toughness properties of the reactor vessel beltline material
will be monitored throughout service life with a material surveillance
program that will comply with Appendix H, 10 CFR 50 (July 17, 1973).
The program is consistent with other surveillance programs that have
been found acceptable for other PWR plants including Arkansas Nuclear
One Unit 1. The copper content of the reactor vessel beltline has been
determined, but the number of capsules provided in the surveillance
program is conservatively based on assuming high values of sensitivity.
Changes in the fracture toughness of material in the reactor
vessel beltline caused by exposure to neutron radiation will be
assessed properly, and adequate safety margins against the possibility
of vessel failure will be provided since the essential material
surveillance requirements of Appendix H, 10 CFR Part 50, are met.
The surveillance program constitutes an acceptable basis for monitoring
radiation induced changes in the fracture toughness uf the reactor
vessel material, and satisfies the requirements of GDC No. 1.
5.2.2
5-4
Although the use of material of known moderate copper content for
the reactor vessel beltline will minimize the possibility that radiation
will cause serious degradation of the toughness properties, should
results of tests indicate that the toughness is not adequate, the reac
tor vessel could be annealed to restore the toughness to.acceptable
levels. We conclude that the reactor vessel material surveillance pro
gram is acceptable.
General Material Considerations
We have reviewed the materials of construction for the RCPB to
ensure that the possibility of serious corrosion or stress corrosion
cracking is minimized. All materials used are compatible with the
expected environment, as proven by extensive testing and satisfactory
service performance. The applicant has shown that the possibility of
intergranular stress corrosion in austenitic stainless steel used for
components of the RCPB was minimized because sensitization was avoided,
and adequate precautions were taken to prevent contamination during
manufacture, shipping, storage, and construction. The means used to
avoid sensitization are in general conformance with Regulatory Guide
1.44, "Control of the Use of Sensitized Stainless Steel," and include
controls on compositions, heat treatments, welding processes, and
cooling rates.
The use of materials with satisfactory service experience, and
the high degree of conformance with Regulatory Guide 1.44, "Control
5.2.3
5-5
of Sensitized Stainless Steel," provide reasonable assurance that
austenitic stainless steel components will be.compatible with the
expected service invironments, and the probability of loss of
structural integrity is minimized.
Water Chemistry Control
Further protection against corrosion problems will be provided
by control of the chemical environment. The reactor coolant chemistry
will be controlled; and the proposed maximum halogen contaminant
levels, as well as the proposed pH, hydrogen overpressure, and boric
acid concentrations, have been shown by tests and service experience
to be adequate to protect against corrosion and stress corrosion
problems.
We have evaluated the proposed requirements for the external
insulation used on austenitic stainless steel components and found that
chloride and silicate content will be adequately controlled.
The possibility that serious corrosion or stress corrosion
problems would occur in the unlikely event that ECCS or containment
spray system operation is required will be minimized because of
the pH of the circulating coolant will be maintained above 9.0
by hydroxide additions.
The secondary water chemistry will be adequately controlled to
prevent stress corrosion of the steam generator tubing, and the adequacy
of the compositional limits used is acceptable.
5.2.4
5-6
The controls on chemical composition that will be imposed on
the reactor coolant, secondary water, emergency core cooling water,
and the use of low chloride external thermal insulation, provide
reasonable assurance that the reactor coolant boundary materials will
be adequately protected from conditions that would lead to loss of
integrity from stress corrosion.
Control of Stainless Steel Welding
We have reviewed the controls used to prevent hot cracking
(fissuring) of austenitic steel welds. These precautions included
·control of weld metal composition and welding processes to ensure
adequate delta ferrite content in the weld metal. The methods used
comply with Section III of the ASME Code, and are in acceptable
conformance with Regulatory Guide 1.31, "Control of Stainless Steel
Welding," dated June 1973. We find there is reasonable assurance that
the austenitic stainless steel welds have been adequately controlled.
5.3 Reactor Vessel Integrity
We have reviewed all factors contributing to the structural
integrity of the reactor vessel and we conclude there are no special
considerations that make it necessary to consider potential vessel
failure for CR-3.
The bases for our conclusion are that the design, material, fabri
cation, inspection, and quality assurance requirements will conform
to the rules of the ASME Boiler and Pressure Vessel Code, Section III,
all addenda through Summer 1972, and all applicable Code Cases.
5-7
The fracture toughness requirements of the ASME Code, Section III,
1965 Edition, have been met. Also, operating limitations on tempera
ture and pressure will be established for this plant in accordance with
Appendix G, "Protection Against Non-:Ductile Failure," of the 1972
Summer Addenda of the ASME Boiler and Pressure Vesse~ Code, Section
III, and Appendix G, 10 CFR Part SO.
The integrity of the reactor vessel is assured because the vessel:
1. Has been designed and fabricated to the high standards of quality
required by the ASME Boiler and Pressure Vessel Code and pertinent
Code Cases listed above.
2. Has been made from materials of controlled and demonstrated high
quality.
3. Will be extensively inspected and tested to provide substantial
assurance that the vessel will not fail because of material or
fabrication deficiencies.
4. Will be operated under conditions and procedures and with protective
devices that provide assurance that the reactor vessel design
conditions will not be exceeded during normal reactor operation
or during most Jpsets in operation, and that the vessel will not
fail under the conditions of any of the postulated accidents.
5. Will be subjected to monitoring and periodic inspection to de
monstrate that the high initial quality of the reactor vessel
has not deteriorated significantly under the service conditions.
5~
6. May be annealed to restore the material toughness properties if
this becomes necessary.
5.4 Reactor Coolant Pressure Boundary (RCPB) Leakage Detection System
Coolant leakage within the containment may be an indication of
a small through-wall flaw in the RCPB.
The leakage detection system proposed for intersystem leakage
is by means of radioactivity monitors and flow and level monitors.
The system proposed to detect direct RCPB leakage to the containment will
include diverse leak detection methods, will have sufficient sensitivity
to measure small leaks, will identify the leakage source to the ex-
tent practical and will be provided with suitable control room alarms
and readouts. The major components of the system are the containment
airborne particulate and gas ·radioactivity monitors~ and containment
sump level and flow indication. Indirect indication of leakage can be
obtained from the containment humidity, pressure, and temperature
indicators.
The leakage detection systems will provide reasonable assurance
that any structural degradation resulting in leaking during service
will be detected in time to permit corrective action satisfying the
requirements of GDC No. 30 and is thus acceptable.
5.5 Inservice Inspection Program
To ensure that no deleterious defects develop during service, all
welds will be inspected periodically. The applicant has stated that
5-9
the design of the reactor coolant system incorporates provisions for
access for inservice inspections in accordance with Section XI of
the ASME Boiler and Pressure Vessel Code, and that methods will be
provided to facilitate the remote inspection of those areas of the
reactor vessel not readily accessible to inspection personnel. The
conduct of periodic inspections and hydrostatic testing of pressure
retaining components in the RCPB in accordance with the requirements of
ASME Section XI Code provides reasonable assurance that evidence of
structural degradation or loss of leaktight-integrity occurring during
service will be detected in time to permit corrective action before
the safety function of a component is compromised. Compliance with the
inservice inspections required by this Code constitutes an acceptable
basis for satisfying the requirements of GDC No. 32.-
5. 6 Plllllp Flywheel
The probability of a loss of pump flywheel integrity can be
minimized by the use of suitable material, adequate design, and
inservice inspection. (Also see Section 5.8.)
The applicant has stated that the integrity of the reactor coolant
pump flywheel is provided by having designed for a 125% overspeed
condition while the maximum anticipated overspeed is 110% o{ normal
speed. In the unlikely event of a 125% overspeed condition the maximum
primary design stress at the bore is approximately 70% of the yield
5-10
strength. The flywheel was purchased prior to the requirements of
Regulatory Guide 1.14 "Reactor Coolant Pump Flywheel Integrity" dated
October 27, 1971, which accepts a lower (67% of yield) strength at the
design overspeed condition. In addition, a 100% ultrasonic volumetric
inspection of the flywheel, using ASME Section III acceptance criteria,
was performed.
Inservice inspections of the flywheel will be performed in
accordance with the provisions of Regulatory Guide 1.14.
We conclude that the provisions for material selection and flywheel
design, and the use of a Regulatory Guide 1.14 inservice inspection
program ensure adequate flywheel integrity.
5.7 Loose Parts Monitor
Occasionally, miscellaneous items such as nuts, bolts, and other
small items have become loose parts within reactor coolant systems. In
addition to causing operational inconvenience, such loose parts can
damage other components within the system or be an indication of undue
wear or vibration. For such reasons, the staff has encouraged appli
cants over the past several years to support programs designed to
develop an effective, on-line loose parts monitoring system. For the
past few years we have required many applicants to initiate a program,
or to participate in an ongoing program, the objective of which was
the development of a functional, loose parts monitoring system within
a reasonable period of time. Recently, prototype loose parts monitoring
5-11
systems have been developed and are presently in operation or being
installed at several plants. The applicant has committed to install
a loose parts monitoring system in CR-3. We will confirm that this
is done prior to issuance of an operating license.
5. 8 Pump Overspeed
The staff is investigating, on a generic basis, the consequences
of an unlikely rupture of a reactor coolant pipe which in certain loca
tions might result in reactor coolant pump overspeed. If this study
indicates that additional protective measures are warranted under specific
circumstances in order to prevent significant pump overspeed or to limit
potential consequences to safety-related equipment, the staff will review
the circumstances applicable to the CR-3 facility to determine what
modifications, if any, are needed to assure that an acceptable level
of safety is maintained.
6-1
6. 0 ENGINEERED SAFETY FEATIJRES
6.1 General
6.2
6.2.1
1he Engineered Safety Features (ESF) consist of the reactor
building and its associated ventilation and isolation systems, the
contairunent cooling system, the containment spray system and the emergency
core cooling system (ECCS). 1he instruments and controls required for these
engineered safety features are discussed in Section 7.0 of this report and
the required electric power systems are discussed in Section 8.0.
Contairunent Systems
Contairunent Functional Design
1he contairunent structure (reactor building) is a free-standing
steel-lined, prestressed concrete structure with a net free volume of
3 2,000,000 ft . 1he structure houses the reactor coolant system
including the reactor, the pressurizer, reactor coolant pumps and
steam generators, as well as certain components of the plant's
engineered safety features. 1he contairunent structure is designed
for an internal pressure of 55 psig and a temperature of 281 °F.
1he applicant has described in the FSAR the results and methods
used to analyze the contairunent pressure response for a number of
designed basis LOG\1 s. 1he applicant has analyzed LOCA's involving a
spectrum of both hot leg and cold leg breaks, up to and including the
double ended rupture of the largest reactor coolant pipe to determine
the contairunent pressure responses. Minimum containment cooling was
6-2
assumed in the analysis of the reactor building i.e., one of the
three emergency building cooling units, and one of the two spray
trains of the Reactor Building Spray System were assumed to operate,
and the core reflood energy and steam generator stored energy were
included, as appropriate, in these analyses. As discussed below,
we reviewed the results of these analyses and verified by our
analyses that the calculational methods used by the applicant to
determine the containment pressure response from postulated loss
of-coolant accidents are conservative.
Mass and energy release rates were calculated using the CRAFT
computer code. These mass and energy addition rates were then used
as inputs in CONTEMPT which is the applicant's computer program to
calculate the containment pressure response.
The CRAFT computer code was used by the applicant to determine
the mass and energy addition rates to the containment for cold leg
breaks during the blowdown phase of the accident; i.e., the phase
of the accident during which most of the energy contained in the
reactor coolant system including the reactor coolant, metal and
core stored energy is released to the containment. The applicant
has, however, increased the energy release rate to the containment
by conservatively extending the time that the core would remain in
nucleate boiling; i.e., the time when the energy removal rate from
the core is highest. By using this method, the core would transfer
6-3
more heat to the containment for containment analysis than for
emergency core cooling analysis. Since this additional energy release
from the core will increase the containment pressure, the calculation
is conservative. The CRAFT computer code is acceptable to the AEC
for calculating energy release during a LOCA.
TI1e applicant has identified the 7.0 ft2 split break at the pump
suction as the cold leg break that results in the highest containment
pressure. The applicant calculates a peak pressure of 49 psig for
this break. The largest break (8.55 ft2) results in a peak calculated
pressure of 49 psig. We have analyzed the containment pressure
response for the 7.0 ft2 break in the suction leg of the reactor coolant
system using the CONTBvWT-LT computer code and included the energy
addition to the containment from the steam generators. We calculated
a peak containment pressure essentially the same as the applicant's.
To determine the mass and energy release to the containment, we used
the applicant's blowdown mass and energy release rates during the
reflood phase of the accident determined by our computer program
FLOODZ.
Blowdown mass ::m.d energy releases for hot leg breaks were also
calculated by the applicant using the CRAFT computer code. The
applicant's analysis indicates that a 14.1 ft2
break of the hot leg
results in the highest hot leg break containment pressure of 49 psig.
The applicant has also analyzed the containment pressure response
due to postulated failures of a main steam line. The applicant
6-4
conservatively assumed that the energy in a steam generator was
instantaneously released and did not take credit for the energy
removal capability of the structural heat sinks. The applicant
calculated a peak containment pr~ssure of 28 psig for this
accident.
We have evaluated the contairunent system in comparison to the
GDC stated in Appendix A to 10 CPR Part 50 of the Commission's
Regulations and, in particular, to Criteria Nos. 16 and 50. As a
result of our evaluation, we have concluded that the calculated
pressure and temperature conditions resulting from any design
basis LOCA will not exceed the design conditions of the containment
structure. The highest calculated accident containment pressure
and temperature were 50 psig and 280°F, respectively. The contain
ment design pressure of 55 psig provides a 10% margin above the peak
calculated accident pressure. We conclude that the maximum contain
ment accident pressure is correctly calculated to be below the design
pressure and that there is sufficient margin between the maximum
containment accident pressure and the design pressure of the con
tainment structure to assure that the health and safety of the public
is adequately protected.
Using the FLASH-2 program, the applicant has analyzed the pressure
response within the reactor vessel cavity, the primary shield pipe
penetration, and the steam generator compartments during loss-of
coolant accidents. 111e applicant calculated peak differential
6-5
pressures of 123.5 psi acting on the reactor cavity structural
elements, .1152.2 psi on the pipe penetration and 17.5 psi on a
steam generator compartment structural elements.
1he applicant's calculated pressures exceed the compartment
initial design pressures, i.e., the design pressure at the con-
struction permit stage of review, in several cases. For these cases,
the applicant has reanalyzed the as-built structural capability of
the compartments as discussed in Section 3.8.3 of this report and
have found them to be acceptable. 1he applicant's design pressures,
as-built capability, and calculated pressures are presented in
Table 6-1.
TABLE 6-1
SUBCOMPARTMENT DESIGN AND CALCULATED PRESSURES
Compartment Initial As-built Applicant's Design Differential Calculated LOCA Differential Pressure Differential Pressure Structural Pressure
Capability*
(psi) (psi) (psi)
Reactor Cavity 170 123,5
Pipe Penetration 1200 2000 1152 .2
South Steam Generator Compartment 15 17.5 17.5
North Steam Generator Compartment 15 17.5 16.8
*Based on Yield of Reinforcing Steel
6.2.2
6-6
We have perfoTI11ed pressure response calculations using the RELAP-3
program and compared our results to the applicant's. Our results
indicate reasonable agreement with the applicant's. We conclude that
the applicant's calculated design differential pressure for the sub
compartments are acceptable.
Containment Heat Removal Systems
The Reactor Building Spray System (RBSS) and the Reactor Building
Emergency Cooling System (RBCS) are provided to remove heat from the
containment following a LOCA. Any of the following combinations of
equipment will provide adequate heat removal capability:
(a) both spray trains of the RBSS,
(b) three fan-cooler units of the RBCS, and
(c) one spray train of the RBSS and one fan-cooler units of the RBCS.
The RBSS serves only as an engineered safety feature and perfoTI11s
no noTIIlal operation function. It is a seismic Category I system
consisting of redundant piping, valves, pumps and spray headers. All
active components of the RBSS are located outside the reactor building.
Missile protection is provided by direct shielding or physical
separation of equipment. TI1e reactor building sump screen assembly
is designed to prevent debris from entering the spray sys~em that
could clog the spray nozzles. NPSH requirements for the reactor
building spray pumps can be met during the post-accident recirculation
phase by throttling pump flow during recirculation from 1500 gpm to
6-7
1200 gpm. The RBSS includes a system for injecting sodium thiosulfate
into the spray water for iodine removal from the containment atmosphere
following an accident, and a system for injecting sodium hydroxide
into the spray water for pH adjustment. The sodium hydroxide will
raise the pH of the spray water into the alkaline range. Both
systems are designed to permit gravity draining of the solutions into
the spray pIBIIp suction piping.
A high reactor building pressure signal from the engineered
safety features actuation system will automatically place the RBSS
in operation. The spray pIBIIps will initially take suction from the
borated water storage tank. When the water in the tank reaches a low
level, the spray pIBIIp suction will be transferred manually to the
reactor building SIBIIp.
The Reactor Building Cooling System (RBCS) will be used during
both normal and accident conditions. Three equal capacity fan-cooler
units are provided. Each unit contains a moisture separator, a
cooling coil, and a ·two-speed fan. Under post-accident emergency
cooling conditions, the unit will operate at a reduced speed. Under
normal plant operatin~ conditions, water from the industrial cooler
will be circulated through the cooling coils. Under accident conditions,
following receipt of an engineered safety features actuation signal~
the high speed portion of the air handling units will be de-energized
and the slow speed portion of the air handling units will be energized.
For emergency cooling, heat will be rejected to the nuclear
services closed cycle cooling system.
6.2.3
6-8
The RBCS is a seismic Category I system. The housings for the
cooling lllits and the supply ducts are designed to withstand an
inward pressure differential of 2 psi. 'Ibe system was analyzed by
the applicant to determine that the duct sizes and outlet locations
are such that a 2 psi differential is not exceeded during.the transient
period. 'Ibe cooling lllits are located outside the secondary concrete
shield for missile protection. 'Ibe RECS equipment is accessible for
periodic testing and inspection during normal plant operation. We
have reviewed the containment heat removal systems for conformance to
the GDC Nos. 38, 39, and 40, and Regulatory Guide 1.1, "Net Positive
Suction Heat for Emergency Core Cooling and Heat Removal System Pumps"
dated November 2, 1970. We conclude that the systems meet the
requirements of these criteria and are acceptable.
Containment Isolation Systems
The Reactor Building Isolation System is designed to isolate the
containment atmosphere from the outside environment under accident con
ditions. Closed systems and isolation valves provide double barrier
protection so that no single, credible failure of malfunction of an
active component can result in loss of containment integrity. Reactor
building penetration piping and the associated isolation valves are
designed as seismic Category I equipment, and are protected against
missiles which could be generated under accident conditions.
0
6.2.4
6-9
Reactor building isolation will automatically occur on a signal
of high reactor building pressure (4 psig) or a high radiation signal.
All fluid penetrations not required for operation of the engineered
safety features equipment will be isolated. Remotely operated
isolation valves will have position indication in the control room.
We have reviewed the containment isolation system for conformance
to GDC Nos. 55, 56, and 57. We conclude that the system meets these
criteria and are acceptable.
Combustible Gas Control Systems
Following a loss of coolant accident (LOCA), hydrogen may
accumulate inside the reactor building. The major sources of hydrogen
generation would be: (1) from a metal water reaction between the fuel
cladding and the steam from.the LOCA, (2) from corrosion of aluminum
by the alkaline spray solution, and (3) from radiolytic decomposition
of water. To prevent generation of sufficient hydrogen to lead to
combustible mixtures, CR-3 has a containment purge system which is
designed to maintain the hydrogen concentration below its lower
flammability limit by introducing outside air into the containment
building and allowing the displaced containment atmosphere to be dis
charged through the purge exhaust filters to the plant vent. The
hydrogen purge system consists of a containment atmosphere monitoring
subsystem (hydrogen and radioactivity), a fresh air makeup subsystem
and a discharge subsystem.
r
6.2.5
6-10
TI1e reactor building cooling fan circulates the atmosphere
within the containment to provide mixing and prevent stratification
following a LOCA.
We have reviewed the system using the guidelines of the supple-
ment to Regulatory Guide 1.7, "Control of Combustible Gas Concentrations
Considerations" dated March 10, 1971. Our independent calculations of
the hydrogen concentrations are essentially in agreement with those of
the applicant. We conclude that the method of purging for control of
combustible gases is acceptable.
We have reviewed the combustible gas control systems for confor
mance to GDC Nos. 41, 42, and 43 and Regulatory Guide 1.7. We find
them in conformance with these criteria and conclude that the systems
are acceptable.
Containment Leakage Testing Program
Tile containment design 1ncludes the provisions and features planned
which satisfy the testing requirements of Appendix J, 10 CPR Part 50.
Tile design of the containment penetrations and isolation valves permits
individual periodic leakage rate testing at the pressure specified in
Appendix J. Included are those penetrations that have resilient seals
and expansion bellows, i.e., air locks, emergency hatches, refueling
tube blind flanges, l1ot process line penetrations, and electrical
penetrations.
Tile proposed reactor containment leakage testing program complies
with the requirements of Appendix J. Such compliance provides adequate
assurance that containment leaktight integrity can be verified
6.2.6
6-11
throughout service lifetime and provides for the leakage rates to be
checked on a timely basis to maintain leakage within the
Technical Specification limits.
