ML20133B946

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Trip Rept of 840703 Site Visit Re Training Organization & Programs.Program Not Implemented Due to Lack of Senior Mgt Support
ML20133B946
Person / Time
Site: 05000000, Davis Besse
Issue date: 07/05/1984
From: Shafer W
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To: Norelius C
NRC
Shared Package
ML20132B273 List:
References
NUDOCS 8510070167
Download: ML20133B946 (7)
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Text

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NUCLEAR REGULATORY COMMisslON

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,p a6 n attym, sLLimoes seest JUL 5 G84 MEMORANDUM FOR:

C. E. Norelius, Director Division of Reactor Projects FROM:

W. D. Shafer, Chief, Reactor Projects Branch 2

SUBJECT:

DAVIS-BESSE, TRIP REPORT On July 3, 1984, Messrs. Jackiw, Rogers, Kosloff and I met with Toledo Edison employees T. Myers. M. Stewart and contract employee C. Thayer to discuss the status of the diagnostic perfome.d on the Davis-Resse Training Organization and Programs.

The major course of action taken by the licensee to date has been program development. Much work has gone into writing new procedures for tie Training DepartmentandtheinitiationofaTrainingSystemsDevelopment(TSD) Action Plan.

The TSD Action Plan development was supported by almost every organization at the Davis-Besse site. Many man-hours were invested in think tank sessions used to identify the needs of each organization. When the TSD Action Plan is completely implemented there is little doubt that many of the training problems identified at Davis-Besse will be resolved.

The implementation of the program however, is blocked by Senior Management.

Clear evidence of lack of Senior Management support is recognized by the following:

1.

Implementation of the TSD Action Plan requires that the Training Department be staffed with 21 personnel.

They presently have 13.

The open positions may not be filled without prior authorization by the President of Toledo Edison.

The proposed organizational needs were recognized as early as January 5,1984. The Vice President, Nuclear R. P. Crouse did not request authorization until July 2, 1984.

Of interest, we were informed that the Training Department has always been budgeted for 21 personnel.

However, even though budgeted, the positions could not be filled without Presidential authorization.

2.

Even if the President authorizes the proposed organization there is little chance that the positions will be filled due to the company's wage and salary structure.

The Training Department Manager feels that he may be able to get some entry level people that can be trained but will move on to better paying jobs within the company as soon as possible.

There is 1

l no possibility that experienced personnel at the plant will accept a pay DR 4

4.

C. E. Norelius 2

JUI 5 E4 loss by transferring into the Training Department.

Evidence that this problem is real is represented by the licensee's failure to fill two positions previously authorized by the Company's President. There has been no internal interest and all outside offers were rejected because they were not competitive.

The root of this problem is centered at the Company's wage and salary structure.

The training positions at Davis-Besse are compared to similar positions within the company and these positions are not considered as important or critical.

Of interest, the Wage and Salary Committee meets once a month to evaluate and determine the salaries for proposed positions.

The frequency of these Committee meetings has not changed.

An increase in frequency would have indicated that some effort was being made to negotiate.

3.

While Senior Management has obviously supported the development of a training program, they have done nothing to support its implementation.

There is no mission policy directing the department Directors, Managers and Supervisors to comply with the newly developed training program. The Training Manager stated that the department supervisors can and do cancel scheduled training courses for their personnel and the Trainirig Department has no say-so in this decision.

While the diagnostic performed on the Training Department is not directly a part of the Performance Enhancement Program (PEP), many of the PEP elements, such as think tank sessions, task analysis and action plan development were used to develope a resolution for each problem identified.- If Senior Management's responsiveness to the training effort is representative of what we can expect when PEP starts recomending actions, there is little reason to believe that PEP will be successful.

The licensee's response to the training issue clearly indicates that while they are capable of identifying their problems, they are not able to resolve them for which the PEP was intended.

In summary, the Davis-Besse Training Program appears to be well developed, but has not and may not be implemented because of the lack of Senior Management support.

The lack of training at Davis-Besse has long been recognized as a prime contributor to the number of personnel errors (more than twice the national average) at Davis-Besse.

It is also the major reason we have an enforcement meeting scheduled to discuss the lack of personnel knowledge regarding the operability requirements of safety systems.

The staff feels that a strong, well implemented training program is a major underpinning of successful safe operations at a ruclear facility.

We recorrnend that Mr. Keppler meet with the highest level in the Toledo Edison organization to resolve this problem.

Q

t 5 '^E4 "fi C. E. Norelius 3

Should you have any questions regarding this memo we will be happy to meet with you.

DAs W

. D. Shafer. Chief Reactor Proj cts Branch 2 cc:

I. N. Jackiw W. Rogers D. Kosloff

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A team of eight inspe:tcrs fr de the C arcting Eaatter Fr:.; rats Branch ccn:ucted an announced inspection at the Cavis-Lesse fU: lear Pc-er Station, and the T: cac Edison Cc ;any cffices esring the peric; of July 3C.

