ML20155G374

Forwards Response to NRC 980619 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions

ML20155G374
Person / Time
Site:	Oyster Creek
Issue date:	10/29/1998
From:	Roche M GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1940-98-20574, GL-96-06, GL-96-6, NUDOCS 9811090039
Download: ML20155G374 (9)

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{ GPU Nuclear, Inc.

A U.S. Route #9 South NUCLEAR Post Office Box 388 Forked River, NJ 087310388 Tel 609-971-4000 October 29,1998 1940-98-20574 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington,DC 20555 Gentlemen:

Subject:

Oyster Creek Nuclear Generating Station (OCNGS)

Docket No. 50-219 Facility Operating License No. DPR-16 Request For Additional Information Concerning Generic Letter 96-06 Pursuant to your letter of June 19,1998 and GPUN's letter of September 17,1998, please find attached the requested information.

If there are any questions or additional information is required, please contact Mr. Joseph D. Lachenmayer of our staff at 973-316-7971.

Very truly yours,

- Sec Michael B. Roche Vice President and Director Oyster Creek Enclosure Attachments cc: Administrator, Region I NRC Senior Resident Inspector NRC Project Manager 'f

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9811090039 DR 981029 ~

ADOCK 05000219 PDR n. -

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l REQUEST FOR ADDITIONAL INFORMATION FOR RESOLUTION OF i l GENERIC LETTER (GL) 96-06 ISSUES AT  !

OYSTER CREEK NUCLEAR GENERATING STATION I

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l GL 96-06," Assurance of Equipment Operability and Containment Integrity During l

Design-Basis Accident Conditions," dated September 30,1996, included a request for licensees to evaluate cooling water systems that serve containment air coolers to assure that they are not vulnerable to waterhammer and two-phase flow conditions. General Public Utilities Corporation (the licensee) provided its assessment of the waterhammer and two-phase flow issues for Oyster Creek in letters dated January 28, and February 26,1997. The licensee indicated that the drywell cooling units and associated reactor building closed cooling water system are not safety related and are not required for accident mitigation.

Ilowever, the Emergency Operating Procedures (EOPs) did allow operators to use the drywell cooling units following an accident if available, and the EOPs were revised to eliminate the potential for waterhammer following a loss-of-coolant accident. In order to assess the licensee's resolution of these issues, the following additional information is requested:

Issuei Describe the revisions that were made to the EOPs to eliminate the potentialfor waterhammer.

ifIso discuss to what extent these revisions eliminate the potentialfor two-phaseflow.

Resnonse GPUN revised the Emergency Operating Procedures (EOPs); specifically EMG-3200.02, Primary Containment Control instructions for " Maximizing Drywell Cooling". The revised procedure specifically prohibits the EOPs from re-establishing RBCCW flow to the Drywell. This eliminates the potential for waterhammer events described in Generic Letter 96-06.

There are, however, some Small Break LOCAs that do not cause isolation of the RBCCW system i in w hich drywell temperature increases above the saturation temperature of the fluid in the l Drywell Cooling Units (297 F). In these cases, the Drywell Cooling Units would continue to operate, however, two-phase conditions do not develop. Continued flow of RBCCW coolant through the coils and the rate of heat transfer to these units are such that the fluid within the system does not change phase. The Drywell Cooling Units and the RBCCW system are not required to mitigate these types of accidents.

lhen though it is not expected that the RBCCW pumps will trip under these conditions, a spurious trip of these pumps must be considered because ofit's potential impact. When this occurs, the fans will continue to operate since there is no interlock between fan and pump operation. With the pumps tripped, RBCCW fluid pressure will drop to approximately 19 psig which has a corresponding saturation temperature of 257 F. If RBCCW were lost within the first ten minutes of the accident (prior to manual initiation of containment sprays) containment conditions would likely exceed the saturation conditions for a large percentage of these break sizes. With the fans in operation significant void formation would be expected to occur. To l

prevent either a water hammer or two-phase condition, the EOPs will be revised to instruct the operator to isolate the RBCCW system from the drvwell during a LOCA or Main Steam Line

, Break.

, issue 2 Implementing measures to assure that waterhammer will not occur, such asprohibitingpost-l accident operation ofthe afected system, is an acceptable approachfor addressing the

waterhammer concern. However, allscenarios must be considered to assure that the vulnerability to waterhammer has been eliminated. Confirm that allscenarios have been considered, including those where the afected containmentpenetrations are not isolated (ifthis is apossibility), such that the measures that have been establishedare adequate to prevent the l occurrence of waterhammer during (andfollowing) allpostulated accident scenarios.

l l Response l When assessing the vulnerability of the drywel! coolers to water hammer, a variety of scenarios are considered as summarized in Attachment 1. There are two basic classifications, those where  ;

RBCCW flow to the drywell coolers isolates automatically and those where automatic isolation  ;

. does not occur. The case where automatic RBCCW flow isolation does occur is addressed by 1 maintaining the RBCCW system in an isolated state. The case where the RBCCW system does not isolate automatically requires further discussion.

