ML20246E701

A Compilation of Reports of the Adivsory Committee on Reactor SAFEGUARDS.1988 Annual

ML20246E701
Person / Time
Issue date:	04/30/1989
From:	Advisory Committee on Reactor Safeguards
To:
References
NUREG-1125, NUREG-1125-V10, NUDOCS 8905110339
Download: ML20246E701 (159)

Text

{{#Wiki_filter:' .. f( _

• , _ [, '

,.L,, _ _ ' l '_ 4- .,-} j. 4,l ' ' 'f,'. ].,' '.',,I ', '. l _? f.*, j'._ { '.~ ;' ;' ' [ '- i?,a. f' '.s 12', '. }. '. '.. !. f p.- :Y'~ ,. ',. '. i 4,q _l .,.l - '.' "., k, '; 'j.. ' ;.. ,fV' . t'. ; ]' .8 M., s,. '., ' ..(I ,./,' f' . ' _. '...l <f l. _. l ' *,0.

f.,%ng,l

':,I. ' l l :.' '.: V '. _ I*','[ T ' I'+ 'k ( Q, \\ 'i,,: l. l.) ' ' - ;,, 6,-' -n. _.'; g., _ l, s,1 (, 's...'l '.,'. {.,; _ c,; '..,, jc..'- , - )- 'Q a,n .?,.,.',,'n,. - ,t t ' s - s \\. c _ S 'e __ y, * ;, d.. c r/ ,...,e ..,s. q, 6,,,i . 3 i, ',,.,.,, _%..'[ :+ 2'8 9, . !~ ,__.,,f . ;.( .. ; >. '. t :,,.. iL...',._ 9', 's.* 7_ 9e: ~_,,y i ; ' * '... ' '.,,, '.,,* =,,\\.: o,' e.' ?. '; y. s '- _ j - ',... ,.. '. f y y (f,.. .g. 9, ' . a.:,..._ ;4 d ',,,'.,i. );,. J->, -.- W,)* a. -,... _,,.,',1.. ,o.,-,. '~..,.,. .. ' _.. '. ',-,'.4 c ? .'l. l ,L4 a,' ."*i,V- ',"- i,, (. ',.-, 'c .>-e v .'g L ,e ,,,,.-;*,,i-',., < ,,i i. > i J..,', ' 'p4* 'O ,9 4 m)' - f f,, '. ','-,..,,' 9,. 33 '..,.i'. 'p' g .i ? f.'h',','.

,',s'.

.\\. y., t,}.'g '.',k.,p, [ ', ,L ' 4 ' (l?',> ,, ; 4 f, (.', v. w.,

'{; -)

.( a' - !I*, ', : - :.,. a 4 ; .v, 9: w ,',( ' - :<~,... ;,,;;. _s

g e -

f ' ,,y ,a ( p,' _,. #r [ e. ,S ,3 e... '. I q '. 3,- 2, .' :. '. ;,, e. .,9 ' l' f n.,. i - i e o, , ' r. l,Or -, p,.. e

• 4 '. 1. -.-

_.;..,.r.-' t _,.- w.;' g - ' ';.,,3 ,. ll., ', l ,',, 4.; ;,. - + - i,. .. s. t ..,c F g ga ' )' ' .. ( ' ' - . [.

I..

[( - ';, i' j.. I+ Wdsn.d...,0 E.,.,.., \\ "

1. JK

• ".,%W,f T Oi 'p.',..-. L,.

.. s.~ ,g,.,,.7 % ,.r. 1 '. ! m , - e '...'y i

i',.,. ','

.f,f,,,j,... e, ,e 1 ...e-. 3E-. a', ' :,e '.,,, -g Y f a o f, ,!',f. ( ,> ' '., z- ,A, 7r '. ', ' A i,( L' f.'.. ,%*R. pe 4 5 ,,. d. - [' ^h pi -,. l 7 . 3 .'r.,L,l i ,,i,.; , :,b,,

J' L-

[L 'Q } ; _ .s

43 c

,;. ;,~ s.. m ' j,' ' ^ [,' ,f ',, , '[ *, E;'; E -,ery 7.g, g,g; ,.[-_ ' ' g, 'l[' y ' i. l ..,'4:: ~ JW@[W8ML.nlReports4thMGM j[}( ggy [ } 44 (. J K P % %, % m M. tt,ee % %. ~ 8-(, e' e . ;$Pxf%i4MMgualmlW7MQ .,...- i s,- . a .g;,s. ...g,,..,. n 3 ~ w. 3.. 3.. . 1,..,, t ' , i +. ,'b g\\g, l g 4 t,' l' o,' f:'. .,.'. '. ',. -..' ) S 084%%W;t][$J Ti! r. w mi A?H -

• ACOMmissichD J

! JrMS. ' / r+ - o,. . ; q. .<>; I .:t' ,t,, .,, ef..' [~_ e ;.,'. 7, g_ 4 ...g .q,,', -.- i, ,L,. (. . (. d D' i-i

f..l ; !. ;....

...l ,, !4...,'. (' - ;L'i ' [q (, . 'c? Sf'. ' t .' '.'..'t.' ' E,. e 'l ,. '.31 j ; l. ',y. ;. i. ;, I: ' l p,.,  ; P n, ' f. , ' i f. i }.', ' '. i, j, f,,. ! l' ;,., fC, l A f,'! c". ' .'.f i 'c.$- . '.'.li '; ; i. ' i.. ....,,..(,.,,,...- ,L, ';,,J. ' ! - - ;... v_, ', ' S',g ,y a., .\\, - ; :.,,- .. i

i.,7 '. f ;.. c. ;

,(;, ..s .o 3 .:)-l. { ;' ,..t 'J 9..'... p,,.....[ e,.'y , f*..',',,';;','1,"/,',* ;, *

l. b.

. r.I......'.. I : d,..' ' - ' qg,. ,7 ,

• j..

..,t.,

• [, %

.I.. * li MF A, ' '1.:,, 8. fi.,., p J: : 4. G j,. 'l, ;. - ' A l ', _,',',;,;;1 A ,J.. 3 q' ..i y. f b ),:' ' ; - ( '.;, ' ;,4 l. '.

