ML17173A304

Forwards Evaluation of Surveillance Requirements for BWR Recirculation Pumps & Discharge Valves in Operation W/Less than All Loops in Svc,Bwr Jet Pump Operating Indications & & Upper Plenum Injection W/Dbes Affectd in LOCA

ML17173A304
Person / Time
Site:	Dresden
Issue date:	10/18/1978
From:	Turbak M COMMONWEALTH EDISON CO.
To:	Desiree Davis Office of Nuclear Reactor Regulation
References
TASK-04-01.A, TASK-4-1.A, TASK-RR NUDOCS 7810240158
Download: ML17173A304 (19)

Text

e Commonwealth Edison One First National Plaza, Chicago, Illinois Address Reply to: Post Office Box 767 Chicago, Illinois 60690 October 18, 1978 Mr. D. K. Davis, Chief Systematic Evaluation Program Branch Division of Operating Reactors - Branch 2 u.s. Nuclear Regulatory Cormnission Washington, DC 20555 fITJg:'11Jr.~~ Fl""""'""'"/

Subject:

Dresden Station Units 1 &'U\\&,C;'.U,JJL] :.,J~L Review of Eight SEP Topics NRC Docket Nos. 50-10/237 Reference (a) :

o. G. Eisenhut letter to c. Reed dated August 17, 1978

Dear Mr. Davis:

Reference (a) requested Commonwealth Edison to make an evaluation of the eight essentially complete topics.

We have completed our examination of the facts upon which the Staff has based its evaluation.

The results of our review were given to Mr. P. O'Connor by telecon on October 11, 1978 and are fo.rmally being docketed by this transmittal. our proposed cormnents/changes are indicated by a marginal line.

Please direct any questions you may have regarding this matter to this office.

One (1) signed original and thirty-nine (39) copies are provided for your use.

attachment Very truly.yours, M. s. Turbak Nuclear Licensing Administrator Boiling Water Reactors l

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i ATTACHMENT. 1 ASSESSMENTS OF.ESSENTIALLY COMPLETE TOPICS TOPIC III - lOC - Surveillance Requirements on BWR Recirculation Pumps and Discharge Valves SEP Plants Affected - Millstone 1, Dresden 2 DBEs Affected - Loss-of-Cqolant Accident Discussion

'This.topic applies to the Low Pressure Coolant Injection System (LPCIS) at Boiling Water Reactors and specifically only to those systems which have undergone the LPCIS modification to remove the LPCIS loop selection logic. This logic network, which is still installed on two of the three applicable SEP Boiling Water Reactors (Millstone Unit No. 1 and Dresden Unit No. 2), is designed to direct LPCIS flow to the intact recirculation loop*in the event of a Loss-of-r.oolant Accident (LOCA).

Oyster Creek has LPCIS.

The logic network also was designed to close the suction and discharge valves of the intact loop to prevent LPCIS flow from bypassing the core and flowing out the.break in the event of. a LOCA.

A modification will be performed on all BWR-3 units (including Dresden Unit No.2*) to allow closure of only the discharge valve.

This is because in the unlikely event of a LOCA occurring between the suction and discharge valves of a recirculation loop with concurrent failure of the loop selection logic, rapid break isolation prior to sufficient reactor depressurization which would allow influx of low pressure, high volume cooling water could result in inc~eased peak clad temperatures.

• Until this modification is installed, Dresden 2 will operate with the suction valves electrically racked out in the open position.

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I I Oq BWR-4 facilities the loop selection logic has been disabled and

• LPCIS flow is now directed*to both recirculation loops, with discharge

. va,ves on both loops directed to ~hut automatically. This topic is directed toward these facilities and concerns.surveillance requirements

• for the discharge valves and recirculation pur11ps bypass valves.

Conclusion This topic does not apply to Phase II SEP facilities.

