ML20064L528
| ML20064L528 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/01/1982 |
| From: | SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML20064L530 | List: |
| References | |
| CON-NRC-03-82-096, CON-NRC-3-82-96, TASK-1.A.2.1, TASK-2.B.4, TASK-TM SAI-186-029-25, SAI-186-29-25, NUDOCS 8207080445 | |
| Download: ML20064L528 (15) | |
Text
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- SAI-186-029-25 TECHNICAL EVALUATION REPORT IMPROVEMENTS IN TRAINING AND REQUALIFICATION PROGRAMS AS REQUIRED BY
' TMI ACTION ITEMS I.A.2.1 AND II.B.4 for the Fort Calhoun 5tation .
(Docket 50-285)
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July 1, 1982
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Prepared By:
Science Applications, Inc. -
(
l 1710 Goodridge Drive i McLean, Virginia 22102 Prepared for:
U.S. Nuclear Regulatory Cornission ashington, D.C. 20555 ~
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' t'l0 Contract NRC-03-82-096
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TABLE OF CONTENTS Section .- Pace I. INTRODUCTION. . . . . . . . . . . . . . . . . . . . . 1 II. SCOPE AND CONTENT OF THE EVALUATION . . . . . . . . . 1 A. I.A.2.1: I. mediate Upgrading of 'R0 and SR0 Training and Qualifications ..... 1 B. II.B.4: Training for Mitigating Core Damage. . 6.
III. LICENSEE SU3MITTALS . . . . . . . . . . . . . . .~ . . 7 IV. EVALUATION. ..................... 8 A. I.A.2.1: Immediate Upgrading of R0 and SRO Training and Qualifications ..... 9 g . . , , ._ .
B. II.B.4: Training for Mitigating Core Damage. . 11 .
7 V.
CONCLUSIONS . . .i;..... . . . . . . . . . . . . . ... 12
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VI. REFERENCES. . . 9.'... . . . . . . . . . . . . . . . 13 l-D e
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3' I. INTRODUCTION Science Applications, Inc. (SAI), as technical assistance contractor to the U.S. Nuclear Regulatory Commission, has evaluated the response by Omaha Public Power District for the Fort Calhoun Station (Docket 50-255) to certain requirements contained in post-TMI Action Items I.A.2.1, Immediate Upgrading of Reactor Operator and Senior Reactor Operator Training and Qualification, and II.B.4, Training for Mitigating Core Damage. These requirements were set forth in NUREG-0660 (Reference 1) and were subse-quently clarified in NUREG-0737 (Reference 2).*
The purpose of the evaluation was to determine whether the the licensee's operator training and requalification programs requirements. The evaluation pertains to Technical Assi' gnment Control satisfy (TAC)
System numbers 44162 (NUREG-0737, I.A.2.1.4) and 44512 (NUREG-0737, 11.3.4.1). As deline ted below, the evaluation covers only some aspects of ites I'.A.2.1.4.
The detailed evaluation of the licensee's submittals is presented in Section IV;. the conclusions are in Section V.
II. SCOPE AND CONTENT OF THE EVALUATION A. I.A.2.1: Imediate Upgrading of R0 and SRO Training and Qualifications j , . c4he clarification of TMI Action Item I.A.2.1 in NUREG-0737 incor-
? porates a letter and f our. enclosures, dated March 28, 1980, f rom' Harold R.
- f. Denten, Director, Office of Nucle'ar Reactor Regulation, USNRC, to all power
} reactor applicants and license'es, concerning qualifi< ations of reactor
~i operators (hereaf ter. referred to as .Denton's letter).
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This . letter and '.
enclosures imposes a number ofc. training requirements on power . reactor y licensees. This evaluation specifically addressed a subset of the require- .
ments stated in Enclosure 1 of Tenton's letter, namely: I tem A.2.c, whi ch . ..
relates to operator training requirements; item A.2.e, which conc' erns .
instructor requalification; and Section C, which addresses operator requal'i .
fication. Some of these r'equirements are elaborated in Enclosures 2, 3, and .