'4_,., t G' s ~

s p ;... J ,. l . :..,; ;;,,.y',. { ' ' ' 'l'b,' \\_ '.y} '.),. } ',9 ' - - V', .,s y .,, l.., y ., l{ t i, f.,.'};:, l,3;. g; i y ;;./ ;,;l] _ f.; -' J. L, 4j, ':;._'.,.zu y l;s:e,, : ".l y.y y:p k: , y; 3 ' /;...., ; q g. _. , 7;e l' 1 3 ' ^ s> :: :;r. ;:,;+:m m.m,'y _ ; g p g;_k.w g;y 9 ~.: :..e r v. : y ( .y ,,?l[ ; p.../ y g y ::, ;,>.;; \\i : v .n. ,f M,[:?N:: p'::. f ?)4l;m;s,w y. y ,:.u,

lM U...Q.n A ~ %, D.. ; O.,..c,';, )

o.. l::' G j 0 ;..'

. y.;.j'L,.:.. e-J:. <M,* c, N:. ;~./. :&e, @ml I....(;.-.;,.'s n,. :,,: !.

• -~uf-

. '.. "... - - 3 li. w:. +G.. n,.. ' <g.;? \\\\ mv .,..; -, :,J...,.7.. :n.. ' '. :.a.,3; ' 4 'o g g

o. ', h'. :.,a s..

.s. 9-

;; -. :. i v..c. :'.. 7n. ql u.-

w ; /..:L ', \\.,7,..u,n.p. f;.a e,i .> f,., a.;. - 9, .. ' :; c, q 3 y ', ! 5' ', l h j' $../i..I,,.. ',$.. I. ;.,,.f.

t..a qa, e

e-j t -; .~..s 1 . v.m ~ :. i '. s '. s,.A $y ,<( .<..;_ n s.' \\. ; ):. ll

a ":

n..': - V, 1, v',,; ' i. t,. ', l.?1. o.r *y .l }.. \\:e: a... l, .I .; l :., F ; ':,'.. '.c,6 t. -... e:

: : ',q;i - : t.':n ; Q.-

....,;..,,:',,,,.: ec r o

n.. <..

q p:

a.,,.."i t ' '. i. 's,.',Y b N.,,, ' > l.. '.,.',,..f.i;; ':\\ [.. i, p f.i [; L, ', ( ^ :. ! [",[n !. l M: 3.,.I,h...., ' ' c. r,,.,,, T, E U,,i. :' ' c i,(7

• 8,k., v,lf;.[, s r.. h - n' D '.,,,;,. : '..' o':

D S, I I , >.,d.a.,.e" '. .Y. 1 s-v. .,f.. i.....q.. t

c. >

,,K.y k z ' 'c j.,.. b -. y.v ,l ;. - v.a rc o .,,,. lG u'J..: {,. ' ':.... .."'t,-.. [,', :.a. ,~y ,d:- &,; ;;,...:, g ;. '.

e. o...'-

? .,.,"i; !. r. z _.', l l,p. e s '\\...'. ,. _.,2. l 1 ;.... y ', ur : - l 9 }... o' '., '!; N, "-l ;' ',i 4 ;, ',9y. ' ' : j s,; _' _.,,J._, _( } o ~,} ,. YD. ' 3PW ;L.,,J.,"hHW: AN.. i .. F.., *,:,, F...Nm.,O ",..m.:;. - ?,..,.% i 2 A,;,,P 2 0- a 9 f. q:i- '...'c -.q.-~,. g' \\.. l. to 3 y [I :.!,. s.,.,4. ';L.1...i; ; M,.f.' * ^ "'. ','. . <.. V

h. '.

.3. n,- ...,n . '..,y'm., .{.,.. j '..., f... e t ;,;.[

• A..,,. j

.., [ f [*,:[_ : j '. 3 .a. ', -. g-f '.; a' _ .g V : a, ; ;^ ' 4',,).. ' _ Q..'. '.', -.. '.p . i ,.\\ .- ;. )[. .. )_c , l1, _( e W

r '. s.

.. *< -L.,, :. ( * - r,; n..'a..... -,.1 .. :.. p%; j,.. :; ;. v,9,. f ~.c, ;. :. : ., c n. 7, ... o. c <. o $.. o. o-l.,..' + LJ g s, '.. - .c x; .. L u,. ', Ma #.. x ',.4.a... U. W 'r :T,s ~. V..,... Y ': q.':

i

cs a.- -..

9..

7,

. :, e

- g ' 3 l *..,,;.,n.,z',

vf,. - p 'l

r c' pc,

' ',. l' .'3. f ~' J :.. ,,\\'- p. ;;, '. j,' n y o : c :. v. l[.. j;,.,., i} ,.} { gi ['. s.1-'

$i. p e

6., Jf 'l .. d. ['., 4,' .... u..M ' WW ' ".. -,; '. "'.?,: i _. ',2. 'y. i - .t tes.-. ; ' ' ',' ' i-j' ' '. '. 2,

• 'f..

) ., 'i' l ',#.8', },,, I ' 4 i h,., );. '., Y y ; p ;: i,-,...' -.,.',... a 4 L ::L i' ;; ) '...v~ s, : : '.: .+ O l i' ~7, j ' ' ' ,' i g. " / - ' ,f, y%._ c,']. N.t. ? '.c, $" < :f. ' ; ' " 'wi. . - ~ l, ' _,.[ '.... c'.h ' i 'o. i :~: r . f9T'--s,.,,f3,.

• O',l

' '",j '5. '.k '.;,. q.,., e,. i, s .)5, ,v. ; .,.;y,.u a \\..;,c,..., %q 8 ,4. [ ' - lf, J;} [ } J, f.. ; j. }.,..; _.; Q J y j,,_ f G 3,; 3 4. Q. f f, p f. y l g~... y. -(. hQ;;,. f.. .,_,, 3, ; . ['[': f; [ c../.. y;_ ,g'._ ,'..,4,', ,'*y,,s. ,,s 4

j. '

D' l ' i.. ) =., s .,,.y. '.',, -, 'l '[. 4,.,

t.

