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MEMORANDUM FOR:
Thomas M. Novak, Assistant Director for
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Division of Licensing FROM:
D. Garner, Project Manager Operating Reactors Branch #4, DL 3
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Robert Reid, Chief g r. i:, <_.
1 Operating Reactors Branch #4, DL
SUBJECT:
TECO's INTENTIONS WITH RESPECT TO THE ADDITION OF A DIVERSE POWERED AUXILIARY FEEDWATER PUMP At your request I spoke to TEco concerning the above subject. The status is as follows.
TECo has received from a contractor a " Scoping Study" which put together the various aTternatives that could be pursued in an attempt to satisfy our concerns in this area. Although we should wait to see the results of the study and a firm proposal by TEco, I think that we should dispal any current thoughts suggesting that they are reluctant to put in a third AFW pump.
Although Fred Miller of TEco has previously stated to us (during the NUREG-0657 report generation) that they do not intend to put in a safety grade pump, it did not take them long to recognize that a non-safety grade pump would seriously degrade their existing system. Therefore, the scoping i
study was perfomed with the direction that all new components would be of the same quality as the existing system. The study has concluded that a direct drive diesel unit be installed at the site which, due to physical layout constraints, would be located in a new dedicated building with its own dedicated water source. The new pump would have its own control and initiation system so that the existing system would not be affected by its addition. They would end uo with a two pump plus one pumo system as dis-tinguished from a three pumo system. That is, they will not integrate this system entirely with tne existing one since that would cause a reanalysis and redesign of an already complex safety grade control system.
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Thomas M. Novak <e? 11 9 0 Three things have held up the study from being completely analyzed by TEco:
1.
The system is estimated to cost around $6 million, and the budget oeople at TECo have challenced__the ancinaariac crouo for some cirosmvceme equfr
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The study endorses a diesel driven unit, and some at TEco would want to propose that the existing 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO in the Davis-Besse Tech Specs not be applied to the unit since its existence does not appear to carry the same importance as the other two pumps. The TECo Licensing Manager (Ted Meyers) responded by saying that NRC would probably reject this request. Since diesels are inherently less reliable than electric or steam driven sources, they may choose not to go this way.
3.
A special comittee, including the Vice-President, Nuclear, and the Station Superintendent, has been established to review the study due to its apparent importance in fulfilling a requirement suggested by the NRC. Since these people are trying desparately to get their plant back on the line after a 5 month outage, they have not had the oppor-tunity to seriously consider the study.
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Ted Meyers has asked for a few weeks more to give us something. This time will'(hopefully) allow them to finish their outage so that they can seriously ccnsider the scoping study and prepare a letter to us stating their comit-rnents. Since the system will be entirely safety grade, Ted has stated that their proposed schedule will be well beyond 1982 for completion. The schedule may also be impacted by our requirements in regard to environmental qualification.
I recommend that we give them the time that they ask and wait until they submit their proposal to pass any judgment on them. There can be no risk involved in the mean time since the plant is not operating.
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6 Daniel J. Garnee Project Manager Coerating Reac/or,s Branch 14 Division of Licensing cc: D. Eisenhut R. Purple D. Ross P. Check R. Tedesco G. Lainas
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Docket No. 50-346 License No. NPF-3 "U[
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3M Serial No. 177 January 23, 1981 o:-a:: :c.a a u.m Director of Nuclear Reactor Regulation Attention:
Mr. Robert W. Reid, Chief Operating Reac:ces 3 ranch No. 4 Division of Licensing i
United Scaces Nuclear Regulatory Commission Washing:en, D.C.
20555
Dear Mr. Reid:
On July 6,1979, the NRC issued a. letter lif ting the Confirmatory Ord' er dated May 16, 1979, allowing the Davis-3 esse Nuclear Power Station Uni: 1 (DB-L) to re: urn to power following the Three Mile Island Unit 2 accident. The safecy evalua: ion at: ached to that letter indicated tha: :he NRC would at some future :ise require :he installation of an additional 100 per cent capacity auxiliary feedwa:er pump 4: Davis-3 esse Unit 1.
Since July of 1979, Toledo Edison has been evalua:ing some options to meet our unders:anding of the NRC requirements of such a sys:em = edification. A review of past actions is appropria:e :o fully understand Toledo Edison's current activi:y.
Davis-Sesse Unit 1 was put in:o operation with a fully safety grade auxiliary feedwater (AFW) system. The Final Safety Analysis Repor: (FSAR) reflected this design. A: : hat ti=e, :he system was unique among its Babcock & Wilcox (3&W) nuclear s:aam supply sys:em (NSSS) counterparts due to the full extension l.
of safe:y grade :ri:eria and requirements to include not only :he mechanical systens but the ins:rumen:2: ion and control systems as well. Davis-Besse Uni: I has installed in its original design an AFW initiation system (the S:eam and Feedwater Rupture Control Sys:e=) and a s:eam generator level control sys:em, be:h comple:ely independent of :he 3&W supplied integrated con:rol system (!CS).
The basic cri:eria of :hese systems was to isola:e the steam generators, pro-vide auxiliary feedwa:er and con:rol level wi:hin 10 seconds of initiation.
'he 35-1 Technical Specif t:a: ions Sec: ion 3/L-3.2 reflects :hese requiremen:s. When pu: in operation :he 33-1 AFW system co= plied in all respec:s :o :he NRC safety grade require =ents.
In the NRC's Safety Evalua: ion Repor:, an addi:ional require =ent tha: one :: sin of AFU vould have to be independen: from AC power requiremen:s was imposed.
This later require =ent was identified as a license condition for nodification that has been met by sys:e= :odifications during the recen: refueling ou: age.
These requirements were met using :wo saf ety grade stea= driven pu=ps.