4 of Denton's letter. The training requirements under evaluation r e sum - .
marized in Figure 1. The elaborations of these requirements in Enclosures 2, 3, and 4 of Denton's letter are shown respectively in Figures 2, 3, and -
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As noted in Figure 1,* Enclosures 2 and 3 indicate minimum require-ments concerning course content in their respective areas. In addition, the Operator Licensing Branch in NRC has taken the position (Reference 3) 1. hat -
- Enclosure 1 of HUREG-0737 and NRC's Technical Assistance Control System distinguish four sub-actions within I.A.2.1 and two sub-actions within II.B.4. These subdivisions are not carried forward to the actual presentation of the reguirements in Enclosure 3 of NUREG-0737. If they had been, the items of concern here would be contained in I.A.2.1.4 and II.B.4.1. .
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Figure 1. Training Requirements from TMI Action Item I.A.2.1*
Fro; ar Ele ent AE" 8ecaire-ents**
Inclesure 1. :ter. A.2.c(2 )
Training p*c;ra s stall be nocifice, as ne:essaey, te seevice trair.ing in test transf er. f1wie flo- a ne the-mocynante s. (Entlesu e 2 pro. ices g.iceltnes for tne minirum centent o' such training.)
02ttat:0N$ Enc 1cture 1. Ite- A.2.c(2) p t:1; .ut t Training cre;rars shall be nocifice, es ne:essary to pe:vice training in the
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wie of insta11ec slant syste s to control or esti;ste er accicer.t in nica tee ccre is se.erely ca agec. (Inc1:swre 3 pr:v4:es gw1:elines for the nir.t .wr centent cf swch trainin;.)
I In:les re 1. Ite- A.2.c. (3)
! Traintes ; c;ra s shall be ro:sfice, as necessary t: pro.ioe increasec e stasis l
on reacter anc plant transients.
i Enclesure 1. Itee A.I.e
- t.iite:102 Instructors shall be enrolled in a::ropriate re: alification
. orc;ra-s to assure
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- g<* ' 3., ; c".'. . es -
they are cegn:: ant of current crerating histo y. arc:1e-s anc ctanges te cro-
. .j tecures an a mir.tstratise limitations.
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- Incloswr e 1. Item C.)
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e ! Centent of tre licensee c:erator re:walif scati:n gr:;ra s stall te ecifice te Q j in'c1.ce instrwe tion ir. heat tra nsfer f1wie fle=, tr.e rner.<r aeics, a nc es tigs-a e tice =f accicerts.f r.olving a cegraced core. (Inc1cswres 2 anc ! service g.1ce.
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I lines fer the rinsnw' c:-tent of such trainteg. ).
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- The criteria for re: wiring a license: trcivie.a1 te cartics:ste in accele atec L* "LII C"..' ION ret.alificati:n stall be rocifice to be cc*ststent -sin :ne ne- :sssing grace for iss.ance of a licerse: 20*. e.erall an:' 7C; ea:h ca tescry.
Enciesure 1. Itee C.3 '
Frograms shculd be noitfice to re wire the certrcI rar.iowlatior.s listec in
[nc1:swre 4 her 41 centrol manipolations. Sucn as :lant er reacter startwes.
nust te perf orres, tentrol ranipulatices e.r.in; at'.:rral er e=ergency ope *a-ttees must be -alser tnrewgh with, anc esalwatec ty a rwe2er of the training staf f at a minines. An a;pecertate sirwlater dy be used to satisfy the retutre-ents for control ranipulations.
- !ne re: wire ents sh:=n are a sutset of these contained in !tre I. A.2.1.
- :e'e ences to Enc 1csu*es are to Denton's letter of March 28,1950. =hich is ccetainec in tne clarifi-cation of ! tee !.A.2.2 in hJEEG-0737.
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Figure 2. Enclosure 2 fron Denton's Letter Tsal=!=*, lh *T TiaWts, f t'J10 FLD. M? TaitCelC5
- 1. fasic fre;erties of Flutes sac Pattee.
This section should cover a basic intrcduction tc rette* and its prose.rties. This section shesis inc1 wee such concepts as tes;er ature seassrerents anc ef f ects, censity anc its ef fects, specific
.et;nt, b.csency, viscesity anc ether prese* ties of flates. A serking tric legge of steam tat:rs sh w1g also be inc1weet. Energy reverent shew 1C te ciscussed in:19 ting sucn f ue:amentals as heat eactange, spe: tite heat, latent heat of es; riaation an: sensible heat.
- 2. F16te Stattes.
