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REGULATORY IN ATION DISTRIBUTION SYSTEM RIDS)

ACCESSION NOR: 80 0 1 290255( DOG ~ DATE! 80/0 1/25'OTARY ZED: 0 DOCKET-,PACIr SO 27S diablo "Canyon Auclear"Power "Plenty Unit 'ig Pacific Ga 050011275.

50 323 diablo( Canyon Nuclear Power Pfantg On'it 2g Pacif'ic- Ga 65000323=AUTH ~ NAME' AUTHOR AFF'IL? AT IOA

CAAAFpV~ A, Pacific Gas 8 h.electric. Co.RECAP ~ NAME AECIPlEAT AFFECT.IATI01I1

DEATON,HARN Office'f NucTear Reactor. Regulation

SUBJECT: Forwards Addendum to: 791231 reptg"Status Rept to NRC frogsutiliResponding to~ Nuieg 0878:TMI 2 lessons Learned TaslcForce Status Rept L Short»i'erm Recommendations"LGla'ri'f>catlons," Ad endum responds to 're 800109 'meeting

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MALCOLMH. FUR BUSHVICE FRESIDEHS AND OEHERAE COUNSEL

ROBERT OHLBACHASSOCIATE OEH ERAL COUNSEL

January 25, 1980CDWASD J, MOCANNEYDAH CCAVSON LU~ SOCKJACK F, FALLIN JaOESNAAO J CELLAOAHTAJO ~ NUA OAS.LCVJD ~ EPN O, ENOLEAT,J1,RUDEST L Hkaal ~RIOHAAD~'ocacDAVIDL LUOVIDSON

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C HARL CS T. VAN DC V 8C N

PHILIP A. CRANCl JR ~

HCNRY J. LOPLANTCJOHN S. OISSON

ARTHUR L. HILLMAN,JR,C HARL CS W. THISSCI. I

DANICL C ~ OISSONASSISTANT Cta ESAL COVattL

Mr. Harold R. Denton, DirectorOffice of Nuclear Reactor RegulationU. S. Nuclear Regulatory CommissionPhillips Building7920 Norfol'k AvenueBethesda, Maryland 20014

DAVIDW, ANDE1 ~ ONDIANA OE1OHAV ~ ENLEION ~ CAS ~ IDYHEATHE1 ~ CI~ SNAOAIAN ~ DENTDNWILLIANH, COWka OSDONALD D C1ICC ~ OHDAYID0 CILDESTJUAN H JATOF RONALD LAVAHENIERHA11YW LON~ Ja,PAULAY RAINS~IOOE1T 0, HCLEHNAHRIOHAAO H, HO ~ 1

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NETS

Re: Docket No. 50-275Docket No. 50-323Diablo Canyon Units 1 and 2

Dear Mr. Denton:

Enclosed are 20 copies of an Addendum dated January 25,1980, to the PGandE report entitled "Status Report to the NuclearRegulatory Commission from Pacific Gas and Electric Company Respond-ing to NUREG-0578: 'TMI-2 Lessons Learned Task Force Status Reportand Short-Term Recommendations'nd Clarifications" which was sub-mitted on December 31, 1979.

This Addendum responds to questions raised at theJanuary 9, 1980 meeting with NRC representatives in Bethesda,Maryland.

Kindly acknowledge receipt of the above material on theenclosed copy of this letter and return it to me in the enclosedaddressed envelope.

Very truly yours,

EnclosuresCC w/enc.: Service List

STATUS REPORT TO THE

NUCLEAR REGULATORY COMMISSION

FROM

THE PACIFIC GAS AND ELECTRIC COMPANY

RESPONDING TO

NUREG-0578

"TMI-2 LESSONS LEARNED TASK FORCE

STATUS REPORT AND

SHORT-TERM RECOMMENDATIONS"

AND CLARIFICATIONS

-Z9's %S.-Dgcket . q~~~0j>o M of Docttmotth

8 "GBL"TGRYQQCKET Ft

December 31, 1979

ADDENDUM January 25, 1980

CROSS REFERENCE INDEX

STATUS REPORT

ITEM BEING,REVISED

STATUS REPORT

PAGE NUMBER

ADDENDUM

PAGE NUMBER

.1.1 (A) (TFP1) ............o...9 .........................22

2.1.1 (A) (TFP2)..........-......9....;....................3II

2.1.1 (A) (TFP3)...........;...10.........................42.1.3.b ,(A) ....................22 .........................5

.1.3.b (B);...................25 .........................92

.1.4........,................2 ~ ~ ~ 27 ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o14

~ 1 ~ 5 ~ 1 ~ ~ ~ ~ ~ "~ ~ ~ ~ ~'

~ ~ ' ~ ~ ~ ~ ~ ~ ' ~2 ~ ~ ~ 29 ~ ~ i ~ ~ ~ v ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~'o

~ ~ ~ o1 7,

.1.5.c.............. ~ ~;.-. ~ ~ ."33" '.-." ~ " ~ " ~ "."~ ""182

~ 1 ~ 6 ~ 6 ~ ~ ~ ~ o' ~ ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ' ~ ~ ~ ~ o35o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~'

~ ~ ~ ~ o192

~ 1 ~ 6 o b o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ 39 ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~'

~ ~ ~ ~ ~ ~ ~ 342

~ 1 7 1 ~ ~ ~ ~ ~ ~'

~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 'o43 ~ ~ ~ ~ ~ « ~ ' ~ ~ ~ ' ~ ~ ~'

~ ~ ~ ~ 362~ ~

~ 1 o 7 o b ~ ~ o' ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ o45 ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ o' i ~ ~ ~~"

~ ~ ~ ~ ~ o392

~ -1 8 1 ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' 53 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o' ~ ~ ~ ' '~ ~ ~ 412=

~ lo8obo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ ~ ~ o69,o ~ ~ ~ ~ . ~ , ~ ~ ~-~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o572

C

2 ~ 1 ~ 9 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ 'o76 ~ ~ <~ ~ ~ ~ ~ 'o ~ ~ ~ ~ ' ~ ~ ~

'~ ~ ~ ~ ~ ~ ~ 58

CRS 84 o ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 'o88 ~ ~ ~ 'o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 'o ~ ~ ~ ~ o 60A

2 ~ 2 ~ 1 o b ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ -~ ~ ~ ' ~ ~ ~ ~ ~ ~ o95 ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ "o 63'

NUREG-0578 ITEM 2.1.1.

A. Task Force Position 1

PGSE Res onse and Status

All of the four pressurizer heater groups can be supplied with power

from the offsite power sources when they are available. In addition.

provisions will be made to provide power to two out of four heater groups

from the. emergency power 'source through the Engineered Safety Reatures

(ESF) buses when offsite power is not available. Sufficient power is

available from the ESF buses to energize enough heaters to establish and

maintain natural circulation at hot standby conditions. Redundancy is .

provided by supplying each group of heaters from a different bus.

All of the equipment associated with pressurizer heater power supply

described in this response will be seismically qualified for the Hosgri

event except for those devices specifically noted as non-safety grade.

Circuit breakers 52-lG-72 and 52-1H-74 will be added to 480 volt

ESF buses 1G and 1H, respectively. These breakers will be seismically

qualifi'ed and installed to meet safety-grade requirements. The seismic

qualification is based on PGSE's vast testing experience which has

previously demonstrated that electromechanical equipment can withstand

numerous seismic tests simulating high seismic events without damage and

that the equipment will be available to perform its safety function after.

the seismic event.

