ML20024B462

Forwards DC Holt 800320 Memo to Gd Qwale Which Contains Summary of Changes That Occurred in Licensing Area Resulting from post-TMI Activities During Past Yr

ML20024B462
Person / Time
Site:	Crane
Issue date:	03/25/1980
From:	Taylor J BABCOCK & WILCOX CO.
To:	BABCOCK & WILCOX CO.
References
TASK-06, TASK-6, TASK-GB GPU-0261, GPU-261, NUDOCS 8307080785
Download: ML20024B462 (6)

Text

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,, " THE SABC0CX & WILCOX COMPANY

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h,' I Distribution From J. H. Taylor, Manager, Licensing (2317) sos. s.s Cast.

File No. int CF

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oate Tectu.ical Pmduct Eva!uation Project Marrt 25, MO

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Ofstribution K. E. !.htte R.E. Qam.a E. A. Womack D. H. Rey.

R. M. 3all E. J Cor414sti A. L Lazar J. McEariaM E. G Ward R. 8. Bomuns D. A. Crowther N. S. Em m y i

Aef: Letter: D. C. Holtto G. D. L'n.le, same subject da tetr "r ch 20,19CJ.

l Attached for ywr infomation is a cegy of the ref.srerced letter. I bellt e 1

l S it contains an excellent sunntry of changes that ant occuring in the

}j h t.icensins area.u u result of the past yeart pst.THI activtties.

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Ma.@ 20, 1980

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Tech.icM Prodact'Er$1uation Project we3

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(Tait it A %ni ta Paper) l

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Purst. ant to ca TPEP Plan, attached is a *vhite paper

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.Cso provided f:r ths secce aad results fcr Task IIA, f.lcansfr:g Requiremn's.TPEP 4

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pavi.St the primary input to 'At whita ;2per.

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Ili CONFIDENTIAL

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a TECHNICAL PRODUCT EVALUATION PROJECT ii lk Licensino Recuirements Survey l'-

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1.0 INT 31000CTION

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-v In supparc of 9.e Technical Pmeect Evalt:atioe Prcject. a survey of nucisar '

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regJ14 tory sourtes was performeti to identif requiresents whica have been

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prossalmtad as a rest.lt of the Faren 13 19{,9 accident at Three il11e Islano, y

1 Items reviewed for t!:Is effort incluc'ad 4

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J Congresstenal Heering Transcripts Press Relessas and Lettan

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4 A'RS Meetiq Trse.scetpts, Letters, vit Repcrts 1RC Meetf rg Iransc.rlpts and Cctrependence 1

NURE3-0560, " staff Aaport on the Canaric Ass 2ssamet of Feedwatar

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Transients in Pressurized Water Reacton Designed by the Babecck

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& Wilcox Company" L

NUREG-0578. "TMI-2 Lessons teamed Task Force Status Report and Shcrt-Tsrm Recoamendations" NUREC-0585, "TM1-2 Larsons Leamed ?e.sk forte Final Report" NUREG-0660 (Craft); " Action Plans for ta:plementing Reco.ncendatices of the Presidant's Cocaissica and Other 5*.udies of TMI-Z Accident-e O*;

"The Report of the President's Cosmissien on the Accident at Three a

Mile !.sland" and CJ2 mission staff.4epc *ts

• Thrae Mile Island: A Report to the Cosm'ssicners and to the Pubitc' UtiiIty Coasttuents to NRC 0

2.0 8-205 FRODUCT CHANGES t.

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2.1 Soecific Oeston Chances I

5pecific design changes identified since TM! which are considered to be s.an-t

, datory for future B-205 plants include:

o Reactor coolant system high point vents Of rect, safety-grada indication of pressurizar relief and safet) i valve position Pressurizer relief and safety valvos qualified for steam and water

• flow conditions Safety-grade pressurizer heatan. relief and safety valves. and ins tmanentation 9

Safety-grade pressurizar relief valve block valve with automatic p

closare on icw RCS pressure 1 '!