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Manage.:ent ccretrci s in tar. at sas we re e.aluated ar.d assign 3: perfc.mance cate; cries as felicas:

C:nmittee Activities, Quality Asserarca, Design Changes and f*.ocificatices, Naintenance, Plant Op.eretic 5, Trs:.ra ar.t, and Enciciogical Contrcis oare rated Cate;;ry Two; Ctrrective A: tion S.s s.ers, C;aretor Trairing, an: N:r. 0;aretor Trainin; were ratec C a t ec.o ry Tr.re e.

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Unresclvec Items:

1.

Apparent failure of the Station Review Board tc revie! procedure codifications within 14 cays of their effective cate.

(Observation 7) 2.

Apparent failure cf the Cc:;any f.'u: lear Review Ecarc to review a rect;nized indica icn of a deficien:y in the cesign of a safety-re;atec c:: enent.c (0:servaticr.El Ouality Assurance:

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iant r.:.:ifi;e:iors and the failure te ;erferr safe:) e'.a wa;; ns f;r o.:rary leti sr:eicir.; irstallet cn safety-relatec ;i:in; : 5:(~s.

The lice sses efft:ter.: tracking and re:ced lee;ir; s s er f or f e:ili;) c t ;e re;uests was ccns':erec a strength.

Unresolvec I ers:

1.

Apparent procedural deficiency which provides the Octential fcr crission of safety evaluations required by 10 CFR 50.59.

(Observation 3) 2.

Apparent failure te update the Final Safety Analysis Report (Otservation i

4) 3.

Ap:arent f ailure te analy:e the loading ef fects of placing ter.p;rary leac snielding cr.. afety-reia ed piping.

(0:ser$ation 5)

Maintenance:

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calibratior. tes e:;i:re.: anc a ia:i of in-the-ftei: r.tir.;e r.te.:e s.:e. a t i:.

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(Chie.ation 1).

2.

A;;arent failure to re.ain recorc.s of ssaluaticns of prior use of cefect u e -sts.:ing a,a :est eo.

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(Obseriation 2).

3.

A;;arent feilure to ;r.' vide the r 5iew and control cwer send;r rarvals used to ccadsc: s af ety-ret e:ed r. air. enar.ce.

(Cis'ers ation 2 ).

4 1;:aren f ailure cf raintenance ~;rt orders te specify cde;; ate

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A;;a er.: f ailure te establish ar, effective ir..te::r t s ter te: seen the control root and the Shif t Supervisor's cf fice as cc cittec to

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(Observation 4) 3.

Apparent failure to use technically trained indivicaals as the plant operations Shif t Acminsitrative Assistant as coanitted to the NRC.

(Observation 5)

Corrective Action Svstems:

Cateoorv Three Ar a:;arer.. t-ealdson of the corrective action syste was found regarcing a ::7::t r de 'ailure of the high pressure irjecticn syster..

Weaknesses ser: tining tc this issue were:

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(Cbser.ation :,)

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Catecory Tso

'leair. esses identified in the area of procurecen: were the inadequate pro:ecural guidance for the preparatien of procurer.er. do:uments and the upgrade of commercial material for nuclear safety related use, the f ailure l

l to comply with pro:edures for station material control, and the procurement of primary piant chemicals as non-nuclear safety related materials.

i A strength was ncted ir the ih:rovemer.ts made since the last performen:e appraisci ir:spection.

i Enresolsed Itees:

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Ap;arent failure tc review purchase creers prior to issuance.

(C:se vatien 3) 1.

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La:L of prcced.res (ticassay program)

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F.e v ' en of cro:3:.res (r: 50,I. review cf precer.res related to tra iyses c, e,. t..+rt sa: : es,

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( ;.. Ji> A N04)CE OF V10LAllCN Toledo Fdiscn F..my Cacket No. 50-M 6 Davis-3 esse 1 As a result of t'.e ir s; action (c. ' o.t-d on.ipril 23-27, hy 1-4, and May 8-11, 1934, and in wrder ce with the h;C infcrcec.ent relicy, I R $987 (March 9, M2),

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1.

10 CFR 50, *;;(ndix B, Criter'on V, as irple ested by Toledo Edison (.nlity As seance F rcgram as de> crit.ed in Section 17.2.5, it.cluding a com.it +nt to ANSI NIE.7-1972, requires that activities af fecting quality be prescrit,ed by dcre e.t-d instructicns and procedures.

Section 5.1.5 of ANSI 18.7-1972, end Er: tion 1.C.6 of NS!G-0737, require inde;.endent verificatirn of t % ;i ; -:tisities e16tive te r m val fron and retsrn to service of plent e. :; e nt.

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  • tQ'.i; ment DUt of 1E sICe.

5 h e r i # i : f '.

inis is a Se.erity Le.el IV violation (Supplerent I).