Emergency power is generated by the onsite emergency diesel. generators

and supplied directly to the 4.16kV ESF buses. Power is then fed through

a step-down transformer to the 480 volt ESF buses.

Design is currently underway to incorporate clarifications 4 and 7

~

~

~

~

~

~into the present design. This modification will be completed by Apri'1 1,

1980.

NUREG-0578 ITEM 2.1.1

.A. Task Force Position 2

PG&E Res onse and Status

PG&E will develop the required procedures and implement the properh

training of the operators. Procedures and training will be completed to

make the operator aware of when and how the required pressurizer heaters

should be connected to the emergency buses. Loading of each ESF:bus:.can be

accomplished from the main control board (see TFP3 below). Procedures/

will identify under what conditions and which selected loads can be shed

from the ESF bus to prevent overloading when the pressurizer heaters are.

connected. The procedures will include provisions to ensure that the

heaters are transferred to the ESF power source as described in TFP3 below.

The time required to transfer the power supplies will be less than 10

minutes and will expose the operator to no more'than '10 mRem.

The procedures will.be written and approved, and 'the operators

trained, by May 1, 1980.

-3-

I


~

~

~ ~A. Task Force Position 3

PGSE Res onse and Status4 ~

I

'The proposed design modifications will provide for simple and rapid

transfer of the heater groups to the ESF power source.'Within 2 minutes

after loss of offsite power; the onsite emergency diesel'enerators willhave started and been connected to any required ESF loads.

When it is determined .that the pressurizer heaters are required;

the Shift Foreman need only to dispatch an operator-to the 100 foot ele-.'.

vation in the Auxiliary Building,.which is just three floors directly

below the main control room (two separate stairwells 'are available).

Once in the area, the operator simply verifies that the 'source breakers

( 52-1H-74, 52-13D-6, 52-13E-2, 52-lG-72) are open and manually throws

the deenergized tr ansfer switches. ,This .action in itself will not connect

the pressurizer heaters onto the ESF buses. Only when the Shift Foreman

is i'nformed that the transfers- have been made, will the heaters again be

controlled using the normal control. devices provided on the'main control

console. Even with manual transfer, there is no problem in meeting the

Westinghouse estimated time requirement o'f providing pressurizer heaters

on emergency power within one hour after the accident. PG&E will provide

control room indication of actual wattage being supplied to each heater

group that has been transferred to the emergency power sources. The

operating and emergency procedures will be changed to reflect these

requirements and changes by May 1, 1980.

-4-

NUREG-0578 ITEM 2.1.3.b

Task Force Position A.

PG&E .Res onse and Status

Procedures and Descri tion of Existin Instrumentation

The Westinghouse Owners'roup, of which PG&E is a member,-has performed

analyses 'as required by Item 2.1.9 to.,study the effects of inadequate

core cooling. These analyses were provided to the NRC "Bulleiins and

Orders Task Force" for review on October 31,:1979. .As part of'he

submittal mlade by the. Owners'roup, an "Instruction to Restore Core

Cooling during a small LOCA" was included. This instruction. provides

the basis for procedure changes and operator training'required to re-

cognize the existence of inadequate core cooling and restore core cooling

based on existing instrumentation. PG&E will incorporate the key con-

siderations of this instruction into the LOCA procedures, and provide

training to the operators in this area.

Subcoolin Meter

PG&E is installing a Combustion Engineering PWR subcooled margin, monitor.

The equipment will be ordered by February 1, 1980. ~ Equipment'delivery

is expected by May 1, 1980 and installation will be completed by June 1,

1980. Details of display, calculator and inputs are as follows:

~Dis la

The display indicates either the temperature or pressure margin

continuously on a digital type display and on an analog meter. The

analog meter provides a redundant display. Both displays are

~ located on the main control board in the control room. There's -a

. low alarm at 30 F and a low-low alarm at 20 F. The uncertainty

ITEM 2.1.3;b continued

in the monitor is 0.5 percent. The overall uncertainty is estimated

to be+10 F. The range of the display is -40 to+200 F. The display".

is qualified to meet the seismic requirements of IEEE 344.

Calculator

The calculator module is a dedicated digital type. The 'plant process

computer is not used. The. calculator module is not redundant. The

selection logic uses the highest temperature and the lowest pressure.

The module is qualified to meet the seismic requirerdents of IEEE 344.

~Iri uts

1. Temperature

The subcooled meter has 10 temperature inputs. Four temperature

signals come from each of the four hot leg wide range RTD's and

six temperatures are taken from core exit thermocouples. All,

temperature signals have a range of 0 to 700 F. The uncertainty

is estimated to be +5 F with no data available to evaluate the

difference between forced flow and natural circulation conditions.

At the present time, none of the temperature inputs are safety

grade, however, PG&E is upgrading these inputs to meet safety

grade requirements.

2; Pressure

Pressure is sensed by two safety grade reactor coolant loop

Barton Model 763 pressure transmitters. Both pressure trans-

mitters are located on the hot leg of loop 4. The pressure

range is 0 to 3000 psi with an uncertainty of +15 psi. These

-6-

ITEM 2.1.3.b continued

transmitters are seismically qualified to the requirements of

IEEE 344 and environmentally qualified to the requirements of

IEEE 323.

INFORMATION RE UIRED ON THE SUBCOOLING METER

~Dis la

Information Displayed (T-Tsat, Tsat, Press, etc.)

Display Type (Analog, Digital, CRT)

Continuous or on Demand

Single or Redundant Display

Location of Display

Alarms (include setpoints)

Overall uncertainty ( F,,PSI)

Range of Display

gualifications (seismic, environmental, IEEE 323)

TSAT - T P - PSAT

Di ital and Analo

Continuous

Redundant

Control Board

30 F 20 F

+10 F

-40, +200 F

Seismic

Calculator

Type (process computer, dedicated digital or analog.calc.) Dedicated Di ital

If process computer is used specify availability. (X of time) N A

Single or redundant calculators

Selection Logic (highest T., lowest press)


Calculational Technique (Steam Tables, Functional Fit,

ranges)

~ln ut

Temperature (RTD's or T/C's)

Sin le

Hi hT- lowP

Seismic

SteamTables'-3200

psi0-999 F

4 RTD's 6 T C's


Temperature (number of sensors and locations)

Range oF temperature sensors

Uncertainty* of temperature sensors ( F at 1)

gualifications (seismic, environmental, -IEEE 323)

Pressure (specify instrument, used)

Pressure (number of sensors and locations)

Range of'ressure sensors

Uncertainty* of pressure sensors (PSI at 1)


Note 1

.0-999 F

+ 0

Note 1

Barton Model 763

2 on Loo 4 Hot Le

0-3000 si

15 si

Seismic environmental

Backu Ca abilit

Availability of Temp & Press

Availability of Steam Tables, etc.

Training of operators

Procedures

Yes

Yes

See below

See below

*Uncertainties must address conditions of forced flow and natural

circulation.

Note 1: See section on Inputs, above.

Procedures to recognize inadequate core cooling are presently being

.written.'he procedures will be approved and the 'operators trained by

May 1, 1980." As described in PG&E's NUREG-0578 response August 27, 1979,

the process computer has b'een prograrwoed to provide margin to saturation

information.

-8-


Task Force Position B


Additional Instrumentation to Indicate Inade uate Core Coolin

The submittal referenced in A above described the capabilities of the

core exit thermocouples in determining the existence of inadequate core

cooling .conditions and their superiority in some instances to the loop

RTDs for measuring true core conditions. Other means of determining the

approach to or existence of inadequate core cooling could be:

1. Reactor vessel water level

2. Incore detectors

3. Excore detectors

4. Reactor coolant pump motor currents

5. Steam generator pressure

A discussion of the possible use of these measurements is addressed below.