Tais section shewld cc.er the pressw e e, tee;eratur e anc volwee effects on flutes. Case;1e cf these paessetric cnanges show1c te illustratei by the instructse an: relatec (alculattens.shesic te perforte:
ty the st. cents anc ciscussed in the traintag sessions. Causes an effects cf pressure anc tee:r' stare cnanges in the eartows coe;onents an: syste=s snosic te discusse: in the training sessions. Casses an:
ef f ects of presswee and tem;eratur e changes in the variews c:r;enents and syste-s show1c de ciscusseo as applicante to the f acility with particular emphasis en saf ety significant featsres. The characteristics of force and pressure. pressure in 11 cutes at rest principles of hycraulics, saturation press 6ce and ter.;erstwre and succooling shesig also be incisced.
- 3. F1wie Dyr.a-ics.
This sectien shew 1c cover the flow of fluids and such conce;ts as $ernow111's principle, ent gy in l -s,ing f1stes, f) . reasure theory sad devices anc ;ressure lesses cae to f riction an: c'sficing.
I Cthe ceaserts anc teres to te discussee in this se:tton are h75* carry coer. Carry 6-cer, kinette j e.e gy. heas.icts relattershi;s enc t=c ;*.ase f1c= f ur.canentals. practical 4;;1icati:-5 relatin; te the rea:ter costant systes ano steas geaerat:rs show1c also te incl. cec.
l 4 heat Tra-sfer tv Csnes: tion. Cea.ection anc ta31stion.
Tnis section showie cover the is,ncamentals of heat transf er ty cinewctions. This section showls ing1wce discussions on such conce;ts and teres as specific heat, heat f1ws and atomic action. Heat transf er . characteristics of fuel rocs and heat enchange's show1c te inclucec in this section.
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. f,, Tats"section shew 1d cover .the f wr. arentals of 1 est transf ee hy ,c:atection. Asteral any forcec cirewla.
i tion snosic tie cis:wstec as a ;1tcatie to the varioss syste s at the f acility. The coa.e: tion escree* heat 7
patterns createt by ensantin; flutes in a confinet area s%:wic te inc19:ec in tnis secticn. a t
treas; ort and flulo fica Te:scticns c* st::: age thewic te cistgssec g.e to steam and/cr n:n::n=e st:1e I
- d gas ferr:aticn caring n:rmal sec acticent cen:1tions.
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This section snos1d cceer the f uniawentals of heat transf ee by the*r al r egiaticr. In ine d ser cf vact a*.t The electroe s;netic er.ergy crittee t>y a tecy as a result cf its teeperature s*cwl: be i' energy. Cce;arisnns sn: wig, ne e.a:e Cis vstec and illustratec ty tne use of et.atices an: sa*;1e cal:wlatt =s.
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cf a 21act bccy 4: sorter an: a er.ste 3:27 eritter,
- 5. D ame cf r* ase - ! citing.
1 This section thesid inc1 wee cestri;tions of the state cf ratter, tsetr innerent characte*istics en 1
the*-o:fr.a'ic prcpectses such as enthal;y aac entropy. Calewisticns shesic be perf orrec in civing steam saality ans voic fractica pre:erties. The types of ceilir'; sho.1: te ciscussec as a;;1ica:Te to i
i the f acility during normal e clutions anc accident con:iticas.
- 6. t.rnest anc flo. !*statt11ty.
f r.is section should ccver descriptions and mechanisms f or calculating such terms as critical flua, critical pc.er, N retto and het channel factors. This section s*cs1C also include irst w:st:ns f oe preventing and conttoeing f or clas or f,e1 gamage and fic. testat:11 ties. Saeple calcw14ttons s*owle in te illustrated by the instructor and calculations snowle be pe*fceret by the stweents anc c?scusse tne training sessions. methocs anc processres f or using tne plant c:t;.ter to deteerine c.antitatt.e values of vertows f actors cu r ing plant c;eration and plant heat talance seterminatiers snesic also te l co.erec in this section.
- 7. Kee t
- Heat Transf er ti-its.
This section shew 1C incluse a discussion of heat trartf er limits by esarining fuei roc and reseter cesign anc 11ritations. The tasis for the lirits newic be cc.erec in inis sectica steng = tin This section sneste tevee recommenced methocs to enssee tnat limits are net aspece:nes or esceesec.
ciscussions of pessing f actors, regial anc asial pc.er distritwtions anc changes of these f act:rs c.,e to the influence cf ciner variables such as rocerator terce atwee, senen and control roc position.