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.I TOPIC IV-lA - Operation with less than all loops in service

• SEP Plants Affected - PWR's and BWR's DBEs Affected - Loss-of-Coolant Accident Discussion The majority of the present,-y operating BWRs and PWRs are designed to operate with less than full reactor coolant flow.

If a PWR reactor coolant pump or a BWR recirculatio~ *pump becomes inoperative, the flow provided by the remaining loops is sufficient for steady state operation

~ at a power level less than full power.

Plants authorized for long term operation with one reactor coolant pump out of service have submitted, and the staff has approved, the necessary~ECCS, steady state, and transient calculations. The remaining

.. PWR and BWR licensees have Technical Specification~ which require a.

reactor s*hutdm*m within a fairly short time i.f one of the operating

... loops* becomes inopj:!rable (with the exception of two which are discussed*

below)~

SEP APPLI°CAB IL ITY The dockete9 material for the 11 *systematic evaluation program plants has been reviewed wit~ respect" to operation with less than all loops in service. One licensee (Dresden 2) has.requested authorization to operate with less than all loops in.service, the staff is reviewing the analyses submi.ttcd with the request and approval will be granted when the staff

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• approves the an..:i lys is~ Fi vc faci 1i ti cs (Yankee Rm*1c, Mil 1 stone 1, Ginna, Palisadcs,and San Onofre) are not authorized to operate with

• Until.th.a~

1t.im.e, Techn.ical.:pe~ifications permit operation for _,

only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with less than all loops in service.

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• J~~~ than all l~qps in service, Technical Specificati.ons restrict this

• mo.d~ to. a. P.eriod of. Z4 hours, a.t*,~~h:ic;h. time the }aciJ i.-t.Y, mus~ _h~v"e the

. 141.i!.. 109p-.re~tored_ :::.e; s.ervice or.-shµ.t.down.

Three faci1 ities (Connecticut

• Yankee, Oyster Cree..<, and Dresden *1) have had an analysis reviewed and.

approved by the staff which authorizes N-1 loop operation.

Two facilities (ACBWR and Big Rod: Point) have had authorization to.operate in the N-1 l~op mode since th;y were licensed, however there is no supporting ECCS

~nalysis to justify operation.

Conclusion

. This ~opic is co:rplete for all the SEP faci1i.ties with the exception of LACBWR and Big Rock eoint, for the latter two if continued authorization is to be permitt;d an ana.lysis will have to be submitted which describes the thermal-hydraulic conditions of N-1 loop operation during ECCS, steady

.state, and tran~ient condition.s. Until such _an analysis is performed

• • • and approved.

1)peration with less than all loops in service should be restricted to a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period at which time the plant should be shutdowr.

unless the idle loop has been made operable

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TOPIC IV BWR Jet Pump Oper~ting Indications SEP Plants Affected - Millstone 1, Dresden 2 DBEs Affected - Loss-of-Coolant Accident Discussion The capability to reflood the core may be precluded in the event of a lOCA if a.11 jet pumps are no_t_operable. A jet pump i~s-trument sensing

.line :failure could result. iii inaccurate core flow mcasurem~nts or the

• inability to detect a jet pump failure.

This topic applies only to Dre.sden Unit 2 and Millstone Unit _l; th.erefore, ft should be removed from the review list for the* nine remain~ng SEP plants.

The review of BWR Jet Pump operating indications has not begun for the

.two applicable facilities. The* SEP staff cannot proceed any further

• • ~ntil additional infonnation *i*s obtained. from th~ licensee. l I&E and HR~ are working closely to determine. the adequacy of present. jet pump operability technical specificatiOns. If resolution cannot be

• made prior to the start of the Design Basis Events (DB~'s) assessments the.topic will be reviewed considering the potential effects on related DBEs.*

References

• 1oresden 2 has not replied to request for information.

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TOPIC V-9 ~ Reactor Core Isolation Cooling System SEP Plants Affected - None DBEs Affected - None Discussion This topic applies to the RCIC system, a BWR system consisting of a steam-driven turbine/pump c~mbination, piping., vaJves, and controls.