'

' ' ~ ;...., s.* ~ 4 pm < _ m o. .. ~ .., < mm. h_.a....,,...~- <. a.c,..., v o.. ~, ' 4. ~. s _s _,.cv';. -,.., - -_,.. - .,, m, -

2. o.

.._.., : y: i - c.. - -

m. e....-.,.s i ; c.

, 4.. 2, ).. [, ; *c. .'...'I, ..iis '..... ' , 'l,'l _ *i, , '

1. I', ' '... ' '.

, ',((,'. 'l -,i., -,, ' ,.y'..'.'x,., k; *, i *. '. e . c,0 [ ',. '[ *[. ', ',' ' j. '.it.... S,' - j( I ., c. --9 g,,

o I AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources: 1. The NRC Public Document Room, 2120 L Street, NW, Lower Level, Washington, DC 20555 2. The Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082 3. The National Technical information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive. Referenced documents available for inspection and copying for a fee from the NRC Public l Document Room include NRC correspondence and internal NRC memoranda: NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investi-i gation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence. The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceed-ings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regula-tions in the Code of Federal Regulations, and Nuclear Regulatory Commission issuances. 7 Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission, j Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries. Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the i publication cited. Single copies of NRC draft reports are available free, to the extent of supply, upon written I request to the Office of Information Resources Management, Distribution Section, U.S. Nuclear Regulatory Commission Washington, DC 20555. I Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and j are available there for reference use by the public. Codes and standards are usually copy-righted and may be purchased from the originating organization or, if they are American I National Standards, from the American National Standards Institute,1430 Broadway, 1 New York, NY 10018. I 1 4

L

f....

n""ai;2s gy,;, A Compilation of Repods of The Advisory Committee on Reactor Safeguards 1988 Annual U.S. Nuclear Regulatory Commission -w April 1989

.c ABSTRACT This compilation contains 47 ACRS reports submitted to the Commission or to the Executive Director for Operations during calendar year 1988. It also includes a report to the Congress on the NRC Safety Research Program. All reports have been made available to the public through the NRC Public Document Room and the U.S. Library of Congress. The reports are divided into two groups: Part 3:. ACRS Reports on Project Reviews, and Part 2: ACRS Reports on Generic Subjects. Part 1 contains ACRS reports alphabetized by project name and within project name by chronological order. Part 2 categorizes the reports by the most appropriate generic subject area and within subject area by chronological order. j L j iii k 1

L PREFACE The enclosed reports represent the recommendations and coments of the U.S. Nuclear Regulatory Comission Advisory Comittee on Reactor Safeguards during calendar year 1988. This publication, Volume 10, 'is an annual supplement to NUREG-1125. Previous issues of NUREG-1125 are as follows: Volume Inclusive Dates 1 through 6 September 1957 through December 1984 7 Calendar Year 1985 8 Calendar Year 1986 9 Calendar Year 1987 v

r ACRS MEMBERSHIP (1988) CHAIRMAN: Dr. William Kerr University of Michigan l VICE CHAIRMAN: Dr. Forrest J. Remick The Pennsylvania State University MEMBERS: Mr. James C. Carroll, Retired Pacific Gas & Electric Company Mr. Jesse C. Ebersole, Retired Tennessee Valley Authority (Term Ended 3/88) I Dr. Harold W. Lewis University of California, Santa Barbara Dr. J. Carson Mark, Retired Los Alamos Scientific Laboratory '(Term Ended 2/88) Mr. Carlyle Michelson, Retired Tennessee Valley Authority and AE00/US Nuclear Regulatory Commission

• Dr. Dade W. Moeller Harvard University Mr. Glenn A. Reed, Retired-Point Beach Nuclear Power Plant Wisconsin Electric Power Company (Term Ended 1/88)

Dr. Paul G. Shewmon Ohio State University Dr. Chester P. Siess, Retired (Prof. Emeritus) University of Illinois

• Dr. Martin J. Steindler Argonne National Laboratory Mr. David A. Ward E.I. du Pont de Nemours & Company Savannah River Laboratory l

Mr. Charles J. Wylie, Retired Duke Power Company

• Membership transferred to the newly established Advisory Committee on Nuclear Waste (ACNW) in June 1988.

i vii

TABLE OF CONTENTS i l Page ABSTRACT.............................................................. iii PREFACE............................................................... y MEMBERSHIP............................................................ vii i i PART 1: ACRS REPORTS ON PROJECT REVIEWS Sequoyah Nuclear Plant.............................................. 1 Babcock & Wilcox Owners Group....................................... 3 Pilgrim Nuclear Power Station....................................... 7 PART 2: ACRS REPORTS ON GENERIC SUBJECTS Advanced Reactors Report on Key Licensing Issues Associated with DOE Sponsored Reactor Designs, July 20, 1988..................... 11 Preapplication Safety Evaluation Report for the Modular High Temperature Gas Cooled Reactor, October 13, 1988........................................ 21 Safety Evaluation Report for the " Power Reactor Inherently Safe Module" (PRISM) Design, November 22, 1988........................................ 29 Class 9 Accidents Report on the Integration Plan for Closure j of Severe Accident Issues (SECY-88-147), ) 35 July 20, 1988.................................................... iX

(s TABLEOFCONTENTS(CONT'D) Page { Decay Heat Removal Systems i See " Generic Issues /USI"........................................ 87,91 Emergency Core Cooling Systems See " Rules and Regulations"...................................... 147 NRC Research Related to Heat Transfer and i ' Fluid Transport in Nuclear Power Plants, { June 7, 1988..................................................... 39 l Comments on the NRC Staff's Dtaft Safety-l Evaluation of the Westinghouse Topical Report, WCAP-10924 " Westinghouse Large-Break LOCA Best-Estimate Methodolo July 20, 1988....................gy, 47 Emergency Planning Rulemaking on Emergency Planning and Preparedness, June 7, 1988....................................... -51 Fire Protection Fire Risk Scoping. Study, May 10, 1988............................ 53 Generic Issues / Unresolved Safety Issues Proposed Resolution of USI A-47, " Safety Implications of Control Systems" -- ACRS Comments, April 12, 1988........................................ 55 Effectiveness of Prograts Relating to Generic and Unresolved Safety Issues - ACRS Comments, April 12, 1988.................................... 59 Proposed Priority Rankings of Generic Issues: Fifth Group, June 7, 1988............................... 67 Proposed Resolution of USI A-17. " Systems Interactions in Nuclear Power Plants," August 16, 1988.................................................. 79 x