The RBCCW system is designed to automatically isolate on a combination of High Drywell Pressure and Low-Low Reactor Water Level, or Low-Low Low Reactor Water Level alone.

Since this issue is associated with the interaction between a hot steam filled containment atmosphere and the Drywell Cooling Units, it is reasonable to expect that the high drywell pressure condition must be present in order for this issue to be a concern. Additionally, these conditions (two phase flow) can only develop when the RBCCW flow is lost (i.e. spurious pump trip, pipe break, etc..) and the system does not isolate because reactor water level is maintained above the isolation setpoint. This may occur for breaks where offsite power is maintained such that a high pressure injection source (i.e., feedwater)is immediately available with suflicient capacity to maintain reactor water level.

The entire evaluation of the cooling coil vulnerability to void formation is predicated on the assumption that the fan motors continue to operate in a steam atmosphere. This is not believed to be likely, however, the assumption is adopted for analysis purposes. When evaluating the failure to isolate scenarios it is necessary to determine if the resulting environment will produce boiling within the cooling units. This leads to a further division of the possible scenarios into those where containment temperature exceeds the saturation temperature of the fluid flowing to the drywell cooling units and those where it does not.

When the cooler is not isolated, fluid flows to the coils at a pressure of 50 psig having a saturation temperature of 297 F. In order for the fluid flowing through the coolers to boil the drywell atmosphere must reach temperatures that exceed the 297 F saturation temperature of the fluid. l This 297*F RBCCW saturation temperature exceeds the temperature of all loss of coolant l accidents within the Oyster Creek design basis. Therefore, it can easily be concluded that for all  ;

' loss of coolant accidents where isolation does not occur and the RBCCW system remains  !

- operational, neither water hammer nor two-phase flow will occur in the drywell cooling units.

The same conclusion is reached regarding large main steam line breaks where containment atmosphere temperature remains saturated and below the 297 F required for boiling to occur in the RBCCW system.

l However, small and intermediate size failures of the steam system inside the drywell may lead to j superheated temperatures that exceed the 297 F RBCCW saturation temperature. For these l l

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' breaks, the cooling coil How and the rate of heat transfer to the coil is such that the possibility of two-phase How does not exist. This was demonstrated by calculation using the GOTilIC (version ,

6.0a) computer code fan cooling coil model. Furthermore, the elevated containment temperature  !

conditions would not persist since the EOP would have the operator trip the drywell cooling fans and initiate drywell sprays (rapidly reducing the temperature in the containment). I When containment sprays are manually initiated in drywell spray mode the drywell cooler fans are manually tripped w hich will significantly reduce the heat transfer to the system. First, the containment atmosphere temperature will be reduced by the drywell spray initiation. Second, the rate at which heat is transferred to the coolers is decreased substantially when the fans are tripped. l The Gnal aspect of the evaluation is the scenarios where there is no isolation of the RBCCW system to the drywell and the RBCCW pumps trip. When this occurs, the drywell fans will continue to function since there is no interlock between fan and pump operation. In addition, the cooling Duid pressure will drop to 19 psig which has a corresponding saturation temperature of 257 F. If RBCCW were lost within the Grst ten minutes of the accident (prior to manual initiation of containment sprays) containment conditions would likely exceed the saturation ,

conditions for a large percentage of the break sizes. With the fans in operation significant void l

formation would be expected to occur, To prevent a water hammer condition, the EOPs will be l revised to have the operator isolate the RBCCW system from the drywell during a LOCA or Main Steam Line Break, issue 3 Ifthe potentialfor two-phasepow has not been eliminated, provide thefollowing information:

a. Identify any computer codes that were med in the two-phasepow analyses anddescribe the methods used to bench mark the codesfor the specific loading conditions involved (see Standard Review Plan Section 3.9.1).

b. Describe andjustify all assumptions and input parameters (including those used in any computer codes) and explain why the values selectedgive conservative results. Also, providejustipcationfor omitting any effects that may be relevant to the analysis (e.g.,

pow induced vibration, erosion).

c. Provide a detailed description of the " worst case" scenariofor two-phasepow, taking into consideration the complete range ofeventpossibilities, system confgurations, parameters, and componentfailures. Additional examples include: 1

• the consequences ofsteamformation. transpo~t, and accumulation; l

• cavitation, resonance, andfatigue effects; and

• erosion considerations.

Licensees maypndNUREG/CR-6031, " Cavitation Guidefor Control Valves,"

\ helpful in addressing some aspects ofthe two-phaseflow analyses. (Note: it is

\ importantfor licensees to realise that in addition to heat transfer considerations.

l two-phasepow also involves structural and system integrity concerns that must be addressed) l

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d. Confirm that the two-phaseflow loading conditions do not exceed any design specifications or recommended service conditionsfor the piping system and components, including those stated by equipment vendors; and confirm that the system will continue to perform its design-basisfunctions as assumedin the safety analysis reportfor thefacility, and that the containment isolation valves will remain operable.

e. Determine the uncertainty in the two-phaseflow analyses, explain how the uncertainty ,

was determined, and how it was accountedfor in the analyses to assure conservative l results.

f Confirm that the two-phaseflow analyses included a completefailure modes and efects \

analysis (FMEA)for all components (including electrical andpneumaticfailures) that could impactperformance ofthe cooling water system and confirm that the FMEA is i

documented and availablefor review, or explain why a complete andfully documented FMEA was notperformed.