RCIC was designed to inject water into the vessel in the case of

~vessel isolation upon loss of both on-site and off-site A-c*power.

In the General Electric Standard Safety Analysis Report (GESSAR), GE took er.edit for RCIC a.s a backup for the High Pressure Coolant Injection System in Loss-of-Coolant Accident (LOCA) analyses for certain small breaks. The NRC concern is that the RCIC system may not have been

.classified as a safety system, although credit was assumed in the safety analyses.

Conclusion This topic docs not apply to the SEP BWRs (Oyster* Cr.eek, Millsone Unit

.No. 1 1 Dresden Unit Nos. 1 and 2, La Cros~e and Big Rock Point) since TIOne of.these facilities has an RCIC system

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TOPIC VI-7.A.2 - Upper Plenum Injection SEP Plants Affected - Ginna DBEs Affected - Loss-of-Coolant Accident

• Discuss ion

• On May 1, 1978, NRC issued Amendment No. 19 to operating license No. DPR-18

.The staff Safety Evaluation Report which supported the l~ce:ise amendrr.ent addressed the upper plenum injection topic.

Ginna submitted ECCS performance analy~es for the Westinghouse and new ExXon Nuc.lear Company '(ENC) fuels. The Westinghuse analysi~ was perfor;.::d for Cycle 7 fuel which the staff believes is a conservative evaluation.

for the Westinghouse fuel during Cycle 8. The ENC analysis was perforr.:ed

• for Cycle 8 using the ENC WREM-.II ECCS evaluation model.

The ENC

.evaluation model *has been reviewed and approved conditionally by the

• .NRC

• The staff has recently considered whether the Westinghouse ge~eric evaluation adequately represented the flow characteristics of Westinghouse two.loop units. The generic evaluation model assumes that a11 safety injection water is. introduced dire*ctly into the lower plenum.

For the.

two loop units, the safety injection water is. injected into the upper plc:"-:t Thus,*the.staff was concerned that the Westinghosue model did not consicer interaction between UPI water and steam flow.

After plant specific submit ta 1 s by 1 i ccnsees operating two lo.op pl ants \\'~ere rcvi ewed, the s ta f,:

conclude~ that the calculations provided by the licensees (with certain r.rod)fications to the staff's model) are acceptable on an interim basis fc;

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~I*~**'.*~-*.'c~ntinued safe operation o~ ~estinghouse t't!O loop pla;ts,.whi1e long.*

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'te'fm.¢ffor.t; co~tinu~*_for cfevelopin9. ~ rn~del specif~~*any ~tea ting. UP..J.

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the modi.fied version of tne s.ta ff. model,' resul t~d "in. *a cha~ge' of only 1 S°F from those using the generic model in which the UPI-core interaction was not specif~cally considered.

In the interim, before these models. are developed, Ginna has pro"!ided a modification to the current Westinghouse model which accounts for UPI-core interaction. It was demonstrated that the modification resulted in the increase of peak clad *tempcrat~re by l S°F.

Since for the Ginna plant both ENC WREM-II and Westinghouse models predict similar PCT'.s (1922°F for ENC WREM-II and 19S7°F for Westinghouse) it can be expected that t.:1e UPI modification, when applied to the ENC WREM-.II

model, would allow abc,ut the same increase* in PCT.

The licensee has drawn a similar conclusion

• Conclusion

• The staff has conch:ded* that al though the Westinghouse and Exxon _two-lcop t;

generic-evaluation ~odels should be changed to consider upper plenum r. :

injection (unless t~e.plant is modifi~d). ari~ly~es" at the specif"ic t..

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• operating cqnditio~s applicable to the Ginna plant demonstrate.that Y.. *.".i ~ : the. effect o/ disregarding upper* plenum injection interaction on refil 1 i?**:.

• and reflood conditions *will not.be signifitant (less than 20°F PCT).