TABLE OF CONTENTS-(CONT'D) Page Generic Issues / Unresolved Safety Issues (cont'd) Proposed Resolution of USI A-47, " Safety Implications of Control Systems," August 16, 1988.................................................. 83 Proposed Resolution of Generic Issue 99, " Improved Reliability of RHR Capability j in PWRs," September 14, 1988..................................... 87 Proposed Resolution of Unresolved Safety Issue A-45, " Shutdown Decay Heat Removal Requirements," September 14,1988(Revised 2/28/89).............. 91 Human Factors Proposed Rule on Fitness for Duty Program -- ACRS Comments, April 12, 1988 (Revised 5/9/88)................... 97 Proposed Commission Policy Statement on the Professional Conduct of Nuclear Power Plant Operators (SECY-88-57), May 10, 1988............................. 101 Report on Proposed Revised Policy Statement on Nuclear Power Plant Staff Working Hours, 105 July 20, 1988.................................................... Metal Components ACRS Comments on Embrittlement of Structural 107 Steel, March 15, 1988............................................ 129 See " Regulatory Guides".......................................... Inservice Inspection of Boiling Water Reactor Pressure Vessels, June 7, 1988................................... 109 ] \\ l Xi

TABLEOFCONTENTS(CONT'D' i Page Operating Organizations i ACRS. Report on the Tennessee Valley Authority's Management Reorganization, February 19, 1988..................... 111 ' Updated Policy Statement on Training and Qualification of Nuclear Power Plant Personnel -- ACRS Comments, April 12, 1988....................... 115 NRC Staff's TVA Lessons Learned Effort, September 13, 1988............................................... 117 Power and Electrical Systems See " Generic Issues /USI"........................................ 55,83 Procedures - ACRS/ Regulatory / Legal ACRS Comments on the Need for Greater Coherence Among New Regulatory Policies, . March 15, 1988................................................... 119 ACRS Review of Shutdown Nuclear Power Plants, April 12, 1988........................................... 121 , Radiological Effects i l See " Rules and Regulations"...................................... 149 Reactor Operations ACRS Comments on SECY-87-314, Interim Policy Statement on Maintenance of Nuclear Power Plants, dated December 30, 1987 February 16, 1988................................................ 123 Proposed Rulemaking Related to Maintenance of Nuclear Power Plants, September 13, 1988...................... 125 l 1 i l xii

TABLE OF CONTENTS (CONT'D) Page Reactor Safety Development NRC Proposed Rule on Early Site Permits, Standard Design Certification, and Combined Licenses for Nuclear Power Reactors, June 7, 1988..................................................... 127 Regulatory Guides ACRS Comments on Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials" dated March 1988, March 15, 1988.................................................... 129 Revision 2 to Regulatory Guide 1.100, " Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants," May 10, 1988............................................ 131 ACRS Action on Proposed Regulatory Guide 1.9, Revision 3, " Selection, Design, Qualification, Testing, and Reliability of Diesel Generator Units as Onsite Electric Power Systems at Nuclear Power Plants," November 22, 1988......................... 133 ACRS Action on Proposed Regulatory Guide, Task No. EE-006-5, " Qualification of Safety-Related Lead Storage Batteries for Nuclear Power Plants," November 22, 1988................................. 135 Reliability and Probabilistic Analysis Program to Implement the Safety Goal Policy -- ACRS Comments, April 12, 1988.................................... 137 Report on NUREG-1150, " Reactor Risk Reference Document," August 16, 1988....................................... 143 153 See " Rules and Regulations"...................................... Equipment Qualification-Risk Scoping Study, December 20, 1988................................................ 145 1 1 xiii i E i

p h 1 i k I-J TA8LEOFCONTENTS(CONT'D). Page l-Rules and Regulations Proposed Revision of the ECCS Rule Contained-in 10 CFR 50.46 and Appendix K, May 10, 1988'..................... 147 Proposed Revisions of 10 CFR 20 " Standards for Protection Against. Radiation," June 7, 1988............................................................. 149 See " Emergency Planning"......................................... 51 See " Reactor Safety Development"................................. 127 ACRS Action on the Proposed Amendment of 10 CFR 50, Appendix J, " Primary Containment Leakage Testing for Water-cooled Power Reactors," October 11,.1988...................................... 153 Safeguards'and Security Licensing of All Chemical Isotope Enrichment, Inc. Facilities, October 13, 1988................................ 155 Safety Research ACRS Comments on Development of a Method to Establish Priorities for Research Activities, February 16,~1988................................................ 157 ACRS Reports to Congress on the Safety Research Program of the Nuclear Regulatory Commission, February 17, 1988................................................ 159 See " Waste Management"........................................... 167 Systematic Evaluation Proposed Generic Letter on Individual Plant Examinations and the Proposed Integrated Safety Assessment Program II, May 10, 1988....................... 163 xiv

l i TABLE OF CONTENTS (CONT'D) Page Waste Management ACRS Comments on Selected FY 1988 NRC Radioactive Waste Management Research Programs, February 17, 1988...................................... 167 ACRS Waste Management Subcommittee Report on Q-List Technical Position, April 12, 1988..................... 169 1 xv ) l J