2.

T,-chni:a1 Spe:ificaticn, Sections 6.5.1.6. d and 0.5.1. '. a, require the Station Ft.it Scard to review all proposed changes or modifications to plant systems or equiprent that af fect nuclear safety and recommend ritten approval or disapproval of changes or modifications to the Station Super-intendent.

Contrary to the above, ttmpoE ry modif_icit3cns. associated with noncon-formance reports NCR-232-81, (Limitorque Valve Modification), NCR 392-81 (Service Water Valve Modification), and NCR 83-01 (Auxiliary feed ater Pump Steam Line Modification) were not reviewed by_th_e StalionJexiecBoArd r-: '.4-and a recorn.endation concerning the modification's acceptability was not made to the Station Superintendent.

This is a Seserity Level V sicletion (Supplement I).

3.

10 CFR 50, Appendix B, Criterion XVII, as implemented by the loledo Edison Operating QA Program and the FSAR Section 17.2, require that the applicant i tert Uth applicatie regulatory r quire-shall pr:. ice record st:-ap :ent t

rte nt s.

The Toledo Ecison QA T rograrr cc-mits te ANSI M5.2.5-1974 and r g d ai c,c hi:e 1.E". Ses'stor 1. Ort:rt-15M..itt a en:er'.ior s:e:i-efyin; a t : har fire ; ::e: tin re.'.; f u t e::.rc ru se f ectiities.

Contrary it the ebeve, recorcs cf audits, aucitor and Q: inspector qualifi-cation /certificatior, a*.d caiitrations were not. pro.ided the recaired prc-

.. ~ ~. _ _

inis is a feverity Lesel V siclatior. (Suplerent I).

NNW&'f 9 2 &g i

T 9

A;;.e nci x 2

4.

10 CFR 50, Ap;endi> B, Criterion XVI, as inplemented by the Toledo Edison's Operational (;slity Assurance Frugram, requires that r.rtsures be esta-blished to ac$ure that conditio~s -Nerse to quality are prceptly identified a.nd corrected.

Contrary to the aS:ve, five epen Nanconforcance '+ ports stre noted with ter.porary f f aes or dispositions for short term use.

T?sse temporary conditions had axisted for a period of one to five ycars.

Additionally, two Corrective Action Requests, written in 1932 with 1983 scheduled comple-

~

tion chles, we_re s.till~open at the time of this inspection.

+

This is a Seserity level V violation (Sepplement I).

5.

10 CFR 50, Appendir B, Criterion XII, as implemented by the Toledo. Edison Operaticnal 0 ality Assu-ance Program, reovires that measures be established te asswre that tr:1 5 962es, instruients and ot'.er rr Ps uring and testing f, cevi:es use: f r e::'vities af f ecting c.ality are trc;triy controlled, caiitrtted 69: h:,b.ited at spe:ified periods te raintair +. curs:y vithin ne;u s a r,s iicits.

Contrary tc ths n :st, tes and easurino ecuil u.nt,e,ai' ele for use ty t*t Quality Contrei 06 6-trent was not controlled ar.c cali:-ated as requirec.

This is a Severity Level V siolation (Supplement I).

6.

Davis-Besse Nuclear Foser Station Technical Specification for Reactor.

Coolant System Chenistry, Section 3.4.7, requires that the reactor coolant i

system chemistry shall be maintained within the steady state limit for' - J

j chlorides of % 0.15 ppm.

1 Contrary to the above, the chloride concentration in the reactor coolant system exceeded the limit, to a maximum of 0.26 ppm, for a total of 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />.

The underlying cause was determined to be the premature breakdown j

of recently installed resin in the purification deminera1{rers as a result of the follesing:

j a.

Pur_ chase of__the resin was uncontrolled as a non-Q itet with_no receipt inspection or testing, allowing an unacceptable type of resin to be installed in the demineralizers, j

b.

FailutA_lo follog_ procedure L14782 which specified the resin chemical and physical requirements.

i, l

Inis is a Ee.e-ity Levei IV siciaticn (Supplement I).

)

7.

10 CFE 50, Appendix E, Criterion 11 states in part, "The prograr snail provide for ir.do:trination and training of personnel perfcering a:tisities_ '

i affecting coality as necessary to assure that suitable prcficiency is achieved and r. air.tained."

i 4

'i

.:'s 3

The Cis is-?+5 s e Updated Ea'ety A i'ysis Report (USAR) in Section 13.2.1 states that the trainir.g prc.grem is drocribad and adTinistered by the

~ aD WS t.- r '.s s o f r ror ed r e s,

f.< tion 13.2.2.2 of the USAR includes

-cc c.it.+r.ts 15 st Cha.ristry a nd " Sith Physics (C8.hP) parsonnel are properly t ained and maintain ;, cficiency in their required job skills thro s.5 tor.t' r ued training.