The use of incore movable detectors to determine the existence of inade-

quate core cool'ing conditions appears doubtful. The detectors could be

driven in to the tops of the incore thimbles, which are located at the

top of the core, following an accident in which concern for inadequate

core cooling exists. The problem comes in the lack of sensitivity of the

detectors to very low neutron levels and changes that would occur due 'to

core uncovery. Gamma detectors could perhaps be employed, but they suffer

from similar sensitivity problems, and the fact that gamma levels in the

fuel region change insignificantly between the covered and uncovered

condition. As a result, it does not appear worthwhile to pursue incore

movable detectors as a means of determining inadequate core cooling

conditions.-9-


The use of excore detectors has been mentioned as a possibility in respon-

ding to core uncovery. The only detectors which would have the required

sensitivity are the source range monitors. Since the intermediate and

power range monitors are not sensitive enough to the low level changes

resulting from vessel voiding. The use of the source range monitors will

be investigated further as part of the more indepth study of inadequate'I

core cooling being performed by the Westinghouse Owners'roup. However,

their use is probably limited to those instances when significant voiding

exists in the downcomer region, since normally water in the downcomer

would effectively shield the detectors from the core region whether voids

existed or not.

Reactor coolant pump motor current, which could be indicative of core

voiding, is inappropriate for a reliable means of determining inadequate

core cooling, since a loss of offsite power or pump trip due to a LOCA

blowdown would shut down the pumps.

Steam generator pressure, which already exists, is useful in the case

where heat transfer from primary to secondary is interrupted due to loss

of natural circulation.'his, however, does not satisfy requirements

to indicate the approach to inadequate core cooling, nor does )t indicate

the true condition o'f the core.

Reactor vessel water level determination is the most promisiqg of the

items discusssed to provide additional capability of determining the

approach to and the existence of inadequate core cooling. Several sys-

tems for. determining water level are under review by the Westinghouse

Owners'roup. A conceptual design of one system is given below:

-10-


Vessel Level S stem Descri tion.

After examining many different methods and. principles for determining

the water level in the reactor vessel, a basic delta pressure measurement

from the bottom of the vessel to the top of the vessel appears to provide

the most meaningful'nd reliable information to the operator. One of

the reasons 'for choosing this system is that the sources of potential

errors are better known for this system than for any other new or untested

system.

The attached figure shows a simplified sketch of the proposed vessel

level instrumentation system. The bottom tap of the instrument would

use a thimble of the incore movable detector system either at the seal

table or in the thimble below the vessel. Use of the thimble as part

of the incore flux monitoring would not be lost. The flux thimble ~

guide tube would be tapped below the vessel and an instrument line

connection made. The instrument line would have an isolation valve and

slope down to a hydraulic coupler connected to a sealed reference leg.

The sealed reference leg would go to. the differential pressure transmitter

located at a higher elevation above the expected level of containment.

flooding. A similar sealed leg would go to the top of the vessel and

penetrate the head using the vent line or a special connection on a spare

RCC mechanism penetration. Two trains of vessel level instrumentation would

be provided.

The behavior of the signal generated by this level instrument under normal

and accident conditions is being evaluated. 'he usefulness of this instru-

ment to provide an unambiguous indication of inadequate core cooling is

being evaluated as part of Item 2.1.9. The potential errors and accuracy

-1 1-

'C7,

Containment Teall

HeadPenetration

'~ Isolator Indicator

I

0 - - = ""'.Hydreulio

H'solator

ITypical)

~ Train 8

I

IhPI,

~ INal Table

~

'nstrument

Train A —, ~-

PROCESS CONNECTION SCHEMATIC (TRAIN A)

-12-


.of a final system configuration are being evaluated to assess its usefulness

to provide information to the operator for proper operation of= a vessel

venting system and for normal water level control during periods when the

primary system is open and a water level may exist in .the vessel.. The

connection of the level system to the vessel head should be designed to',

be compatible with the head vent system. Operation of the vent system

should not upset all indications of vessel level. .This can easily be

avoided by using a separate instrument tap or by using more than one lo-

cation.

PG&E believes that the proposed Westinghouse iridication system .to indicate

adequate core cooling will provide unambiguous indication of adequate core

cooling during all postulated reactor coolant system conditions. In

addition the same indication'system will provide indication of reactor

water level during no flow conditions. Detai.led analyses and system

descriptions will be provided when available from Westinghouse. Delivery

is not scheduled at this time.

PG&E is presently evaluating technical proposals from Westinghouse

for a reactor vessel level indication system. PG&E and other near-term

license applicants have met with the NRC to discuss the requirements

given in the draft Regulatory Guide 1.97, Revision 2. These requirements

in final form may well include additional instrumentation requirements.

PG&E will comply with any new requirements.

-13-


Task Force Position 4

PGLE Res onse and Status

There are two phases of containment isolation at Diablo Canyon. Phase A

isolates all nonessential process lines but does not affect safety injection.

containment spray, component cooling water supplied to the reactor coolant

pumps and containment fan coolers, and steam and auxiliary feedwater lines.

Phase B isolates all process lines except safety. injection, containment

spray, auxiliary feedwater, and the containment fan coolers component

cooling wa'ter system.

r

Phase A isolation is initiated by high containment pressure, high

differential pressure between steam lines, low pressurizer pressure, or

manual initiation. Phase B isolation is initiated by high-high eon-

'

~ ~ ~

~

tainment pressure or manual initiation.

This system fully complies with Section II..6 of the SRP 6.2.4.;.Compli.ance

with all remaining sections of SRP 6.2.4 is documented in Section 6.2.4 of

the Diablo Canyon FSAR.

PGRE has three levels of containment process penetrations. These are defined

as:

1. "Nonessential" process lines are defined as those which do not increase

the potential for damage for in-containment equipment when isolated.

These are isolated on Phase A isolation.

2. "Essential" process lines are those providing cooling watei and seal

water flow through the reactor coolant pumps. These services should

not be interrupted while the reactor coolant pumps are operating

-14-

ITEN 2.1.4 continued

unless absolutely necessary. These are isolated on Phase B isolation.

C

3. Safety system process lines are those required to perform the function

of the Engineered Safety Features System..

Table 2.1.4-1 of PG&E's August 27., 1979 response to NUREG.-'0578 provides the

identification of nonessential,.essential, and safety systems penetrating

containment.

All nonessential systems use'either manually sealed closed valves or

else the valves are automatically isolated on a Phase A containment

isolation signal. Additionally, all essential systems (defined in

response 2 above) are automatical.ly isolated on a Phase B containment

isolation signal.

The Diablo Canyon design complies with the position. There are two basic

actuator methods, one for motor operated valves and once for air operated

valves. For a motor operated valve (NOV), the trip signal operates .the

close coil until the valve closes. If the trip signal is removed, the

valve will not operated unless the open coil is energized by an explicit

operator action. FSAR Figure 7.3-35 shows a typical NOV 'circuit.

For an air operated valve (AOV), the control switch must b'e held in the

open position to open the valve. Once it is opened, a stem mounted

position switch on the valve closes thus closing a latch-in circuit.I

which holds the valve open when the control switch spring-returns to

neutral. The isolation signal contacts are in this latch-in circuit.

