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SEP 3 1950 00CKET FILE ASB READING Docket ? pek-346 ME!'0RANDUM FOR: Thomas M. Novak. Assistant Director for Operating Reactors. DL FROM:
Paul S. Check, Assistant Director for Plant. Systems, DSI
SUBJECT:
DAVIS-BESSE UNIT 1 AUXILIARY FEEDWATER SYSTEM RELIABILITY EVALUATION In order for us to continue our evaluation of the Davis =Besse Unit 1 Auxiliary Feedwater System we must complete the work outlined in NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident," paragraphs II.E.1.1 and II.K.
II.E.1.1 of NUREG-0660 requires all PWR operating plant licensees to reevaluate tneir auxiliary feedwater systemssto include:
(1) perfaming a~ simplified AF.i system reliability analysis; (2) perfoming a deteministic review of the AFi system usino the acceptance criteria of Standard Review Plan Section 10.d.9 and Branch Technical Position 10-1 as principal guidance; and (3) reevaluation of the AF.i system flow rate design bases and criteria.
The licensee has provided the information requested by the staff regarding the simplified AFW system reliability analysis (Item 1 above). The infomation has been evaluated by the Probabilistic Analysis Staff (PAS).and the recom-mendations resulting from that evaluation which must F e addressed by the licensee are included in Enclosure 1, questions 3 through 9,11,13 and 14.
Questions 1, 2,10 and 12 address ASB concerns regarding the licensee's reliability study.
As we stated in our recent memo to you dated August 5,1980, which forwarded the Crystal River Unit 3 questions, NUREG-0660 is not clear regarding perfor-mance of the deterministic review (item 2 above), licensee or staff; conse-quently ASB will undertake this work. However, for us to cerfom the evaluttion, we will need additional infomation from the licensee. Our mquest for infomation is included in Enclosure 1 as question 15.
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Thomas M. Novak SEP 3 1980 Question 16 of Enclosure 1 concerm the reevaluation of the AFW systen flow rate design bases and criteria (itm 3 above).
II.K.1 of NUREG-0660 indicates that some of the review of the I.E. Bu41etins (79.05, 79.35A and 79.053) issued after the TMI-2 accident were not completed.
The licensee has responded to these bulletins. We understand that your project manager will comp 4ete that portion of the review.
II.K.2 of NUREG-0660 discusses the far, that all operating Babcock and Wilcox plants were ordemd to shutdown shortly after the TMI-2 accident. The Orders included both short-tem and long-term actions. The NRR Bulletins and Orders Task Force reviewed the licensee's responses to the short-tem actions in the Orders and issued safety evaluation reports lif ting the Orders.
II.K.2.8 concerns ccmoletion of our review of the long-tem actions of the Orders.
Completion of this item will be perfomed under II.E.1.l.
Tne licensee may have previously responded to some of these items,and, tnere-fore, should be advised that a reference to a previous response (letter, FSAR, amendment, etc.) would be acceptable. The licdnsee should be further advised tnat clarification will be provided regarding questions 10 and 12 when ICSB provides guidance.
You should be aware that this licensee has previously pmvided infomation to the staff regarding AFd pump endurance testing.
We propose to prepare similar enclosures to be fomarded to the other 3&M operating plant (less Rancho Seco and TMI-1). We are prepared to meet with you to discuss any problems which you see in proceeding in this mar 1r.
The licensee should be requested to provide responses by October 1,1930.
Orignal siFM Otan D. Perr 01st 9.' Bhect.CMefstant Director Aufniarjmifspatnm8ran Division of Systems Integrativ.,
Enclosure:
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ENCLOSU9E 1 Davis-Besse Nuclear Power Station Unit 1 Auxiliary Feedwater Sys*.em Reliability Anelysis Evaluation 1.
Section 1.5 cf the auxiliary feedwater system (AFWS) reliability analysis for Davis-3 esse Unit I defines the AFWS mission success criterien as the attainment of AFW flow from at least one pump to at least one steam genera-t0r.
This definition is incanplete.
It should include tne requirement to deliver AFW ficw to the stea.a gen:ratcr in a timelj manner to precluce steam generator crycut. The success criterien should be revised and resub-mitted 3 ordingly.
2.