Safety-grade pressurizer heater cut-off on low pressurizer leve" 8h Inadequate core cooling

• Instrumentation - e.g., reactor coolant jI' water level, fuel assecly exi t temperature, reactor coolant 1

I saturation conditions, natural circulation flow if

- Reactor trip on total loss of pain feedwater i :

Reactor coolant pucp trip on coincident Ica.1CS pressure and RCS l

y) voiding

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Several changes have also been identified which have a high probability of being required for future plants:

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0-Increased pressurizer reifef and/or safety valve capacity P-Dual pressurizer reif ar valve block valves

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.. Reactor trip on loss of heat sink (Iow steam generator level)

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s :. ilardened, leak-tight decay heat mmoval system i

,l i.' !!igh pressure decay ' heat removal system i !

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Contintous monitoring and recording of critical plant parameter *.

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fGP post-event reconstruction - e.g.. upgraded. (rstalled Reactiratey r

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Elininstica cf pesssurizer surge fina leap sean l

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%.?. faneral Oesier and Analytical Charm, h

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In addit.on ts %e above. a rurber of iters have been toentified as a nsu!t i

I of ih! which han m inpact c1 future plant dast ;n and analysis bur. whi.a haver only been r.precsso es gcals, cencral criteria, areas for further can-l sidsration, ate. These inclJdu l

Improved integrated systems apprcach with.egard to operational design.

- Cossnharstre consideration of systems interacticn - e.g.

failure modas and effects analysis for equipmer.t evgnts and a spectrum of operator actions.

Additional analysis if ancmalous trrnsitnu e:d a':Idents with IM.rtased I

<A a*tentice te gt,neral safegaarts (s*Aer tian sp+cific scenatics. Prt.ba-

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bilistic risk cssessment tectnigt.e.s s.cul.t be uniforinly applied.

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multiple failures censidered, and 'various cpe.ratar actions (mittgatir:q j

nd tggrasatiag) esaluated.

Daluation and Ider.ctflution of the basis for automatic.versus manual i

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- Re-evaluation of the design criteria for equiptent currently considered 3

non-safety-related but which may t,e involved in transient or 4::fdent 8

tof tf ation, mitigation, mcovery or ro.11toring.

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Pmvision of additional status socitoriq for equipment important ts g

safety.

Consideration of hi l

mental conditions, gher levels of radioactivity, more adverse environ-longer term operation and recovery / decontamination

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in the design of equipment and systems which may function after an acciden t.

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Technical Specifications which corprehensively address equipment and systems which may be involved in transient or accident initiatlan,.

I mitigation, recovery or monitoring.

The above could be suonarized as the goals of reducing challenges to pisnt j

safety and maximizing the mitigation capatility when such challenges oc:ur.

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The most important general design and analytical change, however, is that increased attention must be directed at the human element in plant design and 1'

opera tion. This " man-machine interface" has been identified as the " primary

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deficiency 11 reactor safety technology ty the NRC (NUREG-0535) in its eval-i 3,

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• u4 tion of TMI. Similarly the Xemeny Corrrifssion has stated in its report that i,

the sos serious "mindset" relative to tar accident was "the preoccupation of i

ji everyone with the safety of equipment, resulting in the down-playing of the 1aportance cf the human element in nuclear power generation.*

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2.3 BW scecific concerns F

There.,continuas to exist in the minds of various individuals involved.in

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nuciear regulation concerns as to the safety of the su NSS product spect-fically. SG plants are viewed as being unusually susceptible to secondary y; 1 system transients; having a significantly shorter t5e period available for 6

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corrective act!ca aftar a feeowater transient; relyino en an extremely cc..' plex

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centrol systut (ICS) whicn has a Icw reliability; anc being pqtentially trade-p[i g

quata relative te t'atural cir31stion core coolirq. hever, except for grass

-l swanfiguratior> of the 353 wch tlat it is no Icnser "different* fmm othe.

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pressurised watsr rta: tor sptm designs. no chenges in addition t these d

given in 2.1 and 2.2 ab6ve hau been identified. 7.e saneral criteria for 3

3 the future licensing ot' the S*Al ;reduct would inc%de, therefore the technical d

and political defent,e of the N13. The sn-going cocawnication of cowrehensive,

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hign-ouality, tarpeed infornation to a variety of diffutnt indivii*uah end I

organizaticos will be r.qu. red.