Compliance with the requirements of Appendix J constitutes an
acceptable basis for satisfying the requirements of GDC Nos. 52, 53,
and 54, .Appendix A of 10 CFR Part 50. We find the containment leakage
test program to be acceptable.
General Material Considerations (Compatibility with Coolant)
We have reviewed the materials selection proposed fbr the contain
ment heat removal and ECCS systems, in conjl.Ulction with the expected
chemistry of the cooling and containment spray system water. The
applicant has shown that the use of sensitized stainless steel has
been avoided, and that the pH of the containment spray and the circulating
coolant will be controlled by sodium hydroxide additions. There are
test data verifying that the proposed chemistry will not cause stress
corrosion cracking of austenitic stainless steel l.Ulder conditions that
would be present during accident conditions.
We have concluded that the material controls provided and cooling
water chemistry proposed will provide assurance that the integrity of
components of these systems will not be impaired by corrosion or stress
corrosion. Welding of austenitic stainless steel for components of
these systems was controlled to prevent deleterious hot cracking. TIJe
control of we~d metal composition and welding procedures described by
the applicant conform with Regulatory Guide 1. 31, "Control of Stainless
6.3
6.3.1
6-12
Steel Welding," dated August 11, 1972, and provide assurance that loss
of function will not result from hot cracking of welds.
Emergen~ Core Cooling System (ECCS)
Desig_n Bases
The ECCS has been designed to provide emergency core cooling
during those postulated accident conditions where it is assumed that
mechanical failures occur in the reactor coolant system piping resulting
in loss of coolant from the reactor vessel greater than the available
coolant makeup capacity using normal operating equipment.
The applicant's design bases are to prevent fuel and cladding
damage that would interfere with adequate emergency core cooling and
to mitigate the amount of clad-water reaction for any size break up to
and including a double ended rupture of the largest reactor coolant
lines. These requirements will be met even with minimum engineered
safeguards available, such as would occur with the loss of one emergency
power bus together with the unavailability of offsite power.
The ECCS subsystems provided are of such diversity, reliability
and redundancy that no single failure of ECCS equipment, occurring during
a LOCA, will result in inadequate cooling of the reactor core. Each
of the ECCS subsystems are designed to function over a specific range of
reactor coolant piping system break sizes, up to and including the
flow area associated with a postulated double-ended break in the
largest reactor coolant pipe.
6.3.2
6-13
System Design
The ECCS consists of two core flooding tanks, two high pressure
injection and low pressure injection systems, with provisions for
recirculation of the borated coolant after the end of the injection
phase. Various combinations of these systems will assure core
cooling for the complete range of postulated break sizes.
Following a postulated LOCA, the ECCS will operate initially in
the passive core flooding tanl( injection mode and the active high
pressure injection mode, then in the active low pressure injection
mode, and subsequently in the recirculation mode.
The high pressure injection system (HPIS) mode of operation, upon
actuation of a safety injection signal, will consist of the operation
of two of three centrifugal charging pumps (rated at 500 gpm each at
a design head of 3000 ft) which provide high pressure injection of
boric acid solution, containing a minimum concentration of 2270 ppm
boron, into the reactor coolant system cold legs. Suction is taken
from the borated water storage tank (BWST) which has a nominal tank
capacity of 420,000 gallons.
The low pressure injection system (LPIS) consists of two decay
heat removal pumps (rated at 3000 gpm each at a design head of 350 ft)
which will take their suction from the borated water storage tank for
short term cooling. The low pressure lines terminate directly in the
reactor vessel through the core flooding nozzles located in the .vessel
wall. The·LPIS lines are equipped with a crossover line inside the
6-14
auxiliary building so that each LPIS plllIIp is connected to both
core flooding tank (CFT) nozzles on the reactor vessel. Manually
operated valves in the crossover line will be arranged so that in
the lllllikely event of the simultaneous occurrence of a break at the
worst location in a CFT line and the loss of one LPIS pIBIIp, half of the
flow of the other LPIS pIBIIp will reach the reactor pressure vessel to
insure adequate long tenn core cooling. One LPIS plllIIp is capable of
providing sufficient water for removing the heat energy generated
after a LOCA.
When a predetennined amol.lllt of water in the borated water storage
tank has been injected, or receipt of a low-level alann for the BWST,
suction will be transferred manually to the containment SlllIIp for the
recirculation mode of operation provided by the LPIS. The ECCS will
then provide the long-tenn core cooling requirements by recirculating
the spilled reactor coolant collected in the containment SlllIIp back
to the reactor vessel through the core flooding line nozzles. However,
should the reactor coolant system pressure be higher than the LP plllIIp
head, the required flow is delivered by the HPIS by aligning the flow
from the discharge of the LP pIBIIps to the suction of the HP pIBIIps.
The passive injection mode of operation is provided oy the core
flooding (CF) system, which protects the core in the event of inter
mediate and large-sized pipe breaks. The coolant is automatically
injected when the RCS pressure drops below the core flooding tank
pressure (600 psig). Each of the two core flooding tanks has a
6.3.3·
6-15
nonnal water volume of 940 ft3 with 410 ft3 of nitrogen gas at a
nonnal operating pressure of 600 psig: Each tank is connected by a
core flooding line directly to a reactor vessel core flooding nozzle.
The driving force for injection of the borated' water, containing 2270
ppm boron, is supplied by pressurized nitrogen. Each core flooding
line is equipped with an electric-motor-operated stop valve for·
isolation of the CFT during reduced reactor coolant pressure non
critical operation and two series inline check valves for isolation
of the CFT during normal reactor coolant pressure operation.
Performance Evaluation
The applicant has stated that the emergency core cooling systems
have been designed to deliver fluid to the reactor coolant system in
order to control the predicted cladding temperature transient following
a postulated pipe break and for removing decay heat in the long-
term, recirculation mode.
On June 29, 1971, the AEC issued an Interim Policy Statement con
taining Interim Acceptance Criteria for the performance of the ECCS
for light-water cooled nuclear power reactors. The Interim Policy
Statement includes a set of conservative assumptions. and procedures
to be used in conjunction with computer codes to analyze and evaluate
the ECCS performance for a pressurized water reactor.
I (
6-16
In accordance with the Interim Policy Statement (IPS), the per
formance of the ECCS is judged to be acceptable because the course of
the LOCA is limited as follows:
1. The calculated maximum fuel element cladding temperature does not
exceed 2300°F.
2. The amount of fuel element cladding that reacts chemically with
water or steam does not exceed one percent of the total amount of
cladding in the reactor.
3. The clad temperature transient is terminated at a time when the
core geometry is still amenable to cooling, and before the· cladding
is so ernbrittled as to fail during or after quenching.
4. The core temperature is reduced and decay heat is removed for an
extended period of time~ as required by the long-lived radioactivity
remaining in the core.
Th.e applicant presented an evaluation of the LOCA in accordance
with the requirements of the IPS in BAW-1397 dated August 1973. This
evaluation resulted in a peak clad temperature of 2299°F for a 17.9
kw/ft peak generation rate using the Interim Acceptance Criteria.
However, this analysis did not adequately consider moderator temperature
coefficient and temperature heat capacity effect? of uo2.
The applicant will submit a LOCA analysis considering these
effects and performed by an acceptable evaluation model under the ECCS
criteria published in the Federal Register on January 4, 1974, and
show that this facility is in compliance with the same criteria. Our
6.3.4
6-17
evaluation of this analysis, including the effects of fuel densifi
cation, .will be provided in a supplement to this report.
Tests and Inspections
1he applicant will demonstrate the operability of the ECCS by
subjecting all components to preoperational tests, periodic testing,
and in-service testing and inspections.
The preoperational tests fall into three categories. One of these
categories consists of system actuation tests to verify the operability
of all ECCS valves initiated by Engineered Safety Feature Actuation
Signal (ESFAS), the operability of all safeguard pump circuitry down
through the pump breaker control circuits, and the proper operation
of all valve interlocks.
Another category is the core flooding system tests. 1he objective
of this test is to check the core flooding system and injection line to
verify that the lines are free of obstructions and that the core flooding
line check valves and isolation valves operate correctly. 1he
applicant will perform a low pressure blowdown of each core flooding
tank to confirm the line is clear and check the operation of the check
valves.
Operational test of all the major pumps comprises the last
category of tests. These pumps consist of the makeup and high pressure
injection pumps, the low pressure and decay heat removal pumps~ and the
containment recirculation pumps. The applicant will use the results
6.3.5
6-18
of these tests to validate the hydraulic and mechanical performance
of these pumps delivering through the flow paths for emergency core
cooling. These pumps will operate under both miniflow (through test
lines) and' full flow (through the actual piping) conditions.
By measuring the flow in each pipe, the applicant will make the
adjustments necessary to assure that no one branch has an unacceptably
low or high resistance. The system will also be checked to assure
there is sufficient total line resistance to prevent excessive runout
of the pump. The system will be accepted only after demonstration of
flow delivery of all components within design requirements.
The applicant will perform routine periodic tests of the ECCS
components and all necessary support systems at power. Valves re
quired to operate after a LOCA will be operated through a complete
cycle, and pumps are operated individually in this test.
Conclusions
On the basis of our evaluation, we have concluded that the per
formance of the ECCS is in accordance with the Connnission's Interim
Acceptance Criteria. Our evaluation of the applicant's LOCA analysis
performed in accordance with the ECCS criteria published
in the Federal Register on January 4, 1974, will be submitted in a
supplement to this· report.
7-1
7.0 INSTRUMENTATION AND CONTROLS
7.1 General
The Connnission's General Design Criteria (GDC), IEEE Standards
including IEEE Criteria for Protection Systems for Nuclear Power
Generating Stations (IEEE Std. 279-1968), and applicable Regulatory
Guides for Power Reactors have been utilized as the bases for
evaluating the adequacy of the protection and control systems. Specific
documents employed in the review are listed in the bibliography of
this report.
The results of our review of the logic and electrical schematics
and site visit are reflected in this evaluation.
7. 2 Reactor Protection System (RPS)
The RPS is essentially the same as that approved for the Arkansas
Nuclear One, Unit 1 except for the absence of the Power/Reactor
Coolant (RC) pump trip function. Since no credit was taken for thls
function in the applicant's safety ;,=cnalysis, deletion of the RC pump
motor power/reactor coolant pump trip was acceptable to the staff ..
We have reviewed all design aspects of the RPS, including logic
schematics, testing. capabilities and control of bypasses, and con
cluded that this system is acceptable.
7. 3 Engineered Safety Feature (ESF) Sys terns
The ESF actuation system is essentially the same as that approved
for the Three Mile Island, Unit 1 nuclear facility. Our review
encompassed all aspects of the protection system that initiates and
7.3.1
7.3.2
7-2
controls the operation of the ESF systems and their vital auxiliary
supporting.systems, including logic schematics, testing capabilities
and control of bypasses. The following sections identify those
aspects of the design that were changed as a result of our review .
. Also, they discuss those design commitments made by the applicant
that must be satisfactorily implemented and reviewed before the ESF
systems are considered to be acceptable.
Core Flooding Tank Isolation Valves
The applicant has elected to open the breakers supplying power to
the core flooding tank motor-operated isolation valves to assure against
accidental closure of these valves during normal reactor operation.
Based on this mode of operation, our review of the valve position in
dication circuits for the core flooding tank isolation valves revealed
that design would not conform to our criteria with regard to providing
redundant and independent indication systems for each core flooding
tank isolation valve. The applicant has committed to modify the design
to conform with our criteria. We will review the design modifications
of the valve position indication circuits to confirm that the final
design is acceptable prior to issuance of an operating license. The
results of our review will be provided in a supplement to this report.
Steam Line Break Isolation (SLBI)
Our review of the proposed SLCI system revealed that the instru
mentation ~ontrol and electrical equipment were not designed in
accordance with the requirements of IEEE Std. 279-1968 and IEEE
7-3
Std. 308-1971. In addition, we have found that a steam line break,
coincident with a single failure of either a feedwater or steam
isolation valve (preventing valve closure by either automatic or manual
means) could result in the uncontrolled continued blowdown of the
steam generator(s).
We require that the capability of the SLBI system design meet the
requirements of IEEE Std. 279-1968 and IEEE Std. 308-1971. The
applicant is making design modifications to make this system acceptable.
We will review these design modifications to confirm that the
design is acceptable prior to issuance of an operating license. The
results of our review will be provided in a supplement to this report.
7.4 Systems Required for Safe Shutdown
We have reviewed the instrumentation, control and electrical systems
being provided for safe reactor shutdown and the design provisions to
place and keep the plant in a safe shutdown condition in the event
that access to the main control room is restricted or lost. We have
concluded that the designs conform to our criteria and are acceptable,
except for the design of the instrumentation, control and electrical
equipment pertaining to the Auxiliary Feedwater System (AFS).
Our evaluation of the proposed AFS indicated that the required
delivery of emergency feedwater to the steam generator(s) could be
inhibited by a number of single failures under normal shutdown and
steam line break conditions. In addition, it was found that the
7-4
instrumentation, control and electrical equipment of the AFS were not
designed in accordance with the requirements of IEEE Std. 270-1968
and IEEE Std. 308-1971. 1he AFS, required for safety, must meet the
single failure criterion and that capability of AFS design must be
demonstrated against t11e requirements of IEEE Std. 279-1968 and
IEEE Std. 308-1971. 1he applicant has agreed to amend the design to
meet the single failure criterion and to demonstrate the capability
of the design against the above stated standards. We will review the
design modifications to confinn that this design corrnnitment has been
satisfactorily implemented prior to issuance of an operating license.
TI1e results of our review will be provided in a supplement to this
report.
7.5 Safety Related Display Instrumentation
We have reviewed the design of the instrumentation systems that
provide infonnation needed by the operator to perform required
safety manual functions and post-accident surveillance. We con
cluded that this safety related display information is acceptable,
conditioned on the satisfactory resolution of the following item:
TI1e design of those parameters available to the operator in the
control room and utilized for post-accident monitoring must provide
for: at least two redlllldant channels of indication of each parameter
monitored wit11 at least one channel to be continuously recorded, and
other(s) indicated, with both channels energized from the Class IE
7-5
power system. The applicant has agreed to modify the design to confonn
with these requirements. We will review the applicable design
modifications to confinn that the design is acceptable prior to issuance
of an operating license. The results of our review will be provided
m a supplement to 'this report.
7.6 Residual Heat Removal (RHR) Interlocks
Our review of the RHR motor-operated suction valve interlocks,
utilized to prevent overpressurization of the RHR system by the reactor
coolant system, revealed that the design would not satisfy our criteria
with regard to providing interlocks of diverse principles to prevent
opening of these valves and interlocks for automatic closure of these
valves. The applicant has agreed to modify the design to conform with
our criteria. We will review the final drawings, including valve
control circuit elementary diagrams, to confinn that the cormnitted
design modification has been satisfactorily accomplished prior to
issuance of an operating license. The results of our review will be
provided in a supplement to this report.
7.7 Control Room Ventilation
Our review of the control room design arrangement revealed that the
ventilation system design provides for exhausting the hydrogen generated
in the battery rooms into the control room through the corrnnon ventilation
system ducts. In addition, we found that the ventilation ducts in the
control room were located in the plenum above the ceiling. The
applicant has agreed to re-evaluate the potential for accumulation of
7-6
an explosive hydrogen mixture in the plenum. Prior to issuance of
an operating license, the applicant must either demonstrate that the
potential problem of a fire or explosion in the control room is
incredible, or modify th.e design to prevent these events from
happening. The results of our review will be provided in a supplement.
to this report.
7.8 Environmental qualifications
The applicant has identified and stated that all safety related
motors, cables, instruments, controls and other equipment located in
side the reactor building will be able to function under the post
accident temperature, pressure, humidity and radiation conditions for
the time periods required. This capability has been acceptably
demonstrated by testing.
7.9 Separation and Identification of Safety Related Equipment
7.9.1
We have reviewed the applicant's criteria used to separate and
identify cables, cable trays, and terminal equipment and have examined
at the site the design arrangement of these as well as other safety
related equipment and systems. We have found that the separation and
identification acceptable, provide that the following items are
satisfactorily resolved.
Reactor Protection System (RPS) Cable Separation
The steel conduits housing the cables that enter the bottom of
the RPS cabinets had been cut short, .thus, exposing redundant cables
to air separation between each other. We infonned the applicant
7.9.2
7-7
that this cable design arrangement appeared to be in violation of the -
separation criteria documented in the FSAR which provide for a minimum
horizontal separation distance of 3 feet and barriers to maintain ver-
tical separation between redundant safety related cable trays. In the
absence of barriers to maintain vertical separation, we will accept in
this case a minimum vertical separation distance of 5 feet between
redundant safety related cables. The applicant has agreed to examine
this cable arrangement and either show that it maintains the minimum
required vertical and horizontal distance separation or provide
barriers when the minimum spatial separation between redundant safety
related cables cannot be maintained. The staff will assure that this action
has been satisfactorily completed prior to issuance of an operating license.
Switchgear Rooms Flooding·
Our review of the safety related switchgear rooms design arrangement
revealed that a main firewater line is located outside but near the
redundant switchgear rooms. The doors separating switchgear rooms
from each other and from the main firewater line are not of the water
tight construction. The applicant has agreed to examine the potential
for flooding both switchgear rooms upon failure of this line and either
demonstrate that_ flooding will. not'result or modify the facility design
to make the consequences of such a failure acceptable. We will review
the results of the applicant's anal,Ysis of this potential problem and
require an acceptable solution prior to issuan~e of an operating license.
The results of our review will be provided in a supplement to this
report.
7.9.3
7.9.4
7-8
Battery Rooms Separation
The redundant safety related battery rooms are directly connected
through the ventilation exhaust duct. TI1e exhaust from one battery
room discharges into the other battery room creating the potential for
a fire or explosion originating in one room propagating to the other
room resulting in the loss of both d-c systems. The battery rooms
also share a connnon wall and door. The applicant has agreed to either
demonstrate the exhaust duct, door and wall designs will confine a
fire or explosion to one of the redundant battery rooms or make
appropriate design modifications to assure complete independence between
the two battery rooms. We will require a satisfactory resolution of
this matter prior to issuance of an operating license. The results
of our review will be provided in ~ supplement to this report.
230 KV Switchyard Breakers Control Power Separation
To satisfy the requirements of GDC 17 as related to offsite power,
the applicant agreed to revise his design to provide two independent
d-c control sources and feeds to the 230 kV switchyard breakers. Our
review of the proposed (not installed) design arrangement revealed
that the CR-3 d-c control power cables emanating from the batteries of
Crystal River Unit 1 and 2 respectively (fossil fueled power plants)
must pass through a connnon walk through tunnel before entering the
switchyard. During our site visit we found this tunnel flooded several
inches deep 1n some areas and the tunnel swnp pwnps were inoperable.
Also, we noted a lack of fire detection and protection in the tunnel.
The applicant must either demonstrate that this proposed
7-9
cable routing through the tunnel cannot result in connnon mode failure
due to flooding or fire or we will require a new design arrange-
ment that meets the requirements of GDC 17. The applicant has agreed
to submit additional information on this matter, including any
modifications. We will assure that this matter is satisfactorily resolved
prior to issuance of an operating license. The results of our review
will be provided in a supplement to this report.
7.10 Control Systems
TI1e control systems are functionally identical to those of the
Arkansas Nuclear One, Unit 1 except for the provisions of the rod
drive control system design to include manual switches for disconnecting
power to each group of rods. In this regard, we have requested from
the applicant inforn1ation tl1at establishes the purpose of this design
feature. In addition, it was found that the non-safety rel2ted
Integrated Control System (ICS) participates in the operation of the
safety related auxiliary feedwater system. This concern is discussed
in Section 7.4 of this report. With the exception of the control
rod drive power disconnect switches and emergency feedwater controls,
we conclude that these control systems are acceptable. However, the
final acceptability of the overall control system scheme is predicated
on the satisfactory resolution of the two aforementioned items prior
to issuance of an operating license. The results of our review will
be provided in a supplement to this report.
7-10
7.11 .Anticipated Transients Without Scram (ATWS)
The applicant is reviewing the staff report, WASH-1270,
"Technical Report on .Anticipated Transients Without Scram (ATWS)
for Water-Cooled Power Reactors". CR-3 has been classified by the
staff as an "IC" facility, and the applicant has been requested
to implement a program to incorporate any design changes necessary
to assure that the consequences of anticipated transients would be
acceptable in the event of a postulated failure to scram in
accordance with Section II.C of Appendix A of WASH-1270. The
applicant will doclllllent the information required by WASH-1270 ·
by October 1, 1974 .. T'ne staff evaluation of this information will
be contained in a supplement to this Safety Evaluation Report.
8--1
8. 0 ELECTRIC POWER
8.1 General
GDC Nos. 17 and 18, IEEE Standards including IEEE Criteria for
Class IE Electric Systems for Nuclear Power Generating Stations
(IEEE Std. 308-1969), and Regulatory Guide (RG) for Power Reactors
including Regulatory Guide 1.6 "Independence Between Redundant Standby
(Onsite) Power Sources and Between Their Distribution Systems" dated
MarchlO, 1971 and Regulatory Guide 1.9 "Selection of Diesel Generator
Set Capacity For Standby Power Supplies" dated March 10, 1971 served
as the bases for evaluating the adequacy of the electric power system.