We also :ansider that the analysis of maintaining adequate core ecoling without a break in the RCS piping using One makeup pump and the startup feed pumo in addition ta the : ening of PCRV within 30 minutes of loss of main feecwater to be unacceptable in meeting this requirement.
Items 3 througn 17 relate to the recommendations resul ting frcm the staff's evaluation of ne information provided by you regarding the simpli#ied AFW system reliacility analysis. The rec mmendations are categcri ed as generic and additional; as well as short-term and icng-term.
Shor: Term 3.
Technical Scecification Administrative Controls of Manual Valves - Lock and Verify Position (35-2)
The licensee should lock : en single valves or multiple valves in series in the AFW system punp suction piping and lock ccen c:her single valves or mul tiple valves in series that could interru t all AFW ficw. Mcntnly inspections should be performed to verify that these vai /es are locked and in :ne open position. These inspections should be croposed for 'nc:r-oration into the surveillance recuirements of the plant tecnnical s:ecifications.
2-4 Emergency Proceduret for Initiating Sac <-ap dater Supplies (35-4)
Emergency procedures for transferring to al ternate scurce of AFA su;; ply snculd be available to the plant operatcrs.
These procedures should include criteria to inform the operators enen, and in wrat order, the transfer to alternate water sources should take place.
The folicwing cases should be covered by *ne prccedures:
(1) The case in wnicn the primary water supply is not initially availacle.
The ;:rocedures for tnis should incluce any operator actions required to cro ect the AF4 system pumps against self-damage before wa*er *iew is initiated, and (2) The case,in wnich the primary water sucaly is being depleted.
~he procedure for this case should provide for transfer to the alternate water sources price to craining of tne primary water supply.
5.
Emergency Procedures for Initiating AF4 Flow Folicwing a Ccmplete Lcss of Alternating Current ?cwer (35-5)
The as-buil* plant snou.d be ca::able of ::roviding the required AFd 'icw for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from one AFd pumo train indecendent of any AC pcwer scurce.
If manual AFA system initiation or ficw control is required fclicwing a complete loss of AC power, the licensee snould estaclisn emergency crcce-dures for manual initia* ion and cuntrol of tne system as reeded.
(See recommendaticn GL-3 for the icnger-term resoluticn of *,nis ccncern',,
.. 6.
AFW System Flow Path '/erification (GS-6)
The licensee should confirm flow path availability of an AFW system train tnat has been out of service to perform periodic testing or maintenance as folicws:
(1) Procedures should be implemented to require an coerator to determine that the AFW system valves are procerly aligned and a second ccerator *o indecendentij verify *na: tne valves are arccerly alignec.
l' 9 T The licensee should propose Technical Specifications *o assure tnat,
- rior to plant start'.p following an extenced cold snutdcwn, a ficw test would be performed to verify the normal ficw path frcm the primary AFW system water source to the steam generators. The ficw test should be conducted with AFW system valves in their ncrmal alignment.
Additicral Short-erm Recommendations
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7.
Interaction of AFW with Integr ated Control System (ICS) and with Steam and Feedwater Line Break Detecticn and Mitigation Sjstems Because ' the pctentially significant interactions witn the AFWS possibly resulting frcm :ne steam and feedwater line creak de*ection and mitigation systems ($FRCS/FCGG) and the !CS, we believe
- hat information sncuid be provided to the opera *ing crews on means *o detect and coce witn AFWS interruptions caused by failures in these systems. Such information may be in *.he form of
- raining and/cr procedures. We note :nat training witn respect to interructions caused by ICS faults may already oe enccmpassed by recuirerents resulting from Cconee event of November 10, 1979 and the Crystal River event of Feoruary 26, 1930. Longer term recommenca*icns rela *ing to this same concern are oiscussed belcw.
4 3.
Hunan Error Curing Tes and Maintenance The licensee snculd assure that plant procedures are wt itten to reduce hunan incuced commen mcde failures of all AFW system trains. For example, the licensee shculd stagger testing of AFW systen trains, i.e., for planned test-ing, no more than one AFW train (or pump) shculd be tested by tne same shift crew.
3.
Flow Blockage by Plugged Strainers The licensee snculc assure that there are no temcorary strainers in : lace in the AFW piping system that may cause fl w bicckage if plugged.
Ccerating ex;erience at several plants has shn', this to be a potential common cause failure mechanism which could fail the entire AFWS. The sucticn strainers between the condensate storage tank and 're pumos are an example.