When an isolation signal is generated, the circuit opens and the valve

is de-energized and closes. As soon as the valve begins to 'close, the

position switch opens so that the circuit will remain open and the valve

-15-

ITEM 2.1.4 continued1

will remain closed even after the isolation signal is reset. FSAR

Figure 7.3-43 shows a typical,AOV circuit.E

a jkwrp 0

Control of.Liquid Radwaste Isolation Valves is somewhat different than

described above. Each valve does not receive an individual isolation

signal. Instead four (4) valves inside the containment are closed by

one signal from Train A and 5 valves outside of containment are closed

by one signal from Train B.

As presently designed some of the valves could open automatically after

manual reset of the isolation signal and the Radwaste Isolation Valves

control circuit for each Train.

This scheme will be modified to require manual control of each valve

after resetting the above signals. - Implementation .of these changes will

be completed by May 1, 1980.

-16-

NUREG-0578 ITEM 2.1.5.a

PGEE Res onse and StatusT

- 4r

Although recombiners located inside the containment are being providedr.

for Diablo Canyon, PGEE has chosen to comply with the Task Force

Position and clarifications concerning dedicated penetrations for ex-

ternal recombiner or post-accident purge system.

Design changes have been made to provide dedicated penetrations and

isolation systems for the hydrogen recombiner and hydrogen purge systems.

It is anticipated, depending upon equipment availability, these mod-

ifications will be completed by July 1, 1980.

The supply lines for the existing Containment Hydrogen Purge Systems use

containment penetrations which are solely dedicated for that purpose.

The exhaust lines previously were teed-off from the 48-inch containment

purge lines. In response to NUREG-0578, PGLE has selected different

containment penetr ations (previously designated as spares) which will be

solely dedicated to post-LOCA hydrogen control. New lines will be from

those penetrations to the existing iodine filters 'and blowers of the

Containment Hydrogen Purge exhaust lines.

The capability of adding external post-LOCA hydrogen recombiners will beK

facilitated by placing piping tees and blank flanges in the post-LOCA

hydrogen purge lines at a place where it would be convenient to install

external recombiners.

The areas selected for the control panels for the external recombiner s

~

~ ~

and for the piping tee connections will be analyzed to evaluate the

possible personnel exposures in connecting and operating the recombiners

during the accident conditions. This analysis will be completed prior to

March 15, 1980.

-17-

NUREG-0578 ITEN .2.1.5. c

PG8E Res onse and Status

Hydrogen recombiners were.not included as a design basis for licensing

Diablo Canyon Units 1 and.2, however, PG&E has-chosen to install hydrogen

recombiners inside containment at Diablo Canyon. Westinghouse hydrogen

recombiners and associated equipment have been purchased and are presently

at the site. Installation of the recombiners will.be completed by May 1,

1980.

The recombiners will be housed inside the containment and therefore will

not present any shielding problems during operation of the recombiners.

Procedures for operation of the installed hydrogen recombiners will be

written and the operators trained prior to May 1, 1980.

-18-

NUREG-0578 .ITEM 2.1.6.a


At intervals not exceeding 18 months, operating pressure. leak tests willbe performed on appropriate portions of SIS. RHR, NSS Sampling, LRW and

GRW Systems. Pressurized systems will be visually inspected for leakage

into the building environment. Any observed leakage will be eliminated.

An attempt will be made during the initial testing to measure system

makeup during the period of pressurization, where the system configuration

permits. System makeup rate provides some indication of system leakage,

but may not be conclusive because of valve seat leakage. Initial testing

will be conducted prior to fuel loading.

I'herefe'asible, liquid containing systems will be pressurized with a

hydro pump and makeup determined by measuring level changes in a grad-

uated tank, or by =normal operating pressure sources. During pressurization

each system will be walked down and visually inspected for leaks. Visual

inspection or other appropriate means will be used to differentiate system

boundary leakage from test boundary leakage to determine actual -leakage

outside the system. There are portions of systems that will be

pressurized with a system pump, and since makeup rate cannot be measured,

any leaks discovered during the walkdown will-be evaluated to determine

the approximate leakage from that system. Baseline leakage rates willbe recorded after appropriate measures to eliminate undesirable leakage

paths have been taken.

For the Gaseous Radwaste System the ihmediate leak reduction plan willconsist of a preoperational measurement, followed by an operational

monitoring program.

-19-

ITEM 2.1.6.a continued

With the waste gas compressors running on recirc to maintain impeller seals .

the system will be maintained at maximum operating pressure using a regulated'.'

nitrogen source. The system makeup rate will be determined with a gas 'flow-

meter at the nitrogen regulator. All non-welded connections in the system

will be checked with a soap and water solution to 'locate leaks and appropriate

means taken to eliminate them. This is the preoperational measurement.

Thereafter .(during operation), a routine radiation monitoring program will be

implemented to detect minute changes in activity of air in the areas occupied

by the system. If leakage is indicated by an increase in airborne activity,

appropriate means, such as explosive gas detectors or soap, will be used to

localize the leak.

Portions of the charging system which are in service during normal operation

(makeup and letdown) are monitored with the rest of the reactor coolant system

by the reactor coolant system water inventory balance. Excessive leakage into

controlled areas will be indicated by abnormally high airborne radioactivity

1 evel s.

Some portions of the CVCS are used for boron recycling during normal operation.

The boron recycling system will not be used under post-accident circumstances,

'nd is not included in this program.

Containment spray recirc piping and valves outside containment are leak rate

tested with precision volumetrics equipment at a 24 Nenth frequency as part

of a containment isolation valve surveillance test.

-20-

ITEM 2.1;6.a continued

Individual Surveillance Test Procedures will be written for each system or .r- C~ a

'"

4

each half system in the case of SIS and RHR. Each STP will list system~ L

r.

boundary valves and specify test. pressures. Each test will be performed

prior to fuel loading, and at refueling outage intervals thereafter.

The portions of systems .that will be monitored are .indicated on the attached

piping schematics.

-21-

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~~

NUREG-0578- ITEM 2.1.6.b

PGEE Res onse and Status',+

A comprehensive site inspection and evaluation was conducted to identifyL

those portions of the systems exposed to high. activity reactor coolant both

during the short-term post-accident cooldown mode and the long-term post-

accident cleanup mode. Plant piping drawings and equipment arrangement

drawings were reviewed and.the systems and components were identified

and labeled to locate all source terms during realistic accident scenarios.

Personnel access pathways, equipment requiring access, and constantly or

frequently manned areas of the plant during accident conditions were

identified for dose rate calculations.

Initial area calculations have been made for the Control Room, Onsite

Technical Support Center, Reactor Coolant Sampling Area and the penetrationI

area between the Containment and the Auxiliary Building.

A computer model of the Auxiliary Building sources and calculation areas

was developed. The initial calculations will be further refined using

this computer model. Dose rates in those areas of the auxiliary'and

turbine bui']dings requiring personnel access will be calculated.

Calculations will be performed,to evaluate exposure to equipment needed

after the accident to insure adequate environmental equipment qualification.

Evaluation of the new sampling equipment (see Section 2.1.8.a) for post-

accident sampling will be conducted to i'nsure personnel protection.

Shielding analysis and environmental qualification studies are currently

in progress and will„be completed by April 1, 1980. When the shielding

studies have been completed (prior to July 1, 1980), a formal design review

report of the results will be delivered to the NRC. If modifications or

-34-

C

ITEM 2.1;6.b continued

procedural changes are required as a result of the findings, details of

the necesssary changes will .be provided to the NRC.C

Redesign and addition of necessary .shielding,=if any, and changes in

operating .and administrative procedures will begin following completion

of the evaluation effort.