Specific doclUilents used in the review are listed in the bibliography
of this report.
8.2 Offsite AC Powt.t system
T1ris plant will be interconnected to the electrical grid system
through two 500 kV and four 230 kV transmission lines emanating
from their respective switchyards. The two 500 kV transmission lines
converge on the 500 kV switchyard through two separate and independent
routes. The four 230 kV transmission lines are arranged in pairs and
each pair is routed to the 230 kV switchyard on a series of transmission
towers which are located on ~eparate and independent rights-of-way
with respect to the other pair of transmission lines. The 500 kV
switchyard is arranged in a ring bus configuration with provisions for
conversion to a breaker-and-a-half configuration to acconnnodate an
additional fourth power plant at the site. The 230 kV switchyard,
8-2
which will serve as the source of offsite power to CR-3, is arranged
in a breaker-and-a-half configuration which is not directly tnter
connected with the 500 kV switchyard. Power from the CR-3 generator
is supplied to the 500 kV switchyard and also to CR-3 auxiliary trans
former. Site fossil units CR-1 and CR-2 supply power to the 230 kV
switchyard. Of;fsite power to CR-3 is from two separate feeders
emanating fr9m different breaker-and-a-half configuration bays in
!_...: the 230 kV switchyard. TI1ese power sources are connected to two "\
N
separate startup transfcrmers of which one startup transformer is
assigned to CR-3 and t,he other is shared between CR-1, CR-2 and CR-3.
T'ne shared startup transformer, feeder line and associated breakers
have suffi~ient capacity to handle all required load demands from
the three units. All of the high voltage circuit breakers in the
230 kV switchyard ~re provided with primary and backup relaying circuits
powered from independent d-c supplies.
The low voltage side of the CR-3 auxiliary transformer and of each
one of th~ startup transformers is provided with two redundant feeder
breakers, each connected to one of the two redundant emergency buses.
The emergency buses are powered from the CR-3 startup transformer during
all modes of plant operation, and upon loss of the normal supply, power
is made available manually from the control room to these buses from
either CR-1 and CR-2 startup transformer or ~R-3 auxiliary transformer.
Each transformer with its attendant distribution system has sufficient
capacity to meet shutdown and emergency load requirements.
8-3
The applicant has conducted electrical grid stability analyses
which show that the simultaneous loss of total generation at the
CR-1, CR-2 and CR-3 site will not adversely affect the stability- of
the remainder of the transmission system or the ability to provide
offsite power to CR-3.
Our review of the offsite power system revealed that the design
provided for only one source of d-c control power to the 230 kV switch-
yard breakers, thus, making the redundant offsite power sources sus
ceptible to single failures. This item and its status are discussed
in Section 7.9.4 of this report.
We have concluded that the offsite power system design satisfies
the requirements of GDC J.7 and 18 and IEEE Std. 308-1969, and is
acceptable subject to satisfactory resolution of the above mentioned
item. This matter will be satisfactorily resolved prior to issuance
of an operating license.
8.3 Onsite AC Power Systems
Redundancy is provided in the a-c emergency onsite power system.
It consists of two independent distribution systems, each powered by
an independent dies.el generator. Each distribution system includes 4160,
480, 240 and 120 volt load centers to accorrunodate the voltage require-
ments of the safety loads. Each 4160 and each 480 volt load center
bus in a distribution system can be connected to its respective
counterpart in the other independent distribution system through
two serially connected bus tie breakers. The safety loads for the ·,
8-4
facility are distributed evenly between the two independent distribution I -
systems with the exception of the third high pressure injection pump that
provides extra redundancy. This pump can be powered from either but not
both distribution systems. The selection of the power feed is limited
and controlled by providing only one circuit breaker which can be
inserted manually in one of the available two switchgear compartments
thus, preventing the interconnection of the power supplies.
TI1ere is a single 480 V motor control center which can be manually
connected to either one of the distribution systems through an
electrically interlocked transfer switch. TI1e applicant, at our request,
had modified the design of the single 480 V motor control center to
delete the automatic transfer feature and to include only the capability
for manual transfer as recorrnnended by Regulatory Guide 1. 6. We,
have determined that the loads connected to this motor control center
have no safety significance and the interlocks provided to prevent
the propagation of faults to the redundant emergency buses are
considered adequate. We conclude that the design of the manual transfer
of this load center is acceptable.
The design also provides for the connection of selected Non-Class
IE loads to one of the Class IE emergency buses through a 4160/480 V
transformer. Our review indicated that in the event of an accident
coincident with the loss of offsite power, a failure ,in the Non-Class IE
8-5
electrical system could result in the unselected connection of Non
Class IE loads to the emergency buses. 1his could result in the
tripping of the associated diesel generator due to overload. The
applicant has agreed to modify the design so that the feeder breaker
connecting the 4160/480 V transformer to one of the emergency buses
will meet Class IE requirements. This breaker will be opened
automatically upon detection of an accident coincident with the loss
of offsite power, and will be prevented from closure during the
transient stabilization period subsequent to this event. We will
review the design modifications submitted for our review to confirm
that the final design is acceptable prior to issuance of an operating
license. 1he results of our review will be provided in a supplement
to this report.
Each diesel generator is rated at 4160 V, 2750 kW continuous,
3000 kW for 2000 hours and 3300 kW for 30 minutes. The diesel generator
units are located in separate seismic Category I structures. Each unit
has independent auxiliary systems and separate seismic Category I
underground fuel storage tank. TI1e total on-site fuel oil storage
capacity provides for at least seven days of diesel generator operation
at full rated load.
The loading of the diesel generators is within the limits suggested
by Regulatory Guide 1.9 except for the voltage dip during the
first loading block which is approximately 28% of nominal
r
8-6
instead of 25% recommended by Regulatory Guide 1.9. To compensate for
this voltage dip, the applicant has provided motor starters that will
hold in during this somewhat lower voltage transient. We have concluded
that this is acceptable. With regard to the diesel generator
qualifications, the ~pplicant has indicated that the diesel generators
for this plant have been previously qualified for use in nuclear power
plant applications. We will review requested information in support of
the diesel generator qualifications to assure applicability prior to
issuance of an operating license.
Each diesel generator will be started automatically on an under-
voltage signal from its respective 4160 V emergency bus, or on an ESP
actuation trip signal. If offsite power is not available, the 4160 V
emergency buses will be isolated automatically from all supply sources.
The diesel generators will be connected automatically to their -respective
4160 V emergency bus. Under accident conditions, the safety loads
will be connecte~ automatically in a predetermined sequence to their
respective diesel ge~erator.
Our review of the electrical schematics revealed a lack of
independence of the redundant emergency buses as a result
of a design feature that provided for paralleling of the redundant
diesel generator~ through the tie breakers connecting redundant 4160 V
buses when the offsite power will not be available. It was also
discovered that the manual controls for the breakers through which
offsite power w~ be supplied to the emergency buses interfered with I
·8-7
the operation of the undervoltage trip signal required to isolate
t11e emergency buses from the offsite power sources when offsite
power is lost. In addition, we found that the tie breakers
connecting redundant e]Ilergency buses at the 480 volt level would '.
not,open automat~cally upon receipt of an ESP actuation trip signal,
thus compromising .the independence of the redundant emergency buses.
1he applicant has agreed to modify the design to resolve these
problems. We will review the revised designs to confirm that they
are acceptable prior to issuance of an operating license.
We have concluded that the a-c emergency onsite power system
satisfies GDC 17 and 18, IEEE Std. 307 and Regulatory Guide 1.6 and
1.9, and are acceptable subject to satisfactory implementation of the
above mentioned design commitments and substantiation of the diesel
generator qualifications.
8.4 D-C Power System
Onsite d-c emergency power is derived from CR-1, CR-2 and CR-3
battery systems. TI1e CR-3 battery system is comprised of two identical
and independent 250/125 volt battery bank-charger units and the attendant
distribution systems. Each distribution system is normally supplied
by its battery charger.and backed up by its floating battery bank which
has been sized to carry all connected loads for two hours upon the loss
of the normal supply. Each 250/125 volt battery charger in a distribution
system is supplied from separate 480 V emergency buses. In addition,
there is an installed 250/125 volt battery charger for each redundant
8-8
battery bank which can be manually connected to either half of its .
corresponding 250/125 volt d-c system. Each 250/125 volt battery
bank is located in a separate seismic Category I room.
Our review of the CR-3 d-c emergency power system revealed that
the design provided for manual cross-connection of the two redundant
main d-c distribution buses in the event of a battery failure. Also,
it was found that these buses could be interconnected tl1rough d-c
distribution circuit panels. Administrative controls were the only
means provided for accomplishing the interconnections and there were
no mechanical or electrical interlocks provided to prevent inadvertent
administrative errors from compromising the independence of the d-c
emergency power system. 'Ihe applicant has agreed to modify the design
to assure that the independence of the two redundant d-c systems is
maintained GDC 17 and IEEE Std. 308-1969 by either supplementing
administrative controls with mechanical or electrical interlocks or
deleting the manual cross-connection-between the redundant d-c system.
We will review the revised design to confirm that it is acceptable prior
to issuance of an operating license. The results of our review will
be provided in a supplement to this report.
Four redundant 120 volt vital a-c distribution buses are provided
to supply power to the plant protection system instrumentation and
associated circuits. Each a-c vital bus is supplied separately from
a static inverter. Each pair of inverters is normally supplied from
separate 480 V emergency buses and backed up from the respective
battery bank.
8-9
Our review of the 120 volt vital a-c system revealed that the
provisions of the design to manually cross-connect the redundant 120
volt vital a-c buses and to supply these buses from the Non-Class
IE regulated instrument buses will make the ESF analog channels
vulnerable to single failures. An acceptable design should preclude
the interconnection of the vital buses during those modes of plant
operation where the plant protection system is required to remain
operable after a single failure. With regard to the vital buses
being supplied from the regulated instrument buses, an acceptable
design should only permit the connection of one vital bus at a time
to the instrument bus and only then for a period not to exceed 8 hours.
Supplying power to one of the vital buses from the instrument bus is
· not a requirement from the standpoint of safety but would be
pennissible for periods up to 8 hours since it could be considered a
desirable feature from the standpoint of preventing spurious signals
from tripping the reactor or initiating the ESFs, while the nonnal
source of power to the vital bus is being repaired. The applicant has
agreed to make the design acceptable and to reconsider the supply of
the vital buses from the Non-Class IE regulated instrument buses. In
addition, we found that a single failure in the transfer control switch
utilized to select the alternate power source for the ESF indicating
lights will compromise the independence of two of the redundant 120 V
vital a-c buses. The applicant has agreed to modify the design so
it would not be vulnerable to single failures. We will review the
8-10
revised designs pertaining to the above mentioned items to confirm
that they are acceptable prior to issuance of an operating license.
The results of our review will be provided in a supplement to this
report.
The CR-1 and CR-2 battery system consists of two separate battery
bank units and attendant distribution systems. TI1ese power sources,
in addition to supplying the d-c loads of the fossil units, provide
control power to all 230 kV switchyard breakers. Our review findings
with regard to this battery system are reported in Section 7.9.4
and 8.2 of this report.
Subject to the satisfactory implementation of the above mentioned
design commitments and satisfactory resolution of the 230 kV switchyard
breakers control power separation (Section 7.9.4) and independence
of the CR-3 battery rooms (Section 7.9.3), we have concluded that
the d-c emergency onsite power system satisfies GDC Nos. 17 and 18,
IEEE Std. 308-1968, and Regulatory Guide 1.6 and is acceptable.
9-1
9.0 AUXILIARY SYSTEMS
The evaluation of safety related auxiliary systems are set forth
in the following subsections. These systems are grouped in the follow
ing paragraphs to indicate those required for safe operation and
shutdown and those required to mitigate radiological releases.
The auxiliary systems necessary to assure safe reactor operation
or shutdown are: (1) decay heat seawater cooling system, (2) decay
heat closed cycle cooling water system, (3) nuclear service seawater
system, (4) nuclear service closed cycle cooling water system, (5)
ultimate heat sink (in conjunction with the nuclear service water
systems, intake canal and intake structure), (6) makeup and chemical
addition system, (7) emergency feedwater system and condensate storage
facility, (8) control room and engineered safety rooms ventilation and
air conditioning systems, and (9) diesel auxiliary systems. These
systems have been designed to seismic Category I requirements.
Other auxiliary systems not required for safe reactor shutdown,
but required to mitigate radiological release to the environment are:
(1) spent fuel pool cooling and cleanup system and (2) new and spent fuel
storage and fuel handling facilities. These systems or essential
portions of the system have also been designed to seismic Category I
requirements.
This facility shares no safety related systems with the
two conventional fossil fired plants at this site; we have
9-2
determined that sharing will be limited to non-safety related
structures and systems such as the intake and discharge canals well
water and water treatment systems, a fire protection system storage
tank makeup and an auxiliary steam system. We find this limited
sharing to be acceptable.
The chilled water cooling system, secondary services cooling water
system, the demineralized water storage tanks, process sampling system,
compressed air system, equipment and floor drainage system, purification
system,
reactor
communication system, and the lighting system
auxili(;y systems that are non-safety related
are additional
and non-seismic
Category I designed systems that have been reviewed. We have deter-
mined that (a) the systems are not required ·to achieve a safe reactor
shutdown during normal or"-accident conditions and are not necessary
to prevent or mitigate the consequences of an accident, (b) the systems
where interfaced or connected to seismic Category I systems or com-
ponents are provided with seismic Category I isolation valves to
physically separate the non-essential portions from the essential sys-
tern or component, and (c) the failure of these non-seismic systems or
portions of the'systems will not have an adverse effect on safety
related systems or components located in close proximity so that their
safety function will not be precluded.
Based on our review, we conclude that these system designs are
acceptable.
9.1
9.1.1
9.1.2
Fuel Storage and Handling
New Fuel Storage
9-3
/
The new fuel storage vault is a separate and protected area for
the dry storage of fuel assemblies in the fuel storage and handling
portion of the auxiliary building. The storage facility is designed
to accommodate 66 new fuel assemblies in storage racks that have been
designed with sufficient spacing between the new fuel assemblies to
assure that, when fully loaded, the effective multiplication factor
of the array (keff) is less than 0.90 even in the flooded condition.
The fuel storage racks and vault have also been designed to seismic
Category I requirements.
We conclude that.the design of the new fuel storage facility is
accept ab le.
Spent Fuel Storage
The spent fuel storage racks provide specially designed underwater
storage space for spent fuel assemblies requiring shielding and cooling
prior to shipment, Pool storage space to accommodate more than one
and two-thirds of the full core fuel load (240 elements) has been
provided. The spent fuel storage racks design assures that the sub
critical multiplication factor (keff) of the array will be less than
0.90 for both normal (borated water) and abnormal (unborated water)
storage conditions. These spent fuel storage racks have been designed
to seismic Category I requirements. These racks have also been
designed to withstand the impact loads resulting from a dropped fuel
9-4
assembly and they are able to withstand uplift forces in excess of
the capacity of the lifting device (fuel handling hoist).
The spent fuel storage facility consists of two separated and
distin¢t spent fuel pools located adjacent to but hydraulically
separated by a watertight gate. Each of the spent fuel storage pools
have been lined with stainless steel to limit the possibility of pool
leakage through seams and penetrations. No inlets, outlets, or drains
have been provided that might allow the pools to be drained below the
normal pool level (23 feet above the top of the stored fuel assemblies).
External lines extending below this level have been equipped with anti-
syphon devices to prevent inadvertent pool drainage. The pools have been
provided with interconnected channel drainage paths behind the liner
welded seams. These channels interconnect to form a series of leak
chase trenches behind the pools and have been designed to (a) provide
detection, measurement, and location of liner leaks, and (b) prevent
uncontrolled loss of contaminated pool water.
A separate spent fuel shipping cask loading_ area has been pro-.. - -- - - -
vided adjacent to one of the spent fuel pools. An interconnecting
-------canal between these areas will permit underwater fuel transfer to the
shipping cask. A watertight gate, located in the canal and above the
top of the fuel assemblies, assures that the watertight integrity of
the pools is maintained. The cask storage area, constructed of rein-
forced concrete and lined with stainless steel, has been designed so
that if the cask drop accident should breach this area the resultant
9-5
drainage would not have an adverse effect on the storage of the spent
fuel or on any safety related equipment located in close proximity
below the pool area. The spent fuel pools and the spent fuel shipping
cask loading area have been designed as Category I seismic .structures.
Our independent evaluation of the spent fuel cask handling indi
cates that transferral of the crane and the shipping cask over the
spent fuel pool will be prohibited during cask transferral by the use
of appropriate interlocks and/or mechanical stops. However, during
handling over the storage area, the potential exists that the -cask
could strike the edge of the pit and roll or tumble in the adjacent
spent fuel pool. To avoid damage to the stored fuel, the fuel assemblies
will be located in the spent fuel pool that is not located adjacent to
the cask loading area whenever the fuel handling crane is operated in
the cask handling mode. For this condition the watertight gate between
the fuel pools is in place and sealed so that an inadvertent cask drop
accident could affect the adjacent pool but would not have an adverse
effect on the fuel pool with the stored fuel assemblies. We find this
acceptable.
Based on our review, we have concluded that the design of the
spent fuel storage facility design meets positions set forth in
Regulatory Guide No. 1.13, "Fuel Storage Facility Design Basis," dated
March 10, 1971, and, therefore, is acceptable.
9.1.3
9-6
Spent Fuel Pool Cooling and Cleanup Systems
The spent fuel pool cooling and cleanup systems have been designed
to maintain the water quality and clarity of the pool water and to
remove the decay heat generated by the stored spent fuel assemblies.
The cooling system has been designed to seismic Category I requirement~
and consists of two spent fuel pool cooling pumps, heat exchangers,
and associated piping, valves and instrumentation. The cooling system
piping is also used to supply the seismic Category I makeup source
from the borated water storage tank to the spent fuel through the
direct valve cross-connection via the decay heat removal system. The
piping from the spent fuel pool to the suction of the fuel pool pumps,
from the fuel pool heat exchangers to the spent fuel pool and all piping
and valves to and from the decay heat removal system are designed to
meet seismic Category I requirements.
The Nuclear Service Closed Cycle Cooling Water System (NSCCCWS)
removes the decay heat during normal operations and has been backed by
direct valve cross-connections to the decay heat removal system for use
during emergency conditions.
Using two pumps, two coolers and the heat from 1/3 of a stored
core while maintaining a pool temperature of 120°F or less; using one
pump and one cooler this same heat can be removed by allowing the pool
temperature to rise to 130°F. The heat from up to 1-1/3 of a core (a
non-typical condition) can be removed by both pumps and coolers without
reaching the boiling temperature of the pool water. Additional cooling
9-7
can be provided by the decay heat removal system.
The cleanup system is a non-safety related system and has been
designed to non-seismic Category I requirements. Isolation capabil-
ities from the Category I portion of the fuel pool cooling system has
been provided by seismic Category I isolation valves.
Based on our review~ we conclude that the design of the spent fuel
pool cooling and cleanup systems are consistent with Regulatory Guide
No. 1.13, "Fuel Storage Facility Design Basis," and are. acceptable.
9.1. 4 Fuel Handling System
The fuel handling system provides the means of transporting and
handling fuel from the time it reaches the plant in an unirradiated
condition until it leaves after post-irradiation cooling. The system
consists of the fuel transfer canal, the fuel transfer system, and
appropriate cranes and handling fixtures. The integrated fuel handling
operations are performed in two locations:. inside the reactor building
and in the spent fuel storage area in the auxiliary building.
Our review of major components necessary for safe fuel handling
operations indicates that the following components have been designed
to seismic Category I requirements: refueling building crane (spent
fuel cask crane); reactor building polar crane; spent fuel pool handling
bridge; and fuel transfer tube and isolation valves. The reactor polar
crane and the spent fuel· cask cran~ including the breaking for the crane
hoists have been designed in accordance with Electric Overhead Crane
Institute Specification No. 61. The cranes and major components
9.2
9.2.l
9-8
provided are of standard design and similar to those we have previously_
found acceptable.
The refueling equipment has been designed to withstand the
associated deadweight, live load and design seismic loads acting without
exceeding the allowable stress of the equipment. Also the crane systems
used for fuel handling have been provided with interlocks or limit
switches or load sensing devices to preclude unsafe fuel handling opera
tions in the auxiliary building. In addition, hoist upper limit switches
limit over-travel of the main and auxiliary hooks in the hoist direction
to preclude any possible inadvertent dropping of the spent fuel cask or
fuel elements during all modes of handling.
On the basis of our review, we have concluded that the fuel
handling system is acceptable.
Water Systems
Nuclear Services Cooling Water System (NSCWS)
The nuclear service cooling water system is a two stage system
consisting of a closed cycle subsystem which rejects its heat to a
seawater subsystem. The entire system has been designed to seismic
Category I requirements. The system provides cooling water to safety
related components essential for safe reactor shutdown durin& normal
and emergency operating conditions. The heat removal from these
safety related components by this system leaves the facility by way of
the seawater intake canal.
9-9
9.2.1.1 Nuclear Service Seawater Subsystem
The seawater subsystem uses four 50 percent capacity heat exchangers
to ensure continuous heat removal from the closed cycle subsystem during
all operating conditions. One normal and two emergency 100 percent
capacity, motor driven·pumps for this subsystem are located in the intake
structure (discussed in Section 9.2.3 of this report)(,and appropriate
valving has been provided to enable any pump to supply seawater to the
headered system. Motor operated valves provide the isolation capabil-
ities so that the pumps and the heat exchangers connected to the system
are capable of being isolated on an individual basis. We found that
the subsystem is capable of providing required cooling in the event of
a single failure in the subsystem.