10.
Indication of AFW Flow to tne 3 team Generators (AS-3)
The licensee snculd implement the folicwing requirements as specified by Item 2.1.7b on page A-32 of NUREG-0573:
(1)
Safety-grade indication of AFW fl:w to each steam generator snculd be orovided in the control rocm.
(2)
The AFW flow instrument channels shculd be pcwered fr:m tne emer;en:y buses consistent with satisfying tne emergency pcwer diversity requirements f:r tne AFW system set forth in Auxiliary Systems Brancn Tecnnical Positien 10-1 of the Standard Review Plan, Sec.icn 10.a.9.
__ Long Term 11.
Elimination of AFW System Dependency on Alternating Current Power Following a Ccmolete Loss of Alternating Current Fcwer (GL-3)
At least one AFW system pump and its associated ficw pain and essential instrumentation should automatically initiate AFW sys em flow and be ca:331e of being operated indecendently of any AC pcwer sourca for at least 2 hcurs.
Conversion of CC power to AC :cwer is accep:3bie.
12.
Ncn-Safety Grade, Non-Recundan: AFW System Automatic :nitiation Signals (GL-5)
The licensee should ucgrace the AFW system automatic initiation signals and circuits t:.neet safety-grade requiremen ts.
additicral Long-Ter, Reccanendations Interaction of AFW with Integrated Control System (ICS) and witn Steam and Feedwater Line Break Detection and Mitigaticn Systems.
13.
The licensee snculd secarate the ICS from AFW initiation and contrci, and reduce the interaction of tne AFWS with Steam and Feecwater Line 3reak Cetection and Mitigation Systems. The potential f:r ccmmen cause failure of ne AFWS due to interactions with these two systems is discussed in NUREG-CS67 (Reference 11).
Speci#ically, reccmmendations 2 and 4 in Table 2.1 of Reference 11 call for (a) the separation of the AFAS ini-iation and c:n:rol frca the ICS, and (b) the reduction in adverse interac-icns of the steam and feedwater line break detection and mitigation systems with ne AFWS.
The license snculd implement those reccamer.dations.
-- Plant-Scecific Recommendations 14.
Diversity in the Motive Pcwer for the AFWS Pumos We are concerned with the dependency of both AFWS pumps in your design on steam from the main steam lines.
Other PWRs are known to have a similar configuration (e.g., Calvert Cliffs); however, because of the more rapid dry-out of the steam system in B&W plants, such a steam dependency is of more concern in Davis-Besse.
State your plans for installing a third AFWS train which will utilize a pump pcwered from a source other than steam.
A schedule of implementation should be provided.
15.
Postulated High Energy Line Break Your design does not appear to meet the high energy line break criterion in SRP 10.4/9 and BTP 10-1; i.e., the AFW system should maintain the capability to supply the required flow to the steam generator (s) assuming a pipe break any where in the AFW pump discharge line current with a sirigle active failure. Evaluate your design in line with these requirements.
1 16. Design Basis for AF',! Systen Flow Requirements The licensee is required to provide the AFWS flow design baiss information required in Enclosure 2 for the Davis-Besse 1 design basis transients and accident conditions.
The response should include tie follcwing:
(1) List all events needing AFW to mitigate the consequences.
(2) Justification that the bounding non-LOCA calculation will serve as a conservative basis i,r sizing the AFW system for non-LOCA core cooling considerations.
In oiner words, show that the calculation will bound all of the non-LOCA evtits requiring AFW, (3) The non-LOCA analysis shc uld include a loss of feedwater event using FSAR type assumptions to maximize heat removal requirements (1.2 At4S decay heat, 2". power level measurement uncertainty, RCP heat input).
The calculation should not take credit for " anticipatory reactor trip" since it will not occur under all conditions.
Lifting of the PORV is not precluded; however, credit for pressure relief through the valve should not be cssumed.
(4) For a small LCCA events, reference may be made to the B&W Report,
" Evaluation of Transient Behavior and Small Reactor Ccolant System 3reaks in the 177 Fuel Assembly Plants dated May 7,1979.
The acceptance criteria for the event will be:
a.
Reactor Coolant Systen pressure remains less than 110% of design pressure s?750 psig).
b.
?io fuel failure (Df1BR greater than 1.30).