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Figure 3. Enclosure 3 from Denton's Letter 4 TI Aln:h3 CE!!!a: A f 03 p.111&Alin; C:At ;A.. ;t A. Inec e teste m tatten
- 1. Use of fiset or eo atle incore detecters to dete*mine estee.1 cf ccre ca a;e se* gt: ret *y charges.
- 2. Use of tSe-e: cow;1es in determining stah te se etwees; etnoes. f or ::s ten:ee car ge re actn;s; a retases f or cirect restings at teerinal ju :tters.
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- 3. Metho s f or calling v: (p*in*ing) incere cata f r oe t .e ple*: co-:.ter,
- s. tacere hwclear lestra e tation (*15)
- 1. t!se cf his f or eeter-tr.atien cf .cte f ermation: .etc locattor. :ssis fe h!5 resroese as a functiori of core te ceratares anc ce-stty cha"5'S.
C. Vital teste n etation e
i 1. Instr ee .tatto. res:: se ir en a::Sce .: eaitree ent; f ail.re te:.ence ; tire :: f ails c. retac: of I f ail.,re), se: .c atica relistility (a:t.a1 vs incicate: le.el).
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- 2. Alte rative r<tho:s f or resswe tr.g fic-s. ;resswees. levels, anc te :e atu es.
- a. Ceteertr.ation of presswrtner level tf all level transmitte*s fail.
- b. *eteceir.ation er letec n flo. =1t% a cic;;ec filter (Ic% fle.l.
- p c. eterrination of other seactor Ccclant Syste= sa ameters if the ;-tr ary et .oc cf meess esent e
- '.- 4 . has feites, g
] :. spet ary cs e-.stry
- 1. ! ace:tec cae-istry results sth severe core tama;e; cc set.e c es cf trarsf errin; s all :.a tities .
. cf Itc.te 4.t ssce cce.t a tment; secor.ance of using less tigst .yste-s.
- 2. . troecte( isete:1: 1.* ease:.n f o- core ca a;e; f or cla: ca a;e.
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- 3. Ccrresten eff erts of estenceC imeesion in peiracy ate *; time tc f ailure.
- i. 3 :iat :cn *:r tt erin-
- 1. Resto*se of Process and Area M:nitors to seve't ca 4;es: Le.avior cf cetectces .'en sat. rates; j r.etnc: f Cr Cetecting rac14 tion rea ings ty direct reaso: cata* 41 ce t ec t or c.t:wt (c.e e' aaje:
cetector); es;e:tec acc. racy of detector- at C1f f erent Ir atices; use of otte:tces tc ce* e-Etne l
- eatent of c re 24*a;L .
- 2. Petnoes of Cele **ining case rett inside containeer1 f rce reasseew e*.ts ta s en o.tsice centainte".t.
F. Oat Generation
- 1. hethces of M2 ge*eratio- crieg ar. acticent; othe* so.'ces of gas (ae. te); tec .r.ic.es f: et tin; or discesa) cf nce.conceastbles.
- 2. M2 flamacility and espicstre lie.it; sos res e of C2 in contain.wnt or Reactor ::clant Systee.
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Figare 4. Control Manipulations Listed in Enclosure 4. , ,
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al. Plant or re ac '.or staetups to include a re ;c that reactietty 'eetta-t fpm owclese heat a:citton is noticeable anc neatup rate is esta211s'ec. /
- 2. Plant shutio=n. l T *l f- s
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+3, Maass) ccr. trol of stese gercratc'ssand/or feeg.ater daring tiart,. a n f s % * *.wr.. g 4
4 Boration and er 451stiry dsrta po.cr eperation. 'f
*5. Any significant (greater than 10%) pp tr chaages in e.anwal
- roc cor. rci or re,ctresistica f irm.
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- 6. Any reactor po.er change of 10: or greater wheet lose change is' pe-f orme '41th ice: 11ett central!-
or .nere flut, tee;eratwre, or speed contreDs on e.an.a1 (for M1:.R). r .f-s
*T. L:ss of cc:lant inc1weing: -
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- 1. sigatficant PJ steam gentr ator leaks
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- 2. inside arc ostside primary contairvnent - i, 1 4
- 3. large anc s'rall, inc1witng less-rate cetemiration 4 sa seate: Eea: tor Coelant ret;cese (F4). g
- 8. Less of instrwnent air (if sim. lated plant specific). ,
i .'l r .,. s,{ 9. ; Less of electrical pc.er (and/org degraced po.er sou ces). ,
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- Lets of ure cociant fic jhaural tirculation.,
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%+? . 11. 't.:ss c( coacerser: :yer.. (
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tess of service water $f re wired f or saf ety.