l 1 Part 1: ACRS Reports on Project Reviews

/ %g UNITED STATES /" NUCLEAR REGULATORY COMMISSION j n 3 E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS e., wAsmNGTON, D. C. 20555 March 15, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS REPORT ON THE RESTART OF THE SEQUOYAH NUCLEAR PLANT During the 335th meeting of the Advisory Committee on Reactor Safeguards, March 10-12, 1988, we met with members of the NRC Staff and the Tennessee Valley Authority (TVA) Staff and reviewed the status of the resolution of the issues relating to the restart of the Sequoyah Nuclear Plant. This subject was also considered during our 328th, 331st, and 334th meetings on August 6-8, 1987; November 5-7, 1987; February 11-13, 1988, respectively. Our Subcommittee on TVA Organizational Issues met on November 5,1987 and Feb-ruary 2-3, 1988 to discuss this matter. We also had the t>enefit of dis-cussions with the NRC Staff and of the documents referenced. The ACRS has previously commented on this matter in letters dated August 12,1986 and February 19, 1988. During our review, we discussed the Sequoyah Nuclear Performance Plan and its review by the NRC Staff. Areas discussed included the plant management organization, plant modf fication and design control, design baseline and i verification program, design calculations program, fire protection, en-1 vironmental qualification of electric equipment, operational readiness program, quality assurance, Integrated Design Inspection, employee concerns program and several related issues, systems integration reviews, safety 1 review programs, diesel generator testing and capability, installation and testing of electric cables, an'd feedback of operating experience. We believe that problems and deficiencies associated with these areas are being ade-quately addressed. We are satisfied with the resolution proposed for the problems with the electric cables that are installed at the Sequoyah Nuclear Plant.

However, we recommend against the continued use of high-voltage testing of installed low-voltage cables.

Nondestructive test methods should be employed for in situ testing of cables. Some of the methods being evaluated in the NRC research programs on plant aging are available commercially and should be considered, l The testing and analysis by TVA of the Sequoyah diesel generators have demonstrated their adequacy to perform their safety functions. I s _ ___o

The Honorable Landn W. Zech, Jr. March 15, 1988 A number of fire protection issues have been identified and are being inves-tigated by the NRC Staff. Some of these appear to have generic implications and will be discussed as generic issues in our subcommittee activities i relating to fire protection. We will report on this at a later date. The Appendix R (10 CFR Part 50) issues relating to the Sequoyah Nuclear Plant startup sho'uld be resolved to the satisfaction of the NRC Staff. Significant improvements have been made in TVA's nuclear power organization. However, careful monitoring of the restart of Sequoyah Nuclear Plant and the other TVA nuclear plants is appropriate. We recomend that the Commission continue to maintain a separate office for the review of TVA nuclear power plants. We recommend also that a plant-specific PRA and ISAP be considered after the restart process has been completed. i We believe that the problems and deficiencies identified at the Sequoyah Nuclear Plant are being addressed adequately, and we see no reason to delay the program for restart. Sincerely, William Kerr Chairman References : 1. Letter dated March 9, 1988 from Wang Lau, Knoxville, Tennessee, to Charles J. Wylie, ACRS,

Subject:

Meeting of ACRS Subcommittee on TVA Organizational Issues on February 2, 1988 2. Letter dated February 17, 1988 from Stewart D. Ebneter, Director Office of Special Projects, to S. A. White, TVA, transmitting Revised Safety Evaluation on the TVA Sequnyah Nuclear Performance Plan, NUREG-1232, Vol. 2, Part 1 I i 2

'o UNITED STATES g / NUCLEAR REGULATORY COMMISSION g j ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wAsmNGTON, D. C. 20565 June 7, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

THE BABC0CK & WILCOX OWNERS GROUP SAFETY AND PERFORMANCE IMPROVEMENT PROGRAM During the 337th meeting of the Advisory Comittee on Reactor Safeguards, May 5-7, 1988, we met with members of the Babcock & Wilcox Owners Group (BWOG) and the NRC Staff to review the BWOG Safety and Performance Improvement Program (SPIP) and the NRC Staff's assessment of this program. This program was also discussed during the 338th meeting of the ACRS, June 2-4, 1988. Our Subcommittee on Babcock & Wilcox Reactor Plants met on May 3 and 4,1988 to discuss this matter. We also had the benefit of the documents referenced. The ACRS' previously comented to the Executive Director for Operations on this matter in a letter dated July 16, 1986. We generally agree with the NRC Staff findings and believe that the B&W SPIP recommendations will improve safety by reducing the frequency and complexity of transients. However, it is important to note that the SPIP and the Staff SER deal principally with trip frequency and transient complexity and not with ' design-basis accidents such as a small-or large-break LOCA or tran-sients as the result of external events. Therefore, one should be careful not to draw any new conclusions from this work concerning the overall safety or response of B&W plants in accident situations. On the basis of this reassessment, the NRC Staff has concluded that no addi-tional regulatory requirements will be needed. Recommendations suggested by the BWOG will be implemented voluntarily. We believe that BWOG should reconsider certain reccmendations made by the NRC Staff. For example, the Staff suggested particular attention to a reexamination of plants as required by IE Bulletin 79-27, " Loss of Non-Class IE Instrumentation and Control Power System Bus During Operation." More than two hundred recommendations resulted from the BWOG study. Although some of the utilities have acted promptly to schedule and implement these recommendations, others have not. The Staff has expressed concern about this. We share this concern. 3

The Honorable Lando W. Zech, Jr. June 7, 1988 A member of *he NRC Staff identified several concerns related to the BWOG reassessment prgram. In 'the Staff's SER each of these is identified and addressed. We reviewed this matter as you requested in Reference 5 and consider the Staff's handling of these concerns appropriate. There are in progress a number of additional NRC Staff activities resulting from this effort. These include a study dealing with the more detailed technical aspects of the Integrated Control System /Non-Nuclear Instrumen-tation (ICS/NNI) reassessment program and recommendations to develop programs to evaluate the balance-of-plant (B0P) systems at nuclear power plants. One objective of the latter effort is to determine if additional rules or regu-lations need to be developed in the BOP area. Where the failure or mal-operation of "non-safety systems" in the BOP can adversely affect the ac-hievement of safety-t elated functions, these systems should be designed, operated, and maintained in such a manner that such adverse effects are eliminated or made highly unlikely. Criteria to achieve this should be developed but with care not to extend further into the BOP than necessary to ensure safety. We wish to single out two additional BWOG programs for special encouragement. The first is the BWOG Valve Task Force which the Owners Group established on its own initiative to address problems with main steam safety volves, mo-tor-operated valves, and check valves. The BWOG should be encouraged to test the valves against a full spectrum of transient and accident conditions. The other is a program by the Owners Group that could result in developing an advanced control system based on modern digital technology for replacing the present ICS/NNI in B&W plants. We will continue to follow the progress of these programs. Sincerely, W. Kerr Chairman