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Therefore, the staff believes that. for the limited range to which the r::odels c not deviate from the requirements of 10 CFR 50 Appendix K item l.D.3, and the calculations are acceptable.

References....,

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l TOPIC VI Long Tenn Cooling Pressure Failures SEP Plants Affected - All PWRs DBEs Affected - Loss-of-Coolant Accidents Discussion This issue was raised by Mr. Ronald M. Fluegge in an October 24, 1976 letter to then Chairman Rowden.

• It was later defined in the Office of Nuclear Reactor ReguTation as follows:

0The General Desig:i Criteria require th~t the Emergency Core Cooling *Systems {ECCS) shall be capable of providing adequate core cooling folk.wing a Loss of Cool ant Accident, assuming a single failure ir Emergency Core Cooling Systems.

The ~taff assumes the single failure to be either an active failure during the injection prase, or an active or pas~ive failure during the

.long-tenn recirculation phase. The physical layouts of engineered safety feature pumps and components on some pressurized water reactors makes them vulnerable to flooding that might result from large passive failures 1n system P.iping, although they are protected for more likely events, such as sudden seal failure.

Lar~e pipe ruptures are not required to be protected against becaLse of their low probability during the ECCS recirculatio-:t mode."

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.e As stated in the*"NRR Reports on Allegations Made by Mr. Ronald M.

Fluegge" (11/76):

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• the General Design Criteria (Appendix A to 10 CFR 50) include the following footnote regarding single failures:"

'single failures of passive components in electrical

.systems should be assumed in designing against a single failure. The :onditions.undcr which*a single failure of a passive component in a fluid system should be considered in designing the system against a single

• failure are under development.'

• thus, the General Oesi.gn Criteria do not provide an explicit requirement for the treatment of failures of passive components.

A~pendix K to 10 CFR 50 pertains to* ECCS performance requir~ments and also does 1ot provide explicit".g.uidelines on the treatment

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. of failures of passive components af~er a loss-of-coolant

• • accident.(LOC.\\). Present plants ar~ reviewed, however, to assure

• *. "that the plar.t arrang.ement and design features provide the

, *necessary pr.:>tec_tion of essential systems and components (such as shLtdo~n cooling and"prcssurized porti~ns of emer~ency core cool in~ systems) due to* potential piping failures as an

• initiating event (not conturrcnt with or consecutive to a LOCA}.

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. ~iP.i.J'l9.. *fa*ilure~ :<?u~t~.~~~*-~.C?"~~-5.nrn.~~~ *. C!.r.~tpo_st.ula;te? ~~-~

I accordance with Brar~h Technical Pas i tions MEB 3-1 and APCSB 3-1 in tre US~RC Standard Review Plan Section 3.6

• Longitudinal or circumferential breaks in high energy

. fluid system pipin*;i or leakage-cracks in a ~oderate energy*

. fluid system pipir; are considered separa~ely as a single postulated event occurring during normal plant* conditions.*

The crack size assumed for a moderate energy pipe is equi-valent to a slot of dimensions (1/2 x pipe thfckness) x**

(1/2 x diameter). The plant must be designed such that the effects of such a postulated piping failure, including the environmental conditions resulting from the escape of container 0

fl*u*i-~s. do.not: affect function~"or" equipment essential to safe shutd~m of the*. reactor.

• iith regard b postulation of faiiure~. in emergency core*

.cooling systems subsequent to a loss-of.:coolant.accfdent, the USNRC St:indard Review Plan on Emergency Core* Cooling

• System (Sect)on 6.3) provides.additional guidance with the statement t'1a t:* 'The ECCS should retain its capability *to

'*Subsequent to a LOCA. all pipes of relevance arc moderate energy pipes

. defined as a pi?ing system carrying fluid at a temperature below 200°F and at a pressure below 275 psig.