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<; 13. . : less of snstd:..s cooling. *d - . .
- tess of c:r p nent cooling system tretccling tc an indiviesal :.r ,enent. . l ),
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- 11. . tds of m: mal f eef.ater or nemal f eed. ster systet. f a11wre.
*16. '. css of all feet.ater (nor al and e .crgency).
Les s of pr:tective syste n cna'n nel, f hlb ,
- 17. f]
I f, . mshsttioaec centrol rod or f s (or rod crers), fF
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- 13. Tea ility to drive centrol roes.
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- 20. Cencttions retwirtrg use of en.e gency torstdr. o* stanity hcde ccetect system.
- 2. Fweb clac ing f ailate or high actidly ir, reactor coolant o- off;#s.
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- 22. broine or generator trw
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- 23. 7;alf wnc* ion of autona.it ecatr..el systee(s) **ilen af f ect reactivity. ,, _ ,/, f
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i' y < 24 Malf un ction of reactor coolant pressareNolor.e Ccetrt1,syste :. , \. j li, teactor trip.
- 26. Pain steam line treak (inside or ostside contai m t).
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j 27.3 huclear instr.ance.tation f allwrt(s).
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* $tereed item to Le pe+f orM annwally, all others tiir.ntally.
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b s the training .in mitigating core damage and related subjeyts should consist of at least 80 contact hours
- in both the initial training and the requali-
, fica, tion programs. The ' NRC considers thermodynamics, fluid flow and heat transfer to be'related subjects, so the 80-hour requirement applies to the
'ccmbined subject areas of Enclosures 2 and 3. The 80 contact hour criterion f is not intended to be applied rigidly; rather, its purpose i I , greater assurance of adequate course content when the licenss to provide ee's training j courses are not described in detail. -
Since thd licensees generaily have their own unique course out-s lines, adequacy of response to these requirements necessarily depends only on whether it is at a level of detail comparable to that specified in the enclosures (and consistent with the 80 contact hour requirement) and whether it can reasonably be concluded from the licensee's descri'ption of his train-ing material that the items in the enclosures are covered. The Institute of Nuclear Power Operatiens (INPO) has developed its own guidelines for training in the subject arear. of Enclosures 2 and 3. These guidelines, given in References 4 and 5, were developed in response to the same requirements and are more than adequate, i.e., training programs based specifically on the complete INP0 documents are expected to satisfy all the requirements pertaining to training material which are addressed in this evaluation. The licensee's response concerning increased emphasis on tran-sients is considered by SAI to be acceptable if it makes explicit reference , to increase! emphasis cn . transients and gives some . indication of,.the nature .,- of the increase, orf if .it . addresses both normal.;and abnormal transients 7 u (without ne'cessarily indicating an increase in emphasi.s.) :and the. -requalifi- . cation ' program satisfies 'the requirements.for control manipulations, Enclo- -
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sure 1, ' Item'C.3. .The.latter requirement calls for all the manipulations .c .~ " listed in Inclosure -4 (Figure-4.in this report) to be perf ormed, :at .the f reque,ncy indicated, unless they are specifically not applicable to 'the .- licc?sefs type of reactor (s). Some of these manipulations may be performed I on a sidulator. Personnel with senior licenses may be credited with these activities if they direct or evaluate control manipulations as they are performed by others, f.lthough these manipulaticns are acceptable for meet-ing the reactivity control manipulations required by Appendix A paragraph . 3.'a of 10 CFR 55, the requirements of Enclosure 4 are more demanding. . inclosute 4 requires about 32 specific manipulations over a two-year. cycle while 10 CFR 55 Appendix A requires only 10 manipulations over.a two-year cyc'e. B. II.B.4: Training for Mitigating Core Damage Item II.B.4 in NUREG-0737 requires that "shif t technical advisors and operating personnel from the plant manager through the operations chain to the licensed operators" receive training on the use of installed systems to control or mitigate accidents in which the core is severely damaged. l
'A contact hour is a one-hour period in which the course instructor is present or available for instructing or assisting students; lectures, i seminars, discussions, problem-solving sessions, and examinations are considered contact periods. This definition is taken from Reference 4.
