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December 23, 1992
U.S. Nuclear Regulatory CommissionAttn: Document Control DeskWashington, D.C. 20555
Subject: Waterford 3 SESDocket No. 50-382License No. NPF-38Reporting of Licensee Event Report
Gentlemen:
Attached is Licensee Event Report Number LER-92-015-00 for Waterford SteamElectric Station Unit 3. This Licensee Event Report -is . submitted inaccordancewith10CFR50.73(a)(2)(ii).
Very truly yours,
$b W0.F. PackerGeneral Manager - Plant Operations
DFP/TWG/ssfAttachmentcc: J.L. M11hoan, NRC Region IV
G.L. FlorreichJ.T. Wheelock - INP0 Records Center
-R.B. McGeheeN.S. ReynoldsNRC Resident Ins)ectors OfficeAdministrator .RPD,
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J.D. llolman, Safety ard Engineering Analysis Manager (504) 739-6265COiM5Ei[ ONI. LINL F OU f ACH COMPONENT I IiiiiiE~DLTURIHLD IN THIS' HLII6Eii (13)
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On November 25, 1992, Waterford 3 confirmed that the previously assumed post-accident dose rate at the manual operators for two containment isolation valveswas non-conservative. Because personnel access to the area is required under.certain low probability accident scenarios, the true radiation levels in the areacould have resulted in a dose to personnel greater than that allowed by GeneralDesign Criteria 19.
The root cause of this event is a calculation which was based on incorrectinputs. Specifically, one of the Waterford 3 Architect / Engineer's calculationsdid not account for the presence of certain piping in the Reactor AuxiliaryBuil. ding which could contain highly radioactive water under certain conditions.
The installation of temporary shielding will allow manual valve operation witha dose less than 5 rem. Additional corrective action will include review ofother dose calculations and investigation of changes to threo containmentisolation valves so that local manual action will not be required. This eventposed no risk to the health and safety of the public.
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05000 382 20F8Waterford Steam Electric Station Unit 3 92 015 00
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REPORTABLE OCCURRENCE
On November 25, 1992, with the plant operating in Mode 1 at 100% power, Waterford
3 confirmed that the FSAR value for the post-Loss of Coolant Accident (LOCA) dose
rate at the manual operators for Component Cooling Water System (CCW; EIIS
Identifier CC) containment isolation valves CC-641 and -713 were non-conservative. Personnel access to the affected area to operate valves isrequired under certain low probability accident scenarios. Because the trueradiation levels in the area were not documented in the FSAR and could haveresulted in a post-accident dose to personnel greater than that allowed byGeneral Design Criteria 19, Waterford 3 was operated outside of its design basisfrom initial startup until shutdown for the fifth refueling outage in September1992. As such, this event is reportable as an LER in accordance with
10CFR50.73(a)(2)(ii).
INITIAL CONDITIONS
Plant Power: -100
Mode: 1
Procedures Being Performed Specific to this Event: None
Technical Specification LC0's in Effect Specific tothis Event: None
Major Equipment Out of Service Specific to this Event: None
H H T SEqMMCI
The Waterford 3 Component Cooling Water System is a closed cooling water system
serving reactor auxiliaries. Heat is removed from the system by dry-cooling
towers and by the CCW heat exchangers. A nonessential seismically qualified
(closed) loop services non-safety related equipment located inside- the
containment including the reactor coolant pump (EIIS Identifier AB-P) seals and I
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the control element drive mechanism (CEDM; EIIS Identifier AA) coolers.
Upon receipt of a Containment Spray Actention Signal (CSAS), the three air i
operated containment isolation valves (CC-641, CC-710, and CC-713) on the CCW
supply and return lines to the Reactor Coolant Pumps and CEDM coolers are
automatically isolated. The CCW supply line to containment also includes a check
valve inside containment (CC-644) located downstream of CC-641. Each of the
three containment isolation valves serving this loop is a pneumatically operated
valve designed to fail open on loss of air. Motive gas for the valves is
normally supplied by the non-safety Instrument Air System (Ells Identifier 1.D).
Each valve is also provided with a safety-related air accumulator to supply
motive gas should the Instrument Air system not be available (the Instrument Air
system may be energized from an emergency diesel generator if off-site power is
lost). The accumulators are sized for a nominal 10 hours beyond which local
manual closure of CC-641 and CC-713 is necessary to isolate containment if the
instrument air system is not available and the CCW piping inside containment has
been damaged during the event and is no longer intact. The manual operator for
CC-710 is inside the containment and is not accessible post-accident.