Following :he 2:cident a: hree Mile Island on March 25, I?79, many ques tions were direc:ed :oward AFV sys:e=s at 35W NSSS units. NRC concerns abou: 35-1ool 5 -s '.s:: I:,s:n ::. r;.w 5:s:. r.:::
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? 4 Dockee No. 50-346 License No. NPF-3 Serial No. 677 January 22, 1981 AFW syste= seemed to concentrate on: 1. Overall system reliability. 2. Potential loss of AFW pump motive power. 3. Steas generator cooling requirements to reduce reactor coolant system pressure af ter s=all break loss of coolant accidents and after complete loss of feedwater transients. In responses :o the staff, each of these items were discussed. Toledo Edison filed submit:als on I:em 1 above on May 23 and July 3, 1979, as well as Topical Report 3AW 1584, " Auxiliary Feedwater System Reliability Analyses - 'A Generic Report for Plants With Sabcock & Wilcox Reactors." Additionally, as repor:ed since these subsi::als, Toledo Edison completed its license con-dicion, in response to Item 2 above, by diversifying power supplies :o the auxiliary feedwater motor-opera:ed valves. This now insures a train operable wi:hout alternating elec:rical current available. On :he poten:ial loss of =ocive power, the s taf f utilized a non-cechanistic, cocplete and instantaneous loss of s:aas pressure in bo:h steam genera: ors (SC) as i:s basis of concern. This ignored completely the safety grade systems tha: provide SG isolation and AFW ini:ia: ion. Toledo Edison's sub=i::al on June 23, 1979, illus::sted by calculation and experience the ability of the design to respond appropria:ely to reduced pressure auxiliary feedwater pump turbine s:ar:s. To further :ompound these assu=ed multiple safety sys tem f ailures, the NRC staf f would not consider any capability to re-establish stea= pressure as motive power to the AFW pu=ps. On May 22, 1979, Toledo Edison's submit:al provided B&W's " Evaluation of Transient Behavior and Small Reactor Coolant Syste= 3:eaks in the 177 Fuel Asse=bly Plant - Volume III - Raised Loop Plant." This illustrated :he addi:ional ti=e delay available for such ac: ion (greater than 30 minu:es) :ha: the 33-1 raised-loop reactor coolant sys:em design would be able to sustain during a c =plete loss of all feedwater, regardless of :he sour:e. Sy ignoring the uni:'s design capabilities, the assu=ptions i= posed by the NRC staf f prede:er=ined : heir conclusion tha: an addi:ional 100 per can: diverse power capabili:7 need be provided. The third YRC concern related :o :he reliance on the safe:y grade AEW sys:em :o depressuri:e :he reactor coolant system to within the ef fective pressure range of :he DS-1 high pressure injection pu=ps. As indica:ed above, the analysis identified :ha: grea:er :han 30 minu:es was required without feedwa:er from any source to provide a po:encial problem. sesynasummuy-mug b * ' ' 7 ts* M.isa,g # ,g b:. r~ rip' %g m,.,a __..,___--w----- fN -te G a. eframe. a ct::ena uf, the same suomi::al provided descrip-
- ions :o depressuri:e the reactor coolant'sys:em even with such a delay.
These scenarios have 'seen factored in:o our operatin; peccedures.
4 Docket No. 50-346 License No. NPF-3 Serial No. 677 January 23, 1981 Your July 6, 1979, letter indicated, regardless of information submitted, an intended requirement was to add an additional 100 per cent capacity AFW pump. Starting from this NRC evaluation, Toledo Edison has proceeded on a feasi-bility study evaluating design options of such a backfit to DB-1. The study was undertaken with criteria preceived by us as adequately addressing the NRC staff's basic concerns. The results are in each case extremely costly and require long lead times, reflecting major plant additions and/or modifications as well as site alterations. In addition, there are AEW systa= operational philosophy differences compared to the present system. In re-viewing the NRC intended purpose for such a modification, and relating it to the magnitude of the physical change required, Toledo Edison has decided to undertake an attempt to quantify the relative risk reduction actually pro-vided by such a modification. To bring this issue to final resolution, it is proposed that, prior to pro-ceeding on any =ajor plant modification, a risk reduction comparison be ecmpleted to provide an evaluation of the acceptable alternatives. This would allow us to optimi:e the plant response results, minimi:e the plant perturbations and still verify that the design provides an appropriate level of,arotection to the public health and safety now and after any such =edifi-cation is complete. A meeting is proposed to identify the performance criteria ar.d risk reduction results behind the proposed modification on the July 6, l979, >RC letter. We expect this information to be available for discussion in February, 1931. Very truly yours, f'/.id',__-- RFC:TJM:aa ec: D3-1 NRC Resident Inspector i 4
..- - o u ~ T ~. W s L 2, f .I GM h'.Ah 0 6 1981 MEM3RANDUM FOR: S. Hanauer, Director, Division of Human Factors Safety D. Ross, Director, Division of Systems Integration R. Vollmer, Director Division of Engineering T. Murley, Director, Division of Safety Technology FROM: Darrell G. Eisenhut, Director, Division of Licensing
SUBJECT:
HUREG 9667 IMPLEMENTATION PLAN contains the NRR plan for inplementation of NUEEG-0557 (Transient Response of Babcock & Wilcox-designed Reactors). This plan was based on DST's recomendations regarding both priorities and extension of applica-bility to non-B&N plants. The approach' generally used was straightforward: Priority 1 efforts identified in the DST evaluation have been given specific implementation actions; Priority 2 and 3 efforts are being deferred because of resource limitations. The only exception to this prioritization scheme is that a few priority' 2 itens CR-3 study) are curren(viz., AFW upgrade. ICS/NNI improvements and IREP tly being implemented through related actions. The development and resolution of the high priority efforts has largely been accomplished by the TMI Action Plan or through other requirements. The attached Table I identifies the high priority items and the referenced requirements document. To ensure that these priority actions are accomplished certain~ activities must be taken, including: (a) verification of Lead Branch responsibility and (b) confirmation that the NUREG-0667/ DST reco::rnendations are within the scope of the referenced requirements document (e.g., NUREG-0737. IE3 79-27). You are requested to review both Table I and Enclosure 1. and'ferward vcur concurrence or coments on these imolementacion acticrs by March 16, 1951. or :.; n:- W" Darrell G.' m G. I,15472... Eisenhut Director Division of Licensing
Enclosure:
As stated cc w/ enclosure: Seeeax t-+ age - WE A i i
TABLEI.-PRIORITYACTIO$5liEMS P r.quirements document and/or through separate programs.ll.csc actiuns items requi e erenced Verification is required by March 16, 1981. 1IfM RECOMMENDATION LEAD BRANCll 1.a REQT. DOCUMENT AfW Seismic Qualification SEPB
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AfW Diverse Power Supplies Seismic Qualification Prog: 1.d ASB AfW Safety Grade CP's NUhEG-0737, II.E.1 ASR 2.1 ATW Auto Start p SRPs -3 3. ICS!! Dil-1 Diverse Drive AFW NUREG-0737 II.E.1.2 5.c ORD#4 Orders Power Separation / Signal Path ICSB S.c Multipie Instrument Failure IED 79-27 Review 5.e 10511 ICB 79-27 Review NNI Power flux Redundancy S.h ICSil IED 79-27 Review followup on Licensee Response ICSB 100 79-27 Review $ 6. Safety Grade Vital Instrument Panel 7. llFEB Improve In-Core Thermocouple Inc. NUREG-0737, I.D.2 fl. CPB Safety Grade Vent / Purge Isolation NUREG-0737, II.F.2 13. CSil Training on CR-3 Event Issue R.G. 1.141 OLB Confirmatory Orders to B&W }t - 14 g plants EPs for loss of NNI/ICS ICSB 16. OLB IEB 79-27 Review 15. Mandatory Simulator Training RCP Restart Criteria. NUREG-0737 I.A.3 IIFEB 17.c,d.e PORV/ Safety System Challenge NUREG-0737, I.C.1 RSn NUREG-0737. II.D.1, II.K.2.1( II.K.3.12-18.a IREP CR-3 Study PAS Comolete Analysis, include ic NUREG-0737, II.C.1 70. RCP Trip During 50 LOCA CPD/RSB NUREG-0737, II.K.3.5 .i