5-ti 2n 33.2.2.3 of the USAR includes

..c Vit-ents tti,t c. air.t er.an. e p.:: s:..el ar e p repa rly t ra ir.ed to pe r f orm their J bs aad that t'.ey re:ain p oficient in the r equired job skills.

Ad ini tratise ProceAre AD H?S.00 (Personnel Trair,ing Frogram) requires s

initial t aining and cor,tinair.g t aining f or C'M p -. < cr.r,el per AD 15?S.12 and fcr all rai. tic.ari:e per>c-.el per AD 1528.11.

Contrary tc the above, t.he Master Trainino Schadule_for_1CS4 did not 1,: e.t.

, b :,,..i f..v t t : t a r,v. i r.ii t i a_l..o r c : 11 n u i. ng. t r.a.s c r.61 or,c r E.9ctrical 1,ir.in 7

...y f:

ai nte.E n:e pu s ; r.ne i.

_L.-

Tr's is a "+.e rity Level V siciation (Sep;.lement 1).

r.E-.

ic : 1+ :.-c, is ions of 3 0 C;F. 2. 201, y ou a-e r e S -1 t c s.d rit ic tr s c ic e. 'i' # r tr.irty cays cf the cate of thi s Net'.c 4 a c r i t *. c r. s t 3 *..; r. - t c- (>:.c.5.C.r i r.

r eply, incisting for each iter of r.: :e ;1'ance:

(1) cct-rectise e: tit r t e.en a*>d the results achiesed; (2) tcrrective 3.:t er. tc be s

i

2;e tc a s :d c ' ' t'.t.r ncr.:o.pl i ar.ce ; and (3) the date

.'.e r. f ul 1 ( t r;.li r.:e iil be 6: ie ec.

C: *.sMe r atier, ray be gis en to o ter ding y c ar re s p:.nse tir e fce gooc c %se 5 %n.

JUL 11 1954 d

6

m. Ac@,- w Dated R. L. Spessard( Director Division of Reactor Safety I

1

-/

.Y 'f ~

(vnoet No 5*'-**.'-

I,i i celu I il t si,n f.nep.ir',

A!TS Wr 4't h.t ril D

'. ' #.4, e dice Pr es u!*eit hur. lear Edison P I.ila i

100 Mod t sa. Avenue Io l eito, 'A 4IM2 Gen t l ewn This confirms our plans as <!iscussed on July 2, 1984 net.een "

4 - t e r s

.?

jour s t.i f f $nd Mr.

I Jacht, c' m, staf? In conduct an (<if orcement Coe ence with you and member. n' four staff at 's. o i + a. m.

(CDT) 'in Jul, ie, l'-4 it th NRC Region !!! of f ire, in Glen E llyn,

!'-ots This infortemet

(. r" ' e r e r r -

relate =, to the tuntinuing equipment o'gerability problems.hich

.e r e.itte it:i.ted to personnel errors.

We will be glad to discuss any questions you may have concerni'1y this matte'

)

SincereIy,

.* I, I C. E. Norelius, Director Div:iion of Reactor Projects 5

i CC w/ encl:

l T. D Murray, Station Superintendent DMB/Occument Control Desk (RIDS)

Resident inspector, Rill Harold W.

Kohn. Ohio EPA I

James W.

Harris, State of Ohio Robert H Quillin, Ohio Department of Health i

i l

E'Ii R!LL "III

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-9407t60224 040709 PDR ADOCM 05000344 R

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,j 'g UNITED STAT s

,s e.g NUCLEAR REGULATORY COMMISSION g

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g wAsecuoics.o.c.zosss (9

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1 SEP 3 1980 Docket No: 50-346 g

i MEMORANDUM FOR:

Thomas M.'Novak, Assistant Director for Operating Reactors. DL FROM:

Paul S. Check, Assistant Director for Plant Systems. DSI i

SUBJECT:

f DAVIS-BESSE UNIT 1 AUXILIARY FEEDWATER SYS EVALUATION 1

Feedwater System we must complete the work outli Plan Developed as a Result of the THI-2 Accident " paragraphs u

"NRC Action

,[

n their auxiliary feedwater systems :to include:II.E.1.1 of NURE q

E e

(1) perfoming a simplified AFW system reliability analysis; I

. j (2) criteria of Standard Review Plan Section 10.4.9 t.:

i Position 10-1 as principal guidance; and

'"1 (3) reevaluation of the AFW system flow rate design bases and criteria.

The licensee has provided the information mquested by the staff r simplified AFW system m11 ability analysis (Item 1 above).

ng the has been evaluated by the Probabilistic Analysis Staff (PAS) and th The information mendations resulting from that evaluation which must be addressed licensee are included in Enclosure 1, questions 3 through 9,11 com-Questions 1, 2,10 and 12 address ASB concerns regarding the licen

,13 and 14.

reliability study.

