HL-4794, Forwards Response to NRC 950213 RAI Re Scope of GL 89-10 for Plant

Forwards Response to NRC 950213 RAI Re Scope of GL 89-10 for Plant

ML20081F089
Person / Time
Site:	Hatch
Issue date:	03/14/1995
From:	Beckham J GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-89-10, HL-4794, NUDOCS 9503210526
Download: ML20081F089 (10)

Text

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o Georgis Power Company i t 40 inr,rness Center Parkway

'- Post Offica Box 1295

Birmingham. Alabama 35201 '

, f ..- Telephone 205 877-7279 '

J. T. Beckham, Jr. . Georgia Pbwer . ,

24, Vice President . Nuclear g March 14, 1995 -

'l Docket Nosi50-321 HL-4794

'50-366' l

U. S. Nuclear Regulatory Commission d

~ ATTN: Document Control Desk .

Washington, D. C. 20555 Edwin I. Hatch Nuclear Plant Units 1 and 2 Response to Request for Additional Information:

Generic Letter 89-10 l

Gentlemen:

In response to your letter dated February 13,1995, " Request for Additional Information

.Regarding Scope of Generic Letter 89-10 for Edwin I. Hatch Nuclear Plant Units 1

]q and 2," Georgia Power Company (GPC) submits the following responses to your '

questionsin Enclosure 1.

If you have questions regarding this matter, please contact this office.

Sincerely, f

UN N

. T. Beckham, Jr.

OCV/eb

Enclosure:

Request for Additional Information - Generic Letter 89-10 cc: Georgia Power Company Mr. H. L. Sumner, Jr., Nuclear Plant General Manager NORMS LI. S. Nuclear Regulatory Commission. Washington. D. C.

Mr. K. Jabbour, Licensing Project Manager - Hatch U S. Nuclear Regulatorv Comminion. Region 11 Mr. S. D. Ebneter, Regional Administrator Mr. B.' L. Holbrook, Senior Resident Inspector - Hatch 9503210526 950314 >  ! -

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.. Enclosure 1 I Edwin I. Hatch Nuclear Plant Request for AdditionalInformation:

Generic Letter 89-10

' 1. Plant Service Water Dilution Line Isolation Valves (P41-F312 for Unit 1 and P41-F310 for Unit 2)

This valve has an active safety function to automatically close. However, its normal position is closed.

Discuss your procedural requirements to ensure that this valve is closed at all times )

and how you ensure that it would not open.

RESPONSE

Both valves for Units 1 and 2 are listed in the valve line-up section of their respective System Operating Procedure as closed. Furthermore, to ensure the valves are not opened by inadvertent operation of the control switch, the system line-ups in these procedures also require that the breakers be turned to the "off' position.

(Ref. 34S0-P41-001-IS and 34SO-P41-001-2S.)

Under the previous guidance, before our re-evaluation was completed early last year, these valves had active safety functions to close. However, the valves were reclassified as having no active safety function since they are closed and remain closed for all design basis events. These valves were designed to be used for the ,

dilution of radwaste with Plant Service Water (PSW). However, radwaste dilution flow, for both Hatch Units 1 and 2, is provided via Circulating Water System valves, therefore, the PSW valves remain closed. These valves are opened during logic .

system functional testing, performed once per eighteen months for each unit.

However, per criterion no. 3 (Reference enclosure 3 of 2-3-94 submittal) of our safety function guidelines, valve operability for GL 89-10 purposes is not required  !

during component testing.

2. Reactor Core Isolation Cooling (RCIC) Valves (E51-F013 and E51-524)

Tojustify removal of the above valves, confirm that they do not have to cycle during a DBA to initiate, isolate, or re-initiate RCIC operation assuming the RCS becomes .

isolated. This should consider both original initiation of the system as well as j recovery from RCIC isolation on high vessel level as the system cycles to maintain ]

RCS level.

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Generic Letter 89-10 RESPONSE .!

Re-injection with RCIC is desirable but not necessary since inventory makeup can be i

accomplished by vessel depresurization using ADS followed by vessel makeup using either LPCI or core spray.

The RCIC injection valve (E51-F013) is normally closed and receives an open signal I on an RCIC system initiation signal. Accordingly, the valve has an active function to open.

Credit is taken for the RCIC system for design basis transient events involving loss of feedwater. The RCIC is not considered an emergency core cooling system (ECCS); 4 therefore, it is not assumed to be available for the loss of coolant accident (LOCA).

The RCIC operation for design basis transient events is initiated on low water level.

Consistent with the requirements of GL 89-10, the steam admission valve (E51-F045) )

and injection valve (E51-F013) are required to open to permit initiation of RCIC. ]

The RCIC is tripped on high water level. The high water level trip signal closes the  !

steam admission valve (not E51-F524) and, following closure of the steam admission . .;

valve, the injection valve is closed. It should be noted that the event has been turned 1 around (core re-flooded with pressure stabilized) prior to RCIC being secured.