NUREG 0667 IMPLEMENTATf0*l PLAN BAEGROUNDANDSCOPE Following the February 26, 1980 Crystal River 3 event, a special Task Force'(i.e., B&W Reactor Transient Response Task Force) was estabifshed to pre ide an assessment of the apparent sensitivity of B&W-designed plants to transients. A final report was issued as NUREG-C667 which contained 22 recommendations of actions in order to (1) reduce the sensitivity of S&W reactors to transients involving undercooling and overcooling events and small break LOCA, and (2) to reduce the likelihood ef reactor transients being initiated by the integrated ~ control systems, non-nuclear instr. mentation and associated power A supplies. Based on the Task Force recommendations and subse:;uent reviews (Memo- ,,randum to H. Denton fron R. Mattson, dated Au;ust 8,1980), the Division - of Licensing has developed this preliminary plan to integrate the 22 recer.:endations into a program for upgrading transient response. The priorities developed and the suggestions for extending certain recommend-actions to other than B&W reactors are being incorporated. Of the original 22 recommendations, there are forty-two (42) individual actions proposed. Of these 42, twenty-one (21) items are currently considered being im::le-cer.ted (througn ';URE3 0737, IEB 79-27, confier.atory orders, anc separate programs). Seventeen (17) items, because of lower priority, will require
-2 ( further evaluation before implementation, and four (4) items are not recommended to be implemented. OEJECTIVES ~ The purpose of this pl'an is to assure that each of the NUREG 0667 recom-tendations have been appropriately considered and acted upon. This plan establishes a method for implementing requirements and/or integrating special programs which will lead to eventual implementation. Additionally, the plan addresses the necessary staff actions in tracking, verificatien-and closing lygt each of the recommended actions. CISCUSSION The implementation status of the individual recommen.dations in NUREG 0557 include items being fully implemented through the existing TMI action plan and IE Bulletins, items being implemented through special orders or programs (e.g., DB-1 orders, evaluation of seismic AFW qualifications), and those items of lower priority which require additional studies and evaluation. Table I contains DL's Implementation plan for the 22 ~ reco=mendations and the required actions. These recommendations have been further itemized into forty-two (42) individual actions. Eecause of manpower constraints and priorities, DL has further classified the S.7.EG 0567 recommendations into near-term and long-term actions. The near-term (Category A) actions are of higher priority and require verifi-cation and assurance of implementation through the referenced recuire-ments documents or through separate ;rograms. It is DL's understanding
s B 3 that the twenty-one (21), Category A items are being implemented,or proposed for implementation,through existing programs. If these f,tems are not currently being implemented the lead technical branch shall identify these exceptions and establish a mechanism for timely implementation. The long-term actions (Category B items) shown in Table I shall not be implemented until resources and scheduling can be established. Eventually, these items will be implemented through special programs which will require more detailed milestones, designated lead individuals and schedules leading to eventual i'mplementation. The' Category C items are not recormended' for implementation RESPONSIBILITIES Within NRR, the_ Operating Reactors Assessment Branch (ORAS) has overall responsibility for assuring implement 5 tion of the NUREG 0667 recommendations. This includes verification of the implementation status of the near-term (Category A) actions and eventual implementation of the Category 8 actions. A lead project manager will be assigned to coordinate efforts with the technical divisions and OI&E to assure the internal reviews and acceptance criteria are established as well as the necessary verification and closecuts are peEformed. Additionally, the development of new requirement criteria, pcsitions/ clarification and implementation schedules will be completed by the technical divisions as requir'ed. Specific evaluations and reviews of licer.sei submittals will als5'be performed by the technical division and, er through contractor assistance. The following highlights some of the divisional assignnents/ responsibilities. 9Y s e=
s 4-The technical divisions / branches indicated on Table I shall verify their . lead branch ' responsibilities and corroborate the implementation status. Additionally, resources and schedules should be established if not in existence for the higher priority (Categcry A) items. As necessary, new or atended positions / clarification and implementation schedules, in tne format of NUREG 0737, shall be developed by the technical divisions. The technical divisions will also have the responsibility to generate SERs and provide technical input to the Standard Technical, Specifications for action items which may require model technical specificatic's. Contract n technical monitors should also being provided for contractor support reviews if not already in existence. 1 The Division of Licensing will have overall management responsibility to assure implementation of the recomended requirements. This will include co:rdination of implementation senedules with the responsible Divisien/ Offices and verification /close out of the required actions. For the loncer-tem actions (Category B) DL will have responsibility to assure their implementation once resources and scheduling can be established. Responsibilities for these items will inclu'de coordinatkon w^ith the technical division in the development of nevi positions, cla'rification, rec,uired documentatien and i..piementation schecules. It is anticjpated that these new requirements will be implemented through NUREG 0737, Item
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T-DST Itca F Reconinenda tion Receninended Lead Action . ' Appl i_cah_i_11_ty_
- Priori ty,
_ Imp 1ementa t ion Branch Rbqu,i ra S Nine specific reconsnendations e to improve ICS/NNI a) Power buses separa tion A 2 signal path channelization ALL b) Evaluate mid-scale instrument IED 79-27 Review ICSB A failure mode B&W c) Mul tiple instrument failure Initiate program indication ,8 B ALL d) Reversion to manual control ALL IED 79-27 Review Initiate Program A e) NNI power bus redundancy ALL f) follow-up on BAW 1564 IEB 79-27 Review B UAW
- 9) NSAC-3/INP0-1 Reconsnendation B&W Ikisting Staf f Action A
h) follow-up on IE Bulletin 79-27 Initiate Study B Responses a ors NT0L i) ACRS Reconinendations lEB 79-27 Review ALL A Initiate Study B 6 Installation of safety-grade panel of vital instruments ALL 1 flVREG 0737 - I.D.2 lifEB A 7 Improve use and display of in-core thermocouple indication ALL 2 NUREG 0737 - II.F.2 CPB A 8 Safety-grade vent / purge isolation ALL 1 Reg Guide 1.141; issue-iupplement to NURLG U/3/, CSD A 9 !!.E.S.2 System response to keep L scale & P above llPI setpoint p p B&W 2' initiate study RSB B 10 Sensitivity studies of operational modifications B&W 2 initiate study RSB B 11 Elimination of post-reactor trip operator actions B&W 2 initiate study RSB B 12 Qualified l&C technician on duty B&W ft/A Not reconinended for N/A C implementation .1