-35-

NUREG-0578 ITEM 2.1.7.aI

PGKE Res onse and StatusIr

"

The auxiliary feedwater system is shown in FSAR Figure 3.2-03, Sheet 2 ofC

4. The pumps .are automatically started by low-low steam generator level,feedwater pump trip, safety injection,.or loss of offsite power. See

FSAR Figures 7.3-8, 7.3-17 and 7.3-18. * As shown on FSAR Figure 7.3-17.

the motor-driven auxiliary feedwater pumps are started by closure of the

Solid State Protection System (SSPS) output relay K633. Relay K633 is

actuated by. safety injection initiation or low-low level in any steam

generator. Each pump is started by a separate relay from redundant

SSPS trains A and B. The motor-driven pumps are also automatically

started by trip of both main feedwater pumps.

The turbine driven auxiliary feedwater pump is started by opening steam

supply valve FCV-95. As shown on FSAR Figure 7.3-18, this valve is opened

by SSPS output relays K632 or K634.- Relay K632 provides for starting on

loss of offsite power and relay K634 provides starting on safety injection

or low-low level in any steam generator. Loss of offsite power is

determined by low voltage on the 12kV=.reactor'coolant pump buses. An

automatic starting signal is provided by redundant SSPS trains A and B.

The system valves are normally open and require no actions for system

operation. The auxiliary feedwater initiation circuitry is part of our

Engineered Safety -Features (ESF) system, and as such, is installed in

accordance with .IEEE Standard 279. This Standard is referenced in 10 CFR

50.55a(h).

The auxiliary feedwater initiation signals and circuitry are testable.

Such testability is included in the surveillance test procedures for the

plant as delineated in the Plant Technical Specifications.

-36-

0

0

0

c aug

'


The initiating signals and circuits are a part of the Plant Engineered

Safety Features. The requirements for the ESF system dictate that the

system shall meet the "single failure criteria". To accomplish this,the initiating signals and circuits must be, and are, powered from

separate emergency buses.

The initiating sensors such as steam genera4er low-low level. are powered

from separate and redundant nuclear instrumentation and control panels,

each of which is supplied by either on-site emergency generators or

station emergency batteries. Each of the two redundant SSPS -trains is

supplied by separate power sour ces. Initiation and flow paths will be

available. upon loss of offsite power.

h

Manual initiation for each train exists in. the control room. The manual

initiation system is installed in the same-manner as the automatic

initiating system. No single failure in the manual initiation portion of

the circuit can result in the loss of auxiliary feedwater system function.

See FSAR Figures 7.3-17 and 7.3-18 for the circuitry.

As with all ESF equipment, the a-c motor-driven pumps and all -valves in the

system are automatically transferred to, and sequentially loaded on, the

emergency buses on loss of offsite power. The sequence is shown on FSAR

Table 8.3-2.

All automatic initiating signals and circuits are installed in accordance

with regulatory requirements and are "safety grade and redundant. No

siqgle failure in the automatic portion of the system will result in loss

of the capability to manually initiate the AFWS from the control room.

-37-

~ I

I

I'

gy as . ~'

a

~ ~ ~ ~

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~ ~ ~

~


As described, the automatic initiating signals presently. meet all safety

grade requirements. No upgrading is required. ~ ~

-38-



The flew from each primary motor-driven auxiliary feedwater pump is

by an indication system powered from .the same primary source as the

+monitored

motor.

This system provides indication, both at the main control board and the

hot shutdown panel.

The flow from the turbine-driven- auxiliary feedwater pump is monitored by

the same indicators which monitor the motor-driven auxiliary feddwater

pump flow.

An additional indication of auxiliary feedwater flow is provided by the

steam generator wide range level indication. This provides recording on

the main control board and indication on the hot shutdown panel. It is

powered from the same bus as powers one of the motor-driven auxiliaryI

feedwater pumps.

-39-

STEAM GENERATORS CB'R

LT ~~ TYPIC+IL

BUS HHSD

LT

BUS H

BUS H

LT

BUS H

FICB

AUXILIARYFEEDWATER

PUMPS

BUS H BUS H BUS F BUS F

FE FE FE FE -'YPICAL

FIHSD

TURBINEDRIVEN

BUS G

S C

MOTORDRIVEN

BUS H

MOTOR

DRIVEl

BUS F

-40-

'

NUREG-0578 ITEM 2.1.8.a

C


~ ~

A review of existing reactor coolant and containment atmosphere sampling

facilities has been conducted for Diablo Canyon Units 1 8 2. The purpose

of this review was to evaluate the stations capability to obtain samples

during accident and post-accident situations. NUREG-0578 Section 2.1.8;a,-

'ImprovedPost-Accident Sampling Station Capability" requirements were used

as the basis for this evaluation.

This review and evaluation has resulted in a conceptual design providing

a tentative Remote Sampling Facility which projected to be located at

the 115 foot elevation on each unit. Shielded sample lines will be

provided for both reactor coolant sampling and containment atmosphere

sampling. A control panel will be provided where controls and indication

may be located for the Remote Sampling System components. The system

includes the capability .for continuous monitoring of containment hydrogen

concentration.

The described Remote Sample Station meets the requirements of NUREG-0578;

Personnel radiation exposures have been minimized by shielding the sample

lines and minimizing sample quantity to as low as practical limits. 'hefinal Sample Station design will allow Diablo Canyon Units 1 5 2 personnel

to obtain RCS and Containment Atmosphere samples during post-accident

conditions with a minimum of effort while also limiting personnel exposure;,.

'I

Final designs for the remote sampling system will be completed by May 1,

1980. Installation of the remote sampling, system, all analysis equipment

and required procedural changes will be completed by January 1, 1981;

-41-

=ITEM 2.1.8.a continued

INTRODUC ION

This report discusses the locat'ion of facilities to accommodate the sampling

of reactor coolant and containment atmosphere during post-accident

conditions. An alternate point, different from the existing sampling

station location, is necessary .due to the high radiation levels existing

from consideration of -source terms in NUREG-0578.

Conceptual designs are presented for the sampling of reactor'coolant and

containment atmosphere. These designs will be used .as the basis for the

final sampling systems to be installed at Diablo Canyon Units 1 5 2.

REMOTE SAMPLE STATION LOCATION

Consideration. was given to a number of locations where a remote sample

station could be installed. A;location was chosen on the 115'levation

because of low potential radiation considerations, spaciousness, and access

to and from grade level.

The remote station will be installed at the 115'levation adjacent to

the steam generater blowdown demineralizers or the prefilter. Sample .

tubing for both containment air. and reactor coolant sampling .will be

routed from the reactor -coolant system (RCS) sample room'(elevation '100'),

vertically upward through an existing pipe chase to the steam generator

demineralizer cubical. The tubing will extend over the top of the wall

separating the demineralizers from the auxiliary building corridor.

Shielding will be provided as required.

Radiation sources in the area result from the Chemical Volume Control

System (CVCS) piping,.the Volume Control Tank (VCT), and containment

spray piping. These potential sources will be discussed in Section 3.0;

-42-

: ITEM 2.1 S.a continued

A ventijaCion hood will be installed at the. remote sample station above

the sample points to remove any potential airborn activity=and will beI

sized to provide a high flow rate. A blower will also be installled in'the hood as opposed to simply branching into an auxiliary building return

duct.. This will insure quick removal and dilution of any airborn

'ontamination.