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- )I q fnclosure 3 of Denton's letter provides guidance on the content of this training. " Plant Manager" is here taken to mean the highest ranking manager at the plant site.
For licensed personnel, this training would be redundant in that it is also required, by I.A.2.1, in the operator requalification program. Hewever, II.B.4 applies also to operations personnel who are not licensed [ and are not candidates for licenses. This may include one or more of the highest levels of management at the plant. These non-licensed personnel are not explicitly required to have training in heat transf er, fluid flow and thermodynamics and are therefore not obligated for the full 80 contact hours of training in mitigating core damage and related subjects. Some non-operating personnel, notably managers and technicians in instrumentation and control, health physics and chemistry departments, are supposed to receive those portions of the training which are commensurate with their responsibilities. Since this imposes no additional demands on the program itself, we do not address it in this evaluation. It would be appropriate for resident inspectors to verify that non-operating personnel receive the proper training. - The required implementation dates for all items have passed.
- w. Hence, this evaluation did not address the dates of implementation.
3 Moreover, the evaluation does not cover training program modifications that 3 might hFle..been_made for other reasons subsequent to the response to ... 2 p Denton's letter b e' "- 3 ._
$ ' C.::
III. LiCENSEESUEMITTALSJ
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g:.; .. s . The licensee (Omaha Public Power District) has submitted to NRC a number of items (letters and various attachments) which explain their train-ing and r qualification programs. These submittals, made in response to Denton't letter, form the information base for this evaluation.. For the Fort Calhoun Station, there were 6 submittals with attachments, for a total of_11 items, which are listed below.
- 1. Letter from W.C. Jones, Division Manager, Production Operations, Omaha Public Power District, t o .P .F . Collins, Chief of Operator Licensing Branch, NRC. July 15,1980.(1 pg, with enclosure: items 2, 3, 4, 5, 6).(Transmittal).
- 2. "C. Department Training". No title, undated. (4 pp, attached to item 1).
- 3. " Training Program for Licensing Senior / Reactor Operator Candidates". Undated. (5 pp, attached to item 1). . .
- 4. "Simulat.or Training Program for Senior / Reactor ;
Operator Candidates", (Course Syllabus).
, Undated. (2 pp, attached to item 1).
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- 5. " Lecture Series Duration". Undated. (1 pg,
) attached to item 1).
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- 6. " Licensed Operator Requalification Program" Undated. (11 pp, attached to item 1).
- 7. Letter from W.C. Jones, Division Manager, Production Operations, Omaha Public, Power District, to D.G. Eisenhut, Director, Division of Licensing, NRC. October 6, 1980. (2 pp, with attachment). (re: Status report on implementation schedule of the training program for mitigating core damage).
- 8. Letter from W.C. Jones, Division Manager, Production Operations, Omaha Public Power District, to D.G. Eisenhut, Director, Division of 1980. (1 pg, with Licensing, NRC. December 31, attachments). (re: Responses to numerous tasks identified in NUREG-0737, " Clarification of TMI Action Plan Requirements", in particular, Item II.B.4).
- 9. Letter from W.C. Jones, Division Manager,
. . Production Operations, Omaha Public Power .
" ' .~~ ~' District, to D.G fisenhut, Director, Division of .
. h' L'i c e n s i n g, NR C. ".Se p t e mb er 14, 198,1. .(0, pg). NRC
.' .. Acc.:No: 8109220234~.:: ( re:- Request for ' deadline extension for the cqtr.aining program to mitigate .
..], core damage). ' :_
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- from W . C. I ..J o~n e s , Division: Manager, ..
Production Operations, Omaha Public Power , . District, to R.A. Clark, Chief of Operating Reactors B anch #3, NRC. December 28,1981.(1 pg, with attachment). (re: Summary Status of Task Action Plan Near-Term Requirements). i 11. Letter from W.C. Jones, D i v i s i o n .M a n a g e r , i District, to R. A. Clark, Chief of. Operating Reactors Branch No. 3, Division of Licensing, NRC. l May 5, 1982. (1 pg, with enclosure). (re: l Response to NRC's RAI dated March 31, 1982). NRC Acc No: 8205100270. IV. EVALUATION SAI's evaluation of the training programs at Omaha Public Power l District's Fort Calhoun Station is presented below. Section A addresses TMI Action Item I.A.2.1 and presents the assessment organized in the manner of Figure 1. Section B addresses TMI Action Item II.B.4.