DST Reconenended I tcu. # Recommendation Applicability _ Priori ty, Implementa tion Lead Action 13 Operator training on CR-3 event Branch Requir B&W 1 Confirmatory Orders i 14 OLB A Emergency procedures for icis of NNI/ICS Covered weth item (2) ALL 1 in t he IER 79-27 review ICSB A 15 Mandatory simulator training for requalification incibded in NUREG 0737 ALL 1 item I.A.3 OLB A 16 Evaluation of RCP restart criteria PWRs 1 i# nclude evaluations in ltem I.C.1 of NUREG 0737 17 IIFS A Alternate solution to PORY unreliability/ safety system challenge rate a) Provide safety grade PORY 2 RSB and position indication Not reconsnended for 04W b) Provide dual safety grade implementation C block valves B&W c) Complete operability test Not to be implemented C program ALL d) Ins tall safety grade reactor Included in II.D.1 included in items A trip on total loss of feedwater B&W/W e) Reset PORV and high pressure II.K.2.10 and ll.K.3.12 A trip implenented through Cat A PWR lessons Learned NUREG 0578 g 18 Complete IRCP CR-3 Study A a) Analysis and Report CR-3 2 b) Plant modifications at CR-3 CR-3 Review 11/6/80 CR-3 response PAS A c) Initiate further IREP NRC Action Other plants NRC Action PAS /0RB B PAS B 19 Development of Performance Criteria for anticipated transients ALL 2 Reg Guide to be i issued RSB .B 20 Continued evaluation of need to i trip RCP's durin9 58 LOCA included in NUREG 0737 - PWRs 1 li.K.3.5 CPB/RSB A 21 Re-evaluate location of AfW injection onto OTSG DAW 23 Initiate study RSB 0 \\
. T Lead Actios Item # Reconsuenda ty.,on Appl icah,i,li ty, ,Pr,lorf.ty, imp _lementa tion firanch Required 22. Study of higher frequency of operator errors in HAW plants llAW N/A Not reconenended N/A C o! Legend Priority 1 - Items should be scheduled and implementation begun as soon as possible. These items may require j rescheduling as NRC staf f and licensee / industry priorities and resources. i Priority ? - Items should be scheduled and implemented.in accordance with existing priorities and resources. I: Actions kcquired A) Verify that the requirements /reconsnendations are being implemented through the referenced requirements l docu.xnt; requires concurrence by February 27, 1981. l P. ) Requires development of special programs / studies prior to issuance of any new requirements. A program plan should be established including workscope, milestones, lead individuals and schedules leading to new positions / requirements and eventual implementation. These items are deferred until resources and schedules can be es tat,Ii shed. s C) No action required. l h l
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April 2, 1981 ,,fe/ 0000* +.- ~. N,s. Occhet No. 50-346 A ~ ', %r,dT*'cer Y' a Mr. Richard P. Crouse Vice President Nuclear Toledo Edison Company 141. r1:n ' i.jgy }CC M. !! ten Av enue,
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Dear Mr Crouse:
We have reviewed the Davis-Besse Auxiliary Feedwater System (A WS) in accordance with NUREG-0737, " Clarification of TM! Action Plan Raouf re-en t s ". I t e, ! !. E.1.1. Our review was based upon your January 4, 1980, s#tttal which enclosed the ANS Reliability Analysis for Davis-Besse: your resoonses to the May 16.1979 Order and the Of fice of Insce: tion and Enforcement Bulletins 79-CSA and 053; and your responses to ac-olicatie sections of NUREG-c572, "T'41-2 Lessons Learned Task rocce: Sta tus Re:oet and Short icem Recomnencations." The results of our review P. ave indicated six additional actions which would provide in-creased reltatility in the Davis-Besse A GS. These actions are identi-fled in the enclosures. s ~ With respect to the sisth iten identifled in Enclosure 1 (N (ntPLORW, we met with your staff in our offices on March 5,1901 to the Davis-Besse AAS could be improved. Your sta f f procosed ciscuss now to reexamine Davis-Besse's AWS and to develop means, including modiff-cations to the systen as necessary, to improve the system relisbility. Your pro:csed criteria for the third AWS train required a safety-grade design that, from your preliminary estimates, would entail substantial costs and time to implement. The princical thrust of your procosed reliability analysis would therefore try to dronstrate the acc20tability g of the reliability of the present two train AWS at the Davis-Besse 1 plant. We believe that you should consider placing more emphasis on uegrading the eristing startue feedwater train to provide diversity from the present steam driven AFWS, and thus im; rove system reliability. In this regard, you should procose methods of upgrading the existing startup feedwater train which provide the increased system reliability, but which may only partially satisfy Engineered Safety Feature standards. It is the staffs opinion that this train of AFW would not have to satisfy the same seismic design regaire-'ents as the presently inst 3lled safety grade EFW. W+ts@&sB Sg p .a p
. Mr. Richard P. Crouse We request that responses to the items identified in the enclosuresJa.- forwarded within 45 days of receipt of this letter. Sincerely, ' U. Jopn F. Stolz Chi eserating Reactors Branch f4 Olvision of Licensing Enclosures : 1. AFdS Reliability Analysis Evaluation 2. Basis for AFWS Flow Requirernents cc w/ enclosures: See next page + eWS e .g i l l l 1 ~~ -,,.7, c-- 9. _m ___m,,.,,_ , - -,e,.,-,-eawyw w -w-wrc---,---.,-e--.
j I DAY!S-BESSE MUCLEAR POWER STATION UNIT NO. 1 AUXILIARY FEEDWATER SYSTEM RELIABILITY ANALYSIS EVALUATION The following itms relate to the recerinendations resulting from the,2taff's evaluatien of the information provided by you regarding the simpli71ed AN system reliability analysis. Technical Specification Administrative Controls of Manual Valves - Lock 1. and Verify Position The licensee should lock open single valves or multiple valves in series % the AW syste ove suction ciping and lock open other single valves N t ! y or multiple valves in series that could interrupt all AM flow. i ins;ections should be perfomed to verify that these valves are locked I These inscections should be proposed for incer- { and in the open position.paration into the surveillance requirements of the plant Techn r ca tions. 2. Local Manual Realignment of Valves The Cavis-Besse plant recuirts local manual realignment of valves to conduct ceriodic i tests on one AN system train and has only one remaining AFW train available for coeration, therefore, the licensee should propose Technical Scecifications to orovide -$ i ,thdt a dedicated individual who is in comunication with the control room be stationed Upon instruction from the control room, this operator =c,ule at the manual valves. re-align the valves in the AN system from the test mode to its operattonal altgnment. 3. AN System Flow Path Verification The licensee should confirm flow path availability of an AN system train ~ that has been out of service to perform periodic testing or maintenance as follows: (1) Procedures should be implemented to require an operator to determine that the AN System valves are properly aligned and a second operator to independently verify that the valves are properly aligned. (2) The licensee should propose Technical Specifications to assure that, prior to plant startup following an extended cold shutdown, a flow test would be performed to vert fy the aor al flow path frem the :-tr.try W system water source to the steam generators. The flow test should be conducted with AFW system velves in their normal alignment.