A control panel containing switches for operating valves, status indication

and direct telephone comnunications with the control room, will be

provided at the remote sample station location. Section 6.0 discusses

the design of the control panel further.

POTENTIAL BACKGROUND SOURCES

The remote sample station sampling tubing, CVCS piping, VCT, and containment

spray piping contribute to personnel exposure, during remote sampling.

Unshielded CVCS piping is visible from the corridor on the ll5'levationadjacent to the prefilter .cubical. Because this pipe chase will also

contain sample tubing, a labyrinth shield at the entrance to the cubical

will be provided.

The volume-control tank, located beneath the S.G. blowdown demineralizers,

will become an important radiation source if a feed and bleed operation

occurs using the assumed TID source terms..

The containment spray piping is located in the penetration area adjacent

to the auxiliary building wall at the proposed sampling location. This

piping will be a significant radiation source in post-accident operations.

-43-


A preliminary evaluation consisting of determining the radiation levels

during %he after the accident situation has been performed to determine

shielding requirements for the area. A further, more specific evaluation

will be conducted to determine the shielding designs.

RCS LI UID SAMPLING SYSTEM

Under normal operation, reactor coolant is continuously let down from

either the hot leg of loop 1 or 4 to the Gross Failed Fuel Detector and

to the hot leg sample vessel, ATSV3. Flow passing through the Gross

Failed Fuel Detector returns to the sample heat exchangers. The flow

continues through the sample line upstream of its connection to thel

CVCS Mixed Bed Demineralizers or the sample sink; depending on valve

line-up. The remote sampling system utilizes this existing system

and adds the capability for remote sampling.

1.. Remote RCS Li uid Sam lin S stem

The Remote Sample System, shown in Figure 1, will utilize new isolation

sample valves downstream of the sample heat exchanger to divert'.flow to

a shielded, removable sample vessel for obtaining a reactor coolant sample.

The shielded sample vessel may be used to obtain coolant samples from

either the RHR pump discharge or hot leg 1 or 4 depending on valve line

up. Capability to control valves 9370A and B must be provided on the

remote sample system panel, due to the fact that the radiation level of

the lines those lines are mounted on would preclude personnel access. .The

reactor coolant returns to the CVCS mixed bed chmineralizers to allow

the sample to flow to the liquid radwaste system. This minimizes

potential gas releases from the sample sink to the auxiliary building

atmosphere.

-44-


The remote sample system will also have the capability to be flushed'r,.

with demineralized water using manual valves controlled at the sample

panel,. This is necessary. in order to flush any high radiation level

sample remaining in the lines prior to taking a new sample. FCV 19 is a

local, manually operated flow control valve which maintains letdown flow

at 1 GPM. This valve does not require control at the remote sample panel

location assuming that it is operating to 'control letdown flow.

The remote, shielded .sample vessel contains flexible lines on one end

to allow the operator to easily connect the vessel to the quick disconnects

by compensating from any minor mis-alignment. Pressure and temperature

indicators for the sample are also provided for system operation and to

allow calculations to be made relating to the sample mass.

2. RCS Li uid Sam le Vessel

The RCS liquid sample vessel must have the capability to withstand

hydrostatic testing to 3750 psig and operating temperatures and

pressures of 130 F and 2250 psig, respectively. It will be constructed

of 304 stainless steel or equivalent. The sample vessel must be capable

of containing about 35 ml of liquid required for analysis purposes.

This total results from requirements for 25 ml for a cloride analysis,

5 ml for boron and 1 ml for activity.

'wo sample container attachments were considered. The 100 ml sample

vessel supplied by the Kerotest Company and used for normal RCS

sampling, could be adapted for use at the remote sampling station.

It would require additional shielding but has the advantage of being

an off-the-shelf item. Figure 3 shows a graph of dose rate vs. shield

thickness for a cylindrical shield placed'round'the kerotest vessel.

-45-


It should be noted that an 8" radius shielded vessel weighs aboutr~

1,000 %m, and has an expected surface dose rate of about 62 mR/hr.

The other alternative allows collection of sample in a length of shielded

~ tubing designed to obtain a 35 ml.sample with shutoff valves and quick

disconnects as shown in Figure 1. The smaller sample volume minimizes

shielding and handling requirements, reduces thd exposure potential, and

would be simpler to fabricate; The connections at'the sample station willbe designed to accept both the 100 ml kerotest vessel and a 35 ml "sample

tube" so that the capability exists to take'ither small or large RCS

samples. The shielded sample tube will be the primary sample collection

method used due to the advantages discussed previously.

CONTAINMENT ATMOSPHERE SAMPLING SYSTEM

Containment atmosphere sampling is normally accomplished by connecting

a sample vessel with a flexible line upstream of the particulate airmonitor (RE ll and 12). The existing particulate monitor is normally

used to determine radiation levels by passing the containment atmosphere

through filter paper and reading levels of particulates which do not

pass through. The existing monitor also measures radiation levels of

any noble gases in the containment atmosphere. For the worst case,

post-accident situation, noble gases, particulates, Halogens and,Hydrogen

may exist along with a sizable quantity of wet steam. 'Modifications must

be made to the existing monitor as it cannot withstand this environment.

In addition, the existing system isolates on high containment pressure

since it is not capable of withstanding it.~ ~ ~

ITEN 2.1.8.a continued

1. Remote Containment Atmos here Sam lin S stemC

The design .for the Containment Atmosphere Sampling System -(%ASS) which

meets NUREG-0578 requirements and clarifications is shown-in 'Figure 2.

The primary design features of the system are the redundant active

components to.meet single failure criteria which'are required for hydrogen

sampling, the capability to obtain's'amples for particulate, iodine analysis

and a permanently connected hydrogen analyzer.

Containment atmosphere is drawn into the system through the existing

FCV 678 and 679 valves by air sample pumps. Redundant valves are

provided to meet single active failure criteria. Valves are also

provided to isolate the existing particulate monitor in'rder.to protect

it from accident environmental conditions. The sample continues through

to either the containment iodine,'particulate and atmospheric sampling

'ectionor to the hydrogen analyzer, depending on valve lineup. It is

not expected that both atmospheric sampling and hydrogen monitoring will

be conducted concurrently; thus a coranon air pump and supply line is

provided.

A shielded charcoal/paper filter is provided upstream of the sample vessel

to remove particulates and iodine for separate analysis. Both the paper

and charcoal filters require a moisture-free atmospheric sample.to avoid

deterioration. In addition iodine will plate out on the sample lines ifthe temperature of the sample drops below the saturation point. To

minimize deterioration of the sample and plate out, heat tracing is used

on all CASS lines prior to the iodine:sampling jpoint. Temperature and

pressure indicators between the filter and the sample vessel may be used to

calculate the air'ensity and, when used in'conjunction with the flow

-47-


indicator, the mass flow rate passing through the filter assembly (or~ ',r

hydrogen analyzer) may be determined.

Any required corrections for the volume fraction of steam (if'resent)can be determined by cooling the sample vessel- and measuring the liquid

content. guick disconnects are provided for both the filter assembly

and sample vessel to allow removal of that equipment for analysis.

Flexible lines are provided on the sample vessel and filter assembly

as described in the RCS sampling section. The sample continues to the

hydrogen analyzer where hydrogen content within the containment atmosphere

is determined.

A connection to instrument air is provided near the remote sampling

system locations. A flexible hose may be connected from the instrument,

air supply ot the instrument air purge connection on the sample system

to allow this air flow to remove any sample plate out or particulate

mattter. Bypass lines containing quick disconnects will be attached

to the sample points to allow purge air to flow through the containment

atmosphere/iodine-particulate sampling section.