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_.;toor=. the:*core *.in~ the: eve'nt of :a: ~fail ~r~*~ofr any.s~.ngl.e *~~tiv~

.*.~n* paS-sive.. failure *a~-ring: the.long-term recirculation**cooli~g*

phase following an accident.' Based on this guidance, the

• Jtaff assures the ECCS design and layout satisfies the requirement for redundancy in such systems. The imple-mentation of the passive failure statement" does not require significant ruptures of moderate-energy piping subsequent to LOCA~ as this combined e'lent would be extremely unlikely.

The more credible passive failure is at pump or valve seals, or measurement devices*. The staff review of the effects "of such a postulated leak rate in.cludes.consideration of: (l) the flow paths of the radioactive fluid through floor drains, sump pump discharge piping,.and the auxiliary building; (2)

. the operation of the auxiliary systems that would receive this radioactive fluid; (3) the ability.of the leakage detection system to detect the passive failure; and (4) the ability of the operate~ to isolate the ECCS pa~sive failure.

• Therefore, the ECCS passive f~ilure criterion being implemented by the staff requires the consideration of additional leakage but not pipe breaks beyond the ini.tiating LOCA.

The basis for this is the staff's judgment that the probability of I

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.... -1'eed not be~ ~onsidered-a. d~s.ign..ba~.is.~_ev,c11t,1.s.ince;.when operating in the long-term recirculation mode, the ECCS is subjected to temperatures and pressu~es much less than those for which the system is* designed. In addit~*on, after long-term cooling has been initiated, the.need for recirculation diminishes due to the decrease in available core decay heat.

For ~xamp 1 e, for a 3500 M~*lt reactor, the amount of core decay heatwhich is being produced at the beginning of a normal shutdown is 203 MWt; atter.one week it has decreas~d t?

13 MWt; an~ after eight weeks it is only 5.7 MWt.

This means that significantly less coolant redrculation would.be necessary after several wsek~.* -The r.eeded cooling*\\1ater to prevent core overheating can be.provided by the RHR system even ci>nsidering leakage in the suction or discharge side of the piping.

In addition, should recirculation cooling be temporarily interrupted at the end of one week,* the core would be adequately cooled by the. heat transfer* effected by vessel boiloff.

To maintain vessel level, a makeup of only about 100 gpm weuld be necessary."*

CONCLUSYONS

• He consider this* issue to be closed. The effect of ECCS leakage will be assessed on the SEP plants during the DBE cv~luation of LOCAs.

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TOPIC VII-1.B - Trip Uncertainty and Setpoint Analysis Review of Operating Data *sase SEP Plants Affected - All SE? Plants DBEs Affected - Al 1 transi er.ts Discuss ion

• lhis issue was identified in September 1976 by the Electrical, lnstru~entatio and Control System Branch of the Division of Systems Safety, Office of Nuclear Reactor Regulatic-n. The issue. was defiped as. follo\\'ls:

• inclusion is needEd in Technical Specifications of instrument errors in determining instrurr;ent tr.ip setpoin~s in relafi~n to allowable va1uEs of th-measured variable. Operating and under review LWRs are likely to have trip setpoints set at unsafe levels. The mar:gin between trip --~~tP~ints. and "a.1lm*1ab1e v~lues" has not ':Jeen reviewed. Standard Tec~nica1 Specifications for BWRs for instrarce do not even define "allowable values.n Nu:n~rical

• values listed i, the Standard Technical Specifications far* trip setpoints and *ra 1 lcwable values" a*re. identical."

Staff consideratior of instrument errors in the evaluation and approval.

of trip setpoints.-;'or s*afety *related instrumentation has been *perforr.:ed by either of two rr.ethods. Operating licenses issued on plants after the Spring of 1977 co1tain trip setpoints in their technical specifications whose values hav~ been evaluated and approved based upon consideration of the individua1 factors used to assure an adequate margin of safety for each safety related channel. The informJtion upon \\*1hich our evaluations

• arc made.is con:oincd fo the dctailpd Rcgulato.ry.P~sitio.ns o(Regulatory*:.