~~

As we stated in our recen't memo to you dated August 5,1980 which the Crystal River Unit 3 questions NUREG-0660 is not clear regardi marded mance of the deterministic review

--quently ASB will undertake this wor (k..However,)for us.to perf ng perfor-evaluation, we will need additiohal infomation from the licensee request for information is included in Enclosure 1 as question 15 Our

~

6

Contact:

V. Leung x28241 i

)

Thomas M. Novak

. SEP 3 1980 "

Question 16 of Enclosure 1 concerns the reevaluation of th rate design bases and criteria (item 3 above).

4 (79.05, 79.05A and 79.05B) issued after the THI

.. Bulletins The licensee has responded to these bulletins.

re not completed.

project manager will complete that portion of the reviewWe understand tha plants were ordered to shutdown shortly after th g

c and Wilcox included both short-term and long-tem actions.

xi The Orders Orders and issued safety evaluation reports liftin actions in the concerns completion of our review of the long-tem actions of the II.K.2.8 Completion of this; item will be perfomed under II E 1 1 s.

pi The licensee may have previously responded to some of these it ti fore, should be advised that a reference to a previous resp

((

ems and, there-amendment, etc.) would be acceptable.

onse (letter, FSAR.

that clarification will be pmvided regarding questions 10The licensee s h

provides guidance.

4 hj and 12 when ICSB You should be aware that this licensee has previously provided g

the staff regarding AFW pump endurance testing.

n omation to We propose to prepare similar enclosures to be fontarded to th d

operating plant you to discuss an(y problems which you see in proce lessRanchoSecoandTMI-1).

j e other B&W We are prepared to meet with The Itcensee should be requested to provide responses by Oct b manner.

o er 1,1980.

ekIAs for' Plant Systems nt Director 1

Enclosure:

Division of Systems Integration As stated cc:

D. Eisenhut D. Ross i

1 R. Reid

~ ' ~ ' -

O. Parr G. Edison 6

V. Leung 6

.s.=e-

)

\\

y ENCLOSURE 1 Davis-Besse Nuclear Power Station Unit 1 Auxiliary Feedwater System Reliability Analysis Evaluation 1.

Section 1.5 of the auxiliary feedwater system (AFWS) reliability analysis for Davis-Besse Unit 1 defines the AFWS mission success criterion as the attaiment of AFW flow from at least one pump to at least one steam genera-tor. This definition is incomplete. I.t should include the requirement to deliver AFW flow to the steam generator in a. timely manner to preclude ;

steam generator' dryout. The succes's criterion should be revised and resub-mitted accordingly,s 2.

We also consider that the analysis of maintaining adequate core cooling

{#

without a break in the RCS piping using one makeup punp and the startup m

feed punp in addition to the opening of PORY within 30 minutes of loss of' l

. ~

{

main feedwater to be unacceptable in meeting this requirement.

Items 3 through 17 relate to the reconsnendations resulting from the staff's evaluation of the information provided by you regarding the simplified AFW system reliability analysis. The reconsnendations are categorized as generic and additional; as well as short-term and long-term.

Short Term 3.

Technical Specifiestion Administrative Controls of Manual Valves - Lock and Verify position (GS-2) l The licensee shoul'd, lock open single valves or multiple valves'in series in the AFW system pump suction piping and lock open other single valves or multiple valves in series that could interrupt all AFW ficw. Monthly inspections should be performed to verify that these valves are locked and in the open position. These inspections should be proposed for incor-poration into the survei11ance requirenents of the plant technical specifications.

I L

\\

2-4.

Emergency Procedures for Initiating Back-up Water Supplies (GS-4)

Emergency procedures for transferring to alternate source of AFW supply should be available to the plant operators. These procedures should include criteria to inform the operators when, and in what order, the transfer to alternate water sources should take place. The following cases should be i

,=

covered by the procedures:

P l

(1) The case 1,n which the primary tater supply is not initially available.

The procedures for this should include any operator actions required to I

protect the AFW system peps against self-damage before water flow is initiated, and (2) The case,in which the primary water supply is being depleted. The procedure for this case should provide for transfer to the alternate water sources prior to draining of the primary water supply.

5.

Emergency Procedures for Initiating AFW Flow Following a Complete Loss of Alternating Current Power (GS-5)

The as-built plant should be capable of providing the required AFW flow 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 AFW pump train independent of any AC power source.

If manual AFW system initiation or flow control is required following a-complete loss of AC power, the licensee should establish emergency proce-dures for manual initiation and control of the system as needed. (See reconnendation GL-3 for the longer-term resolution of this congern).

o

O E 6.

AFW Systen Flow Path Verification (GS-6)

The licensee should confirm flow path availability of an AFW systen train that has been out of service to perfonn periodic testing or maintenance as follows:

(1)

Procedures should be implemented to require an operator to determine that the AFW system valves are properly aligned and a second operator to I

independently verify that the valves are properly aligned.

l (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 verify the normal flow path from the primary AFW system water source to the steam generators. The flow test should be conducted with AFW system valves in their nonnal alignment.