Subsequently, only inventory makeup necessary to compensate for steaming is required. Numerous systems, such as HPCI, CRD, Feedwater and ADS followed by low pressure system injection, can accomplish this function.

3. Outboard Low Pressure Coolant Injection (LPCI) Valve (E11-F017 A & B) k The staff questions the basis for removal of the above valves from the GL 89-10 ,

program, because they may have to perform an active safety function in response to a i DBA. If one or both of these valves are closed to direct flow to containment cooling, they must be reopened to restore LPCI to the vessel for long-term accident recovery.

The February 3,1994, submittal stated that ifone valve does not open, the other loop _

could be opened to provide injection to the reactor pressure vessel (RPV). This assumption does not account for the possibility that both valves could malfunction due to a common cause defect. If these valves do not need to function in response to a design-basis event, then you should demonstrate that the restoration of RPV flow is not safety significant.

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RESPONSE

A common cause defect which simultaneously affects both valves in the "A" and "B" loop is beyond the licensing basis of the plant. ' Initial core cooling and core 3 reflooding for the design basis LOCA is accomplished by the operation of LPCI and core spray systems, which involves the opening of the inboard injection valves.

E Adequate core cooling can be accomplished assuming the limiting single failure.

l Diversion to suppression pool cooling is not needed until approximately 30 minutes .

post DBA LOCA. At this point in time, the core has been reflooded, and only the make-up necessary to maintain water level is required. Any single low pressure ECCS system is capable of providing makeup.

Accordingly, Georgia Power Company does not intend to include the El1-F017 ,

valves in the scope of the GL 89-10 program. 4

4. Test Valves -

It appears that the category of safety system test valves is absent from your GL 89-10 program. These valves co defeat safety system response when in the test position.

Many of these valves receive auto reposition signals upon safety injection. Some safety systems are assumed to remain operable even when in test position due to the capability of these test valves to return to their " safety position" under DBAs. Since -

repositioning of these valves are relied upon to justify the availability of various safety system responses, they should be included in the GL 89-10 program. l

RESPONSE

i System test valves are not included in the GL 89-10 program because these valves, consistent with Technical Specifications philosophy, are not assumed to be called upon to operate during the short periods of time the system is in test.

Criterion 3 of the GPC submittal of February 3,1994, submittal states, " Valve operability requirements, as defined by GL 89-10, is not required during periods of  :

system or component testing." This is consistent with the reliability assumptions that form the basis for the Unit I and Unit 2 Technical Specifications. This criteria is ,

consistent with the original GE specifications for the ECCS which states that it is not necessary for the test valves to re-align within the accident response times. This approach is considered acceptable because the period of time that standby systems are in the test mode during normal plant operation is small in comparison with the time ,

that the systems are in the standby mode of operation during normal plant operation. l Thus, these valves were not intended to recover from the test mode in a timely HL-4794 El-3 l

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manner to mitigate the consequences of design basis accidents. . Consequently, current design basis event analyses do not consider an additional single failure for l event mitigation when a system is in the test mode in accordance with Technical ~

Specification testing requirements. Therefore, GPC does not intend to place these l valves in the scope of the GL 89-10 program. j

. 5. Containment Sorav Valves (E11-F016 A & B. E11-F021 A & B. E11-F027 A & BF Georgia Power's analyses indicate that the drywell and suppression pool spray functions are not needed to maintain the containment below its design temperature and pressure. ' The licensee should also confum that the spray function serves no  : i radiological safety function and is not needed for environmental qualification purposes. Based on these analyses and subject to these confirmations, the spray valves, for purposes of GL 89-10, may be considered not to have an active safety function to open. Since the spray valves are normally closed, it is unlikely that they would be required to reposition to the closed position for purposes of containment isolation during an accident. Thus, for purposes of GL 89-10, the spray valves may be considered not to have an active safety function to close.

RESPONSE

No credit is taken for the operation of the containment spray system in the performance of the radiological evaluations for design basis events.

The drywell and wetwell sprays are assumed to function for environmental.  ;

qualification purposes.

t However, the spray valves are not required to be considered for GL 89-10 purposes for the following reasons:  ;

For design basis events, small steamline breaks inside the primary containment

(<0.5 ft 2) result in the highest containment temperature. For these breaks, only the _ ,

ADS and low pressure injection systems are realistically required to function after ,

containment isolation. The emergency operating procedures require that the drywell sprays be initiated prior to the drywell reaching its design temperature. Ifit is determined that the drywell sprays cannot be initiated, emergency depressurization will be initiated. Thus, if the containment spray valves fail to open, the operators . ,

t would proceed to initiate emergency depressurization of the reactor vessel, directing steam flow to the suppression pool. The ADS components are qualified to higher

< temperatures than the qualification envelope. Further, the ADS qualification envelope includes a duration of about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for temperatures in excess of 330 F, i which is the peak calculated drywell temperature for the limiting break size. Core l

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Generic Letter 89-10 cooling will be accomplished with the injection oflow pressure ECCS systems, _

following the emergency depressurization. Both injection valves for the LPCI and Core Spray systems are located outside the primary containment and are therefore not subjected to the harsh environment inside the drywell.