2. H dro en Monitorin

An existing hydrogen analyzer detects between 0-100$ of the concentrations

of combustible gases or vapors required for an explosion to occur.

The detector, a Mine Safety Apparatus $5A) model 501 system, utilizes

a Wheatstone Bridge connected,to heated filaments. Burning the

combustible portion of an air sample produces heating of the filaments

and causes a proportional change in electrical resistance.:This device

has two limitations: (1).it will not operate in'the presence of steam

-48-


or inert atmospheres, and.(2) it.can only measure combustible mixtures

up to. the, explosive limit.(for hydrogen 5 - 10Ã). This device does not

meet the .intent. of the .clarifications of ACRS Coranent 83, which

requires a measurement capability of hydrogen concentrations over a

0-105 range.

The modified CASS is desi:gned to include a new hydrogen analyzer with the

capability for monitoring the 0-10K,range. The hydrogen'nalyzer would

also operate in the presence of high temperature steam-air and inert gas

mixtures. Investigations are continuing with vendors to determine ifhydrogen analyzers with this capability are available.

Redundant components are provided to meet single active failure criteriafor hydrogen analysis. For example, Figure 2 shows dual FCV's on the

inlet and outlet sample lines, air sample pumps and dual hydrogen analyzers.

3. CASS Sam le Vessel

The shielded sample vessel for use in obtaining containment atmospheric

samples will consist of a pipe with a septum closure nipple installed on

the sidewall. After opening the 'shutoff valve and allowing sample to

flow through the pipe, a shielded hypodermic will be used to withdraw 1 cc

of atmospheric sample from the septum closure. This can be transported

in a shielded carrier for analysis using existing laboratory methods and

analytical techniques.

The shielded particulate and charcoal filter is located upstream of the

atmospheric sample. Sample would flow through the filter upon opening

of the shutoff valve. The filter would then be transported in a shielded

container to the..chemistry'laboratory for analysis to determine iodine.

-49-


and particulate levels using standard techniques.

REMOTE SAMPLE SYSTEM CONTROL PANEL

Due to the high radiation levels which could exist on the sample lines,

remote control of certain valves and equipment is required at the remote

sample station. In addition, temperature and pressure, valve status and

other indications of system performance are required. .A panel containing

these components will be located at the point where remote sample station

tubing is being routed as discussed in the location'section.

1. Panel Mounted Instrumentation and Controls Re uired forReactor Coolant Sam lin S stem

The following control and indication will be added and/or mounted on the~ ~ ~ ~ ~

panel:

a. Valves

(1) Manual valves, shown in Figure 2, will be mounted such that

these handles are shielded from any sample line radiation level.

(2) Control valves which close on receipt. of a containment isolation

signal.

(3) Valves 9370B,,9370A .will have solenoid operators added to allow

remote operation.

(4) Valves 9351A, 9356A and 9356B will have operators added to allow,

. operation from the remote sample panel.

(5) Sample isolation valves.

-50-


b. Indication~ c

(T) .All new pressure and temperature indication.

(2) Status indication for all remotely operated valves.

2. Panel Mounted Instrumentation and Controls Re uired for ContainmentAtmos heroic Sam lan S stem

Indication and controls for the following devices will be mounted on the

panel:

a. Valves

(1) „Valves receiving a safety isolation signal which require reopening

to allow sampling during post-accident conditions: FCV-678, 679

and 681.

(2) New control valves installed in parallel with FCV-678, 679and'81.

(3) New manual valves, shown in Figure 1 will be mounted such that

the handles are shielded from any sample line radiation.

b. Indication

(1) New pressure, temperature and flow indication.

(2) Status lights for heat tracing'.

(3) Hydrogen. analyzer status and monitoring indicators.

(4) Valve status ligh'ts for. all.remote, operating valves.

(5) Sample air pump's status.

-51-


c. Contr ol s

'1} Sample air pump controls.P ~

(2) Fume hood ventilation fan.

REMOTE SAMPLE STATION SHIELDING

Two alternatives were considered for shielding sample tubing at the remote

sample station. The first'of these was to place the tubing coaxially inside

larger diameter lead-filled pipes. The second was to place. it behin'd sheet

steel. Sheet steel will be used because of the ease of fabrication and

it afforts greater design flexibility. The shielding should be removable

for maintenance and inspection purposes by placing all penetrations for

valve stems, instrument lines,'etc., along the seam between movable

shield plates.

Preliminary calculations by Radiation Research Associates (RRA) indicate

that in order to limit external contact dose rates to below 1 R/hr for an

8 ft. run of sample tubing, 6.7" and 2.8" of iron are required for reactor

coolant system liquid samples and containment atmosphere air sample lines,

respectively. The upper limit for the thickness of iron plate is then

about 7". After a detailed design tubing layout is approved, shielding

requirements will be finalized. The only unshielded portion of the

sampling system will be the quick disconnects which must protrude from

the steel plates. Movable shielding 'will be provided to 'cover the couplings

after the sample vessels are removed. Personnel exposure will be minimized

by using long-sterned valves and instrumentation which does not create a

ra'diation field at the gauge faces.

-52-


REFERENCES't C

1. Letter 12/14/79, K; Warkentin;(RRA) to,S. Skidmore (PGSE)

RRA.Letter O'RL4273-09

2. MSA Series 500 Combustible Gas Detection Systems;

Instruction Manual

3. TMI-2 Lessons Learn'ed Task Force Status Report and Short Term

Recommendations,. U. S. NRC, NUREG-0578, July 1979

4. Report to the Nuclear Regulatory Coranission from PG&E Responding

to NUREG-0578: TMI-2 Lessons Learned Task Force Status Report andE

Short Term Recommendations, PGRE,=-August 27, 1979

5. Clarifications of NUREG-0578, Enclosure 1, USNRC, November 9, 1979.

-53-

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Reactor Coolent LiquidSurface Dose Rate vs Radial Shield Thickness

ec102

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-56-


Extended Ran e Radiation Monitors

Each containment radiation monitor (there are two) is manufactured by

Victoreen and consists of an ionization chamber inside containment andK

a readout in the control room. The unit can respond to photon energies

as low as 60 KeV and has a range of 10 to':10. R/hr.

The plant vent monitor is manufactured by Victoreen, and consists of a

particulate prefilter, a silver zeolite filter, a noble gas monitoring

chamber, and appropriate air moving components. The iodine filter has a .

sodium iodide crystal monitor with a range of 10 uCi/cc to 10 uCi/cc,'nd

the noble gas chamber has a sodium iodide crystal monitor with a range

of 10 uCi/cc to 10 uCi/cc based on Xe-133.-5 5

Readouts are in the 'control room.

All parameters are recorded in the control room.

If our supplier meets current delivery date cormitments, installation

is expected to be completed'y August 1, 1980. For operation prior to

August 1, 1980, PG&E will provide the necessary instrumentation and methods

as required by the clarifications.

-57-

NUREG-0578 ITSY 2.1.9

PGGE Res onse and Statusr,~

Analyses of small break loss-of-coolant accidents, symptoms of inadequate

core cooling and required actions to restore core cooling, and analysisof transient and accident scenarios including operator actions not pre-viously analyzed are being performed on a generic basis by the Westing-

house Owners'roup, of which PG&E is a member. The small break analyses

have been completed and were reported in WCAP-9600, which was submitted

to the Bulletins and Orders Task Force by the Owners'roup'on June 29

1979. Incorporated in that report were guidelines that were developed

as a result of small break analyses. These guidelines have been re-viewed and approved by the B&0 Task Force and have been presented tothe Owners'roup utility representatives in a seminar held on October

16-19, 1979. Following this seminar, each utilityhas developed plantspecific procedures and trained their personnel on the new procedures.