Additional Short-Term Reconnendations

~

~

7.

Interaction of AFW with Integrated Control System (ICS) and with Steam and Feedwater Line Break Detection and Mitigation Systems Because of the potentially significant interactions with the AFWS possibly resulting from the steam and feedwater line break detection and mitigation systems (SFRCS/FOGG) and the ICS, we believe that information should be provided to the operating crews on means to detect and cope with AFWS interruptions caused by failures in these systens.

Such infonnation may be in the form of training arid /or procedures.

We note that training with respect to interruptions caused by ICS faults may already be artcompassed by requirenen'ts' resulting from Oconee event of November 10, 1979 and the Crystal River event of February 26, 1980.

Longer term reconnendations relating to this same concern are discussed below.

W 4-8.

Human Error During Test and Maintenance The licensee should assure that plant procedures are written to re uce haan induced comon mode failures of all AFW system trains.

For example, the licensee should stagger testing of AFW system trains, i e

.., for planned g

testing, no more than one AFW train (or pump) should be tested e

same shift crew..

9.

Flow Blockage hy Plugged Strainers.

r h '

The licensee should assure that there are no temporary strai in the AFW piping systen that may cause flow blockage if plugge ace i

Operating experience at several plants has shown this to be a potential comon cause failure mechanism which could fail the entire AFWS.

1.:

The suction strainers between the condensate storage tank and the peps I

10.

Indication of AFW Flow to the Steam Generators (AS-3)

The licensee should implement the following requirements as s Item 2.1.7b on page A-32 of NUREG-0578:

e y

(1) Safety-grade indication of AFW flow to each steam generator s 1,

be provided in the control room.

(2)

The AFW flow instraent channels should be powered from th buses consistent with satisfying the emergency power diversit ency requirements for the AFW systen set forth in Auxiliary Systems B Technical Position 10-1 of the Standard Review Plan, Section 1 e

e l

l

- 5.-

lono Term 11 Elimination of AFW Syste Dependency on Al a Complete Loss of Alternating Current Po ternating Curren At least one AFW system pump and its ass wer (GL-3) instrumentation should automatically initiociated flow path g

of being operated indepe'ndently of any AC ate AFW system b-*

Conversion of DC power to AC power is power source for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> s

acceptable.

12..Non-Safety Grade, Non-Redundant AFW

'EI

{ l!

The licensee should upgrade the AFW sSysten Automa E

and circuits to meet safety-grade requireystem automatic in men ts.

E*h Additional Lono-Term Reconnendatio 4

ns f

Interaction of AFW with Integrated Cont Feedwater Line Break Detection and Mitigatirol System (IC d,

on Systens.

13.

The licensee should separate the ICS f I

reduce the interaction of the AFWS with Strom AFW initiatio Detection and Mitigation Systems eam and Feedwater Line Break of the AFWS due to interactions with these tThe pot NUREG-0667 (Reference 11).

wo systems is discussed in 2.1 of Reference 11 call for (a) theSpecifically, recomm n Table control from the ICS, and (b) the reductiseparation of the AFW i

steam and feedwater ifne break detection on in adve AFWS.

The license should implement those reand mitigation sy connenda tions.

I l

6-

?lant-Specific Recomendations 14.

Diversity in the Motive Power for the AFWS Pumps We are concerned with the dependency of both AFWS pumps in yo j

s gn on steam from the main steam lines. Other FWRs are known to have a similar configuration (e.g., Calvert Cliffs); however, because of the mor t

dry-out of the steam system in B&W plants, such a steam depen s of more concern in Davis-Bess'e.'Weste your plans for installing a third AFWS train which will utilize a pump powered from a source other t steam.

A schedule of implementation should be provided.

p; (lI f

15.

_ Postulated Hich Energy Line Break I

Yque design does not appear to meet the high energy line break crit

[

erion in SRP 10.4/9 and BTP 10-1; i.e., the AFW systen should maintai ty 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 single active failure. Evaluate your design in line with these requirements ee e

J 8

e a

e O

m

. 16.

Design Basis for AFW System Flow Requirement The license'e is required to provide the AFWS fl s

required in Enclosure 2 for the Davis-Besse 1 dow design baiss accident conditions.

esign basis transients and The response should include the following:

(1)

List all events needing AFW to mitigate the con

\\

sequences.

(2)

Justification that the bounding non-LOCA calc l ti ua on will serve as a conservative basis for sizing the AFW systen for non LO ii considerations.

