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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

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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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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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