When offsite power is lost under any operating or accident con-
dition, the seawater pumps will be powered by the emergency diesel
generators. Only one of the emergency seawater pumps is required to
supply the minimum essential cooling requirements.
We conclude that the Nulcear Service Seawater Subsystem is acceptable.
9.2.1.2 Nuclear Service Closed Cycle Cooling Water Subsystem
The closed cycle cooling water subsystem including pumps, heat
exchangers and associated equipment have been designed to seismic Category
I requirements. This subsystem acts as an intermediate heat sink for all
vital components and receives its cooling water supply from the seawater
subsystem described above. Non-essential components or systems normally
9.2.2
9-10
cooled by this subsystem are automatically isolated during an accident
condition by seismic Category I valves. This subsystem provides an
additional barrier between systems that may contain radioactivity and
the seawater intake canal to prevent accidental release of radioactivity.
A radiation monitor will detect the.accidental in-leakage of radioactivity
in this subsystem. Redundant component trains (motor driven pumps and heat
exchangers) are protected against missiles and appropriate valving enables
any pump to provide cooling water to the heat exchangers connected to the
headered subsystem. In addition, the subsystem components can be isolated
on an individual basis. We find the subsystem is capable of providing
required cooling in the event of any single active failure in the subsystem
and is acceptable.
Decay Heat Services Cooling System
The decay heat service cooling system (DHSCS) also utilizes two
independent subsystem functions consisting of the decay heat service
seawater cooling system (DHSSCS) and the decay heat closed cycle cooling
water system (DHCCCWS). This system has been designed to meet
seismic Category I requirements and provides cooling water to safety
related components. These components include the decay heat .removal
heat exchangers, decay heat service seawater pump motor, DHCCCWS pump
motor air handling units, decay heat pumps and motors, reactor building
spray pump and motors, and the makeup (HPIS) pumps. The heat removed
from the safety related components leaves the facility by way of the
seawater intake canal. In the event that off-site power is lost
9-11
during normal or accident conditions the pumps in both subsystems will
be powered by the emergency diesel generators. Any single header in
each subsystem can supply the minimum required cooling.
9.2.2.1 Decay Heat Service Seawater Cooling System (DHSSCS)
The DHSSC System consists of two independent (split header) full
capacity, 100 percent redundant headers to ensure continuous transfer of
heat from the DHCCCWS during all operating and accident conditions. Each
header of the DHSSC system contains a full capacity motor driven pump
and heat exchanger and appropriate valving has been provided to enable
isolation of the pump or heat e~changer. It is concluded that the
system is capable of providing required cooling in the event of a
single failure in any part of the DHSSC system.
9.2.2.2 Decay Heat Closed Cycle Cooling Water Subsystem (DHCCCWS)
The decay heat closed cycle cooling water subsystem (CHCCCWS)
cooling water pumps, heat exchange.rs and associated equipment have been
designed to seismic Category I requirements. This subsystem acts as
an intermediate heat sink for the decay heat removal system and.for
safety related components. We find that this system can provide the
required cooling in the event of a single failure in any part of the
system.
9.2.2.3 Conclusions
We conclude that the decay heat services cooling system is acceptable.
9.2.3
9-12
Ultimate Heat Sink (UHS)
The ultimate heat sink consists of the seawater intake and discharge
canals connected to the Gulf of Mexico, the intake structure and
openings, the intake and discharge conduits and the seawater pump sump
pit. The nuclear service seawater system and the decay heat service
seawater cooling system provide the means of supplying cooling water to
reactor equipment. The ultimate heat sink has been designed to seismic
Category I requirements and will be used for safe reactor shutdown
during normal and emergency operations.
The intake canal seawater is conveyed from the intake structure
to the seawater pump sump pit by two separate underground conduits.
Each conduit connects individually to separate sump pit compartments.
A manually operated sluice gate, normally open, connects the two com
partments. The seawater pump sump pit is located in a seismic Category
I portion of the auxiliary building that has been designed to withstand
the effects of a probable maximum hurricane, an SSE, tornadic wind forces,
and missiles discussed in Section 3.5 of this report •. After being
circulated through the nuclear service and decay heat seawater systems
the water is returned to the Gulf of Mexico by way of the discharge
canals.
Based on our evaluation of the ultimate heat sink, we conclude
that the design meets the positions set forth in AEC Regulatory Guide
No. 1.27, "Ultimate Heat Sink" dated March 23, 1972, and is acceptable.
9.2.4
9.3
9.3.l
Condensate Storage Facility
The condensate storage tank in conjunction with the auxiliary
feedwater system has been designed to provide a seismic Category I
auxiliary feedwater source to the steam generators for required heat
removal from the reactor coolant system during the loss of off-site
power conditions. The outdoor storage tank has been designed to seismic
Category I requirements. In addition, this tank can withstand the
effects of design level tornadic wind forces and associated missiles
so that a minimum of 112,000 gallons of condensate will be available
for removal of the reactor coolant system heat to achieve a safe shut-
down condition.
'Initially, the piping from the condensate storage tank to the
suction. side of the auxiliary feedwater pumps was designed to non
seismic Category I requirements. In response to our request, the
applicant has agreed to modify the design of the condensate storage
system so that all piping utilized in conjunction with the emergency
feedwater system will be designed in accordance with seismic Category I
requirements.
Based on our review and the applicant's conunitment to modify their
system design which we will verify prior to issuance of an operating
license, we conclude that the design of the condensate storage facility
is acceptable.
Process Auxiliaries
Chemical Addition and Makeup Systems
The chemical addition system has been designed to: (1) adjust the
9-14
concentration of boric acid for reactivity control; (2) regulate the
reactor coolant system inventory; (3) control the concentration of
hydrogen, oxygen and corrosion inhibiting chemicals in the reactor
coolant; (4) supply seal injection water for the reactor coolant
pumps; (5) supply borated makeup water to the core flooding tanks and
(6) provide emergency high pressure injection coolant to the reactor
cooling system following a LOCA. Accordingly, the portion of the
chemical addition and makeup system used for emergency core cooling
has been designed to seismic Category I requirements.
During normal reactor operation, one of three makeup (HPIS) pumps
takes suction from the makeup tank to return demineralized reactor
coolant to·the reactor coolant system and to provide seal injection
water for the reactor coolant pumps. During emergency operation two
of the three pumps will inject borated water into the reactor coolant
system from the borated water storage tank. A low reactor coolant
system pressure signal or a high containment pressure safety injection
actuation signal (SIAS) will automatically start the makeup pumps.
The SIA signal will also function to transfer the makeup pump suction
from the makeup tank to the borated water storage tank. The safety
related portion of the system has been provided with sufficient
component redundancy to make the consequences of a single active
failure acceptable.
We conclude that the system design is acceptable.
9.3.2
9-15
Storage of Compressed Gases
The storage of containers containing gases under pressure, such
as nitrogen, hydrogen, oxygen, compressed air, and carbon dioxide
tanks, is necessitated by the use of the gases in the operation of the
facility.
The applicant has evaluated the potential hazards from fail~re of
components pressurized by gases. Protection of the facility from
missiles is based on the following: (a) the containers will be designed,
constructed and tested to rigid· specifications; (b) relief valves will
be provided on tanks with set points below the design pressure of the
tanks, (c) tanks will be located in limited access areas; (d) tanks and
cylinders will be anchored to minimize potential for missiles in the
event of failure of attached piping, and (e) the remote location of gas
storage facilities and/or the location of missile proof walls in
relation to equipment essential for initiating and maintaining a safe
reactor shutdown precludes the possibility of interaction in the event
of an incident. The applicant has also stated that it meets all the
requirements of the Hazardous Material Section of Occupational Safety
and Health Administration OSHA 29 CFR 1910 Subpart H. Based on our
review and the above considerations, we conclude that the protection
provided is acceptable.
9.4
9.4.1
9-16
Air-Conditioning, Heating, Cooling and Ventilation Systems
Control Complex Building
The control complex air-conditioning and ventilation systems
provide a continuous supply of cooled air to the control room, and
other areas containing safety related equipment during normal, shutdown
and accident conditions. The control complex air-conditioning system
consists of two full capacity, 100 percent redundant seismic Category
I air handling and chilled water cooling units. Each air-conditioning
system train has been provided with necessary dampers and controls for
automatic transfer to the emergency recirculation mode (ERM). The ERM
system consists of two 100 percent redundant, seismic Category I HEPA
filters, charcoal filters, control dampers and recirculation fan units.
The two full capacity trains for the air-conditioning and ERM system
has been provided with two 100 percent capacity return air system fan
units and all components of the. independent trains are powered by the
standby A-C power system in the event of loss of offsite power under
any operating or accident condition. We find that the air-conditioning
and ventilation system, including the ERM and return systems, meet the
single failure criterion and is acceptable.
During an accident condition or upon receipt of a high
radiation or an engineered safeguards signal, the ventilation system
control dampers are automatically placed into the complete recirculation
mode of operation so that all air is filtered through the emergency
filter bank. During this recirculation mode all outside air dampers
9.4.2
9-17
are closed to minimize intake of contaminated air into the control
room. Dampers may be regulated so that system air or outside air
can be directed through the control room filtration charcoal filtera
prior to being discharged into the control room.
We conclude that the control complex normal and emergency air
conditioning and ventilation systems are acceptable conditioned on
satisfactory resolution of the battery room ventilation system discussed
in Section 717 of this report.
Fuel Handling Area
The fuel handling area (FHA) ventilation system has been designed
to function during normal operation. The fuel building system has been
designed as a once-through ventilation system and will provide ventila
tion to the fuel handling area and pump room area to maintain the fuel
handling area at a negative pressure with respect to the surrounding
areas so that all leakage will be to the fuel handling ventilation
system.
During normal plant operations, the supply fans to and exhaust
fans from the auxiliary building and fuel handling area operate
continuously. The FHA air supply ventilation system consists of a
particulate filter, inlet ventilation fan and heating unit. The
exhaust from the fuel handling area during normal and accident opera
tion is discharged through the station vent by the auxiliary building's
main exhaust system. The exhaust system consists of four 50 percent
capacity fans, and four 25 percent capacity HEPA and charcoal filter
9-18
plenums. No credit was taken for the FHA ventilation to miti ate ~~~~~~~~~~~~~
the consequences of a fuel handling accident because of the seismic
Category II design of the fan motor. The doses resulting from the
fuel handling accident were calculated by the staff and were well below
the guideline exposure indicated in 10 CFR Part 100. Additional infor-
mation on this matter is provided in Table 15.1 of this report.
In the event of a fuel handling accident, high radiation sensors
located in the auxiliary building exhaust vents will automatically
stop the auxiliary building and FHA supply fans to maintain a negative
pressure within the fuel handling area to minimize outleakage of
contaminated air. We have concluded that the fuel handling ventilation
system is acceptable.
9.4.3 Engineered Safety Feature and Other Essential Equipment
The engineered safety feature and other essential equipment com-
partment ventilation and air-conditioning systems have been designed
to provide the required supply of air to areas containing safety
related equipment. These areas include the ECCS pump rooms (the HPIS,
the LPIS, and the containment spray pumps), emergency feedwater pump
area, vital electrical and switchgear rooms, and diesel generator
rooms.
These areas have been provided with redundant 100 percent capacity,
seismic Category I air-conditioning and ventilation systems that have
the capability of being powered from the emergency buses. We find the
9-19
design of these safety rooms ventilation and air conditioning systems
meet our single failure criterion.
We conclude that the design of the engineered safety feature and
other essential equipment rooms, air-conditioning and ventilation
systems are acceptable.
9.5 Other Auxiliary Systems
9.5.1 Fire Protection System
The Fire Protection System (FPS) has been designed to meet the
requirements of the National Fire Protection Association (NFPA), Factory
Mutual Research Corporation, and the Nuclear Energy Property Insurance
Association (NEPIA). This includes inspection and approval of the
fire protection system and its equipment by appropriate insJ?eCtors.
The FPS has been designed to non-seismic Category I requirements. J/ '\
However, in response to our request, the applicant has ~the FPS
so that isolation valves provided for each fire hydrant, sprinkler, or
deluge system, and at various other locations throughout the system can
be isolated to protect areas housing safety related equipment and that
a preaction sprinkler system utilizing a dry pipe system will preclude
flooding the Category I equipment.
External fire protection is provided around the station complex
by a full-capacity motor-driven pump and two diesel-engine driven
pumps. A jockey makeup pump will maintain the fire protection piping
full and pressurized. The internal fire protection for general plant
9-20
areas is provided by water hose stations and strategically located
portable dry chemical, pressurized water and carbon dioxide fire
extinguishers. The fire .protection for specific plant areas utilize
carbon dioxide and special extinguishing agents are as follows:
(1) deluge water spray systems protect the main power transformers,
startup and auxiliary transformers, th.e charcoal filters in the
auxiliary building and control complex, the hydrogen seal oil unit area
and the turbine lube oil reservoir and purifier. This system consists
of dry pipe, open head sprinkler arrangement activated automatically
or remote manually controlled, (2) an automatic wet pipe sprinkler
system has been designed to provide protection for the turbine generator
building, the fire pump house, and the control complex floors, (3) total
flooding carbon dioxide will protect the oil lubricated bearings of the
turbine generator and the main feedwater pumps, (4) a special freon
FE-1301 system will protect the cable spreading room areas located in
the control complex, and (5) the emergency diesel generator rooms will
be protected by an automatic preaction sprinkler extinguishing system.
A fire detection system utilizing product of combustion (ion
ization) and heat actuated detection devices provides protection for
the cable spreading rooms, electrical chases and tunnels, switchgear
room, diesel generator rooms, feedwater pump areas and other areas
where fixed fire protection is required. This detection system will
initiate an alarm in the main equipment control room panel. Actuation
9.5.2
9-21
of all sprinkler and deluge systems in the sy$tem activates a local
alarm and an audible-visual alarm in the control room.
We conclude that the design of the station fire protection system
is acceptable.
Diesel Generator Fuel Oil Storage, Transfer and Auxiliary Systems
The standby A-C power system consists of two separate diesel
generator sets and associated auxiliary equipment. The diesel gener
ators are housed in separate diesel generator rooms located in seismic
Category I, tornado protected portions of the auxiliary building. The
.diesel generators are located at an elevation above the probable maxi-
mum flood level established for this facility. Each diesel generator
room is self-sufficient and protected from the other for fire,
flooding and internally generated missiles. A seismic Category I
diesel generator fuel oil storage and transfer system has been provided
for each of the two diesels and consists of an underground emergency
storage tank, an AC and DC engine driven fuel oil transfer pump, and
associated piping and valves. Each emergency storage tank has been
designed to seismic Category I requirements and are protected against
tornado missiles and flooding. Appropriate piping, cross connections
and valving in the fuel oil transfer system enable either or both
storage tanks to supply the fuel oil transfer pumps. The cross
connecting piping has been provided with two seismic Category I valves
in series to ase11re system isolation. We find that this system meets
9-22
our single failure requirements and provides a minimum of at least
seven days of diesel oil inventory for each diesel generator.
Each diesel generator is provided with an independent cooling
water system, starting system, lubrication system and air intake
system. The design and location of these subsystems meets the single
failure criterion and is acceptable.
We conclude that the diesel generator fuel oil storage, transfer
and supporting systems are acceptable.
10-1
10.0 STEAM AND POWER CONVERSION SYSTEM
10.1 Sunuila.ry Description
The steam and power.conversion system is of a conventional design
similar to those of previously approved plants i~cluding Oconee-1. The
system·. ha.s been designed for the maximum expected energy from the
nuclear steam supply system. Upon loss of full load, the system
dissipates the energy in the reactor coolant through turbine bypass
valves to the condenser or through atmospheric steam dump valves and/or
main steam safety valves to the atmosphere.
Based on our review of the steam and power conversion systems, we
have determined that other than the circulating water system intake
and discharge canals there will be no significant sharing of systems
with the onsite fossil units.
10.2 Turbine Generator
The. turbine generator is a tandem compound, three element turbine
consisting of a high pressure turbine and two low pressure stages.
The turbine generator is provided with two independent overspeed
protection systems. During normal operation overspeed is precluded
by the speed governor action of the electrohydraulic control system
that ~s-designed to fully terminate steam admission to the turbine at
approximately 103 percent of rated turbine shaft speed by closing the
turbine stop, control and intercept valves. Speed sensing for this
control is provided by two magnetic pickups.
10.3
10 .4
10.4.1
10-2
A mechanical overspeed trip device is also provided which is
actuated at 111 percent of rated speed by centrifugal force causing
a reduction in the system's hydraulic pressure forcing turbine stop
control and intercept valves to close.
We conclude that the design for the turbine generator and its
two overspeed protection systems are acceptable.
Main Steam Supply System
The main steam supply lines have been designed to transport steam,
generated in the once-through steam generators, to the high pressure
turbine.
'lhe steam lines from each steam generator.have been headered
between the turbine stop valves and the control valve in the turbine
steam chest. Each steam line has been provided with a main steam line
isolation valve designed to seismic Category I requirements.
·Based on our review of the main steam supply system design, we
conclude that it is acceptable.
Steam and Power Conversion Subsystems
General
The following sections discuss subsystems of the steam and power
conversion system that are used during the process of converting
thermal energy to electrical energy. Other non-safety related
subsystems of the steam and power conversion system have been reviewed
but not discussed in detail. On the basis that the failure of these
10 .4.2
10-3
other systems will not have an adverse effect on safety related systems, .or
components and are similar to those provided on previously approved
facilities, we conclude they are acceptable.
Turbine Bypass System
The turbine bypas!? system will discharge steam directly to the
condenser during load transient and turbine trip. The turbine bypass
system has been designed for a total steam flow capacity equivalent to
30 percent of the turbine design steam flow. The bypass system con
sists of four.automatically actuated regulating valves mounted on a
manifold. The manifold is connected to the main steam lines between
the steam line .isolation valves and the turbine stop valve. Each of
the bypass valves individually discharge to the main condenser and are
provided with manual isolation valves upstream of the bypass control
valves for isolation in the event of malfunction of the bypass control
system.
The turbine bypass system allows a large, sudden load decrease
(turbine trip) from full power without adverse effect to the reactor
system. The bypass valves are fully opened within three seconds after
a turbine trip to avoid lifting of the safety valves. If the condenser
is unavailable, .the bypass valves close automatically and the safety
and atmospheric dump valves exhaust the steam generated to the atmo
sphere. The dump capacity (7.5 percent of reactor power) is sufficient
to cool the .reactor .coolant system to safe shutdown.
We conclude· that the deslgn of the turbine bypass system is
acceptable.
10.4.3
10.4.4
10-4
Circtilation Water System
The circulating water system has been designed to provide cooling
water to the main condensers and the secondary service cooling water
system. The system has been designed to serve as a heat sink to
dissipate rejected heat from the power conversion system.
In response to our request, the applicant reevaluated this to
determine that a failure of any component in the circulating water
system such as pipe breaks, pump failure, or expansion joint ruptures
will not result in the loss of any safety related components or systems
necessary for safe shutdown due to resultant flooding. The applicant
also determined that cableways, pipe chases, or passageways interconnectin~
other spaces in the vicinity of the circulating water system will not I
be flooded.
On the bases of our review we have concluded that the design of
the circulating water system is acceptable.
Auxiliary Feedwater System (AFS)
The AFS provides feedwater to the steam generators for the removal
of decay heat from the reactor coolant system during emergency operations.
The applicant has-agreed to be redesign the AFS to seismic Category I
requirements and provide additional values and piping to meet the single
failure criteria coincident with postulated failure of a high energy
pipe outside the reactor building.
We will review the applicant's design modifications to assure that
the consequences of a failure in the AFS and a concurrent failure in a
10-5
high energy ~ipe outside the reactor building are acceptable. The
results of our review will be provided in a supplement to this report.
11.0
11.l
11-1
RADIOACTIVE WASTE MANAGEMENT
Summary Description
The radioactive waste management system for CR-3 is designed
to provide for the controlled handling and treatment of radioactive
liquid, gaseous and solid wastes. The design objective for these
systems is to restrict the amount of radioactivity released from normal
plant operation to unrestricted areas to levels that are as low as
practicable.
The Technical Specifications will require the applicant to maintain
and use existing plant equipment to achieve the lowest practicable releases
of radioactive materials to the environment in accordance with the require
ments of 10 CFR Part 20 and 10 CFR Part 50. The applicant will also be
required to maintain radiation exposures to inplant personnel and the
general public "as low as practicable" in conformance with the requirements
of 10 CFR Part 20.
Our evaluation of the design and expected performance of the
waste management system for CR-3, is based on the following design objectives:
Liquids
1. Provisions to treat liquid waste to limit the expected releases
of radioactive materials in liquid effluents to the environment
to less than 5 Ci/yr, excluding tritium and noble gases.
11-2
2. The calculated annual average exposure to the whole body or any
organ of an individual at or beyond the site boundary not exceed
5 mrem for expected releases.
3. Concentration of radioactive materials in liquid effluents not
to exceed the limits in 10 CFR Part 20, Appendix B, Table II,
Column 2, for the expected and design releases.