The work required to address the other two areas —inadequate core cool-ing and other transient and accident scenarios —has been performed in

'onjunctionwith schedules and requirements established by the Bulletinsand Orders Task Force . Analysis related to the definition of inadequatecore cooling and guidelines for recognizing the symptoms of inadequatecore cooling based on existing plant instrumentation and for restoringcore cooling following a small break LOCA were submitted on October 31

1979. This analysis is a less detailed analysis than was originallyproposed, and will be followed up with a more extensive and detailedanalysis which will be available during the first quarter of 1980.

With respect to other transient accidents contained in Chapter 15 of theDiablo Canyon FSAR, the Westinghouse Owners'roup has performed an

evaluation of the actions which occur during an event by constructingsequence of event trees for each of the non-LOCA and LOCA transients .

From these event trees a list of decision points for operator action has

been prepared, along with a list of information available to the operatorat each decision point. Following this, criteria have been set forcredible misoperation, and time available for operator decisions have been

qualitatively assessed. The information developed was then used to testAbnormal and Energy Operating Procedures against the event sequences and

determine if inadequacies exist in the @Ps and EOPs. The results of this-58-

study will be provided to the Bulletins and Orders Task Force on

March Pl, 1980, as required.

The 9A1ers'roup has also provided test predictions analysis of the LOFZ

L3-1 nuclear small break experiment. This analysis was provided on

December 15, 1979, in accordance with the schedule established mutuallywith the Bulletins and Orders Task Force.

Procedures identified by these analyses are presently being written.The procedures and operator training will be completed by May 1, 1980.

-59-

CRS ITEM k4

PGEE Res onse.and Status

A head vent s~tem meeting the requirements of this comment will be installed at

Diablo Canyon... The syst: em design is being purchased from the nuclear steam supply

system manufacturer, Westinghouse. The remotely operated valves, which are the

item controlling the schedule, have been on order for several months and will be

delivered in September, 1980.

The head vent system will use the existing manual vent connection to the reactor

head, which is three-quarter inch diameter. Four one-inch electrically operated

valves meeting the requirements of IEEE-323-74 and IEEE-344-75 will be installed

in a series-parallel arrangement satisfying the single failure criterion. These

valves, will be operable from and position indicated in the control room. They willfail closed. A flow diagram showing the arrangement schematically is attached.~

~

~ ~F

three-eights inch diameter orifice will be installed in each of the parallel vent

lines between the existing manual valve and the new electrically operated valves.

The new valves, orifices, and connecting pi~ will be supported'from the head

lifting structure and will be qualified seismically for this location. Instrumen-

tation and control will meet the requirements of IEEE-279.

The orifices in the vent lines would restrict the flow thru a postulated break in

the new vent system to that which normal makeup can maintain. This would preclude

the possibility of the occurrence of a small loss of coolant accident in the head

vent system. (A rupture in the existing manual head vent line is covered by the

small break LOCA discussion in Chapter 15 of the FSAR and by Westinghouse WCAP-9600

on the same subject which has been submitted to the NRC staff by Westinghouse.)

In a Westinghouse reactor coolant system, the head vent will also serve to remove

~

~

non-condensible gases from the "hot leg" and'Infold leg" reactor coolant system

iping. Should the existence of non-condensible gases in the tubes of a steam

generator be suspected, it is proposed to."bump" the related reactor coolant pump

so as to move the gases

head vent system.

to the reactor vessel, where they may be removed by the

-60-

ACRS ITEN,84 continued

ipment and piping for React. or Coolant System venting willbe installed, pro-

cedures finalifed and operator training completed by January 1, 1981. Preliminary

calculations cbnfirm that a gas'volume greater than one-half the reactor coolant

system volume Could be vented by either train in less than one hour. .On the other

hand, the procedure will call for venting to be terminated before the hydrogen

concentration in the containment could exceed 4% under the most adverse assumptions.

Venting would not be continued until the actual containment hydrogen concentration

achieved had been determined.

The head vent system described would vent into the area immediately ''surrounding

the reactor vessel head. This area is swept by ventilating air flow and gases

released would be mixed into the containment volume by this air flow. Consideration

has been given to piping the vented gases to other locations, for example, the

~

~

~ressurizer relief tank. At this time, the identification or advantage of such an

alternate vent release location is not known. However,, should an alternate location

be required in the future, an extension of the described system could be added.

Inadvertent opening of the head vent system would be detected by the valve position

lights in the control room and by the leakage monitoring system which is described

in Section 5.2.4 of the FSAR.

Venting of the pressurizer, should that be necessary, can be carried out by the

use of the pressurizer power operated relief valves, which at Diablo Canyon are

seismically qualified and have qualified control circuits.

-61-

TO CONTAINMENT

S S

S

SEE NOTE 1

I

3/8" OR IF ICES

REACTORVESSEL

HEAD

NOTES:1. EXISTiNG VENT LINg

Ftow Diagram of the Reactor Vessel Head Vent +stem

NUREG-0578 ITEM 2.2.l.b

~ ~ ~

PGGE Res onse and Status

The "employment positions have been identified and authorized by PGGE. PGGE

attempting to locate six qualified individuals to hire for these posit'ions.~ ~

is

Every

effort will be made to fillthese positions as rapidly as possible. In any event,

the~e positions will be filled and the personnel trained to allow operation by

May 1, 1980. The STA job description is as follows:

Basic Res nsibilitTo ensure increased plant safety by providing an accident assessment function in

the control room at all times and to provide on-shift operating experience assess-

ment capability.

ecific Duties

1. Provides sound technical advice to the Shift Foreman during transient or

accident conditions.

2. Evaluates plant equipment performance and makes recommendations for improve-

ment.

3. Reviews operating experience and equipment performance at other nuclear plants.

4. Ensures that operating and emergency procedures are revised and operators

trained based on his review of operating experience.

5. Provides assistance to the Shift Foreman in reviewing surveillance test results.

6. Assists the Shift Foreman in first aid and fire protection training.

7. Advises the Shift Foreman on the plant Technical Specifications.

8. Conducts operator training as specified by the Training Coordinator when his

shift. is on a training assignment.

9. Assist the Shift Foreman in the preparation of Special Work Permits.

10. Nay perform independent audits of operating practices and procedures.

Provides a critique of fire and other emergency drills.Provides liaison between the operating and technical groups at the plant.

'

-63-

~ 'TEM 2.2.l.b .continuedCp

lationshi s

Reports directly to the Shift Foreman and is responsible and accountableC-

to him in his advisory capacity.

2. Reports indirectly to the Senior Power Production Engineer (Operations

Engineer) and is responsible and accountable to him in administrative

and technical matters.

3. Maintains close working relations with the Maintenance and Technical Depart-

ments.

4. At the direction of the Shift Foreman or Operations Engineer, confers with

other departments and agencies both inside or outside the Company.

Authorities

l. To carry out his assigned duties.

To take action in an emergency to the full extent justified. Such action willbe reported to the Shift Foreman and Operations Engineer as soon as practical.

To consult with and/or advise other plant supervisors on activities related to

plant operations.

-64-

forwards addendum to 791231 rept, 'status rept to nrc from ... · ad $ ktei aal'isis...

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