References:

1. Letter dated July 22, 1987 from G. R. Skillman, BWOG, to D. M. Crutch-field, NRC, transmitting BWOG Safety and Performance Improvement Pro-gram, BAW-1919, Revision 5. July 1987. 2. NUREG-1231, SER Related to Babcock & Wilcox Owners Group Plant Reassess-ment Program, November 1987. 3. NUREG-1231, Supplement No. 1, March 1988. 4. Memcrandum dated August 14, 1987 from D. L. Basdekas, NRC, Office of Nuclear Regulatory Research, to W. Kerr, Chairman, ACRS, SubDet: Chainnan Zech's Memorandum to D. L. Basdekas, dated August 3,1987. 5. Memorandum dated August 3, 1987 from L. W. Zech, NRC, to D. L. Basdekas, Office of Nuclear Regulatory Research,

Subject:

Your Request of July 20, 1987. 4

The' Honorable Lando W. Zech, Jr. June 7, 1988 6. Letter dated August' 14, 1986 from H. R. Denton, NRC Director, Office of Nuclear Reactor Regulation for David A. Ward, Chairman, ACRS,

Subject:

ACRS Comments on the Babcock & Wilcox (B&W) Owners Group Safety and Perfonnance Lnprovement Program. 7. Letter dated January 24, 1986 from V. Stello, NRC, to H. Tucker, Chair-man,BWOG,

Subject:

Reassessment of B&W Reactors. I' i 5

p tro / o UNITED STATES

1. g 8"

NUCLEAR REGULATORY COMMISSION n {, ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4 WASHINGTON, D. C. 20555 September 14, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

PROPOSED RESTART OF THE PILGRIM NUCLEAR POWER STATION During the 341st meeting of the Advisory.Comittee on Reactor Safe-guards, September 8-10, 1968, we reviewed the proposed restart of the Pilgrim Nuclear Power Station. Members of the Comittee visited the plant on August 25, 1988, and a meeting of our subcommittee on Pilgrim Restart was held in Plymouth, Massachusetts, with represen-tatives of Boston Edison Company and the NRC staff on August 26, 1988. During our 341st meeting, we had further discussions with members of the NRC staff and with representatives of the Boston Edison Company. We received coments from representatives of the governments of the Commonwealth of Massachusetts and nearby loca-lities, as well as individuals. We also had the benefit of the documents referenced. By the time of the Pilgrim plant shutdown in April of 1986, the NRC staff had developed serious reservations about the management of the plant. We therefore gave special attention to this issue. More than half of the upper and middle level management personnel of the plant I have been replaced since the plant last operated. The new management l group is made up of people of demonstrated competence. A new vice i president with responsibility for Boston Edison's nuclear power program was employed about a year ago. He has assembled a team whose members have a variety of experience in naval and comercial nuclear l power plant operation. We are favorably impressed by their creden-tials and by the changes in the physical plant and the organizational approach to operation that has occurred since the new group has been in place. Although there is a preponderance of navy as contrasted with comercial nuclear power plant experience, we found no reason for concern in the backgrounds of the present team and in the ap-proach to management and to operation that has been inaugurated, i i 1 7 l

The Honorable Lando W. Zech, Jr. September 14, 1988 Those members of the Committee who toured the plant were favorably impressed with decontamination of the plant which has occurred, with the way in which the operational staff is organized, and with the way in which operation and maintenance were being performed. We also examined a number of systems that have been installed in an effort to improve the plant's and the operational staff's capability to avoid and to mitigate the consequences of severe accidents. During our visit to the training center, we observed the operation of the simulator that has been installed since plant shutdown. It is a modern and versatile system, and appears to have been effectively integrated into their training program. Its capabilities are being used to train the operational staff in the application of the newly installed systems mentioned above. The simulator can be expected to have a marked positive influence on the readiness of the operational staff to deal with both normal and off-normal events. Because of the history of weather-related loss of off-site power at this site, an assured source of emergency power or a demonstrated capability to provide emergency cooling in the absence of electric l power is of special importance for the Pilgrim plant. Since the shutdown, a number of features have been added that contribute to increused safety of the plant. Among d.ese are an additional emer-gency diesel generator and a system that permits water from the fire l i protection system to be made available to the decay heat removal I system. These added features should decrease the risk associated with the loss of electric power. However, we are not sure, and the staff and licensee are not certain, that the plant systems now satisfy the station blackout rule. We recommend that the staff give particular attention to this item as the rule is being implemented. We understand that use of the hardened vent, for relieving possible torus overpressure during a severe accident, will be reviewed by the NRC staff as a generic issue for all Mark I containments. We intend to review the matter in that context. A report, dated August 6, 1987, from the Federal Emergency Management Agency, states that the Emergency Plen which existed at the time of plant shutdown has a number of inadequacies. Information provided to us indicates that significant progress has been made in correcting i ) these deficiencies. We recomend that before startup is approved, a clearly defined pronram for eerly correction of these inadequacies be i available end be approved by the NRC staff. 8 l L__

l i I The Honorable Lando W. Zech, Jr. September 14, 1988 I i We believe that, subject to the comments above, restart of the Pilgrim plant will not lead to undue risk to the public health and safety. Sincerely, William Kerr Chairman

References:

1. NRC Confirmatory Action Letter to Boston Edison Company dated April 12, 1986 and Supplementary Confirmatory Action Letter dated August 1 27, 1986 1 2. NRC Memorandum dated July 27, 1988 transmitting Systematic Assess-ment of Licensee Performance (SALP) Board Report No. 50-293/87-99 3. Memorandum dated June 17, 1987 from W. Russell, Regional Adminis-trator, NRC, to R. Bird, Boston Edison Company, transmitting Systematic Assessment of Licensee Performance (SALP) Report No. 50-273/86-99 4. Memorandum dated May 23, 1986 from T. Murley, Regional Administra-tor, NRC, to W. Harrington, Boston Edison Company, transmitting Systematic Assessment of Licensee Performance (SALP) Report No. 50-293/85-99 5. Memorandum dated September 7, 1988 from T. Martin;, NRC Region 1 Office, to R. Bird, Bosto-Edison Company,

Subject:

Pilgrim Nuclear Power Station Power Ascension Program 6. Memorandum dated September 7, 1988 from S. Collins, NRC Region I Office, to R. Bird. Boston Edison Company, transmitting NRC Region I Inspection Report No. 50-293/88-21, Integrated Assessment Team ) Inspection 7. Memoranda dated May 17, 1988 and July 22, 1988 from R. Gallo, NRC Region I Office, to R. Bird, Boston Edison Company,

Subject:

Inspection No. 50-293/88-11 8. Memorandum dated December 31,1986 from W. Kane, NRC Region 1 Office to J. Lydon, Boston Edison Company, Subject, Management Meeting 50/293/86-4 9. Boston Edison Company, Pilgrim Nuclear Power Station, ACRS Brief-ing books dated August 2, 1988: Volume 1, " Introduction and Restart Plan," Volume 2, " Appendices to Restart Plan," Volume 3, "Self-Assessment of Readiness for Restart," Volume 4, " Power Ascension Program (PAP)," Volume 5, " Safety Enhancement Program (SEP)" l 9 l 1

The Honorable Lando W. Zech, Jr. September 14, 1988

10. Testimony dated August 26, 1988 of Representative Lawrence R.

Alexander, House Chairman of Massachusetts' Joint Comittee on Energy to the Advisory Comittee on Reactor Safeguards

11. Statement dated August 26, 1968 of Douglas Hadfield, Director, Civil Defense for the Town of Plymouth, before the ACRS, at Memor-tal Hall, Plymouth, Massachusetts 12.

Press Release dated August 26, 1988 by Steven Comley representing "We the People, Inc."

13. Statement dated August 26, 1988 by J. Kriesberg, Pesearch Director, Massachusetts Citizens for Safe Energy to the ACRS Hearing on the Restart of the Pilgrim Nuclear Plant 14.

" Question for Inclusion in Congressional Record of January 7,1988 Hearing on the Restart of the Pilgrim Nuclear Power Plant" (un-dated); Submitted to Ad Hoc Subcommittee on Pilgrim Restart at its August 26, 1988 meeting

15. Testimony dated September 7, 1988 of Representative Lawrence R.

Alexander, House Chairman of Massachusetts' Joint Comittee on Energy, to the Advisory Comittee on Reactor Safeguards 16. Letter dated September 6, 1987 from C. Barry, Secretary of Public Safety, Commonwealth of Massachusetts, to W. Kerr, Chairman, ACRS regarding readiness [to restart] of Pilgrim Station

17. Memorandum dated August 6,1987 from R. Krim, FEMA, to F. Congel, NRC, transr.itting:

" Report cf Self-Initiated Review and Interim Finding [of off-site emergency planning] for the Pilgrim Nuclear Power Station," dated August 4, 1987

18. Report dated December 16, 1986 entitled, " Report to the Governor on Emergency Preparedness for an Accident at the Pilgrim Nuclear Power Station" and Supplemental Report dated December 1987, submitted t'y C.

V. Barry, Secretary of Public Safety, Commonwealth of Massachusetts 10

Part 2: ACRS Reports on Generic Subjects t

_ - _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ ___ _________=___________ -___________ _____ ___ [pKEIp 'o UNITED STATES P g NUCLEAR REGULATORY COMMISSION o U E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Y WASHINGTON, D. C. 20555 s.,...../ July 20, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

REPORT ON KEY LICENSING ISSUES ASSOCIATED WITH DOE SPONSORED REACTOR DESIGNS During the 339th meeting of the Advisory Committee on Reactor Safe-guards, July 14-16, 1988, we met with members of the NRC Staff and the Department of Energy (DOE) Staff and reviewed a draft Commission Paper on " Key Licensing Issues Associated with DOE Sponsored Reactor De-signs," dated February 9, 1988. This subject was also considered during our 334th, 335th, 336th, and 337th meetings on February 11-13, 1988; March 10-12,1988; April 7-9,1988; and May 5-7, 1988, respec-tively. Our Subcommittee on Advanced Reactor Designs met on January 6, 1988 to discuss this matter. We also had the benefit of the documents referenced to this letter. The Commission, in a letter dated July 9,1987, instructed the staff to develop such a key-issues paper in advance of projected safety evaluation reports on each of the three conceptual designs being proposed by DOE and its contractors. The Committee believes this was a wise decision; it is appropriate to confront and attempt to resolve the most important safety and licensing issues in a general and direct way, rather than only by reacting to design proposals. In doing this, the NRC Staff has undertaker. an important and difficult task. It can be viewed as an attempt to create, from the top down, a comprehensive rationale for licensing requirements. This would be very different from the existing body of regulations for light water reactors (LWRs), which has grown an elerr,ent at a time in a more reactive and pragmatic fashion. The nation has more than thirty years of experience in the development and realization of practical nuclear power. The DOE sponsored de-signers have made use of this experience and of associated research 11