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3.0 INTGFACE SCCPE MD IGVit! Ul/hGC5 f

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Sectica 2.0 preceeding wts presented in the context of the 8-105 product. The contents of 2.2 and 2.3 howaer should adcitionally be evaluated relative to h*

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.huges it: %f-80P interface criterf t revisions to the curren; NSS secpe of b

sup".ly, a.nst stadifications to separably s:encrac*ad services. Otter regulator /

I mnsiderations whir.'s have been i&.tified sitca Tat which sh4vM also be eval-f nated in this centsxt are:

4 Assurad.emer-ency pswcr for 1.resn.ri.ur acetsrs, val ses and instrrents l

tion.

01versecca*.atnuent isolation.

Two. in.hnndent, full-coaci ty, s4 fety-grad.: emrgency feed ater traf as.

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4 with automatic actuaciore en loss of leadwater and flow control to opti-i mizs ' ansient mitigation t

Igrored secondary system reliability and cperation. particularly main

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i feedwater Additional secondary system statas mont toring I

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Expanded (range and function) post-accident monitoring capability j

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Igre.ved post-accident sampling capability a

Improved collection, evaluation ard feedback of operating experience, j

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g including ongoing NSS vendor suppcrt Additional off-site technical support capability and preparations -

t e.g., data link with NSS vendor and NSS vendor response team capability.

Improved control room design-e.g.. identification and physical concen-j j

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tration of a minimum set of critical plant parameters frdt: which to l

identify and mitigate off-normal conditions j,*l Upgraded operating and erergency procedures - e. g., provide for designer /

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analyst review, improve clarity, and address symptoms and actions in

~I context of information provided to operator.

7 Upgraded operator training - e.g., additional-understanding of principles 8

. I of reactor safety, increased sirulator capability and use, designer 1

k scenarios. and improved documentation.

involvement. e@hasis on generalized safeguards rather than specific j

NRC accreditation of operator training facilities, prograes, fastructors,

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,A 4.0 f.ICENSING PROCESS CiWIGES 8

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Oue to a lack of management control or overall policy direction. Identiri-cation and assessmene of a currect or future regulator / status is speculative.

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l Certain changes in attitude are evident, however, in tr.e licensing process l

as a nsult of TNI. These include:

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' t Accidents art to be ccasidered credible, up to and in:1uding events j

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involving significs:t cart damage.

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• !afety-related' will be consf ormd to be much more eaccupass ng g

than previously, and a formal regulatory definition swy be prec.elt-i lt.

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i Acy release of racioactive maverici, actual or acst. accident. wil h

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i te viewed negativa1y tnd may r.ot ta psrv.itted as a dasign basis.

I lj The dep h and detail of NP.C review wit,1 incrtasc and r.kare crcss-I disciplican som:urmnf.e will be enuired - e.g., NRC/CNRR review r

of plant pmeerister cza be expectec.

Exped!ticus resolucics of serwric safety issues will te eghesized with a es.;ran% urate decavase in 2nf incastry flexibilit/ relathe j

to cosplistet.e.

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);l Plant cperation will recsive adgiltiotal acte?tica; aa.s it can be expected that there w!11 be in ecanstoa in reporting crit.eria, 1

acre requests for inf3n.atir t EM an fr. create 16 directions fur

l remedial actions.

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Dc4 of' the'abcve has the sost prMble ef fect or caking the if censing process

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stat be densene and t*.e adverstry ntiationship betwees industry ard the regu-I

• j la tors e:t-a pronounced. hre ts s general recqqaition however that *.he f

fuadamental deficiency in the tictnsing precass is tr+e lack of an *prticulata j[ j aid videly rzticed r.aticaa? safecy pcifcy with whic't t? bind together the i'

g' narrm and highly technical licer: sing requirerr.eits." T'.erettsra, whfie a t

9t.anta:n increase in current regulatory requiruents af t) be unc:unteredt if a, c.efinitive statement of tha Jafety otjactive of nucicar power can be realized a certain degree of future stability in t.% licensing process is t

possible.

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