- 4. Flow Blockage by Plugged $ trainers The licensee should assure that there are no temporary strainers in place in the AN piping system that may cause flow blockale if plunced. Operating experience at several plants has shown this to be a potential corrion cause failure mechanism which could fall the entire ANS.
The suction strainers between the condensate storage tank and the pumps are an example, s. Design desls /or AN Systein Flow Requir ements The licensee is requested to provide the ANS flow design basis information required in Enclosure 2 for the Davis-Besse 1 design basis transients and acefdent conditions.
~ Enclosum 1 . 6. Diversity in the Petive Power 'or the AFWS Ptsmas We are concemed with the dependency of both AFWS pumps on stem from Other PWRs are known to have a similar configura-the main steam lines. tien (e.g., Calvert Cliffs); however, because of tse more rapid dn-out of the steam system in B&W plants, such a steam deoendency is of mor concern in Davis-Besse. a third AFWS train which will utilize a pume %. ; frtyi a source other A schedule of imolmentation snould be provided. than steam. dB89 e Sem 9 U
_= I l Esclosun 2 & asis for Assiliary Feemsster 1-hystas Fisw Assuirumenu l As a result of reca::t suff reviers of eserstin; slant Ansailiary feet-f aid-untre Systans (APC). tas sisf" sensivees tnat taa sesip bases f c-itrMa yrveidad by licensass for asun11shing Ard ensvirements fde j i fles to the staas gamerstsr(s) is assum aseguata emeral of reactse escar est are not well de*ined se secamented. n i f ) We mowest sat you Frevise tne following Arc flow tr. rip basis infer-autan as apolicable to tan easip basis t nasianu and ac=ident ann-etttans for your plant. Immatify tne plant t-unsient and acciennt cumtittsas amsidered 1. a. % satantisming Arc fler recuirements, inclueing me following i eumsu:
- 1) Lass of Main Famd (LW)
C ufV m/less of aft /stia M power
- 3) L9V w/issa of onsin and offsita M posen-
- 4) Plant contdown
- 5) Turtina trip wits and vitanut typass
- 5) Main staan isolation valve clasere 71 Main feed line trook Il Rain staan line trenc 1
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- 9) Small besak W I
Otaer transient or ac:taant canditions not listad aseve .i 10) f Das:rtha tne plant pressetion ac=sotanca cMtaria and cer=es-5. pending tecnical tasas used for es=h initiating event identi-The as:semanca cMtaria should address plant I find shove. i finits ascs as: P00R BM; e 1 - -...... ~V'meem-.g.,_ -~^*"N*wh,,.
.g. \\ - Iturian X 3 pmssure (PCRY er safety valve actuation) . Puni temperture er damage limits (gns, kT, maxima fuel samtr:1 temerrturs) .- RC3 sosling reta limit to avoid ancessive coolant shrinkage 2- - Mtatasa stass generstar level to assure sufficient stans generstar heat trnasfer surfaca to reseve decay heat and/or smet down the primary systas. .iencrite sne er.alyses ar.J us:.;;;;;stions ud err'sspending tecnnical L incation used with plant condition senstdered in 1.a. aheve testadtsg: maissa rescur power (including instr.mant er=or allowanca) a. at the time o# the initiating transient or accident. Tim delar from initiating event ta rescue trip. 2. Flest paramstar(s) unich initiatas Anis flaw and time delay c. between initiating event and introduction of AFC flow into I staan generstr(s). Minima stass senerter votar level vnen initiating ever.: s. setsfrs. Initial staas genertur vntar inventry and decletion rata befort e. l and aftar AF*C flow m. css - ident1% rescur decay haa j
- l I
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3 Maximse pressure at wMen staan is released fra steam geiersur(s) f. J' and against which the AN puss mast develop sufficient head. g. Rinisms number of staan ganartten that must rective AW flow; e.g.1 out of 27, 2 out af 47 tc flow condition - can:,inued operstian of AC pures or naturst h. cinulation. h Maziam AN inlet tapersture. J. 7ellowing a postulated staan er feed line break. time delay ass.ased in isolata treak and direc. AM flow to intact staan generatar(s). ATs* pues flow capacity allements ts ac::mmodata the tfse delay and mintain minimum staan generatar utar level. Alas identify credit taken foe primary systas heat removat { ese to bloodeun. Volume and assima tamperaturn of water in usin feed lines k. between staan generator (s) and ANS connection sa main feed.line. Operating condition of staas generatar noriuit blowdown following l 1. inittating event. Primary and secondary systas watar and estal sensible heat a. med for cosidsun and AN fisw string. i Itas at het standby and thus to cooldsen RC3 to RHR systs: cut. n. in taneerature to si:n AFV water saurce inventory. l l PDDR ORIGINAL ,,._______,_,---s-- a. e.,. .,7.. .,ym-,,m__w,-,-.w,,,,n, ,.,w, p.m,,.-- m.,- .-,,mw-_..pe,
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4= Verify that the AMi puets in your plant will supely W mesa 3. flow to the staan genentar(s) si,detamined try ites 1 and 2 ' aten considering a single failun. Identify the martin in si:ing the pump f3em ts allow for P8's recirculation flow. teal leakage and pump wear. 8 l
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~ TCLEco EDISON Docket No. 50-346 ~ go.4ao p cacx.u 1.icense No. NPF-3 Serial No. 717 we mswsa sw 4 May 22, 1981 , t' /4 y '. jVN i*SN d p ic85 :[, Director of Nuclesr R3 actor Regulation C \\ v.6. M # Attention: Mr. John F. Scola "\\ c, Operating Reactor Branch No. 4 -g,' Division of Operating Reactors \\*D g % N. ' United States Nuclear Regulatory Cocmission MQQt \\ N ~' Washington, DC 2055j
Dear Mr. Stolz:
This is in response to your letter dated April 2, 1981 (Iog No. 687) relating to the Auxiliary Feedvater System reliability analysis evalua-tion for Davis-Besse Nuclear Power Station Unit 1. to your 2, letter listed six items requiring To1 do Edison response. Attachnent 1 9 to this letter sarizes our response to itens 1 through 4 and 6. Attachnent 2 provides our response to item 5. Very truly yours, 5 T ^^ ^ RPC:SCJ ds a/5 Attach =ents cc: D7-1 NRC Resident Inspector pu o.A, y j) S b 00l // rse reuco scisen ecwany scisen puu aco vacisen avenue teucc. cro ceen p 8tofr02Tgj /$f ~ I.