A core cooling In other words, show that the calculation will b all of the non;LOCA events requiring AFW

i. I ound Ei 1

(3) The non-LOCA analysis should include a loss FSAR type asstanptions to maximize heat remo of feedwa "j

E.r decay heat. 2% power level measurement unce t ival re The calculation should not take credit for "a tir a nty, RCP i

^I.

since it will not occur under all conditions cipatory reactor trip" n

is not precluded; however, credit for Lifting of the PORY valve should not be assumed. pressure relief through the (4)

For a small LOCA events, reference may be

" Evaluation of Transient Behavior and Small Rmade to the in the 177 Fuel Assembly P1 ants. dated May 7 eactor Coolant System Breaks

, 1979.

The acceptance criteria for the event will b e:

a.

Reactor Coolant S'ystem pressure remains le (2750 psig),

ss than 110% of design pressure r

b.

No fuel failure (DNBR greater 'than 130) i

Basis for Auxiliary Feedwater System Flow Recuirements

. As a result of recent staff revices of cperatin; plant A water Syster.s (AF45), the staff concludes that th uxiliary Feed-criteria provided by licensees for establishing AP45e desii flow to the steam generator (s) to assure ad requirements for decay heat are not well defined or docume t dequate removal

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ne.

\\

We require that you provide the following AT nation as applicable to the design basis transie td5 flowi ditions for your p1dt.

n s and accident con-I

-a 1.

a.

t [ i.

Identify the plant transient and accident co di i 6 0 in establishing An's flow requirements, including th n

t ons considered L

. x f.

wou:

e following v=

1) f Loss of Main Feed (LMN) 2)

LMFW w/ loss of offsite AC power 3) h LMFV w/ loss of onsita and offsite AC power

)'

4).Plantcooldown

~

5) Turbine trip with and without bypass 6)

Main staam isolation valve closure 7)

Main feed line break

8) Main staam line break
9) SmallbreakbCA i

10)

Other transient or accident conditions not li stad abeve b.

Describe the plant protaction acceptance critaria and corres-ponding technical. bases used for each initiati fied above.

ng" event identi-The acceptanca critaria abould addruss pl lisits such as:

ant I

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2

- Maxima RC3 pressure (PORY or safety valve Fuel tamperature or damage Jimits (DNB central ta@erature) *

, PCT. maxima fuel RCS cooling rata Itait to avoid excessive cool Minimum stasm generator level to assure suffici ant shrinkage ent stas:

generator heat transfer surface to reseve decay heat I

cool down the primary systsN.

and/cr i

2.

Describe the analyses and assur9tions and corp '

justification used with plant condition considssponding techni including:

ered in 1.a. above f4

a. Maxima reactar power (including instnment err or allowance) at the tfee of the initiating transient or accident E.-

y b.

Tiair delay from initiating event to reactor tHp e

E n

c. plant parameter (s) which initistas Fd5 flow and ti i

between initiating event and introduction of AF43 me delay flow into stasm generator (s)'.

nian steam senerator,,,,,,

occurs.

when inftfating event Initial steam generator water inventory ahT de l ti e.

pe on fata before and after Fds flow commencas - identify reactor deca rata used.

y heat l

e e

I

4 3-f.

Maximum pressure at which steam is ret sased fra and against which the AN pump aust develop s ff s u

icient head g.

Mir. fan nabar of steam generators that m e.g.1 out of 27, 2 out of 47 ust receive AFV flow; h.

RC flow condition - continued operation

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f.

circulation.

of RC pumps or natural

1. Maxima AN inlet tauperaturg.

g o

y J. Following a postulated stama or feed line b L

ji n

kf, assumed to isolata break and direct AN flow treak. tim

. e(

generator (s). AN, pop flow capacity allowanc o intact staan the time delay and'asintain m'inima staan a to accamodata E

. ;_g 1

Also identify credit'taken for prf'aa'ry systgenerator un

'JU due to blowdown.

an heat removal 3 ;;

, k.

Yolme and maxima temperature of water i u'

between staan generator (s) and ANS conn n main feed t

ection to main feed line.

1. Operating condition of stess generato t

initiating event.

r normal blowdown following m.

Primary and secondary system water and used for cooldown and AN flow sfzing. antal sensible heat Tfas at het standby and time to cooldown RC3 a.

in temperature to size AN water source in to RHR system cut-ventory.

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-m y,,

-4.

3.

Ver.

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ify that the AN purgs in

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flow to the stasm generete ( )your plant will supply i

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rs necessary above considering a single failuas,detamined by it 1

the pump f3cw to allow for pIdentify the margin in sizin re.

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and pump wear.

ump recirculation flow. seal l g

eakage

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A he N..

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NUREG-0737 j;

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f.,

s C,ar...ncation of f

TMI Action Plan Requirements

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Manuscript Completed: November 1980 Date Published: November 1980 Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l'

/

i.

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d y'-N-r-3P 8

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l I

I ABSTRACT This document, NUREG-0737, is a letter from D. G. Eisenhut, Director of the Division of Licensing, NRR, to licensees of operating power reactors and applicants for operating licenses forwarding post-TMI requirements which have been approved for implementation. Following the accident at Three Mile Island Unit 2, the NRC staff developed the Action Plan, NUREG-0660, to provide a comprehensive and integrated plan to improve safety at power reactors.