An evaluation of the small steamline break assuming no sprays with ADS initiation has not been formally done at Plant Hatch. However, a realistic, qualitative, review shows that once ADS is initiated, the driving force for the containment heatup is ,

eliminated prior to exceeding the environmental qualification for qualified components. Although it is possible, that at some point later in the event qualification

- temperatures for some components may be exceeded, the components necessary to .

mitigate the consequences of the event have already performed their functions and are no longer needed. For example, containment isolation valves have closed by the time the ADS is initiated; even in the unlikely event that these valves go open at some later .

time in the scenario, no unacceptable consequences will result since the design basis small/ intermediate steamline break does not result in fuel failures. Consequently, ,

10 CFR 100 limits are not in danger of being exceeded.

In summary, although the drywell and wetwell sprays are assumed in the Plant Hatch environmemal qualification analysis, an assessment of the limiting temperature event  :

! in the drywell, shows that the containment spray valves need not be included in the GL 89-10 scope.  ;

6. Recombiner Valves Since the recombiner valves are normally closed and the recombiners are " closed systems outside containment," it is unlikely that they would be required to reposition ,

to the closed position for purposes of containment isolation during an accident. The recombiner valves, for purposes of GL 89-10, may thus be considered not to have a closing safety function.

The hydrogen recombiner system is provided as a hydrogen control system in the '

Unit 2 licensing basis. From a licensing perspective the recombiners serve a safety

• function and on that basis, the recombiner valves should have an active safety function to open. However, if Unit 2 has a safety-grade vent / purge capability (as does Unit 1) which provides a similar hydrogen control capability using fully-tested  ;

valves, the vent / purge capability provides a satisfactory basis for excluding the i recombiner valves from the more comprehensive capability evaluations. >

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RESPONSE .

Unit 2 has a safety grade vent / purge capability.

. As we mentioned in our submittal of February 1994, the recombiner system was

- implemented as a condition for license for Unit 2. GPC does not believe these valves belong in the GL 89-10 program, because per the Nuclear Safety Operational Assessment (NSOA) report, the system is not required to function for any design basis event and tb NSOA is the governing document for defining design basis events -

for our GL 89-10 program. The design basis LOCA event for Hatch does not result-

' in sufficient buildup of hydrogen and oxygen to reach combustible concentrations, thus the recombiners are not necessary for the mitigation of design basis events, and thus, by criteria nos. 4 and no. 5 (Ref. 2-3-94 GL 89-10 submittal) the system is not required to be considered for GL 89-10 purposes.

Unit 2 does have a vent and purge capability for the primary containment and, like Unit 1, vent and purge is the primary strategy for combustible gas control. Both Unit's systems use air operated valves. Per the Hatch System Evaluation Documents, f the post accident vent and purge function for Unit 2 is not safety related. However, primarily due to the containment isolation function, the valves used for post accident vent and purge on Unit 2, as well as Unit 1, are safety related.

Accordingly, including the recombiner valves in the GL 89-10 program is not warranted.

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7. HP_CI valves During a February 22,1994, meeting, GPC's representatives stated that all the high-priority valves will have their active safety functions intact. For Hatch, there are ten valves listed under this category which includes HPCI, RCIC, RHR and RWCU valves. It is not clear how some of these valves can be removed from the program, regardless of their normal positions, given their importance and that the operators may be required to manipulate them to close, open, or throttle during a design basis event such as a Small Break Loss-of-Coolant Accident. Particular' emphasis is placed on the HPCI injection valve which might be required to automatically open or close in response to reactor vessel water level.

RESPONSE

All of the 14 high priority valves for Hatch are included in the GL 89-10 program.

None of these have been removed from the program, and there are no immediate plans to remove any from the scope of GL 89-10.

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- Re-injection with HPCI is desirable but not necessary since the vessel can be depressurized using ADS and inventory maintained with any low pressure system. .

As was stated in the February 22,1994, meeting, the active safety function for all 14 high priority valves is included in the GL 89-10 program. Further, these valves are tested in accordance with the Technical Specifications and applicable code requirements to provide additional assurance of their reliability for other functions. -

Thus, these valves are considered highly reliable and would be expected to have a  ;

very high probability ofbeiag available for all design basis events.