Desk:t No. 50-346 Licenso No. NPT-3 Serial No. 717 May 22, 1981
- to Toledo Edison letter to the NRC on Auxiliarv Feedwater System Reliability Analvsis Evaluations Item 1.
Technical Specification Administrative Control of Manual Valves - Lock and Verify Position The licensee should lock open single valves or multiple valves in series in the ATV system pump suction piping and lock open other single valves or multiple valves in series that could interrupt all ATV flow. Monthly inspections should be performed to vnef fy that thess 'rtliss ars locked and in the open position. These inspections should be proposed for incorporation into the surveillance requirements of the ilant Technical Specifica-tions. Response: All manual valves in the suction and discharge of the auxiliary feedwater (ATV) pumps at Davis Besse 1 (DB-1) are locked and administrative 1y controlled per the existing administrative procedure AD 1839.02, " Operation and Control of Locked Valves". This procedure also requires independent reverification of res toration of a valve position if the position was changed. In addition, PT 5186.01, " Locked Valve Verification Periodic Test" ensures, on a monthly basis, that these valves are leaked in their correct position. The above controls provide aduquate assurance that the manual valves are not inadvertently positioned to interrupt ATV flow to the steam generators. As a previous commitment (Toledo Edison Letter, Serial No. 1-56 dated April 11, 1979) and an approved station procedure, these s controls are fully auditable and subject to inspection and enforcement action, if not complied with. Consequently, no changes to plant Technical Specification surveillance require-ments are necessary. Item 2. Local Manual Realignment of Valves The Davis-Besse plant requires local manual realignment of valves to conduct periodic tests on one ATV system train and has only one remaining ATV train available for operation.
- Lerofac3,... e 1. censes shculd propoc4 Technical Spec.4.ea;icas to provide that a dedicated invididual who is in communication with the control room be stationed at the manual valves. Upon instruction from the control room, this operator would re-align the valves in the ATV system from the test mode to its operational alignment.
Response: The valves that need manual realignment to conduct periodic tests on one ATV system train and could affect availability of the train are AT21 for train 1 ( AT22 for train 2), AT23 and AT50 (or AT51). These valves are in series. See T$AR figures 10-5 and 10 6. The testing is conducted by recirculating the ATV flow to the condensate storage tanks. The surveillance test ST 5071.01
, Docket No. 50-346 Litenso No. NPT-3 S:rici No. 717 May 22, 1981 "ATW System Monthly Test" requires that in modes 1, 2 and 3 an operator (located by the ATP to be tested) be in direct communi-cation with the control room when AF23 is open. If the affected train is required to be operated as demanded by the Stgam and Feedwater Rupture Control System, the control roos Obsediately instructs the operator to close AF23. Since this valve is in series with the other above mentioned valves, closure of this valve sakes the train available for feeding the steam genera-tors by closing the path for ATV flow diversion. Thus this ites is already covered by existing surveillance test require-seats. It should be noted that only one train of AFVS is tested at a time. The redundant 100% capacity train is tvailabin for feeding AFV to the stais 3enee9tnes if needed. The above provides adequate resolution of your concern. No additional Technical Specifications are therefore proposed. Ites 3: ATV Systes Flow Path Verification The licensee should confirm flow path availability of an ATV systes train that has been out of service to perform periedic testing or maintenance as follows: (1) Procedures should be implemented to require an operator to determine that the ATV systes valves are properly aligned and a second operator to independently verify that the valves are properly aligned. (2) The licensee should propose Technical Specifications to assure that, prior to plant startup following an extended j' cold shutdown, a flow test would be performed to verify the normal flow path from the primary ATV systes water source to the steam generators. The flow test should be conducted with ATV systen valves in their normal align-sent. Responses (1) The manual valves in the ATWS are controlled under AD 1839.02 (as stated in response to ites 1 above) and are restored to their required position fo11cwing periodic testing or maintenance. This procedure also requires independent reverification of any valve, whae posit un s aa shar.;W Jutir.J n.W. ',', u.c i testing or maintenance. Following performance of saintenance on any AFV train causing inoperability of the entire train (e.g. esintenance on the pump, turbine or associated controls), the operability of entire train is demonstrated by performance of any one of the ATVS surveillance tests (ST 3071.01, 3071.03 or 5071.04), Following performance of maintenance on a portion of an ATV train (e.g. valve repacking, torque / limit switch repair), the surveil-lance test is performed to demonstrate the operability of the affected equipment only. All of the above e t 2
3 Dock;t No. 50-346 License No. NPF-3 Serial No. 717 May 22, 1981 surveillance tests ensure the restoration and inde-pendent verification of the affected valves to their required position. Therefore Toledo Edison is in compliance with this item. (2) Toledo Edison believes that requirements of a flow test prior to plant startup following an extended cold shutdown is restrictive and is not required, keeping in view the already stringent surveillance requirements and other administrative controls on the AFV System. Ve believe that a verification of valve pusitions sad paap opera'oility is adequate to assure the operability of the train. In addition, performance of a flew test has an adverse i= pact on the steam generator water chemistry. This may require additional time to restore the steam generator chemistry within acceptable limits. Therefore, such a flow test is not desirable. Item 4: Flow Blockage by Plugged Strainers The licenses should assure that there are no temporary strainers in place in the ATV piping system that may cause flow blockage if plugged. Operating experience at several plants has shown this to be a potential common cause failure mechanism which could f ail the entire AFVS. The suction strainers between the condensate storage tank and the pumps are an example. 5' Response: There are no temporary strainers in the ATV suction piping at the present time that may cause flow blockage when plugged. ~ ~~ Item 3: Design Basis for ATV System Flow Requirements The licensas is requested to provide the AFVS flow design basis information required in Enclosure 2 for the Davis-Besse 1 design basis transients and accident conditions. Response: The design basis information required in Enclosure 2 of your letter for DB-1 design basis transients and accident conditions is provided in Attachment 2 to this letter. We are concerned with the dependency of both ATVS pumps on steam from the main steam lines. Other PVRs are known to have a similar configuration (e.g., Calvert Clif fs); however, because of the more rapid dry out of the steam system in B&V plants, such a steam dependency is of more concern in Davis-Besse. The licensee should state plans for providing a third ATWS train which will utilize a pump powered from a source other than steam. A schedule of implementation should be prov esd. S 3