Specific items from NUREG-0660 have been approved by the Commission for implementation at reactors.

In this NRC report, these specific items comprise a single document which includes additional information about schedules, applicability, method of implementation review, submittal dates, and clarification of technical positions.

It should be noted that the total set of THI-related actions have been collected in NUREG-0660, but only those items that the Commission has approved for implementation to date are included in this document, NUREG-0737.

iii

f, 8 884g 6

o, UNITED STATES

?-

NUCLEAR REGULATORY COMMISSION

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i wasHWGTON, D. C. 20086

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OCT 3 I se TO ALL LICENSEES OF OPERATING PLANTS AND APPLICANTS FOR OPERATING LICENSES AND HOLDERS OF CONSTRUCTION PERMITS Gentlemen:

SUBJECT:

POST-TMI REQUIREMENTS On September 5,1980, the NRC staff sent you a draft clarification letter regarding approved TMI Action Plan items. During the week of September 22, 1980,. four regional meetings were held to provide a more detailed explanation of these requirements and to obtain industry coments concerning these items.

Based on these discussions and other coments received, the NRC has revised its requirements regarding these items. It is the purpose of this letter to set forth those requirements.

This letter incorporates in one document, all TMI-related items approved for implementation by the Comission at this time. This document is beina published as r!UREG-0737.

Enclosures 1 and 2 contain an itemized listing of OR and OL requirements including implementation schedules, applicability, method of implementation review and licensee submittal dates. Enclosure 3 contains more detailed clarifications of most of the NRC positions including the identifi-cation of any changes from previous requirements and guidance.

Most of the items in the attached document have already been issued as requirements by previous correspondence. Those items that are being issued as requirements for the first time by this letter are identified by an asterisk in Enclosures 1 and 2.

Additional guidance on the Emergency Response Facilities,Section III.A.I.2, will be fonvarded separately in the near future.

Licensees and applicants should note that the set of requirements identified in the enclosures do not constitute the total set of TMI-related actions in the TMI-2 Action Plan, NUREG-0660. Rather, as noted above, the enclosures are a compilation of those items that have been specifically approved by the Comission for implementation. Upon further staff development of criteria and planning, additional items will be issued. For example, in the relatively near future, the staff expects to issue further criteria on emergency operational facilities (NUREG-0696), auxiliary feedwater system improvements (derived from NUREG-0667), and instrumentation (Regulatory Guide 1.97, Revision 2).

In general, the implementation of those requirements will be carefully examined to ensure that they do not unnecessarily impact any of the requirements in this letter, g$rd

F i

2-The requirerrents herein (which include the requirements from NUREG-0694) are applicable to applicants for operating licenses and such applicants are expected to meet the same schedule of implementation as indicated for operating reactors.

Operating license reviews being finalized over the next few months will be handled on a case-by-case basis. Any item for which the implementation date is prior to the expected date of issuance of an orerating license will be considered to be a prerequisite to obtaining that license. For such items, applicants must submit information or documentation four months prior to the staff's scheduled issuance of its Safety Evaluation Report or four months prior to the listed implementation date. whichever is later.

A large number of post-TMI requirements require the installation of a number of control room indications.

It is important that licensees and applicants give consideration to human factor engineering considerations in planning for the installation of such new control room equipment.

In the coming months, the NRC will be requiring human factors engineering reviews of control room designs as part of Action Plan Item I.D.1, and such an effort 4

at this time may reduce the potential for later modifications. As an example of possible considerations, licensees and applicants might well consider at this time whether some control panel indications are of lesser safety significance and can be moved to other locations in the control room.

It is expected that the requirements contained herein will be met. However, it is recognized that licensees have proceeded with implementation of some of these items prior to issuance of these clarifying criteria. The staff will consider requests for relief from various aspects of these criteria.

Such requests should explain the need for relief, include a clear description of design features of the proposed installation, and provide a safety rationale supporting the adequacy of the proposed installation. A licensee or applicant seeking relief from any element of our criteria should submit a request for relief, along with supporting justification, in response to this letter, Accordingly, pursuant to 550.54(f) operating reactor licensees are requested to furnish, within forty-five (45) days of this letter, confirmation that the implementation dates indicated in Enclosure 1 will be met. For any date that cannot be met, furnish a proposed revised date, justification for the delay, and any planned compensating safety actions during the interim. After our evaluation of your response the NRC staff will take action, as necessary to assure that such reqirements and commitments are appropriately enforceable.

This may include, as needed, issuance of a Confirmatory or Show-Cause Order.

Sincerely, t

M Nrer G

Eisen u Cirector Divisiono{' Licensing Office of Nuclear Reactor Regulation

Enclosures:

As stated viii o

M

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