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During the fifth refueling outage, pressur e testing performed on the accumulator
associated with valve CC-710 indicated that the accumulator was leaking at a rate
that would exhaust the backup gas supply in less than 10 hours. In response to.
that finding, the Waterford 3 Design Engineering group began reviewing the basis
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of the requirement that the CC-710 accumulator last for 10 hours. A central part
of that review included determining whether the radiation dose near the manual
operator for CC-713 would allow access by personnel before 10 hours in order to ,
shut the valve. Since CC-710 and -713 are the inside and outside containment
isolation valves for the CCW return line from containment (respectively), the
ability to manually close C0-713 before the CC-710 accumulator was exhausted was
necessary to ensure containment isolation in the event that the CCW line inside.
containment was ruptured due to a LOCA and the Instrument Air system had not yet
been restored.
On November 3, 1992, while the plant was in its fifth refueling outage, the
Design Engineering review of radiation dose calculations in the area of the'
manual operators identified the fact that the dose calculations in the FSAR did
not account for the presence of several Containment Spray and Safety injection
lines which pass through the Reactor Auxiliaries Building (RAB; ElIS Identifier
| NF) -4 Wing area. These lines can contain highly radioactive water under certain
accident conditions.
On November 7, 1992, the Waterford 3 Architect / Engineer completed a draft|
calculation for the ' radiation levels at the manual operators for valves CC-641
and -713. The reported radiation levels in the area were approximately 180
Rem / hour at 6 hours post-LOCA, unshielded. The dose rate at 10 hours was
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previously estimated to be on the order of 200 mrom/hr.
On November 8,1992, shielding was erected around the Engineered Safety feature:
piping in the -4 RAB Wing area. This additional shielding is sufficient to1
reduce the dose to operators entering this area at 6 hours post-LOCA to less than
the 5 Rem requirement of General Design Criterion (GDC) 19.
On November 25, 1992, the final Architect / Engineer calculations regarding the
unshielded radiation levels on the -4 RAB Wing area were received at Waterford
3. The calculations indicated that the post-LOCA dose rate at the manual1
operators for containment isolation valves CC-641 and -713 was 231 Rem / hour
thereby confirming that the values included in the FSAR were non-conservative.
110 wever, the previously installed shielding is still sufficient to satisfy the
requirements of GDC 19, given the revised radiation level in the area.
CAUSAL FACTORS
The root cause of the incorrect post-accident radiation levels in the in the -4
RAB Wing area is a calculation using incorrect inputs. Specifically, one of the
Waterford 3 Architect / Engineer's calculations, performed as part of the post-
accident shielding study required by NUREG-0737, " Clarification of Post-THIu
Action. Plan Requirements," did not account for the presence of Safety Injection i
and Containment Spray piping in the -4 RAB Wing area. This piping is expected-!
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to contain highly radioactive water from the Safety Injection Sump once the
recirculation phase of the accident recovery process begins.
C9RRECTIVE ACTION
First, because personnel access to the operators for CC-641 and CC-713 might be
required in certain accident scenarios, shielding has been erected around the
Containment Spray and Safety Injection system lines in the -4 RAB Wing area.
This shielding will reduce the dose to operators entering the wing area to close
CC-641 and CC-713 to less thaq the 5 Rem requirement of GDC 19.
Second, calculations for the NUREG-0737 post-accident shielding study will be
revised to reflect the existence of the Containment Spray and Safety Injection
piping in the -4 RAB Wing arca. Confirmatory analyses will be performed to
demonstrate that the dose rate for operator access to the +21 RAB Wing area is
within the criteria established by GDC 19. In addition, the Waterford 3
Architect / Engineer has reviewed the piping model throughout the remainder of the
RAB and judged it to be correct.
Third, calculations to determine Equipment Qualification (EQ) dose requirements
for the -4 and +21 RAB Wing areas,will be reviewed and, if necessary, updated,
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entry to the -4 RAB Wing area to close valves CC-641 and CC-713 in order to
maintain containment integrity after a LOCA.
The analyses described above will be complete by July 13, 1993.
SAFETY SIGNIFICANCE
Probabilistic Risk Analysis (PRA) results indicate that the probability of an
event sequence that could result in the need to shut CC-641 and -713 to maintain
the containment isolated is less than IE-7 per year. If an event had occurred
which required CC-641 and -713 to be shut and radiation levels in the -4 RAB Wing
area prevented operators from manually closing the two valves without exceeding
GDC 19 requirements, alternative actions were available which could have
maintained the containment isolated. One example would be the restoration of the
Instrument Air system to supply the motive force necessary to maintain the
containment isolation valves closed. A second alternative which would have
maintained the containment isolated would have included isolating the CCW supply
and return lines by closing CCW valves outside of the -4 RAB Wing area.
In the event that no other alternative was available, estimates indicate that
before the shielding was installed, two entries into the -4 RAB Wing area to shut
CC-713 and'-641 could have been made with an accumulated dose of less than 25 Rem,
for each entry. While not inconsequential, this dose is within standards
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established for the emergency operation of equipment.
Given these factors, this event posed no risk to the health and safety of the
public.
SIMILAR EVENTS
None.
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