Gaseous
1. Provisions to treat gaseous waste to limit the expected release
of radioactive materials in gaseous effluent from principal
release points so that the annual average exposure to the whole
body or any organ of an individual at or beyond the site boundary
not exe.ead 5 mrem.
2. Provision to treat expected and design radioiodine released in
ga:seous effluent from principal release points so that the annual
average exposure to the thyroid of a child through the pasture-cow
milk pathway not exceed 15 mrem.
3. Concentration of radioactive materials in gaseous effluents not to
exceed the limits in 10 CFR Part 20, Appendix B, Table II,
Column 1, for the expected and design releases.
Solid
1. Provisions to solidify all liquid waste from normal operation
including anticipated operational occurrences prior to shipment
to a licensed burial ground.
' ' I 'l I
11 Ii
11.2
11-3
2. Containers and method of packing to meet the requirements of 10 CFR
Part 71 and applicable Department of Transportation regulations.
Liquid Wastes
Treatment of the waste is dependent on the source, activity and
composition of the particular liquid waste and on the intended disposal
procedure. The liquid waste treatment system is divided into two
subsystems, i.e., the makeup and purification subsystem and the
miscellaneous waste processing subsystem. The wastes in these two
subsystems will normally be collected and processed through separate
evaporators; the condensates from each evaporator will be passed
through common demineralizers, and collected in the evaporator
condensate storage tanks. The two subsystems are normally isolated
from each other; however, cross connection between the subsystems
provides flexibility for processing by alternate methods. Treated
wastes will be handled on a batch basis as required to permit optimum
control and rele~se of radioactive waste. Prior to the release of
any treated liquid wastes, samples will be analyzed to determine the
type and amount of radioactivity in a batch. Based on the analytical
results, these wastes will either be recycled, reprocessed, or
released. Radiation monitoring equipment will automatically trip a
valve on the discharge pipe terminating the release of liquid waste
if the levels of activity are above a predetermined value.
11-4
The makeup and purification subsystem will maintain the quality
of the reactor coolant. A normal 45 gpm stream will be continuously
let down, cooled, passed through a mixed-bed demineralizer, filtered
and fed to the makeup tank from which it will be returned to the
reactor or discharged. The boron concentration will be maintained by
diverting a portion of the letdown stream to one of the three 76,000
gallon bleed tanks. Equipment drains and miscellaneous high purity
liquid wastes will also be collected in the bleed tanks. From the
bleed tanks the radioactive liquid wastes will be processed through a
cation demineralizer and the 12.5 gpm reactor coolant evaporator. The
condensate from the evaporator will be processed through a mixed-bed
demineralizer and collected in one of the two 8,230 gal evaporator
condensate storage tanks. After sampling, this liquid will be sent
to the reactor coolant storage tanks for recycling in the reactor or
discharged to the river with the circulating water system.
Our evaluation assumed that 425 gal/day of deaerated wastes from
the reactor coolant drain tank and 320 gal/day from the shim bleed for
boron control will be processed by the makeup and purification system
and that 90% will be recycled and 10% discharged.
We estimate that approximately 0.5 Ci/yr excluding tritium and
noble gases will be discharged from this source. The applicant did
~ot estimate the release of radioactive material in liquid effluents
by subsystems.
11-5
Aerated liquid wastes from the containment and auxiliary buildings,
laboratory drains and sampling sources, demineralizer regeneration
solutions and sluice, and other wastes will be collected in the 20,500
gal miscellaneous waste storage tank and processed by the 12.5 gpJ!l
miscellaneous waste evaporatoro The condensate is processed through
a mixed-bed demineralizer and collected in one of the evaporator
condensate storage tanks. After sampling, this liquid wi'll be sent
to the primary waste storage tanks for recycling in the reactor or
discharged to the river with the circulating water system..
Our evaluation assumed that 375 gal/day of aerated wastes will
be processed by the evaporator and polishing demineralizer and that
100% of the distillate will be discharged. We estimate that
approximately 0.12 Ci/yr excluding tritiuJ!l and noble gases wi~l b~
discharged from this source. The applicant did not estimate the
release of radioactive material in liquid effluents by subsystems.
In addition to the sources listed above, we estimate slightly
less than 0.1 Ci/yr will be released in untreated effluent from the
turbine building drains and about 0.04 Ci/yr will be discharged from
the laundry system. The applicant did not estimate the release of
radioactive material in liquid effluents by subsystem.
We calculate that approximately 0.6 Ci/yr excluding tritium and
dissolved gases will be discharged to the unrestricted area from the
plant. To compensate for equipment downtime and expected operational
11-6
occurrences, we have normalized our calculated release rate of radio
activity release in liquid effluents to 5 Ci/yr, excluding tritium and
dissolved gases. The applicant estimated 0.0025 Ci/yr of mixed radio
isotopes will be dischared. Based on operating experience of other
pressurized water reactors, we estimate that tritium releases will be
approximately 350 Ci/yr. The applicant has also estimated that approxi
mately 350 Ci/yr of tritium will be released.
Our estimates are based on a revised version of ORIGEN code which
is adjusted to apply to this plant. The ORIGEN code is described in
ORNL 4628, Oak Ridge Isotope Generation and Depletion Code." The
·waste stream activities and flows used in our evaluation are based on
experience and data provided from operating reactors. The model uses
somewhat different values for the parameters than those of the
applicant. The applicant has considered less volume being processed
and discharged from the system, a lower release fraction of fission
products and uses higher decontamination factors. Our calculated
radioactive release doses therefore differ from those of the applicant's.
From our evaluation of the expected liquid radioactive releases
we calculate a total whole body and organ dose of less than 5 mrem/yr.
The applicant calculates a total whole body dose of 0.08 mrem/yr and
a total critical organ dose of 0.016 mrem/yr. We conclude that the
liquid radwaste system will reduce radioactive effluents to as low as
practicable in accordance with 10 CFR Part 20 and 10 CFR Part 50.
11. 3
11-7
Gaseous Wastes
The gaseous waste treatment and ventilation systems will process
waste gases from degassing of the reactor coolant, auxiliary building
ventilation, containment purging, and sweep gas for the various liquid
tanks. The primary source of gaseous radioactive waste will be from
the degassing of the reactor coolant during letdown of the reactor
coolant into the various holding tanks.
Gases stripped from the reactor coolant letdown flow in the
makeup and purification system, and from the deaerated wastes and
shim bleed in the bleed tanks will be processed by the waste gas
vent header system. This system is divided into two subsystems; one
subsystem in the auxiliary building and one subsystem in the reactor
building. Waste gases from the reactor coolant drain tank in the
containment building flow into the miscellaneous waste storage tank.
The waste gases from the three reactor coolant bleed tanks and the
miscellaneous waste storage tank discharge to the waste gas surge tank.
The gas that will be removed from the circulating stream will be removed
from the surge tank to one of the three decay tanks for holdup before
processing through the charcoal and HEPA filters (gaseous waste
disposal filters) and released to the reactor building. All releases
of the environment from th·e decay tanks will be monitored twice,
once as it leaves the decay tanks and after it mixes with exhaust
ventilation from the auxiliary building. Either monitor will
11-8
terminate the gas discharge automatically when release of radioactive
material reaches that specified in the Techni_cal Specifications.
Considering an annual average gas flow of 144 cu ft/day and a
total storage capacity of approximately 20,000 cu ft in two of the
three delay tanks, we calculate that approximately 135 days decay
will be provided. In our evaluation, we assum~d a_~in~mum of 90 days
holdup- will b-e provided prior fo releas~ to the environment; as used
by the applicant. The difference in the activity released after a
delay of 90 days or 135 days is negligible since Kr-85 becomes the
predominant isotope after 90 days delay which has a half-life of
10.7 yrs. We estimate that approximately 650 Ci/yr of noble gases and
less than negligible amounts of I-131 will be discharged from this
source. The applicant estimates approximately 340 Ci/yr of noble
gases and 5. 3 x 10-.13 Ci/yr of I-131 will be discharged.
The condenser air ejectors will remove gases which collect in
the condenser. These gases will be vented directly to the atmosphere
without treatment. There is no blowdown from the once-through type
steam generators used in this plant. We calculate that approximately
970 Ci/yr of noble gases and 0.01 Ci/yr of I-131 will be discharged
from the air ejector exhaust. The applicant estimates approximately
210 Ci/yr of noble gases and 4.6 x 10-s Ci/yr of I-131 will be
released.
Radioactive gases may be released in the auxiliary and turbine
buildings due to equipment leaks. The ventilation system for the
11-9
auxiliary building has been designed to insure that air flow will be
from areas of low potential to areas having a greater potential for
the release of airborne radioactivity. During normal operation the
auxiliary building ventilation will draw air from the equipment rooms
and open areas of the building through HEPA filter and charcoal
adsorbers and discharge to the atmosphere through the facility vent.
The turbine building is open and therefore not amenable to treat gaseous
releases.
We calculate that approximately 970 Ci/yr of noble gases and
0.008 Ci/yr of I-131 from the auxiliary building, and small amounts of
noble gases and 0.09 Ci/yr of I-131 from the turbine building, will
be discharged from these sources. The applicant made no estimates of
the release of radioactive material from the auxiliary building and
the turbine building.
Radioactive gases may be r~leased inside the reactor building
when components of the reactpr coolant system are opened to the building
atmosphere or when minor leaks occur in the primary coolant system. The
reactor building atmosphere will be purged through charcoal and
HEPA filters, and discharged to the facility vent. Based on a composite
leak of 40 gal/day in the containment building, we calculate that
approximately 470 Ci/yr of noble gases and 0.11 Ci/yr of I-131 will be
discharged from this source. The gaseous source terms are calculated
by means of STEFFEG code as described in the F. T. Binford, et. al.,
11.4
11-10
report, "Analysis of Power Reactor. Gaseous Waste Systems," 12th Air
Cleaning Conference'. The applicant estimates approximately 70 Ci/yr
of noble gases and 0.0017 Ci/yr of I-131 will be discharged.
We calculate that a total of 3050 Ci/yr of noble gases and 0.13
Ci/yr of iodine-131 will be released to the unrestricted area by this
facility. The applicant estimates about 620 Ci/yr of noble gases and
0.0017 Ci/yr of iodine-131 will be released from the facility. These
differences in estimates can be explained by the applicant not
considering a source for auxiliary building leakage assuming a reactor
building leak rate of 10 gpd rather than the 40 gpd used in our evaluation,
a fission product release of 0.1% rather than the 0.25% used in our
evaluation on an removal efficiency of 95% rather than the 90% used in
our evaluation.
From our evaluation of the gaseous radioactive releases to
unrestricted areas, we calculate a total whole body and critical
organ dose of less than 5 mrem/yr and less than 15 mrem/yr to a child's
thyroid due to the pasture-cow-milk chain with the cow at the nearest
dairy cow located 4 miles ENE of the facility. Based on our evaluation,
we conclude that the gaseous radwaste system will meet the low as
practicable requirements of 10 CFR Part 50 and 10 CFR Part 20.
Solid Wastes
The solid radwaste system will be designed to collect, monitor,
process, package, and provide temporary storage for radioactive solid
11-11
wastes prior to off site shipment and disposal in accordance with
applicable regulations.
Spent demineralizer resins from the various treatment systems will
be transferred to a spent resin storage tank. The resins will then
be dewatered. The resin sluice water will be processed later by the
aerated waste system. The spent resins will be discharged into a
truck mounted shipping cask.
Evaporator concentrates are stored in the concentrated waste
storage or concentrated boric acid tanks. From these tanks the
concentrates will be pumped to an evaporator concentrates shipping
container where it will be mixed with a solidifying absorbent.
Expended filter cartridges will be placed into a shielded drum
for storage and offsite shipment. Other dry solid wastes consisting
of contaminated rags, paper, protective clothing and miscellaneous
contaminated items will be packaged in drtlllls or other suitable
containers for disposal.
Containers will be filled and sealed by remote control when the
radiation levels so require. All containers will be contained and
shipped in accordance with AEC and Department of Transportation (DOT)
regulations.
The staff estimates approximately 15,000 Ci/yr of solid wastes
will be shipped offsite. We find the proposed system acceptable.
11.5
11.6
11-12
Design
Decay tanks and surge tanks are designed to meet ASME Class III,
Section C and seismic Category I requirements. The t:a1iks, deminera;I.~zers,
ind evaporators in the liquid radwaste system are designed to meet
ASME, Class III, Section C and seismic Category I requirements. All
piping is designed to USAS B31.l-1967, but is fabricated and installed
in accordance with USAS B31.7 Class N3.
We conclude that the radwaste system design codes are in accordance
with appropriate codes and standards and are acceptable.
Process and Area Radiation Monitoring Systems
The process radiation monitoring system is designed to provide
information on radioactivity levels of systems throughout the plant,
on leakage from one system to another, and on leveis of radioactivity
released to the environment. The system will consist of particulate,
iodine, and gross activity monitors and samplers for auxiliary building
exhaust, fuel building exhaust, nuclear sample room exhaust~ radio
chemical laboratory exhaust, and spent fuel area exhaust. The final
discharge point for all gaseous releases from the facil1ty will be through
the containment purge exhaust or the auxiliary building exhaust.
Other gaseous process monitors ~ocated within the facility measure the
activity for control room ventilation intake, containment, waste gas
11. 7
11-13
decay tank, condenser vacuum pump exhaust, and gas sampling station.
The liquid process monitors located within the facility measure the
activity for the reactor coolant letdown spent fuel cooling water,
decay heat closed cooling water, nuclear services closed cooling water,
and the final monitor on the facility discharge line.
The area radiation monitoring system is designed to provide
information on radioactivity fields in various areas with the facility.
The system will consist of 19 monitors at the following locations in
the facility: control room, radiochemical laboratory, sample room,
auxiliary building, and the reactor building.
The system will detect, indicate, annunciate and/or record the
levels or fields of activity to verify compliance with 10 CFR Part 20
and keep the radiation levels as low as practicable. We conclude
that the facility is adequately provided with process and area monitoring
equipment.
Radiation Protection Management
The objective of radiation protection is to ensure that radiation
exposure to station personnel is as low as practicable. The applicant
will establish health physics procedures under the direction of the
health physics supervisor which will assure that all requirements
releated to radiation protection are followed by all station personnel.
These procedures will provide rules for personnel monitoring, use of
protective clothing and equipment and will require that a radiation
11-14
work permit be obtained for c~rtain areas of potential exposure.
Supporting data regarding the effectiveness of the heal th physics
program will be obtained through the collection of bioassay samples~
comprehensive medical examinations and film badge or thermal
luminescence dosimeter (TLD) data.
All areas within the facility will be identified by different
radiation zones in accordance with the expected maximum occupancy.
The applicant will provide four areas of radiation control within the
facility during full power operation according to maximum design radiation
·dose rate. These are: Zone O, continuous access, 0.5 mrem/hr or
less; Zone I, periodic access, 2.5 mrem/hr or less; Zone II, limited
access, 15 mrem/hr or less; and Zone III, controlled access, general,
25 mrem/hr or less, and Zone IV, restricted access, greater than 25
mrem/hr. These areas will be identified by radiation caution signs.
Personnel monitoring equipment shall be provided for all personnel
at the facility. Records showing the radiation exposures of all personnel
at the facility will be maintained by the applicant. Neutron film badges
will be provided whenever neutron exposures are expected. Bioassays
will be made as necessary to determine internal exposures to facility
personnel. Protective clothing and respiratory protective equipment
will be available for the protection of personnel, when required.
Portable radiation monitoring instruments will be available to
determine exposure rates and contamination levels in the facility.
11.8
11-15
The applicant's design objective for radiation shielding for
normal operation is to maintain whole body dose rates for all
controlled access areas of the facility to less than 1.25 rem per
calendar quarter, considering occupancy of each controlled access
area. For areas outside the facility, the shielding design objective is
to maintain whole body rates to less than 0.5 rem per calender year.
The principal shielding material used in the facility is ordinary
concrete. Other material will be used by the applicant for special
situations. Equipment, pumps, valves, and pipes that will contain
significant levels of radioactive material will be segregated into
modules by shield walls to minimize radiation exposures from mainten
ance of these items. We conclude that precautions taken for personnel
protection satisfy the requirements of existing regulations as
pertains to exposure of individuals to radiation, and are ~cceptable.
Conclusions
Based on our model and assumptions, we calculate an expected whole
body and critical organ dose of less than 10 mrem/yr to an individual
from gases and less than 5 mrem/yr from liquids at or beyond the site
boundary. We calculate the potential dose to a child's thyroid from
the pasture-cow-milk chain to be less than 15 mrem. Therefore, we
conclude that the liquid., gas and solid waste treatment systems meet
the requirements of "as low as practicable."
11-16
We also conclude that the system is designed in accordance with
acceptable codes and ·standards, that the process monitoring system is
adequate for monitoring effluent discharge paths as specified in
GDC No. 64 and personnel protection systems satisfy the requirements
of existing regulations as pertain to exposure .of individuals to radiation.
12.0
12.1
12-1
RADIATION PROTECTION
This section presents an evaluation of the adequacy of the
shielding, ventilation and health physics program to control
radiation exposures within 10 CFR Parts 20 and SO.
Shielding
The radiation shielding provided has been designed based on a
criterion that during normal operation the radiation dose to
operating personnel and to the general ~ublic is within the limits set
forth in 10 CFR 20. Standard methods and recognized computer codes
(SDC, QAD) were used by the applicant to evaluate the shield design.
Staff calculations at selected locations (using SDC) indicate that the
shielding provided will be adequate to meet designated radiation zone
requirements.
Information provided in the FSAR as well as observations made
during a site visit. show that the general principle of shielding
compartmentalization for major components which are expected to
contain radioactivity has been employed. In general, enough room
has been provided to allow for maintenance and temporary shielding if
necessary. The solid radioactive waste packaging system has been
designed to minimize the radiation exposure of personnel percforming
12.2
12-2
the packaging operations. Piping of radioactive process system is not
"field-run" but is routed by the architect-engineer. The applicant
estimates the total exposure of on-site personnel to be about 220
man-rem per year based on the operating experience of similar facilities.
In addition it estimates approximately 75 man-rem would accrue to con
tractor personnel during a projected annual six week outage.
We conclude that adequate consideration has been given to shielding
design to keep exposures within applicable limits and to reduce
unnecessary exposures during normal operation of the plant. During
startup of the facility and when full power operation is attained, the
facility will be mapped for dose levels and these will be compared with
anticipated levels.
Ventilation
The building ventilation system has been designed to continuously
supply a fresh air flow from the normally occupied areas of the build
ing, through rooms containing radioactive waste equipment to the
ventilation exhaust discharge system. The capacity of the exhaust
system is higher than that of the fresh air supply system so that the
air pressure inside the building will be slightly below the outside.
All tanks and processing equipment which may evolve radioactive gases
are vented to the waste gas vent header system to prevent radioactive
gases from escaping to the building atmosphere. The spent fuel pit
12.3
12-3
ventilation system design provides for a continuous sweep of air across
the top of the spent fuel pits and cask loading pit.
Based on the description of the ventilation system in the FSAR,
the monitoring of airborne contamination, and the planned procedures
for inhalation exposure control we conclude that the ventilation
system will be adequate.
Health Physics Program
The health physics program is the responsibility of the Chemical
and Radiation Protection Department. It is plant policy to keep radiation
exposure to personnel as low as possible and to adhere to pertinent
regulations. This department is responsible for the orientation and
training of personnel in radiation protection principles and procedures
to maintain exposures as low as practicable.
Personnel protection will be accomplished through administrative
controls and procedures, through the use of protective equipment and
will be verified through an extensive personnel monitoring program.
Administrative exposure limits and the use of Radiation Work Permits (RWP)
enable the Chemistry and Radiation Protection Engineer to ensure
compliance with 10 CFR 20. The issuance of a RWP allows for prejob
surveillance and specification of protective measures such as protective
equipment and radiation monitoring.
Special protective equipment includes a full array of protective
clothing, temporary shielding, respirators and self-contained
12-4
breathing apparatus. Personnel decontamination facilities are also
provided. The flow of traffic to the Radiation Controlled Area from
the secondary side is through the Health Physics area where monitors
and change rooms are available.
All CR-3 personnel will wear TLD dosimeters and neutron sensitive
film badges~ Pocket chambers or special TLD badges will be issued to
personnel working in relatively high radiation areas. Whole body
counts will be routinely made on selected employees and in special
cases as needed. Bioassays fo~ tritium will be performed on an as
needed basis. Further, the applicant has made a verbal commitment
to provide two air particulate and iodine monitors and at least six
TLD badges in the conventional units.
Based on our review, we conclude that the applicant's health phy
sics program is acceptable.
13.0
13.1
13-1
CONDUCT OF OPERATIONS
Plant Organization, Staff Qualifications and Training
The CR-3 staff will consist of approximately 80 full time employees,
not including clerical and security force personnel. CR-3 is under the
onsite supervision of the Nuclear Plant Superintendent who reports to a
General Plant Superintendent who in turn reports through the Production
Superintendent to the Assistant Vice President - System Operations.
The Nuclear Plant Superintendent is directly responsible for the safe
operation of the facility; he has an Assistant Plant Superintendent
responsible for operations, maintenance, technical support and nuclear
engineering,; a Compliance Engineer responsible for auditing of
operational and maintenance quality; a Chemistry and Radiation Protec
tion Engineer responsible for the health physics program and plant
.water chemistry~_and an Administrative Supervisor in charge of clerical
help, security and building servicemen.