1 ' Dock:t No. 50-346 Licensa No. NPF-3 Serial No. 717 May 22, 1981 g?j' 19jQ@ Response: To11owing our meeting with your staff on March 5,1981, we have initiated a detailed probabilistic risk assessment study on the W system. This study considers the Pre-TMI-2 DB-1 N /\\ M S, Post TMI-2 M S and the Post TMI-2 M S with a. third / source of W consisting of the existing electric motor driven startup feed pump discharging into the W nozzles. This study is expected to be complete in late July 1981, and will develop the probabilities for system unavailabilities for the above configurations. This study will also identify dominant failure contributors to system unavailability. Based on the conclusions and recommendations of this study, upgrades to the existing M system and/or the startup feedwater pump will be planned. We will be willing to discuss our plans and senedule with you at the time that this study is completed. ds a/1-4 eP
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Docket Ns. 50-346 License No. NPF-3 Serial No. 717 May 22, 1981 to Toledo Edison Letter to the NRC on Auxiliary Feedwater System Reliability Analysis Evaluation s-Item 1.a. Identify the plant transient and accident conditions considered in establishing AFWS flow requirements, including the following events: 1) Loss of Main Feed (LMFW) 2) LMTW w/ loss of offsite AC power 3) LMFW w/ loss of onsite and offsite AC power 4) Plant cooldown 5) Turbine trip with and without bypass 6) Main steam isolation valve closure 7) Main feed line break 8) Main steam line break 9) Small break LOCA 10) Other transient or accident conditions not listed above Response to Item 1.a. The original design of the Aasiliary Feedwater System (AFWS) established a requirement for a minimum flow sufficient to remove heat load equal to about 5% full power. This flow rate was not based on any specific transient. However, where appropriate, this flow rate is used as part of the transient analysis for the accidents considered in the FSAR. Table 1 contains a list of those transients considered in the FSAR along 5 with their acceptance criteria. A value of 800 gym has been determined to be an acceptable flow for the AFW system taking into account a single ~ failure. This value has been shown to be acceptable for a loss of main feedwater (LMFW) event from 112% full power (including instrument error allowance). This event requires the maximum heat removal capacity (including RC pump heat) for the AFW system, therefore, is considered the design basis event. For the above parameters and assuming 1.2 times the ANS 5.1 decay heat, analysis showed that the primary acceptance criteria were met (with a reactor trip on high RC pressure). These criteria are: RCS peak pressure less than 110% DNBR greater than minimum allowable Site boundary doses less than 10CFR100 limits In addition to meeting these criteria, with a minimum AFW flow of 800 gps, the pressurizer did not go solid. Accidents 7, 8,12 and 13 of Table 1 specifically require AFW for mitiga-tion. As reported in the FSAR, the results are acceptable with an 800 gpa AFV flow. The other accidents listed in Table 1 do not require AFW for mitigation although the availability of the AFWS is assumed. 'the otner eveats listed in the questions but not included in Table 1 are discussed below.
Docket Ns. 50-346 License No. NPF-3 1 Serial No. 717 [. May 22, 1981 Loss of Onsite and Offsite AC Power - This event is not a design basis for Davis-Besse 1 since it requires a failure of both emergency diesel generators. However, one train of AFWS is capable of supplying IEW to the steam generator even with loss of both onsite and offsite AC. power. Plant Cooldown - Plant cooldown with AFW and with a loss of offsite power is a controlled event with decay heat levels equal to or lower i than the loss of feedwater event identified as the design basis event. The design basis event bounds this case for the AFW flow required. 1 i i Turbine Trip With and Without Bvpass - This event does not affect the AFWS unless MFW fails. In which case, the loss of MFW event previously addressed would bound the AFWS design. i Main Steam Isolation Valve Closure - Again, this event does not directly affect the AFWS unless MFW is lost as discussed above. A Small Break LOCA - The AFW criteria assumed for this event are described in the B&W report entitled, " Evaluation of Transient Behavior and Small Reactor Coolant Systes Breaks in the 177 FA Plant, Volume 3." This report was submitted to the NRC on May 22, 1979 with a letter serial No. ~$ 506 and demonstrated that an AFW flow of. 800 gym for Davis-Besse will not lead to the violation of the acceptance criteria. a. Ites 1.b. Describe the plant protection acceptance criteria and corresponding technical bases used fo,e each' initiating events identified above. The e-
- I acceptance criteria should address plant limits such as
Maximum RCS pressure (PORV or safety valve actuation) Fuel temperature or damage limits (DNB, PCI, maximum fuel central temperature) RCS cooling rate limit to avoid excessive shrinkage Minimum steam generator level to assure sufficient steam generator heat transfer surface to remove decay heat and/or cooldown the primary system Resnense to Item 1.b. 4 The design basis event for sizing the AFWS is the Loss of Feedwater i event discussed in the response to Item 1.a. The acceptance criteria for the other transients which assume the availability of AFW are given 4 in Table 1. The acceptance criteria for these accidents include RCS pressure limits, ensuring RC pressure boundary integrity, fuel limits and offsite dose limits. The RCS cooling rate is not an acceptance criterion for accident i analyses. An overcooling event that drains the pressurizer is not l desirable, however, it does not violate any of the accident analysis j acceptance criteria. 4 3 I 4 2 1 --. _,~._._... _. ___[._._,.. .___-...-_,_...-.m..