1
Basis for Auxilia y Fe%nster Syste Flet: Re:wi re ents As a result of recent suff revie:S ed c eratin; plant Auxilia y Fee:-
water Syste s (ATd3), the staff c:n:ludes that tne design bases and criteHa provided by licensees for esublishine AF45 retuireents fer flew to the sten generat:r(s) to assure ade:uate reeval of rea: :P de:aj neat art not weil defined er d::'sented.
We escuire that you provide the following AF45 ficw design basis infer-
- atien as a:plicable to the design basis t ansients and a :ident =n-ditiens for ycur plant.
Idan*f fy the plant transient and at:1 dent ::nditions =nside-ed 1.
a.
in esublishing AF'4 flew require ents, including the fo11 ewing events:
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- 1) Less of Main Feed (LMF4)
- 2) L*Fi w/lcss of cffsita AC pcwer
- 3) Lw?4 w/lcss of ensita and offsite A; pcwer
- 4) Plant c:cli:wn
- 5) Turbine trip with and without bypass
- 6) Main stem isciation valve closure
- 7) Main feed line break
- 8) Main stan line break
- 9) Sr.all break LOCA
- 10) Other transient er a::ident =ndittens net listed abcve b.
Describe tte piant pr:ta:ti:n a::ectance :-itaMa and ::r es-pending ta:nnical bases used for ea:n initiating event identi-fied above. The acceptance cH taria shculd address plant limits such as:
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Maxi =u.- R 3 pi?ssu-e (PORY cr safety valve actuatien)
Fuel te::e stun Or tamage Jimits ( W, PC, maxi.r-fuei central te perature)
R;S cooling rata limit t: avoid excessive c:clant shrinkage Mini =um stac generat:r level t: assure su'ficient stee generst:r heat innsfer sur' ace t: remove decay heat and/:r ceci dcwn the prf=ary systa=.
2.
Des: ribe the nRalyses and assu=:ti:ns and c:r esponding technical justification used with plant c:ndition c:nside ed in 1.2. abeve including:
a.
Maxt::x:s rea:t:r pcwer (including inst: :.ent er cr allcwanca) at the ti=e :# the initiating tnnsient cr. accident.
b.
Ti=e delay fr::m initiating event : react:r t:ip.
c.
Plant parretar(s) whi:5 initiatas AFu flew and ti=e deiay betwen initiating event and intnducti:n Of A~G 'l:w inte stem generstcr(s).
d.
Minimu:n staa gene-st:r water level when initiating ever.:
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e.
Initial staam generser water invent:ry an2' depletion rate bef:re and after ATn ficw c:cmencas - identi'y reacte decay liaat rata used.
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Maximum pressure at whi:5 stac is released fr= ster gene-st:-(s) and against vnich the ATd pt.p must develop sufficient head.
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Mini:=.:m nu:.cer of stem genent:rs that =ust rt:sive AF4 C:w; e.g.1 out of 27, 2 cut of 47 h.
R flew condition - c:ntinued,c;eration =f RC pu :s :r natuni circulation.
1.
Maxt:u:n A7i inlet tar::eraturs.
- j. Following a postulated sten or feed line break, time delay assu:eed t: is:Tata break anc direct AFi flew to ints:: stam generater(s). AFa* pt.: p flow capacity allowance t: ac::medata the ti::e delay and maintain mini::::3 stan generat:r watar level.
Also identify edit taken fcr pri=ary systan: heat receval due to blewdewn, t.
Volu:se and mxt::x.:n tm:>erature =f water in =ain feed lines between s. tan gene-at:r(s) and AFiS c:nnection to main feed line.
1.
Coerating :::natien of stem generat:r normal blewd:wn fo11 ewing inttf ating event.
Pri=ary and sec::ndary systs water and utal sensible heat m.
used for =c1dcwn and A7d f1cw si:ing.
n.
Time at het standby and time to cocidewn R03 to R.9 systs et-in tem:e*ature to si:e A74 watar seu :e inver. ::ry.
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Verify that tne AT4 pc s in your ; Tant will succly the ne: essa y ficw t: the stans gener.t:r(s) as,detsmined by ite s 1 and 2 above c:nsidering a single failun. Identify the urgin in si:in; the cu: : flew t= alle for pu: : recirculatten flow, seal leakage and pu::P wear.
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