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.-.:. *tu{d-ff'*f-}'los, Revfsion 11 ~.tnstrument Setpoints:,!'~.re*issue~*.*in~: Novemb,er*:..

... 1976, and in the NRC Staneard ~eview Plan

• Most operating licenses issued prior to this were evaluated in the more generalized manner~ In this approach, the di~crete components of each*

of the margins to safet) in trip setpoint values are not *evaluated

• on an individual basis-Jut are included in an overall safety margin.

• _Each set point value b based upon the most limiting transient or

. postulated accident co:idition associated wi-th the bases for that set

• point. The magnitude of this safety margin and the resulting set points are establ~shed ~o e""lSure that there is a lo\\~ probability of the margin being removed by an adverse combination of instrum~n~ calibration error,

.. ~.instrument error an:f instrument drift. The staff br::lieves that this

.'*method is acceptable.

The staff has,howe,.er, changed from a genera 1 i zed method of trip setpofnt

. evaluation to a mathod that considers each of the discrete factors that

~ake up the marg:ns of safety for. each safety related "instrumentation channel. Either method contains conservatism; however, the newer method allows the safety m~rgin in the trip setpoints to be quantified in a more detailed manncrt In addition, consideration of instrument error is explicit in* the newer method, whereas previously it was an imp1 icit assumption prcs~~~d to be considered as plrt of the overa11 margin.

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• * '* _:~.* 1~'ii1J.. be. i"ncluded fn.FSARs :relati.ng*to:instrument drift and error beca:ise of the guidance now pro1tided int eh NRC's Standard Review Plan and

~n Regulatory Guide 1.1~5. Accordingly, ali ~echnical Specifications

• t~at are iss\\,led with neM op~r.ating licenses after the Spring of 1977 wi-11 have the instrumer.t drift allowance factored into the trip setpoint:

.specificatfons. The rtaff is reviewing this more detailed information on instrument errors and draft to.evaluate its impact, if any, ~pan the safety margins of the trip setpojnts being used in older plants.

Independent of the SS?, appropriate action will be taken to assure that the sctpoitns in use retain an adequate degree of conservatism in maintaining

.sa.fety margins as a result of this staff effort.

• conclusions Adequate safety irargins have been provfded by the trip setpoints now in use for SEP plan-:s, and this Topic does.not warrant additional review apart from that for Topic XV!, Technical Specifications

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TOPIC XVII*- Operational QA Program

.SEP Plants Affected - All DBEs Affected - All

. Discussion Since 1973 new guidance for operational quality assurance programs have*

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been issued in the form of Regulatory Guides and WASH documents descr1bing methods to comply with criteria of 10 CFR 50 Appendix B. The objective*

of this guidance is to assure that operation, _maintenance, modificatior.s and test activities do not d<!grade the capc:b"ility of safety-related.

equipment to perform their intended function

• This "topic has been completed for all SEP plants. Attached is a listing of the dates and specific reports* containing the basis for their acceptance.

Ten of the facilities were :-reviewed by the* Quality Assur~nc2 Sranchi

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• • t~e last (LACBWR) ~as reviewed by the Plant Systems Branch of DOR.

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• Revision 5 Re!erences *

.ATTACHMENT SEP PLAUT Big Rock Point Connecticut Yankee Dresden 1

. Dresden 2 Ginna Lacrosse Millstone 1 Oyster Creek Palisades San Onofre Yankee Rowe.

DOCUMENT 9Topica1 Report Evaluation, 4/21i76 letter, Swit~er to Purple, 2/28/75

• *_Topi ca 1. Rep_ort_ Evaluation_: 8/31/75* r Topical Rep*ort Evaluation, 8/31/75*

Safety Evaluation Report, 9/30/74 Memorandum, Eisenhut to Stello, 2/2/78 Amendment 35 to SAR, 7/16/76 Safety Evaluation.Report, 11/22/76 Topical Report Evaluation, 4/21/76 Safety Evaluation Report, 4/8/75

• Topical Report Evaluation, 4/4/77

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