The Operations Engineer who reports to the Assistant Plant
Superintendent is responsible for directing the day-to-day operation
of the operating shifts. The minimum operating shift complement
is one Shift .Supervisor licensed as a Senior Reactor Operator, one
Chief Operator and one Control Center Operator licensed as R~actor
Operators, two Assistant Control Center Operators and one Equipment
Operator. The Chemistry and Radiation Protection Engineer who reports
l
13-2
to the Nuclear Plant Superintendent has reporting to him a staff of
approximately eight persons including a Health Physics Supervisor and
an Assistant Chemistry and Radiation Protection Engineer. The
Maintenance Engineer who reports to the Assistant Plant superintendent
has a staff of approximately 18 persons and is responsible for
mechanical and non-instrument related electrical equipment. The
Technical Support Engineer reports to the Assistant Plant Superintendent
and has a staff of approximately 12 persons including a Computer and
Controls Engineer, a Technical Support Supervisor and a Results
Engineer and is responsible for all control and information systems in
the nuclear plant. The Nuclear Engineer reports to the Assistant
Plant Superintendent and has a staff of three persons reporting to
him and is responsible for core performance and core analysis.
The applicant has conducted a training program for most operating
personnel which consists of six phases: (1) academic training,
(2) nuclear instrumentation training and research reactor training,
(3) nuclear plant observation and participatory experience, (4)
nuclear plant design training, (5) nuclear plant simulator training,
and (6) on-site training and testing. Selected members of the CR-3
staff technical support groups completed formal training specifically
oriented to their assigned responsibilities.
The qualifications of key supervisory personnel with regard to
educational background, experience, training and technical specialties
13.2
13-3
have been reviewed and conform to those defined in Regulatory Guide
1.8, Personnel Selection and Training (ANSI Nl8.l, "Selection and
Training of Nuclear Power Plant Personnel").
Technical support for the CR-3 staff is primarily provided by the
Production Departments Plant Performance Section, Maintenance Section,
Chemical and Environmental Surveillance Section, Nuclear Section, and
Fuels Manager. In addition, assistance can also be obtained from
approximately 79 employees from the Generation Engineering and
Construction Management Departments.
We have concluded that the organizational structure, the training
and qualifications of the CR-3 staff are adequate to provide an
acceptable operating staff and technical support for the safe operation
of the facility. During initial startup, the CR-3 staff will be aug
mented in the areas of operations management, technical support,
chemistry and radiation protection and shift operations. In addition,
technical assistance for the startup will be provided by Babcock &
Wilcox, relative to the Nuclear Steam Supply System. The applicant's
operating personnel requalification training program, which must meet
the provisions of Appendix A to 10 CFR Part SS, is currently under
review by the staff. The results of our evaluation of this matter will
be provided in a supplement to this report prior to issuance of an
operating license.
Safety Review and Audit
The safety review and audit will be conducted by the Plant Review
Committee and the General Review Committee. The Plant Review Committee
13.3
13-4
is advisory to the Nuclear Plant Superintendent and will review all
safety related procedures and design modifications. The General Review
Committee provides corporate management with a review and audit
capability to verify that organizational checks and balances are
functioning to assure continued safe operation and design adequacy
of the plant. The General Review Committee will function in accord
with Regulatory Guide 1. 33 (ANSI Nl8. 7, "Standard for the Admini
strative Controls for Nuclear Power Pla?-ts," Section 4.). Detailed
features of the review and audit program will be incorporated in the
Administrative Controls Section of the applicant's Technical
Spee if ica t ions .
We conclude that the provisions for the review and audit of
plant operations are acceptable.
Plant Procedures and Records
Facility operations are to be performed in accordance with written
and approved operating and emergency procedures. Areas include normal
startup, operation and shutdown, abnormal conditions and emergencies,
refueling, safety related maintenance, surveillance and testing, and
radiation control. All procedures and changes thereto will be reviewed
by the Plant Review Committee and approved by the Nuclear Plant
Superintendent prior to implementation. Facility records to document
appropriate station operations and activities will be maintained by
the applicant. Facility procedures and record keeping have been
reviewed against Regulatory Guide 1.33.
13.4
13-5
We conclude that the provisions for preparation, review, approval,
and use of written procedures and record keeping are satisfactory.
Detailed features regarding Facility procedures and the records
management program will be incorporated in the Administrative Controls
Section of the Technical Specifications.
Emergency Planning
The applicant has established an emergency plan that describes
those elements necessary for coping with emergencies at the facility.
The plan includes the organization for coping with emergencies.
Agreements, liaison and communications have been made with appropriate
agencies that have responsibilities for coping with emergencies. The
applicant has defined categories of incidents, including criteria for
determining when protective measures should be considered and for the
notification of offsite support groups. Arrangements have been made
by the applicant to provide for medical support in the event of a
radiological incident or other emergencies. Provisions for periodic
drills for both plant personnel and offsite emergency organizations
have been included in the Emergency Plan.
We have reviewed the Emergency Plan and conclude that it meets the
criteria of Appendix E of 10 CFR 50, and that adequate arrangements have
been made to cope with the possible consequences of the accidents at
the site, and that there is reasonable assurance that such arrangements
will be satisfactorily implemented in the unlikely event that they are
needed.
13.5
13-6
Industrial Security
The applicant has submitted a description of its Industrial Security
Plan for protection of CR-3 from industrial sabotage. The information
was submitted as proprietary information pursuant to Section 2.790 of
the Commission's regulations. We have reviewed the plan and conclude
that it conforms to the requirements of 10 CFR 50.34(c), 10 CFR 73.40,
and to the provisions of Regulatory Guide 1.17, June 1973, and that
adequate security provisions have been made for CR-3.
14.0
14.1
14-1
INITIAL TEST AND OPERATION
Test Startup Program
Florida Power Corporation has overall responsibility for the
supervision and performance of the .. test and startup program. The
development, planning, scheduling and execution of the test program
is administered by a Test Working Group (TWG). The test working group
is presently comprised of six members of FPC, one B&W member, and
one member of FAI A/E organization. The FPC Manager Power Testing
is Chariman of the TWG. The Nuclear Plant Superintendent is responsible
for all plant operational activities.
The preoperational and startup test procedures are initially
developed by FPC, Babcock and Wilcox and other independent agents as
required. They are then distributed by the Manager Power Testing to
various groups for review and then finalization by a Test Procedure
Review Group. The Plant Review Committee will review safety related
procedures. Test procedures are valid·for distribution and use only
after approval by the Director - Generation Engineering. Test results
are reviewed by the Test Working Group and safety related test results
by the Plant Review Committee. The Manager - Power Testing makes the
determination that a system test is complete and acceptable.
We conclude that the applicant's preoperational and startup
testing program is in general accord with the AEC publications "Guide
for the Planning of Preoperational Testing Programs". This program
provides an adequate basis to confirm the safe operation' of the plant
and is therefore acceptable.
15.0
15.1
15-1
ACCIDENT ANALYSIS
General
We artd the applicant have evaluated the offsite radiological
consequences for postulated design basis accidents. These accidents
are the same as those analyzed for previously licensed PWR plants and
include a steam line break accident, a steam generator tube rupture
accident, a loss-of-coolant accident, a fuel-handling accident, and
a rupture of a radioactive gas storage tank in the gaseous radioactive
waste treatment system.
The applicant has evaluated the loss-of-coolant accident, the
fuel handling accident, the rod ejection accident, and the radioactive
gas decay tank rupture. The offsite doses we calculated for these
accidents are presented in Table 15.1 of this report, and the
assumptions we used are listed in Section 15.2.1. All potential doses
calculated by the applicants and by us for the postulated accidents
are within the 10 CFR_Part 100 guideline values.
On the basis of· our experience with the evaluation of the
steam line break and the steam generator tube rupture accidents
for PWR plants of similar design, we have concluded that the conse
quences of these accidents can be controlled by limiting the ·permissible
reactor ·.coolant and secondary coolant radioactivity concentrations
so that potential offsite doses are small. We will include
appropriate limits in the Technical Specifications on these
4
.15-2
TABLE 15.1
I POTENTIAL OFFSITE DOSES DUE TO DESIGN BASIS ACCIDENTS
I I I· i
Accident
Loss of Coolant
Post-LOCA. Hydrogen Purge Dose
Fuel Handling
Two Hour Exclusion Boundary
(1340 Meters) Thyroid Whole Body
(Rem) (Rem)
88 5
(with filters) 9 <l
Fuel Handling* (without filters) 57 <l
Gas Decay Tank Rupture Negligible 1
Rod Ejection Case I 38 Case II 43
<l <l
Course of Accidents Low Population Zone
· (8047 Meters) Thyroid Whole Body
(Rem) (Rem)
7
<l
<l
3
Negligible
15 4
<l
<l
<l
<l
<l
<l <l
*We conclude that the offsite thyroid dose due to a coincident failure of the non-seismic Class I filter train used inthe spent fuel building ventilation system to reduce the iodine activity released to the environment from a1refueling accident is acceptable as this postulated dose is well within the guideline exposure indicated in 10 CFR Part 100.
15.2
15.2.1
15-3
coolant activity concentrations. Similarly, we will include
appropriate limits in the Technical Specifications on gas decay
tank activity so that a single failure (such as sticking and lifting
of a relief valve) does not result in doses that are more than a
small fraction of the 10 CFR 100 guidelines.
Hydrogen Purge Dose Analysis
Using Regulatory Guide 1.7 assumptions, the applicant has
calculated a hydrogen purge dose of approximately 0.1 Rem at the
Low Population Zone. Our independent calculations are in substantial
agreement with this incremental dose.
Design Basis Accident Assuniptions
Loss-of-Coolant Accident (LOCA)
1. Power level of 2544 Mwt.
2. Regulatory Guide No. l~, "Assumptions Used for Evaluating the
Potential Radiological Consequences of a Loss-of-Coolant Accident
for Pressurized Water Reactors," Revision 1, June 1973.
3. Design containment leak rate of 0.25% for the first 24-hours
and 0.125%/day thereafter.
4. Iodine removal by the containment quench spray system was based
on:
Reactor Building Volume
Spray Fall Height
Spray Flow Rate
15-4
Elemental Mass Transfer Velocity
Organic Mass Transfer Velocity
Spray Drop Diameter
Spray Terminal Velocity
Factor of Conservatism
Spray Reduction Limits
Elemental
Organic
Particulate
Spray Removal Rates
Elemental
Organic
Particulate
2.0 x 106 ft3
96 feet
1500 gpm
5. 72 cm/sec
0.081 cm/sec
1500 micron
480 cm/sec
1.11
1000
1000
100
7.56 hrs- 1
0.107 hr- 1
0 .45 hr-1
5. Ground level release with Pasquill type "F" conditions with wind
speed of 1.3 meters per second for short-term releases based on
the meteorological data discussed in Section 2.3.4 of this report.
Our evaluation of the iodine removal.effectiveness of the
containment sprays is dis cus.sed further in Section 6. 2 of this
report.
(
15.2.2
15.2.3
15.2.4
15-5
Fuel Handling Accident
The assumptiOI1S used to calculate of fsite doses from a fuel
handling accioent {Regulatory Guide 1.25) are:
1. Rupture of all fuel rods in one assembly.·
2. All gap activity in the rods, ass'umed to be 10% of the noble
gases and 10% of the iodine (with a peaking factor of 1.65), is
released.
3. The accident occurs 72 hours after shutdown.
4. 99% of the iodine is retained in the pool water.
5. Iodine above the pool is 75(. inorganic and 25% organic species.
6. Standard ground release meteorology and dose conversion factors.
7. Iodine removal factor of 90% and 70% for the charcoal filter for
elemental and organic iodines respectively.
Gas Decay Tank Rupture
The assumptions used to calculate ·the offsite doses from a gas
decay tank rupture were:
1. Gas decay tank contains one complete reactor coolant loop inventory
of noble gases resulting from operation with 1% failed fuel
(100,000 curies of noble gases).
2. Standard ground level release meteorology and dose conversion
factors.
Control Rod Ejection Accident
The assumptions used to calculate offsite doses from a control
rod ejection accident are:
/ ,/
15-6
Case I
1. Power level of 2544 Mwt.
2. 28% fuel failed in transient.
3. 10% of iodine and noble gas inventory in gap of failed fuel.
4. Release of total gap activity in failed fuel to containment
building.
5. 50% plate-out of radioactive iodines.
6. Containment building sprays are not initiated.
7. Containment building leak rate of 0.25%/day for 24 hours and
one-half this value thereafter.
8. Standard ground level release meteorology and dose conversion
factors.
Case II
1. Power level of 2544 Mwt.
2. 28% fuel failed in transient,
3. 10% of iodine and noble gas activity in gap of failed fuel.
4. Release of total gap activity in failed fuel to reactor coolant.
5. Reactor coolant to secondary coolant operational leakage is 1 gpm.
6, Loss of off-site power so that steam is released from secondary
side relief valve.
7. Reactor coolant-secondary coolant equilibrium reached at 16 minutes
after the accident.
8. Standard ground level and dose conversion factors.
·15.2.5
15-7
Hydrogen Ptirge Dose
The assumptions used to calculate the low population zone doses
due to post-loss-of-coolant accident hydrogen purging are:
Power Level: 2544 Mwt
Containment Volume: 2.0 x 10 6 ft 3
Purg~ Time: 30 days
Holdup Time Prior to Purging: 11 days
Purge Rate: 32.5 cfm
Sodium Th:tosulfate Spray Reduction Factor for Iodine: 1000
Charcoal filter efficiency of 90% and 70% for elemental and
organic iodine, respectively
X/Q Value: 4 - 30 days (4.3 x 10- 7 sec/m3)
16.0
16-1
TECHNICAL SPECIFICATIONS
The Technical Specifications in a license define certain
features, characteristics, and conditions governing operation of a
facility that cannot be changed without prior approval of the AEC.
We have reviewed the proposed Technical Specifications and have held
a number of meetings with the applicant to discuss their contents
and bases. Modifications to the proposed Technical Specifications
submitted by the applicant were made to describe more clearly the
allowed conditions for plant operation. The finally approved Technical
Specifications will be made part of the operating license. Included
are sections covering safety limits and limiting safety system settings,
limiting conditions for operation, surveillance requirements, design
features, and administrative controls. On the basis of our review,
we conclude that normal plant operation within the limits of the
Technical Specifications will not result in potential offsite
exposures in excess of the 10 CFR Part 20 limits. Furthermore, the
limiting conditions for operation and surveillance requirements will
assure that necessary engineered safety features will be available
in the event of malfunctions within the plant.
17.0
17.1
17.2
QUALITY ASSURANCE
General
17-1
The Quality Assurance (QA) Program for CR-3 is described in
Section 1.7 of the FSAR, as amended. Our evaluation of the descrip
tion of the QA Program for operation of CR-3 is based on a review of
this information and detailed discussions with the applicant to determine
the ability of FPC to comply with the requirements of Appendix B to
10 CFR Part 50.
Organization
As described in the FSAR, the Senior Vice President of Systems
Engineering and Operations (see Figure 17.1) has the total responsibility
and authority to plan, organize, staff, execute and control the CR-3
QA Program. Reporting directly to him are the Assistant Vice President -
Generation Engineering and the Assistant Vice President - Systems
Operations.
The authority to plan, organize, staff, execute and control the
applicant's total QA program has been delegated to the Assistant Vice
President - Generation Engineering. He reviews and approves all
sections and subsequent revisions of FPC's QA Manual. The Director -
Generation Quality and Standards (Director) reports directly to the
Assistant Vice President - Generation Engineering and is assigned
the management of the QA program. The Director is responsible for:
1. assessing that all QA functions are being implemented,
17-2
2. conducting audits to verify and evaluate the QA program's
effectiveness,
3. the review, updating and co-approval for all sections of the
QA Manual for FPC, and
4. reporting to FPC management on the effectiveness and implementation
of the QA program.
The Director has the authority and responsibility to stop work
which, in his opinion, adversely affects the end use of safety related
structures, systems, and components during the operation of CR-3.
The Director reports to a management level that has broad
res,ponsibilities in the areas of design, procurement and operation
and that is independent of the organization directly responsible for
operational costs and schedules (Asst. Vice President - Systems Operations)
for CR-3. Therefore, we conclude that sufficient independence and authority
exists in the QA organization to establish an effective QA program for
CR-3 and to assure, through audits, that the program is carried out in
accordance with Appendix B to 10 CFR Part 50.
The Assistant Vice President - Systems Operations is responsible
for the implementation of the QA functions associated with operation,
maintenance, repair and refueling. The Superintendent of CR-3
reports to and receives technical direction from the General Plant
Superintendent who reports through the Production Superintendent to
17-3'
the Assistant Vice President - System Operations. The Superintendent
of CR-3 is directly responsible for implementing the QA program
defined in the QA manual and established by the Director of Generation
Quality and Standards. Reporting directly to the Superintendent is
the Compliance Engineer, who has responsibilities for verifying that
quality related activities performed at CR-3 comply with the QA program.
He participates in the control of nonconformance and corrective action
reports; the control of quality records; and the review of maintenance
and modification procedures to assure that adequate quality and
inspection requirements and qualified inspec~ion personnel are identified.
The Compliance Engineer is organizationally independent from the operation
and. maintenance departments thus preclud_ing undue influence on his
activities based on operational costs and schedules.
Based on our evaluation of applicant's organizational assignments
and responsibilities, we find that sufficient separation exists between
the positions responsible for quality assurance and those responsible
for operational cost.s and schedules. We conclude that the, applicant's
QA personnel, both at the offsite office and at the plant operating
level, have suffici~nt authority and organizational freedom to perform
their ,QA functions effectively and without r~servation and that the
requirements of Criterion I of Appendix B are met.
17-4
17.3 QA Program
The QA program, described in the FSAR, encompasses procedural
controls necessary to satisfy each of the eighteen criteria of
Appendix B to 10 CFR Part 50. It provides for the involvement of
all participants in quality activities in a controlled and systematic
manner during the operation of CR-3. The QA program also provides
for the verification and inspection of quality requirements of safety
related structures, systems, and components by individuals or groups
independent of those who performed the work being verified or inspected.
Provisions in the QA program require indoctrination and training
programs to be established and conducted for those personnel perform-
ing quality related activities. This training is to assure that they
are knowledgeable of the requiremen~0 .::~-;.~·-procedures,
. ···"· --- ,_ - . r~vLiClent in implementing them.
Design activities, including design changes, are procedurally
controlled in the applicant's QA program. The program requires that
applicable design bases and AEC Regulatory requirements are correctly
translated into specifications, drawings, proc<>rl11..-0 ~ ..:.. .. ~ : .. ~; '"""'=:.::::.::;;
!:~.'.!t .~.-_...;.gn verification or checking is performed; and that the
individuals or groups responsible for design verification or checking
are other than those who performed the design activity. The applicant's
Generation Engineering Department participates in the design activity,
17-5
providing independent reviews of design and drawings changes. The
Quality and Standards Departments audits the design activity, including
the design documents, to assure they are in compliance with the QA
program requirements.
For procurement activities, the QA program provides for .the
Compliance Engineer to review procurement documents, prior to·
purchase, to assure that all quality requirements are addressed.
Both the Generation Engineering and the Generation - Quality and
Standards Departments conduct vendor evaluation surveys of potential
bidders to determine. the ability of suppliers to provide acceptable
quality products. Audits and surveillance of suppliers during
fabrication, inspection, testing, and shipment of materials,
equipment, and components will be determined in advance and performed
by the applicant in accordance with written procedures. The Generation -
Quality and Standards Department and a delegated qualified QA
consultant will participate in the surveillance and audit activities
of suppliers of safety related equipment for CR-3.
The applicant's QA program also provides measures to assure that
special processes are performed by qualified personnel and accomplished
by written procedures. These procedures require recorded evidence of
verification and, if applicable, inspection and process results. In
spection operations are performed by inspection personnel who are
independent from those performing the activity being inspected, arid
17-6
in accordance with predetermined. inspection procedures. The applicant
states that testing activities will oe conducted at the CR-3 to verify
the compliance of components with design requirements •. The QA program
requires that such testing oe identified, documented, and accomplished
in accordance with written, controlled procedures and that the
inspection, tes.t and operating status of structures, systems, and
components be clearly indicated. The.QA program also provides
measures to control.the identification, documentation, segregation,
review and disposition of nonconforming materials, parts,. components
or services, including the initiating and verification of corrective
action. The .applicant ·states that a system for permanent .retention of
records in the Quality Files has been established which contain results of
personnel, procedures and equipment; drawing specifications, procurement
document, calibration procedures, calibration reports and nonconforming
and corrective ac·.tion reports •
. . The QA program requires comprehensive documented audits of all ·
quality related activities of CR-3 and at suppliers faciiities. The~e
audits are to be conducted by the Generation - Quality and Standards
Department or by a delegated qualified QA consultant to assure compliance
with all aspects of the QA program. The applicant will utilize qualified
QA.consultants having specialized skills .in the audit activity. Consultant
17-7
activities will be audited by the Generation - Quality and Standards
Department·for conformance to the applicant's overall QA policy.
The audit activity will include an objective evaluation·of ·
quality related practices, procedures, and instruction; the
effectiveness of implementation, and the conformance with policy
directives. The applicant states that audit results will be docu-
mented and r~viewed with management having responsibility in the area
audited. Deficient areas are required to be re-audited until corrections
have been adequately accomplished.