With respect to the HPCI injection valve, credit is taken for the HPCI system for design basis events involving loss of feedwater and/or a small break LOCA. .The HPCI operation for design basis events is initiated on either low water level (L2) or high drywell pressure. Consistent with the requirements of GL 89-10, the steam admission valve (E41-F001) and injection valve (E41-F006) are required to open to permit initiation of the HPCI. The HPCI is tripped on high water level. Th: high water level trip signal closes the hydraulically operated turbine stop valve and, following closure of the stop valve, the injection valve is closed. If these valves are .  :

closed as a result of the trip signal, the GL 89-10 requirement to open provides additional assurance that the injection valve would again open on demand. The closure requirement for the injection valve is not considered necessary for design basis events because the system can be manually tripped if the turbine stop valve does not close on high water level. Should the HPCI system be manually tripped and fail to restart on low water level, its function can be fulfilled by the RCIC system for events involving loss of feedwater and by the ADS for a small break LOCA. In the unlikely event that the RCIC system is not available for events involving a loss of +

feedwater, adequate core cooling, in accordance with the emergency operating procedures, can be provided by the combination of the automatic depressurization system (ADS) and the low pressure core cooling systems, as discussed previously.

A close active safety function is therefore not needed for the HPCI injection valve.

Toms and Shutdown Coolina Suction Valves In addition to the responses to your questions, GPC is offering this information on the ,

RHR torus and shutdown cooling valves 1/2E11-F004A-D and 1/2El1-F006A-D, which were removed from the scope of GL 89-10 subsequent to Plant Hatch's February 3,1994 submittal. Thejustification for removal of these valves is detailed below:

Formerly, the RHR torus suction valves, El1-F004, and the RHR shutdown cooling -

valves, El1-F006, were defined as having active safety functions to open and close, HL-4794 El-7 L

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Generic Letter 89-10 .

respectively. They are being changed to have no active safety function. The fonner classification'was thought to be necessary to support placing alternate shutdown -

cooling in service, on a loss of the normal shutdown cooling common suction path.'

. . (This was so because it was originally assumed that the shutdown cooling loop would . .

be placed in suppression pool cooling.) Alternate shutdown cooling involves placing Core Spray in service (with suction from the toms), and raising level to the steam lines. Coolant is then being recirculated through the SRVs to the torus and back to I the vessel.- RHR is then placed in suppression pool cooling. To place the loop. l i

previously in shutdown cooling in the suppression pool cooling mode, the F004s-would have to be opened and the F006s would have to be closed.

The loss of shutdown cooling event is described in the Nuclear Safety Operational

< Analysis (NSOA) report. However, criterion no. 2 of our guidelines states that valve :

operation requirements are limited to changing from the normal operating position required to mitigate dedgn basis events. Upon closer review, therefore, criteria no. 2 -;

may be used to remove the El1-F004 and F006 valves from the GL 89-10 active - .

' J safety function list as described below.

In States C and D (as defined in the NSOA), the reactor is pressurized below -

135 psig (Ref. figure G.6-16, Unit 1 FSAR); the Core Spray and LPCI systems are required to be operable. Therefore, the standby RHR loop is aligned for LPCI with the El1-F004 valves already open and the El1-F006 valves already closed.

Assuming a loss of the common shutdown cooling path, and assuming the F004 or F006 valves in the loop originally in shutdown cooling are unable to change position, -

the standby RHR loop could be used for suppression pool cooling'since the F006 and l F004 valves are properly aligned. This could be accomplished by opening the El1-F024 and F028 valves, or by opening the El1-F015 valve, and using LPCI l instead of Core Spray to recirculate water through the SRVs. The NSOA also  !

defmes States A and B (Ref. figure G6-16, Unit 1 FSAR). This is essentially the reactor in the cold condition and vented.- In this situation, all that is required per the NSOA is to maintain core cooling. This can be accomplished either by the availability.

of a low pressure ECCS system, or having the. cavity flooded.

m In evaluating single failure, the two RHR pumps in one loop are off different power supplies. The F004, F006, and the suppression pool injection valves (El1-F024 and El1-F028) are all powered from the same motor control center, R24-S011 for the "A" loop and R24-S012 for the "B" loop. However, if this power supply should fail, or if for some other reason one of the suppression pool cooling valves fails to open, then the LPCI injection valve, El1-F015, can be opened. Since water is being recirculated between the reactor vessel and the suppression pool, it is immaterial whether the return path is to the RPV or the suppression pool. At any rate, cooling is i

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Generic Letter 89-10 being provided by the RHR heat exchangers. Thus, no single failure can prevent ,

ultimate cooling of the suppression pool /RPV. ,

The applicable abnormal operating procedures have been revised to accomm~Iate the .

change.

t The F004 and F006 valves are listed as low priority MOVs in the " Application of Probabilistic Safety Assessment to GL 89-10 Implementation" topical report developed by GE for the BWR Owner's Group.

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