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4 I J 2.1: Immediate Upgrading of Reactor Operator and Senior ; I Reactor Operator Training and Qualifications. fnclosure 1, Item A.2.c(1) The basic requirements are that the training programs given to ! i f actor operator and senior reactor operator candidates cover the subjects of heat transfer, fluid flow and thermodynamics at the level of detail specified in Enclosure 2 of Denton's letter. In July of 1930, the licensee provided the NRC with submittal item 3, a description of their training program for licensing reactor operator and senior reactor operator candidates. This program d,escription provides for lectures on heat transfer, fluid flow and thermodynamics. No details were provided with this submittal on the content of these lectures. In submittal item 11, the licensee stated that they had reviewed the content of the lectures on these three subjects and that they believed the lectures sufficiently addressed the topics detailed in Enclosure 2. It appears the NRC requirements for this training are met at Fort Calhoun. Because no cutlines were'provided, an NRC auditor wishing to verify the content of the program should audit against the guideline of Denton's Enclosure 2. Enclosure 1, Item A.2.c(2) The requirements are that the training programs for reactor and. senior reactor operator candidates cover the . subject of accident mitigation h at the leTel -of detail specified -in Enclosure 3 of Denton's letter-(see .M-1 Figure . 3 o'f. 't'his report); :t: -
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E . .. '. 3 The. licerisee submittal. of. July 1980 (submittal . item 3) described .. 4 the reactor trainingeprogram which calls for a training ' lecture ;on the use . :.. 4 of installed plant systems t to control or mitigate accidents involving severe '
-t core damage. No additional detail's. on the specifics of the course ceritent .-
were provided with this July 1980 submittal. In a later submittal (. item 11), the licensee stated that lectures addressed the topic of Denton's Enclosure 3. This meets the NRC requirements for accident mitigation training according to the guidelines of Enclosure 3. Omaha Public Power District also responded to. NRC's question concerning the number of contact hours involved in training program elements which address accident mitigation, heat transfer, fluid. flow, and thermodynamics. In submittal item 11 they stated that more than 80 contact hours were involved in these subject areas. Details supporting this claim w'ere also provided and are presented below. Instruction Area Contact Hours NUS tapes on accident mitigation 10 hours Raactor and plant transients 50 hours Emergency procedures Teview 15 hours Recognition and mitigation of accidents 15 hours provided by Combustion Engineering - Simulator transients and accidents ~ 40 hours Total 130 hours
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On this basis, SAI judges that Fort Calhoun meets this NRC criterion. Enclosure 1, Item A.2.c(3) The requircment is that there be an increased emphasis in the training program on dealing with reactor transients. In the licensee's submittal item 11, it was' stated that the training effort relative to dealing with transients was increased when the program was revised in July of 1980. The program currently addresses both normal and abnormal plant conditions. Furthermore, the program as detailed in submittal item 4 involves both classroom lectures and simulator training. This area of the Fort Calhoun training program exceeds th'e NRC requirements. Enclosure 1, Item A.2.e The requirement is that instructors for reactor operator trainirig programs be enrolled in appropriate requalification programs to assure they are cognizant of current operating history, problems and changes to procedures and administrative limitations. Submittal item C is the training program for licensed opertors and in Section 1.1.1 the program states that (1) The training coordinator will review all completed plant modifications, additions and Plant Review Committee niinutes for significant items (current operating history, -
. problems,;;p.r. ocedure changes ,and administrative limits) applicable to training, . and (2) wilb. disseminate this information to training instructors. , ..
- From this informationiit -is reasonable to conclude that this program meets N -
the NRC requirement?f,or: instructor requalification.. , c l - l Er.clasure '1,7 tem C.lg,, ,, _? . , .