Based on our review ~f the QA program description of CR-3 as
contained in the FSAR, we conclude that the program.provides for
sufficiently detailed quality assurance requirements and controls to
fully comply with the requirements of Appendix B to 10 CFR Part 50.
7.4 Conclusion
We have performed a detailed review and evaluation of the applicant's
QA Program description, and conducted a series of discussions and meetings
with the applicant. Based on these, we conclude that the Florida Power
Corporation organization provides sufficient independence and authority
to effectively carry out the QA program for CR-3 without undue influence
ana pressure from those organization elements responsible for cost and
schedules. We further conclude that the QA program description in the
FSAR contains adequate QA provisions, requirements and controls demonstrating
17-8
compliance with Appendix B of 10 CFR Part 50 throughout the
operational phase, which includes the maintenance modification and
repair activities, of CR-3.
Asst. Vice President -
Generation Engineering
Director -Generation Quality &
Stai1dards
Senior Vice President System Engineering
& Operations
A~st. Vice President -
System Operation~
Production
Superintendent
General Plant
Superintendent
Superintendent
Crystal River Unit 3
Compliance Engineer
Fiqure 17-1 Florida Power Corporation Organization Chart
__, ....;.J I
l.O
18-1
18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARD~ (ACRS)
The report of the ACRS on the operating license review of Crystal
River Unit 3 will be placed in the Commission's Public Document Room
and at the Crystal River Public Library, Crystal River, Florida, and
will be published in a supplement to this Safety Evaluation. The
supplement will be published prior to the final determination regarding
issuance of an operating license.
19.0
19-1
COMMON DEFENSE AND SECURITY
The application reflects that the activities to be conducte0 will
be within the jurisdiction of the United States and that all ri the.
directors and principal officers of the applicant are Unite.r: States
citizens. The applicant is not owned, dominated, or cont-_·olled r y
an alien, a foreign corporation, or a foreign government. Th~
activities to be conducted do not involve any restricted ~~~a, but
the applicant has agreed to safeguard any such data wh'..r.1 might
become involved in accordance with the requirementP ri 10 CFR Part 50 •
. The applicant will rely upon obtaining fuel as -~· is needed from
sources of supply available for civilian pur~,ses, so that no
diversion of special nuclear material frr military purposes is
involved. For these reasons and in t-.;e absence of any information to
the contrary, we find that the ?~·' .ivities to be performed will
not be inimical to the connnC'.t ..iefense and security.
20-1
20.0 FINANCIAL QUALIFICATIONS
The Commission's regulations which relate to financial data and
information required to establish the financial qualifications of an
applicant for a facility operating license are 10 CFR. Section 33(f)
and 10 CFR 50, Appendix C. We have reviewed the financial information
presented in the application and have concluded that the applicant is
financially qualified to operate CR-3. We have also examined the
Annual Report for Florida Power Corporation for 1973; our examination
does not cause us to change our judgment of the applicant's financial
qualifications. A detailed discussion of the basis for our conclusion
is presented in Appendix C.
21.0
21.1
21-1
FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS
Pursuant.to the financial protection and indemnification pro
visions of the Atomic Energy Act of 1954, as amended (Section 170 and
related sections), the Commission has issued regulations in 10 CFR
Part 140. These regulations set forth the Commission's requirements
with regard to proof of financial protection by, and indemnification
of, licensees for facilities such as power reactors under 10 CFR
Part 50.
Preoperational Storage of Nuclear Fuel
The Commission's regulations in Part 140 require that each holder
of a construction permit under 10 CFR Part 50, who is also the holder
of a license under 10 CFR Part 70 authorizing the ownership and pos
session for storage only of ~pecial nuclear material at the reactor
construction site for future use as fuel in the reactor (after
issuance of an operating license under 10 CFR Part 50), shall, during
the interim storage period prior to licensed operation, have and
maintain financial protection in the amount of $1,000,000 and execute
an indemnity agreement with the Commission. Proof of financial
protection is to be furnished prior to, and the indemnity agreement
executed as of, the effective date of the 10 CFR Part 70 license •.
Payment of an annual indemnity fee is required.
Florida Power Corporation has furnished to the Commission proof
of financial protection in the amount of $1,000,000 in the form of
21. 2
21-2
a Nuclear Energy Liability Insurance Association policy (Nuclear Energy
Liability Policy, facility form) No. NF-195.
Further, Florida Power Corporation executed Indemnity Agreement
B-54 with the Comiilission as of June 20, 1973, the effective date of its
preopPrational fuel storage license, SNM-1275. Florida Power Corporation
has paid the annual indemnity fee applicable to preoperational fuel
storage.
Operating License
Under the CoIIllllission's regulations, 10 CFR Part 140, a license
authorizing the operation of a reactor may not be issued until proof
of financial protection in the amount required for such operation
has been furnished, and an indemnity agreement covering such operation
(as distinguished from preoperational fuel storage only) has been
executed. The amount of financial protection which must be maintained
for CR-3 (which has a rated capacity of more than 8&~,000 electrical
kilowatts) is the maximum amount available from private sources, i.e.,
the combined capacity of the two nuclear liability insurance pools,
which amount is currently $110 million. Accordingly, no license
authorizing operation CR-3 will be issued until proof of financial
protection in the requisite amount has been received and the requisite
indemnity agreement executed.
21.3
21-3
We expect that, in accordance with the usual procedure, the nuclear
liability insurance pools will. provide, several days in advance of
anticipated issuance of the operating license document, evidence in
writing, on behalf of the applicant, that the present coverage has been
appropriately amended so that the policy limits have been increased, to
meet the requirements of the Commission's regulations for reactor operation.
Similarly, no operating license will be issued until an appro
priate amendment to the present indemnity agreement ·has been executed.
Florida Power Corporation will be required to pay an annual fee for
operating license indemnity as provided in our regulations, at the rate
of $30 per each thousand kilowatts of thermal capacity authorized in
its operating license.
Conclusions
On the basis of the above considerations, we conclude that the
presently applicable requirements of 10 CFR Part 140 have been satisfied
and that, prior to issuance of the operaitng licenses, the applicant
will be required to comply with the provisions of 10 CFR Part 140
applicable to operating licenses, including those as to proof of
financial protection in the requisite amount and as to execution of an
appropriate indemnity agreement with the Commission.
22.0
22-1
CONCLUSIONS
Based on our evaluation of the application as set forth above,
we have concluded that:
1. The application for a facility license filed by the Florida
Power Corporation dated February 8, 1971, as amended (Amendments
Nos. 1 through 39) complies with the requirements of the Atomic
Energy Act of 1954, as amended (Act), and the Commission's regulations
set forth in 10 CFR Chapter l; and
2. Construction of Crystal River Unit 3 (the facility) has proceeded
and there is reasonable assurance that it will be substantially
completed, in conformity with Provisional Construction Permit No.
CPPR-57, the application as amended, the provisions of the Act,
and the rules and regulations of the Commission; and
3. The facility will operate in conformity with the application as
amended, the provisions of the Act, and the rules and regulations
of the Commission; and
4. There is reasonable assurance (i) that the activities authorized
by the operating license can be conducted without endangering
the health and safety of the public, and (ii) that such activities
will be conducted in compliance with the regul.ations of the
Connnission set forth in 10 CFR Chapter l; and
5. The applicant is technically and financially qualified to engage
in the activities authorized by this license, in accordance with
the regulations of the Commission set forth in 10 CFR Chapter l;
and
I
22-2
6. The issuance of this license will not be inimical to the connnon
defense and security or to the health and safety of the public.
Before an operating license will be issued to the Florida Power
Corporation for operation of Crystal River Unit 3, the unit must be
completed in conformity with the provisional construction permit,
the application, the Act, and the rules. and regulations of the Connnission.
Such completeness of construction as is required for safe operation
at the authorized power level must be verified by the Counnission's
Directorate of Regulatory Operations prior to license issuance. In
addition, satisfactory resolution of outstanding matters discussed herein
will be required.
Further, before an operating license is issued, the applicant will
be required to satisfy the applicable provisions of 10 CFR Part 140.
APPENDIX A
CHRONOLOGY
REGULATORY REVIEW OF FLORIDA POWER CORPORATION
CRYSTAL RIVER, UNIT 3
1. February 8, 1971 ·
2. April 13, 1971
3. May 1, 1971 ''
4. June .28, 1971
5. August 9, 1971
6. September 27, 1971
7. October 15, 1971
8. November 11, 1971
9. November 11, 1971
10. January 17, 1972
Submittal of Amendment No. 11 replacing PSAR in
its entirety.
Letter from Consultant, John A. Blume regarding
review of seismic analysis.
Letter to applicant transmitting Federal Register
Notice regarding Application for Operating License.
Submittal of Amendment No. 12 consisting of
answers to questions requested by Attorney General.
Letter to appliCant transmitting adopted interim
acceptance criteria for .the performance of ECCS.
Submittal ·Of Amendment No. 13 consisting of
revised pages of the PSAR.
Letter from the applicant regarding SHOW CAUSE.
Submittal of Amendment No. 14 consisting of
revised· and additional pages to the PSAR.
Submittal of Amendment No. 15 consisting of answers
to questions regarding SHOW CAUSE.
Letter to the applicant requesting additional
information.
11. February 1, 1972
12. February 8, 1972
13. February.10, 1972
14. March 20, 197Z
15. Marc.h 30, 1972
16. April 10, 1972
. '' ~, .
17. Apri.l 10, 1972
18 •. April 11, 1972
19 •. April 25, 1972
20. May 2, 1972
21. May 8, 1972
22. May 8, 1972
- 2 -
Letter from the applicant furnishing additional
information.
Letter from applicant furnishing additional information.
L~tter to the applicant transmitting PWR Inservice
Inspection Program •
. Letter from applicant regarding PWR Inservice In
spection Program.
Submittal of Amendment No. 16 consisting of revised
and additional pages to the PSAR, Statistical Report,
and the Annual Financial Report for 1971.
Submittal of Amendment No. 17 consisting of revised
and additional pages to the PSAR .
Letter to applicant requesting additional information.
Letter .to applicant transmitting Sr.aft ,Criteria on
Industrial Security.
Letter from th~ applicant furnishing information
for-the FSAR.
Letter to the applicant requesting additional
information.
Submittal of Amendment No. 18 consisting of three
reports.
Submittal of Amendment ~o. 19 consisting of
responsed to AEC's April 10, 1972 letter.
23. May 18, 1972
24. May 19, 1972
25. June 1, 1972
26. June 1, 1972
27. June 9, 1972
28. June 15, ·1972
29. June 16, 1972
30. July 6, 1972
31. July 7, 1972
32. July 17; 1972
33. July 24, 1972
34. August 15, 1972
- 3 -
Submittal of Amendment No. 20 consisting of
Supplement No• 1 to Environmental Report. ·
Le.tter from applicant in response to AEC 's
April 10 and May 2, 1972 letters.
Letter to applicant requesting compilation of all
applicable B&W Topical Reports.
Letter to applicant transmitting order extending
construction completion date. .. ;'[.
Letter· from applicant advising that info regarding
B&W Topical· Report's to be submitted by 7/3/72.
Subni.ittal'of Amendment No. 21 consisting of revised
arid ·additional pages to th~ FSAR.
Letter to ACRS transmitting copies of revised pages
to the FSAR and Supplement No. 1.
Letter from applicant transmitting B&W Topical
Report~
Letter to applicant advising of postponing
continuation of review of FSAR.
Letter to applicant transmitting B&W letter requesting
additional information on thei~ topical report.
Letter to applicant requesting additional information.
Letter from applicant regarding earliest response date.
35. August 23, 1972
36. August 29, 1972
37. September 21, 1972
38. September 26, 1972
39. October 12, 1972
40. October 12, 1972
41. October 13, 1972
42. October 16, 1972
43. October 18, 1972
44. October 20, 1972
- 4 -
Letter to ACRS transmitting B&W Topical Reports.
Submittal of Amendment No. 22 consisting of
new Volume 5 containing Supplement No. 2 to
Environmental Report and also Supplement No. 3
to Environmental Report ..
Letter to applicant requesting action to be taken
regarding outstanding safety issues.
Letter to applicant requesting information
relative to design regarding non-Category I
(seismic) equipment.
Letter to applicant transmitting Federal Register
Notice of Consideration of Issuance of Facility
Operating License and Notice of Opportunity of
Hearing.
Letter to applicant regarding Topical Report
BAW-10047, Revision 1.
Letter to applicant confirming meeting of 10/27/72. . .
Letter from applicant submitting report on delays
in schedule.
Letter to applicant regarding Topical Report
BAW-10013.
Letter to applicant regarding Topical Reports
BAW-10037, BAW-10038, BAW-10050 and BAW-10051.
- 5 -
45. October 31, 1972 · Letter from applicant regarding water and fire
protection system.
46. November 20, 1972 Letter to applicant transmitting Technical Report
" on Densification of Light Water Reactor Fuels.
47. December 15, 1972 Letter to applicant regarding postulated pipe
failures and main steam.or feedwater lines.
48. December 21, 1972 Letter from applicant regarding non-category I
(seismic) equipment.
49. December 27, 1972 Submittal of Amendment No. 23 consisting of
revised and additional pages and-. Supplement Ill.
50. January 2, 1973 Letter from applicant furnishing information on
postulated pipe failures.
51. January 24, 1973 Letter to applicant regarding Topical Report
BAW-10029.
52. February 27, 1973 Letter from applicant regarding current position
on the re-evaluation of the probable maximum hurricane
surge height.
53. March 7, 1973 Letter to applicant regarding deficiency in control
circuit design.
54. March 9, 1973 Submittal of Amendment No. 24 consisting of New
Volllllle No. 6 to FSAR.
55. March 12, 1973 Letter to applicant requesting additional information.
56. March 15, 1973 Lette~ from applicant regarding fuel densification.
57. March 19, 1973
58. March 26, .1973
59. April 2, 1973
60. April 2, 1973
61. April 2, 1973
62. April 4, 1973
63. April 12, 1973
. 64. ·April 18, 1973
65. April 23, 1973
66. April 24, 1973.
67. April·27, 1973
- 6 -
Letter from applicant advising that info not
included in Amendment 25.
Letter from applicant furnishing info regarding
hurricane protection.
Submittal of Amendment No. 25 consisting of New
Volume No. 7 to FSAR.
Letter from applicant furnishing additional info
regarding hurricane protection.
Letter to applicant -regarding fuel densification.
Letter to applicant regarding cirrent review
schedule.
Letter to applicant regarding hurricane-related
design requirements .
Letter from applicant adivsing of intent to co11DDence
excavation of extensions of present intake canal
and discharge canal.
Letter from applicant regarding inadvertent
disabling of components by racking out of circuits
breakers.
Letter from applicant advising that requested info
to be submitted as amendment.
Letter to applicant requesting additional info.
- 7 -
68. May 15, 1973 Letter from applicant advising of earliest possible
date for completely adequate response.
69. May 25, 1973 Submittal of Amendment No. 26 consisting of revised
and additional pages to Supplement No. 1.
70. June 29, 1973 Submittal of Amendment No. 27 consisting of answers
to questions of 4-20 and 4-21-73.
71. July 9, 1973 Letter to applicant regarding our review of BAW-1403.
72. July 13, 1973 Letter to applicant requesting additional info.
73. July 30, 1973 Letter to applicant requesting additional informa-
tion and additional financial information,
74. July 31, 1973 Letter to applicant transmitting Regulatory Staff
Positions.
75. August 1, 1973 Submittal of Amendment No. 28 consisting of revised
and added pages to the FSAR and Dames & Moore Report
on Hurricane Study.
76. August 10, 1973 Letter from Dames & Moore transmitting reports,
77. August 15, 1973 Submittal of Amendment No. 29 consisting of revised
and additional pages to the FSAR.
78. August 22, 1973 Letter to applicant transmitting Amendment to 10
CFR Parts 50 and 55. I
79. August 30, 1973 Submittal of Amendment No. 30 consisting of revised
and additional pages to the FSAR and Topical Report
BAW-1397.
80. August 31, 1973
81. September 1, 1973
82. September 12, 1973
83. September 24, 1973
84. October 1, 1973
85. October 1, 1973
86. October 4, 1973
87. October 9, 1973
88. October 9, 1973
89. October 9, 1973
- 8 -
Submittal of Amendment No. 31 consisting of revised
and additional pages to the FSAR and the Annual
financial report and Technical qualifications.
Letter to applicant requesting additional info.
Letter from ~pplicant requesting info. on Founda
tion problems and transmitting report on foundation
problems.
Letter from applicant.concerning the design of the
borated water storage tanks.
Submittal of Amendment No. 32 consisting of revised
pages to the FSAR and report on piping system breaks.
Letter from applicant transmitting report on high
energy piping systems.
Letter from Coastal Engineering Research Center
submitting reconunendations for the FSAR.
Submittal of Amendment No. 33 consisting of revised
and additional pages to the FSAR.
Letter from Department of the Army regarding review
of Dames & Moore's V~rification Study of Hurricane
Storm Suge Model.
Letter to applicant transmitting Technical Report on
ATWS for Wat~r-Cooled Power Reactors.
90. October 12, 1973
91. October 15, 1973
92. October 19, 1973
93. October 23, 1973
94. November 12, 1973
95. November 15, 1973
96. November 30, 1973
97. December 5, 1973
98. December 18, 1973
99. December 21, 1973
- 9 -
Letter to applicant requesting additional informa
tion regarding hydrologic engineering and hurricane
verification.
Letter from applicant transmitting Security Plan.
Letter from Department of the Army regarding review
of Dames & Moore's Verification Study of Hurricane
Storm Surge Model.
Letter from applicant furnishing irifo regarding new
meteorological data acquisition system.
Letter from applicant submitting Revision No. 1 to
report on Effect of High Energy Piping Systems.
Submittal of Amendment No. 34 consisting of revised
and additional pages to the FSAR.
Submittal of Amendment No. 35 consisting of revised
and additional pages to the FSAR.
Letter dated 12-5-73 regarding inspection of hydrau
lic shock suppressors (snubbers).
Letter to applicant reaffirming position for need
to modify Industrial Security Plan.
Submittal of Amendment No. 36 consisting of revised
and additional pages to the FSAR and revision to
Section 2.6 Site Environmental Radiological ~onitoring
Program.
100. December 27, 1973
101. December 28, 1973
102. January 2, 1974
103. January 22, 1974
104. February 4, 1974
105. February 12, 1974
106. February 14, 1974
107. February 15, 1974
108. February 19, 1974
109. February 19, 1974
110. March 18, 1974
111. March 25, 1974
- 10 -
Letter from applicant transmitting Revision No. 1
to the Security Plan.
Letter from applicant regarding WASH-1270, ATWS,
advising that Crystal River is a Category C Plant.
Letter from Department of the Army furnishing info.
regarding site hydrology.
Letter to applicant requesting additional info.
Submittal of Amendment No. 37 consisting of revised
pages to the FSAR and Revision of Appendix 14C to
the Meteorological analysis of periodic venting.
Letter from applicant regarding snubbers.
Letter from applicant transmitting Revision No. 2
to report regarding effect of high energy piping.
Letter to applicant regarding section of TS being
unacceptable.
Letter from applicant furnishing info on quality
assurance.
Letter to applicant regarding the Nuclear Service
Sea Water Piping.
Letter from applicant furnishing info regarding
Byproduct Material License.
Submittal of Amendment No. 38 consisting of revised
and additional pages to the FSAR.
112. April 8, 1974
113. April 30, 1974
114. May 3, 1974
- 11 -
Letter from applicant regarding inspection of snub
bers.
Letter from applicant regarding trip delay time.
Letter to applicant regarding quality assurance
personnel.
CEREN-DE
Mr. Harold R. Denton
APPENDIX B
DEPARTMENT OF THE ARMY COASTAL ENGINEERING RESEARCH CENTER
KINGMAN BUILDING
FORT BELVOIR, VIRGINIA 22060
Assistant Director for Site Safety U. S. Atomic Energy Commission Washington, D.C. 20545
Dear Mr. Denton:
50-302
2 O DEC 1973
Reference is made to your letter of 28 August 1967 initiating review of Docket No. 50-302, the Florida Power Corporation's, Crystal River Nuclear Generating Plant, Unit No. 3 including Amendments thereto through No. 34.
In accordance with our arrangements, an engineer on my staff has reviewed pertinent information in the report leading to the establishment of the maximum and minimum design water levels for the Crystal River Plant. This review covered the applicant's determination of the surge associated with the occurrence of the Probable Maximum Hurricane at the site as well as wind-wave generation, wave runup and overtopping also associated therewith.
It is his opinion, with which I concur, that the maximum and minimum stillwater levels proposed by the applicant as the result of the passage of the Probable Maximum Hurricane are El. 33.4 feet, MLW datum (El. 121.4 ft., plant datum) and El. -9.0 feet, MLW datum (El. 79.0 ft., plant datum) respectively.
I also concur with the applicant's analysis that a flood protection level of El. 41.0 feet, MLW datum (El. 129.0 ft., plant datum) ·for essential facilities is sufficient to withstand the maximum limit of wave runup coincident with the hurricane induced water levels.
CEREN-DE Mr. Harold R. Denton
2 0 DEC 1973
It is further pointed out that those structures exposed to direct wave attack are subject to dynamic forces. The dynamic forces can greatly exceed those static forces determined by assuming a static water level associated with the flood protection level.