1 Docket No. 50-346 License No. NPF-3 Serial No. 717 May 22, 1981 Maintaining a minimum steam generator level is not an acceptance criterion for accident analyses. It is desirable that the reactor be tripped and AFW initiated prior to steam generator dryout, but this is not-required in order to obtain acceptable results. After AFW has been initiated, the high injection point in the steam generator reduces system dependence on a specific level for adequate heat transfer. The steam generator level control is set low for decay heat removal with forced circulation and for natural circulation without a small break LOCA. The level is set high for small break LOCA event. For a detailed analysis of this dual level control, see Serial No. 475 dated 12/22/78 and Serial No. 471 dated 12/11/78. D O y mun e e ~ ~ 3 - ~_, ,,,-.3 ,..,_y.,,,-y9-- -,w.n -v----
7 Docket Ns. 50-346 License No. NPF-3 Serial No. 717 May 22, 1981 TABLE 1 ~ Accident Description FSAR Section Acceptance Criteria 1. Startup Accident 15.2.1 A,B 2. Uncontrolled Control Rod 15.2.2 A,B Assembly Group Withdrawal at Power 3. Cuatrol Rod Assembly 15.2.3 A,8 Misalignment 4. Makeup and Purification 15.2.4 A,B,C Malfunction 5. Loss of Forced Reactor 15.2.5 D,E Coolant Flow 6. Reactor Coolant Pump 15.2.6 A,3 Startup Accident 7. Loss of Normal Feedwater 15.2.8 B,E Due to Closure of Feedwater Valve, Pump Failure, or a Feedwater Line Break I 8. Loss of All AC Power 15.2.9 B,E (Station Blackout) ~ ~~ 9. Excessive Heat Removal 15.2.10 B,E Due to Feedvater System Malfunction 10. Steam Generator Tube 15.4.2 F,G Rupture 11. Control Rod Assembly 15.4.3 F,G Ejection Accident 12. Steam Line Break 15.4.4 E F,G 13. Loss of Coolant Accident 15.4.6 F d' 4
l Docket Ns. 50-346 Liccnsa Ns. NPF-3 Serial No. 717 May 22, 1981 KEY ACCEPTANCE CRITERIA A Reactor Thermal Power Jese than 112% of Rated Power B Reactor Coolant Pressure Less than Code Pressure Limits (110% of Design Pressure) C Minimum Shutdown Margin of 1% k/k during Refueling Conditicas D Minimum DNB Ratio Greater Than 1.3 E Fuel Cladding. Temperature Less than 2200*F F Resultant Doses Less than 10CFR 100 Limits G No Loss of Reactor Coolant Pressure Boundary Integrity O l eP + M ee i 0 49 5 1 i
l Docket Ns. 50-346 License No. NPF-3 Serial No. 717 May 22,1981 Ites 2 Describe the analyses and assumptions and corresponding techniEal~justi-fication used with plant condition considered in 1.a. above. Response to Ites 2 As discussed in Response to Item 1.a., the design basis event which verifies the AFWS design flowrate is loss of main feedwater. The analysis assumptions for this event are listed below. Corresponding technical justification where not specifically listed, is based on licensing requirements and prudent engineering judgment at the time of the analysis. The infonmation is not provided for the other events identified in Ites 1.a. and Table 1 because the LMDi event is the most limiting. The LMFW analysis used for this response has not been done for Davis-Besse although it has been done for the Sacramento Municipal Utility District's Rancho Seco plant. These two plants are very similar in design and performance, therefore, the results for the Rancho Seco analysis are also applicable to Davis-Besse, and the numbers given' below are taken from that analysis. A comparison of the applicable parameters was made for these two paints. The results of the comparison demonstrated that the Davis-Besse ATW system is adequate for cooling following a LMFW transient. a) Maxisus reactor power (including instrument error allowance) at the time of the initiating transient or accident. 5 112% full power (including instrument error allowance) b) Time delay from the initiating event to reactor trip. The reactor will trip on high reactor coolant pressure approximately 14 to 15 seconds after the loss of main feedvater event. c) Plant parameter (s) which initiates ATWS flow and time delay between initiating event and introduction of ATWS flow into steam generator. The AFWS is initiated by a low steam generator level signal from the STRCS. It is assumed that the time delay between receiving the initiate signal and full ATV flow to the steam generators is 40 seconds for a case without loss of offsite power. This is a total delay of approximately 55 seconds from the loss of main feedwater event. d) Minimum steam generator water level when initiating event occurs. O m 6
,e } Docket Ns. 50-346 License No. NPF-3 Serial No. 717 May 22, 1981 Steam generator inventory rather than water level is used as an input to this analysis. e)' Initial steam generator water inventory and depletion gate before and after AFWS flow commences - identify reactor decay heat rate used. The initial steam generator inventory is dependent on power level. For this case, a liquid inventory of 39,600 lba per steam generator was used. The depletion rate of the inventory before initiation of AFW averages about 248 lba/sec. Following initiation of AfW, there is not enough 1evel to determine the depletion rate (since AFW is initiated on low steam generator level). Following com-plate depletion of the liquid inventory, the entire AFW flow is vaporized dutil decay heat plus RC pump heat drops below the capability of the AIV system. At that time, stema generator inventory begins increasing again. The decay heat used in this calculation was 1.2 times the ANS 5.1 decay heat. f) Maximum pressure at which steam is released from the steam generator (s) and against which the AFW pump must develop sufficient head. The peak steam pressure occurs shortly after AFWS initia-tion and is about 1075 psig. Soon after this peak, however, the steam pressure is controlled by the first bank of I steam safety valves to a pressure of 1050 psig. g) Min 4=u= number of steam generators that must receive AFW flow. This analysis was run assuming both steam generators were available, however, the heat load can be removed with one AFW pump and one OTSG. See FSAR Section 15.2.2 for details. h) RC flow condition - continued use of LC pumps or natural circu-lation. Continued operation of RC pumps was used for this analysis. i) Maximum AIV inlet temperature. I An inlet temperature of 120 F was used. j) Following a postulated steam or feedline break, time delay assumed to isolate break and direct ATW flow to intact steam i generator (s). AFW pump flow capacity allowance to accommodate the time delay and maintain minimum steam generator water level. Also, identify credit for primary system heat removal due to blowdown. t. 7
.t Docket No. 50-346 License No. NPF-3 Serial No. 717 May 22, 1981 FSAR Sections 15.4.4 and 15.2.8 contain the details of the assumptions used in the main steam line and main feedwater line break analysis, respectively. The time deiaf assumed (to isolate the isolable steam or feedling break and direct AFW flow to intact steam generator at full flow) is 40 seconds from the AFW initiating signal assuming no loss of offsite power. k) Volume and maximum temperature of water in main feedlines between steam generator (s) and AFWS connection to main feed-line. There are no piping connections between the AFWS to the main feed line at Davis-Besse 1. 1) Operating condition of steam generator normal blowdown follow-ing initiatihg event. Davis-Besse steam generators do not have a blowdown system at the present time, ~$ a) Primary and secondary system water and metal sensible heat used for cooldown and AFW flow sizing. 6 A heat capacity of 1.256 x 10 BTU /*F is used for calculat-ing the volume of feedwater required to cool the RCS to decay heat system parameters. I n) Time at hot standby and time to cool'down the RCS to DER System cut-in temperature to size AFV water source inventory. The condensate storage tank is sized to accommodate the plant at hot shutdo g for thirteen hours followed by a six hour cooldown to 280 F as reported in Section 9.2.7.2 of the FSAR. ds d/10-17 ~ 8 i}}