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The primary requirement is that the requalification programs have l instruction in the areas of heat transfer, fluid flow, thermodynamics and i accident mitigation. The level of detail required in the requalification l program is that of Enclosures 2 and 3 of Denton's letter. In addition, i these instructions must involve. an adequate number of contact hours. i In their submittal item 11, the licensee stated that their requalification program covered the subjects of heat transfer, fluid flow, l thermodynamics and accident mitigation at a level of detail compatible with i E.~1osures 2 and 3 of Denton's letter. While no further details on the l specifics of course content were provided by the licensee, information was l provided on the title of lectures and their number of contract hours which i are part of the requalification program. These details are the same as l those described in the training program analyses (item I.A.2.c(2)) with a l total of 130 contact hours being covered. l ! On this basis it is reasonable to conclude that the Fort Calhoun. l requalification program meets the NRC requirements relative to course content and number of contact hours. W
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p 4' Enclosure 1, Item C.2 ; 1 The requirement for licensed operators to participate in the accelerated requalification program i..ust be based on passing scores of S0% overall, 70% in each category. Submittal item 6 is a description of the operator requalification program for Fort Calhoun. The-evaluation procedores for the program are defined in Section I and they state that an individual receiving a score of less than 80% overall or 80% on particular category is required to undergo an accelerated requalification program. This particular aspect of the Fort Calhoun requalification program exceeds NRC requirements. Enclosure 1, Item C.3 TMI Action Item I.A.2.1 calls for the licensed operator requalifi-cation program to include performance of control manipulations involving both normal and abnormal situations. The specific manipulaticns required and their letterp(erformance see Figure 4frequency are identified in Enclosure 4 of the Denton of this report). - The description of the Fort Calhoun requalification program, S'ec-tion H, lists control manipulations which are to be part of the program. The program description states that the first priority is for performance at the plant followed by performance on the vendor simulator. Only one control }'. manipulation, " Loss of Instrument Air" is not handled by means of actual plant opePition # or . simulator operation.. 4t is not handled by the simulator because the-Combustion Engineering simulator does not . provide this function. V It is insteid handled:by classroom.. lecture. The manipulation frequencies of (,; .. performance are in'. compliance With the NRC requirements of Enclosure 4. " .'. Based on this', SAI., judges - that . Fort,;.Calhoun meets the control manipulation , requirements., - '6 ply .) v.~ . , . . . , . B. II.B.4 Training for Mitigating Core Damage , Item II.B.4 requires that training for mitigating core damage, as indicated in Enclosure 3 of Denton's letter, be given to shif t technical advisors and operating personnel from the plant manager to the licensed operators. This includes both licensed and non-licensed personnel - The TMI Action Item II.B.4 requirement of training licensed personnel is met by implementing the training discussed under Enclosure 1 Item A.2.c(2) or under Enclosure 1 Item C.1 for Action Item I.A.2.1. The requirement for. training non-licensed operating personnel and shift technical advisors has also been evaluated. In submittal item 11 the licensee provided information on the plant persor.nel being trained in the area of accident mitigation. Specifically the people trained are: plant manager, shif t technical advisors, operations supervisor, shift supervisor . and licensed operators. The licensee also described other activities'such as procedure review, meeting with the vendor and participation on the Plant Acceptance Committee which increase their exposure to accident mitigation concerns. SAI judges that Fort Calhoun meets the NRC requirement of training non-licensed operating personnel and shif t technical advisors in the area of accident mitigation.
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'V. CONCLUSIONS SAI has evaluated the programs at Omaha Public Power District's prt Calhoun Station relative to the requirements of TMI Action Items 1.A.2.1 and II.B.4. The evaluation focused on the establishment and content of the programs relative to the NRC requirements.
The training and requalification programs at Fort Calhoun Station meet all the requirements of TMI Action Item I.A.2.1. I The training programs for both licensed and non-licensed personnel l are in compliance with the requirements of TMI Action Item II.S.4 at Fort Calhoun Station. . l
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' , .T V. REFEREtiCES f i
- 1. "NRC Action Plan Developed as a Result of the TMI-2 Accident." NUREG-0560, United States Nuclear Regulatory Commission. May 1980.
- 2. " Clarification of TMI Action Plan Requirements," NUREG-0737, United States Nuclear Regulatory Comission. November 1980
- 3. The NRC requirement for 80 contact hours is an Operator Licensing Sranch technical position. It was included with the. acceptance criteria provided by NRC to SAI for use in the present evaluation. See letter, Harley Silver, Technical Assistance Program Management Group, Division of 1.icensing, USNRC to Bryce Johnson, P; ogram Manager, Science Applications, Inc.,
Subject:
Contract No. NRC-03-82-095, Final Work
, Assignment 2, December 23, 1981.
- 4. " Guidelines for Heat Transfer, Fluid Flow and Thermodynamics I nstruction," STG-02, The Institute of Nuclear Power Operations.
December ~12, 1980. S. " Guidelines for Trair.ing to Re ognize and Mitigate the Consequences of Core Damage," STG-01, The Ir.titute of Nuclear Power Operations. January 15, 1981.
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