If I can be of assistance in the safety evaluation of this plant please let me know.
CF: Mr. L. G. Hulman/AEC
2
Sincerely yours,
1/) ~· ~ R~~TE~ Lieutenant Colonel, Corps of Engineers Acting Commander and Director
--------------- B-1 APPENDIX B
. October 17~ 1974
Honornblc Dixy Lee: Ray Cb.'.l ir~::a•: U. S. Ato~ic [ncr;~ Cc~~ission lfashingt:o:-t, D. C. 20.J-'i) . \ . Sub j(:C: 1:: l~~L'O?~"i' o:; Cf\.";'ST:\L l:.l\.'ER ~~UCLEli..l{ GE:-;r:?J~:u:~G Pll~::T, U:iIT 3
Dear Dr:. Ray:
At :i.ts 174th :::ce:t:in;, O;::to:J;:;r 10·-12, 197!;, the· ,..,_~~viso::-y Co::-c::5.tte:c on Reactor Safegu~rd.s ec~~pl\:tc=d its rcvic\1 of the ~~p;~1 i.ct1Li~n c-[ !:!~e Flo!'.":id;, l'u.-;.::r Co::~~o::.'."!tion fo:: .:i l:i.c0nsc to 01:.:.:::.::~c: th.:': Cryst.:-:l ::.:.·;er lluclc.::.r Gc:nc::.-.:tti.w_: i.>}:-!:-it, t:n:!.t 3, 3.t FC'.·li.!r h'v0l~ ~~i' .to 2:.s.: :-:::(t)". 'fhis p;:oj8ct t-:::s coas!d(:t-ed c:i..~ri.~1~·~ .:i S1J:.1cG::~:~·it:t{!c ::cet~n::: in t:..J~;:1i.r.;to~, D C 0 " l L· 1 v '> q 1 (-; -1 l· .., "cl •• "' ; t .. "i· s -i t· \ ·.., '"" •·• •• •1 ·• ·• n J · ' 1 · · c~ , c• ., t, I ·1 e • ) •, .._ ...... J - .. ) ..,. r , c.... ... '.. .:J .l. \.. "' ,. ... .. t.l. ~ 1 •• ~ u \..: ,,,. l u - ) ..,.. ) .L .> I . • . I
the course of tl:c rc:-.1i.c:·.-:, U:.:!-CC.':;-.:~itt..::c !1.:-;d liw tlc:~·~~:.t o:' di.:>cu:~~;i0:::, 1"tl --. r"\C::"1 ~"'I~·-·~~ -':"'"(~ 0'"'<"··-1·---'""'•-("' or t;,,.., -·10-: .. =·1 ~)c··.r.-,- c ......... ~,- ___ :.:."'., \,: J ) 1. l. p ~ t_ ._... t..: n Lt .. L 1 \I '. d c..&. •' - C. L l •-' \. l- 'I.• ... ~•I 1. - •. "- ! ,.. - '-J '-• - •• ..._. - ._ .._ :· .. > • ... "- • '- ~ • )
t·l1"' B-.1-~'J'"!r ..,...,r, i•':1,.c, .. Cr-.·-·1··--·· c:it...·.-·t .'.-c-··c1·.., ..... ,. I-c "'r··1 r- 1· .• -. ·''C . .... ) -- • : '-; -· • .. (.... :....! 10 - .. - ••• • - . ;, ••.•• • ' , .l. .. J \.. •. .. "'_') ~·... .. ....... l. L.: .,., ). . • ~ • , <.. .. ... • .. •• - .• '\ •• J
Rq;ul.1tc,r;· St.:i.ff;. The C.::-::::15.ttce: '1lso i:.::d .. tl1c b.:~1H.'f!t o'.: the c.!c:::::..::::cats
listed bol0·.1.
The plo.n~ is locnlcd on tl:c Gulf o( ~-~e:dco i~ Citn;s Ccv.;~1ty, Flo:-i.~J.,
about 7-1/'2 r.:i.lC!s nort!:· .. :.:::s:: of tbc tO\·;n o[ Cryst.:i.l ~~i·,·cr. i.:hc site cc-::;priscs 473S acr2s and incl~~cs Units land 2 ~hie~ .::.r2 oil-fir~d. Tha 1nini~u::t c:-:c lt:s f.oi.1 ci :.~ t:'.:!::.::..~ is ·!;~~co £e'2 t, ~r-.d t:1c r.:;d ius of tl!c lo: .. " f-O?!J·· .. lation zcne has been selected as 5 ~ilcs. The cooling 0atcr int~k~ cn~~l cxten~s ~bout 14 ~ilcJ into th~ Gulf, abou:: 10 f:l"i.lcs •.
and t ;," ··- di.s~h~1r~·~ c.::.:::tl c.:·: t •2 !'"-~ 5 C'...! t
Ihe prot~ction a~ainst flooding h~s been cxpn~ded to meet tbe csti~3ted tnaxir..u;:i hun:ic.:inc-i.r.ct.:ced SLlrgc level with a~ditic:12l w:.ve ru::-t.ip heis~ts, using p~r~2etcrs more conservative than those idcntif icd at the constn1ction st'-lge.
·• L"&i·.
or Lit:.:li~I n~ JI .\·Ii .j D ;J {C.C--..!.I_:.'.-,;.../..__ 1 _ __,_....,.. ... . ...... 'J.'ilnC--.-'-·--- .;...,..,..r-.-u
. " I . ( 1
·--' ....... - ,. ") •-f _ _,
/ '/.!" • • • • /,: /.'..I! 1 l d ! ( ; d I . . ~ . I , \
-,.
' (,
< ...
C-1
APPENDIX C
FINANCIAL QUALIFICATIONS
The Commission's regulations which relate to financial data
and information required to establish financial qualifications for
applicants for operating licenses are Section 50.33(f) of 10 CFR
Part 50 and Appendix C to 10 CFR Part 50. We have reviewed the
financial information presented in the application and amendments
thereto regarding financial qualifications. Based on this review
we have concluded that Florida Power Corporation possesses or
can obtain the necessary funds to meet the requirements of 10 CFR
50.33(f) to operate the Crystal River Unit 3 Nuclear Generating
Plant and if necessary permanently shut down the facility and main
tain it in a safe shutdown condition.
Crystal River Unit 3 Nuclear Generating Plant will be used
to augment the applicant's present electrical generating capacity.
At the time CR-3 is proposed to be placed in service in 1974, it
will represent, according to the applicant, 24.5% of the applicant's
total net maximum dependable generating capability of 3,365,000 KW
including CR-3 and assuming no facility retirements. Operation
and maintenance costs, including fuel costs, during the first five
full years of commercial operation of Unit 3 (1975-1979) are
presently estimated by the applicant to be (in millions of dollars)
C-2
$11.4; $12.1; $12.4; $15.1; and $15.5 in that order. Annual
fixed costs on a levelized basis for the same period are estimated
by the applicant at $51.0 million. Assuming a plant factor of 80%,
total annual costs to operate CR-3 range from 1.01¢ per kwh for
1975 to 1.07¢ per kwh for 1979. Estimated fuel costs range from
0 .130¢ per kwh for 1975 to 0 .177¢, per kwh for 1979.
The applicant's levelized annual fixed costs of 17.01% (actually
17.0088%), which was applied to the estimated $300 million cost of
CR-3 excluding the cost of the initial core, consist of the following
components: equivalent return on investment: 7.6748% (cost of
capital as computed by applicant plus the depreciation annuity and
less straight-line depreciation); Federal income taxes (FIT):
1.5512%; FIT credit for liberalized depreciation: - 1.4435%; FIT
credit for the investment tax credit: - 0.3070%; depreciation on
the straight-line basis: 3.333% (30 years); insurance: 0.7000%; and
other taxes: 2.5000%.
Revenues from the sale of electric power to retail and wholesale
customers are expected to provide funds necessary to cover the total
costs estimated to be applicable to CR-3. Assuming a plant factor
of 80%, total annual costs to operate CR-3, as noted above, range
from 1.01¢ per kwh for 1975 to 1.07¢ per kwh for 1979. These unit
costs are substantially.below the unit price of 1.75¢ per kwh ex
perienced by the applicant on its 1972 system-wide sales of electric
power to retail and wholesale customers.
C-3
The cost of permanently shutting down CR-3 is estimated by
the applicant at $750,000 based upon leaving the reactor and its
associated nuclear systems in place and salvaging the secondary
side of the plant, with all nuclear fuel removed from the plant
and sent off site for final reprocessing. The annual cost of
maintaining the subject facility in a safe shutdown condition
is estimated by the applicant at $50,000 based upon isolating
the plant area by suitable fencing and monitoring the area
periodically by guards. The source of funds to cover these costs
is expected to be obtained from revenues derived from sale of
electric power to retail and wholesale customers.
The applicant states that uranium for the first core and for
subsequent core will be purchased on the open market and then
toll enriched.
We have examined the financial information submitted by Florida
Power Corporation to determine whether it is financially qualified
to meet the above estimated costs. The information presented in
Florida Power Corporation's annual report for 1972 indicates that
operating revenues totaled $201.9 million. Operating expenses were
stated at $148.8 million, of wh~ch $22.8 million represented
depreciation. Interest on long-term debt was earned 2.8 times. Net
income totaled $42.0 million, of which $22.9 million was distributed
as dividends to stockholders with the remaining $19.1 million
c..,.4 ..
retained .. for use in- the, ,business •. As o.f. December 31, 197.2, the ~ • • 1 -·
. , Company's. assets. totaled. $9,81. 2: million, most of :which. vr~s iI).yes ted - • • • • • • • • • : J • • - • - ... .' ' -~~- ••••
in.utility ·plant. ($937. 2 _ ~illion) • . Retai:ried earnings .. aw.ouI)._t~~
to $143. 8 mill.io:1;1. .Financial ra.tios computed from the 1972 statements '· . . - ... '··: ... ; - .. : .. • .. . ·' .- ' - ·~ _. .
indica t;e an ./ildequa te financial_ C()nd;i. ti on, e.g. , . lor:i.g-term deb.t ... ' . ._, - . . -· ;_.. \ ·. . •'
to to.ta! capit_alization - 56_%, ... a~d to. n~~- u_tflity. p_lant._~ .. 50%;
net pl.ant to,,c~p_j,ta~ization.:- 1.1_~; ~he .. p_per.;itin~ :r.:at,io -:-J4%,; and
the rates of return on common equity ,~ 13. 3.%,. ,on stockh~lders'
investment_ :- 1],. 5%, . \ind on total_ investment, - 7 ._0% •.. The re~pi:-d of
the Company's .operations during 19 7 07"" 72 shows that opera ting, . . . - . . . ' .. - . ' .. '
revenues increas_ed from $15~.-~1 ~illion in :19~0. tp ~201.,9" ~~ll,ion
in 197_2; net income increased, .fro~: $31.3 .millio~ .to. -~42 .~"-million;
and net investment in utility._plant,.from $640.2 million to. $937 .2
million. However, the number of times interest ear-q.ed. ~~c,lined
from 3. 2 to 2. 8 •. , .Moody.' .s. Inves,tors Service rates the Copipany' s
first mortgage bonds as Aa (high grad~ bonds) and its convertible • - ... • •. . '~ ,, .· • ' ,: : . • f:_·. ·:· •. ,
debentures as _A_ ,_(upper middle grade obligations) •. The Company's
~.current Dun a~d~ Br,adstreet rating is 5Al,, the highe~t r,ating .
. A summ<lry -:nalysis reflectir:i.g these ratios and _other. pertinent
data is attached.
C-5 FLORIDA POWER CORPORATION
FINANCIAL ANALYSIS DOCKET NO. 50-302
(dollars Calendar Year Ended
in millions) December 31
Long-term debt Utility plant (net
Ratio - debt to fixed plant
Utility plant (net) Capitalization
Ratio of net plant to capitalization
Stockholders; equity Total assets
Proprietary ratio
Earnings available to connnon equity Connnon equity
Rate of earnings on connnon equity
Net income Stockholders; equity
Rate of earnings on stockholders; equity
Net income before interest Liabilities and capital
Rate of earnings on total investment
Net income before interest Interest on long-term debt
No. of times long-term interest earned
Net income Total revenues
Net income ratio
Total utility operating expenses Total utility operating revenues
Operating ratio
Utility plant (gross) Utility operating revenues
Ratio of plant investment to revenues
Capitalization: Long-term debt Preferred stock Common stock & surplus Total
1972 . 1971 $ 469.3 $ 374.2
937~2 778.4 .50 .48
937.2 778.4 834.5 690.3
1.12 1.13
365.2 316.1 981.2 817.5
.37 .39
37.5 31.9 281.7 262.6
13.3% 12.1%
42.0 35.2 365.2 316.1
11.5% 11.1%
68 .6 55.9 981.2 817.5
7.0% 6.8%
68.6 55.9 24.2 18.8 2.8 3.0
42.0 35.2 217.4 185.7
.19 .19
148.8 129.8 201.9 176.5
.74 .74
1,105.6 927.7 201.9 176.5
5.48 5.26 1972
1970 $ 329.8
640.2 .52
640.2 588.3
1.09
258.5 676.9
.38
29 .8 205.0 14.5%
31.3 258.5 12.1%
48.5 676.9
7.2%
48.5 15.2
3.2
31.3 162.8
.19
114.3 158.1
.72
777 .5 158.1
4.92 1971
Amount % of Total Amount % of Total $469.3 56.2% $374. 2 54.2%
83.5 10 .o 53.5 7.8 281.7 33.8 262.6 38.9
$834.5 100.0% $690.3 100.0%
Moody's Bond Rating: Mortgage Aa, Debentures A Dun & Bradstreet Credit Rating: 5Al
Meteorology
APPENDIX D BIBLIOGRAPHY
Alaka, M. A. 1968: Climatology of Atlantic Tropical Storms and Hurricanes. ESSA Technical Report, WB-6, Techniques Development Laboratory, Silver Spring, Maryland.
Cry, G. W., 1965: Tropical Cyclones of the North Atlantic Ocean. Technical Paper No. 55, U.S. Department of Connnerce, Weather Bureau, Washington, D. C.
Gross, E., 1970: The National Air Pollution Potential Forecast Program. ESSA Technical Memorandum WBTM NMC 47, National Meteorological Center, Washington, D. C.
Holzworth, G. C., 1972: Mixing Heights, Wind Speeds, and Potential for Urban Air Pollution Throughout the Contiguous United States. AP-101, Environmental Protection Agency, Office of Air Programs, Research Triangle Park, North Carolina.
Huschke, R. E., 1959: Glossary of Meteorology. American Meteorological Society, Boston, Massachusetts.
Korshover, J., 1967: Climatology of Stagnating Anticyclones East of the Rocky Mountains, 1936-1965. Public Health Service
.Publication No. 999-AP-34, Cincinnati, Ohio.
List, R. J. (ed.), 1971: Smithsonian Meteorological Tables. Smithsonian Institution, Washington, D. C.
Memorandum HUR 7~97, 1968: Interim Report - Meteorological Characteristics of the Probable Maximum Hurricane, Atlantic and Gulf Coasts of the United States. From the Hydrometeorological Branch, Office of Hydrology, U.S. Weather Bureau to the Corps of Engineers.
Hydology
Memorandum HUR 7-97A, 1968: Asymptotic and Peripheral Pressures for Probable Maximum Hurricanes. From the Hydrometeorological Branch, Office of Hydrology, U.S. Weather Bureau to the Corps of Engineers.
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Pasquill, F and Smith, F. B., 1970: The Physical and Meteorological Basis for the Estimation of Dispersion. Paper presented at the Second International Clean Air Congress of the International Union of Air Pollution Prevention Associations, Washington, D. C.
SELS Unit Staff, National Severe Storms Forecast Center, 1969: Severe Local Storm Occurrences, 1955-1967. ESSA Technical· Memorandum WBTM FCST 12, Office of Meterological Operations, Silver Spring, Maryland·.
Simpson, R. H. and Lawrence, M. B., 1971: Atlantic Hurricane Frequ.encies Along the U.S. Coastline. NOAA Technical Memorandum NWS SR-58, Southern Re·gion~ National Weather Service, Fort Worth, Texas.
Slade, D. H. (ed.), 1968: Meteorology and Atomic Energy-1968. TID-24190, National Technical Information Service, Springfield, Virginia.
Thom, H. C. S., 1963: Tornado Probabilities. Monthly Weather Review, October-December 1963, pp 730-737.
Thom, H. C. S., 1968: New Distributions of Extreme Winds in the United States. Journal of the Structural Division, Proceedings of the American Society ·of Civil Engineers - July 1968, pp 1787 - 1801.
Turner, D. B., 1970: Workbook of Atmospheric Dispersion Estimates. Public Health Service Publication No. 999-AP-26, Cincinnati, Ohio.
U.S. Department of Connnerce, Environmental Data Service: Local Climatological Data, Annual Sunnnary with Comparative Data -Tampa, Florida. Published annually through 1972.
The Computer program used to calculate the X/Q values may be referenced as follows: Nuclear Power Station Evaluation Program (in FORTRAN code), J. R. Safendorf, ARL/NOAA progrannner. Program available at the USAEC, Directorate of Licensing, Bethesda, Md., or at the Air Resources Laboratory, NOAA, Field Research Office, Idaho Falls, Idaho.
Structural Engineering
"Wind Forces on Structures," Final Report of the Task Connnittee on Wind Forces of the Connnittee on Load and Stresses of the Structural Division, Transactions of the American Society of Civil Engineers, 345 East 47th Street, New York; N. Y. 10017, Paper No. 3269, Vol. 126, Part II, 1961, p. 1124-1198.
D-3
A. Amirikian, "Design of Protective Structures," Bureau of Yards and Docks, Publication No. NAVDOCKS P-51, Department of the Navy, Washington, D. C., August 1950.
National Defense Research Committee, Effects of Impact and Explosion, Summary Technical Report of Division 2, Vol. 1, Washington, D. C., 1946.
R. C. Gwaltney, "Missile Generation and Protection in Light-WaterCooled Power Reactor Plants," USAEC Report ORNL-NSIC-22, September 1968.
"Structures to Resist the Effects of Accidental Explosions," TM 5-1300, NAVFAC P-397, or AFM 88-22, Departments of the Army, the Navy and the Air Force, June 1969.
American Institute of Steel Construction, "Specification for Design, Fabrication and Erection of Structural Steel for Buildings," 101 Part Avenue, New York, N. Y. 10017, 1963.
American Concrete Institute, "Building Code Requirements for Reinforced Concrete (ACI 318-63 and -71),: P. 0. Box 4754, Redford Station, Detroit,·Michigan 48219.
American Society of Mechanical Engineers and the American Concrete Institute, "Proposed Standard Code for Concrete Reactor Vessels and Containments," United Engineering Center, 345 East 47th Street, New York, N. Y. 10017.
American Society of Mechanical Engineers, "ASME Boiler and Pressure Vessel Code," Section III, United Engineering Center, 345 East 47th Street, New York, N. Y. 10017.
Instrumentation, Controls and Electrical Systems
Sections 6, 7, 8, 9 and 10 of FSAR for Arkansas Nuclear One, Unit 1.
Operating License Safety Evaluation Report for Arkansas Nuclear One, Unit 1, issued June 6, 1973.
Babcock & Wilcox (B&W) Schematic Diagrams for the Reactor Protection System.
Gilbert Associates, Inc. (GAI) Elementary Diagrams for the Engineered Safety Features Actuation System.
GAI Elementary and Single Line Diagrams for the Electric Power System and Safety Related Actuation Devices Control Circuits.
Institute of Electrical and Electronic Engineers (IEEE) Standards: IEEE Std-279-1968 - "Proposed IEEE Criteria for Nuclear Power
Plant Protection Systems."
l
D-4
IEEE Std 308-1969 - "IEEE Criteria for Class IE Electric Systems for Nuclear Power Generating Stations."
IEEE Std 317-1971 - "IEEE Standard for Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations."
IEEE Std 323-1971 - "IEEE Trial-Use Standard: General Guide for Qualifying Class I Electric Equipment for Nuclear Power Generating Stations."
IEEE Std 334-1971 - "IEEE Trial-Use Guide for Type Tests of Continuous Duty Class I Motors Installed Inside the Containment of Nuclear Power Generating Stations."
IEEE Std 336-1971 - "IEEE Standard I~stallation, Inspection, and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stat ions."
IEEE Std 338-1971 - "IEEE Trial Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems."
IEEE Std 344-19 71 - "IEEE Trial-Use Guide for Seismic Qualification of Class I Electric Equipment for Nuclear Power Generating Stations."
IEEE Std 382-1972 - "IEEE Trial-Use Guide for Tupe Test of Class I Electric Valve Operators for Nuclear Power Generating Stations."
IEEE Std 387-1972 - "IEEE Trial-Use Standard: Criteria for DieselGenerator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations."
Radiological Assessment
1969 Census of Agriculture U.S. Department of Commerce Social and Economic Statistica Administration Bureau of the Census.
Radiological Control Procedure RP-1601 Florida Power Corporation Crystal River Unit 3. Radiation Protection Manual.
SDC, A Shielding-Design Calculation Code for Fuel-Handling Facilities, ORNL-3041.
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Operational Safety
ANSI NlS.7-1972 (ANS 3.2), "Standard for Administrative Controls for Nuclear Power Plants."
ANSI NlS.17-1973 (ANS 3.3), "Industrial Security for Nuclear Power Plants."