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g. Explain andjustify all uses of "engineeringjudgement. "

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Response

We believe the potential for two-phase Dow has been eliminated per the discussions above.

Please note that, except for the Containment isolation feature, the Drywell Cooling Units and

, RBCCW system do not serve Nuclear Safety Related functions and are not required to mitigate

! design basis accidents. Therefore even though the revised procedure eliminates its possibility, if j two-phase Dow were to somehow occur, the resulting heat exchanger degradation or system Cow i degradation would not directly correspond to a reduction in the capability of Safety Related l Systems to mitigate design basis accidents.

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i Issue 4 Provide a simphpeddiagram ofthe afectedsystem, showing major components, active components, relative elevations, lengths ofpiping runs, and the location ofany orifices andjow restrictions.

Response

! A simplified diagram of the RBCCW system is provided as Attachment 2.

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l ATTACllMENT I l

SUMM ARY OF SCENARIOS

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l Event Assumptions Fan Cooler Status Operator Actions Result l Large Break below With offsite Fan Coolers Trip Maintain the system No water hammer core LOCA power RBCCW Isolates isolated in the system l Without Fan Coolers Trip Maintain the system No water hammer ofTsite power RBCCW lsolates isolated in the system Small Break below Without Fan Coolers Trip Maintain the system No water hammer core LOCA offsite power RBCCW isolates isolated in the system

. With offsite Fan Coolers don't Trip Fan Coolers if No water hammer l

) power trip and RBCCW Drywell Sprays are in the system. No I i remains operational required. Manually two phase flow in isolate RBCCW to the the system DW.

With otTsite Fan Coolers don't Trip Fan Coolers if No water hammer power and trip and RBCCW Drywell Sprays are in the system.  !

loss of trips. required. Manually l

RBCCW flow isolate RBCCW to the l

DW. '

Large Break above Without Fan Coolers Trip Maintain the system No water hammer the core MSLB offsite power and RBCCW isolated in the system. No isolates two phase flow With offsite Fan Coolers don't Trip Fan Coolers if No water hammer power trip and RBCCW Drywell Sprays are in the system. No remains operational required. Manually two-phase flow.

isolate RBCCW to the DW.

With offsite Fan Coolers don't Trip Fan Coolers if No water hammer power and trip and RBCCW Drywell Sprays are in the system.

l loss of trips. required. Manually RBCCW tiow isolate RBCCW to the D W.

Small Break above Without Fan Coolers Trip Maintain the system No water hammer the core MSLB offsite power and RBCCW isolated in the system.

Isolates With offsite Fan Coolers don't Trip Fan Coolers No water hammer power trip and RBCCW Drywell Sprays are in Ee system. No remains operational. required. Manually Two phase flow.

isolate RBCCW to the l DW.

With offsite Fan Coolers don't Trip Fan Coolers No water hammer j power and trip and RBCCW Drywell Sprays are in the system.

{ loss of trips. required. Manually j RBCCW llow isolate RBCCW to the DW.

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I ATTACIIMENT 2 SIMPLIFIED DIAGR.AM REACTOR CLOSED COOLING WATER SYSTEM 1

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9 REACTOR BUILDING CLOSED COOLING SYSTEM SUPPLY TO DRYWELL COOLING UNITS 6  % MAKE UP NON DRYWELL SURGE Ik LC TANK Lt V-5-102 i

R8 EL 95'-3 D Cmc l WATER PUMP 1-1 12' CC-4 (12" )

i RB EL 51'-3* V-5-130 HEAT d \l (L N lx: l\l EXCHANGER l-l

-H\W V-5-151 V-5-131 V-5-149

_ (12" )

i CLOSED C00UNG V-5-685 8 WATER PUMP Il2")

• _ 2* CC-4 e R8 EL Sr..y V-5-132 HEAT

( -

N  :\l \H- EXCHANGER -H\W M'V-5-150 V-5-152

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V-5-133 (12" ) X i 33 _ o-V-5-684 8 TOTAL LENGTH -

(6* ) V-5-165 @ l DW EL 46*-0*

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V-5-148

._ - RB 23'- 6*

8 TOTAL LENGTH JL X

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11 *-0* N

'[OTOTAL 3 g

' LENCTH -

8 - -- - -- --

DRYWEU.

U g3C d y C

y C C --

C 0 g e Cg D EL 46 0*( ) g 2 5*- 6* E 'e *.

-O TOTAL *e i I LENGTH Q Q V-5-711[ [ [

"C i V-5-167 V-5-166

! A RB EL 23'-6= DW EL 46'-0*

I A N-

! - NON

' DRYWELL f

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