The Honorable Lando W. Zech, Jr. July 20, 1988 and analytical development to create three conceptual designs which they believe offer significant advantages over existing LWR plants. Similarly, the NRC should take advantage of experience in the regu-lation and safety analysis of plants to create an improved approach to j the specification of safety requirements. In doing this, care must be i taken that regulatory requirements do not unnecessarily frustrate the development of advanced reactors. The regulations should permit the application of innovative reactor concepts while protecting the health and safety of the public. We believe this can be done, but additional effort on the part of the Commissioners and the NRC Staff will be required. False urgency should be aveided; it is more important to do the job right than to do it socn. The staff effort so far has been thoughtful and productive, and pro-vides appropriate preliminary guidance. They have identified four key issues as a basis for review of the design proposals: Accident selection Siting source term selection and use Adequacy of containment systems Adequacy of off-site emergency planning. We believe these are important issues, but they do not adequately encompass the full set of concerns. We comment below on these issues and then discuss several additional issues that we believe are also important and deserve further development. We suagest that the staff's key-issues paper be regarded as preliminary guidance and that a continuing program of development and dialogue is necessary before criteria are considered final. ACCIDENT SELECTION The staff has proposed four event categories for selection of design basis events based on estimates of the probability of events that tr.ight challenge a given system and on past practice and engineering judgment. I For the second of these event categories (EC-II), the staff would require that there be tolerance for single failures, that only safety-i i grade systems should be credited in meeting the event challenge, and that reactor plant systems should continue to operate normally in response to the challenge. We believe this general approach is sound, but requires two caveats: 12

The Honorable Lando W. Zech, Jr. July 20, 1988

• Credit for performance of nonsafety grade equipment in this class of events should be permitted when this can be justified.

Designation of a component or system as safety grade is intended to ensure it has certain specific attributes. Among these are the ability to resist certain seismic events, ability to function within certain harsh environments, and a high level of reliabil-ity (supposedly guaranteed by a quality assurance program). Not all postblated initiating events are challenges to all of these attributes. Selectivity should be permitted when sufficient information is available about the nature of the design basis event. We agree there should not be complete dependence on probabilistic arguments. Although estimates of probability are a proper first-cut approach to the definition of event categories, uncertainty in these estimates is large. Judgments are needed about whether and how to include as design criteria the capability to accommo-date phenomena and seauences that are not specifically indicated to be necessary by probabilistic estimates. CONTAINMENT SYSTEMS Containment structures clearly are intended to restrict release to the environment of radioactive materials resulting from a severe accident. For LWRs, although the design bases for containments have included a source term related to severe accidents, the design pressures and temperatures have been those related to a large-break.LOCA rather than those resulting from an accident involving severe core damage. Whether this seemingly inconsistent but pragmatic approach has served the nuclear pcwer enterprise well can be debated. On the one hand, some of the severe accident issues facing the NRC and the industry today are a legacy of that approach. On the other hand, such a containment performed very well in the TMI-2 accident. Research over the past few years indicates that most existing containments would be rea'anably effective in reducing the consequences of severe accidents. i The staf f proposal for severe accident and containment requirements for advanced reactors seems to be taking e different, but not neces-sarily better approach, then that used for LWRs. Their contention is that, if the early lines of defense, namely: - prevention of challenges to protection systems, and prevention of core damage by protection systems

The Honorable Lando W. Zech, Jr. July 20, 1988 are effective enough, then the next two lines of defense, namely: - a conventional containment structure, and - an emergency plan for the area around the. site, are not necessary. The so-called prevention and protection attributes of the three designs being proposed by D0F end its contractors are indeed im-pressive. The modular high temperature gas cooled reactor (MHTGR) has no conventional containment structure, but relies instead on the capacity of its' unique fuel particles to retain fission products, even at abnormally high temperatures, with high reliability. The two liquid metal reactor (LMR) designs have containers around the reactor vessels, but these have low volume and pressure capacity. It is unclear how they would-accommodate a challenge greater than minor leakage of sodium coolant. Accidents car be postulated that would challenge the defense-in-depth concepts being advanced. For the LMRs, a contemporaneous failure of the gua'rd vessel and the reactor vessel, coupled with a sodium fire, would seem to lead to severe consequences. For the MHTGR, a fire in I the graphite moderator, perhaps permitted by massive failures of the I reactor vessel and core support, might also have severe consequences. Whether these or other accidents could be effectively mitigated by a containment enclosure, or a filtered vent, has not been determined. We note that in all three designs, absence of containment helps to make feasible one of the major safety advantages, passive systems for removing decay heat. In each case, the reactor vessel surroundings are designed so that air from outside the plant will flow by natural . buoyancy through the reactor vessel cavity and thereby remove decay heat.- This seems to be a highly effective heat transfer rceans if the reactor vessel and core are intact. If they are not, this ready supply of oxygen and access to the environment might be a problem. This seems to be a major safety trade-off. We are not prepared at the present time to acr.ept these approaches to i defense in depth as being completely edcquate. Further, we are not j prepared at this time to accept the arguments that increased preven-tion of core melt or increased retention capacity of the fuel provide j adequate defense in depth to justify the elimination of the necd for conventional containment structures. This is not to say that we could not decide otherwise in the future, in response to an unusually persuasive argument. 1 I 14 I i

The Honorable Lando W. Zech, Jr. July 20, 1988 EMERGENCY PLANNING We agree seith the present approach of the staff's proposal.

However, we believe that emergency planning should be reexamined in an effort to describe an approach that would be applicable to all types of reactors.

ADDITIONAL ISSUES How safe should these plants be? We believe the debate about how safe is safe enough is concluded. The safety geal policy is in place. That should stand as the definition of how nafe these advanced reactors, as well as future LWRs, should i be. There are, of course, matters of interpretation and implementa-tion with regard to safety goal policy. These need to be dealt with for all types of reactor plant designs. The focus of licensing and regulation for advanced reactors should be consistent with the safety goal policy; no more, no less, no enhancements, no compromises. The Advanced Reactor Policy states that advanced reactors must be at least as safe as the current generation of LWRs. The staff interprets this to mean the " evolutionary" generation of LWRs now being reviewed by the NRC for preliminary design certification. We believe the Advanced Reactor Policy requires no more than, and should require no more than, the level of safety called for in the safety goal policy. Reactor developers, i.e., DOE and the industry,. ocy seek a design that is safer than the safety goal would suggest as necessary, or whose safety is more readily apparent to the public. Those are not unreasonable goals for a developer in seeking public acceptance or more economic operation. However, it seems to us inappropriate for the NRC to ratchet on the standard of safety it has established as necessary ar.d sufficient. To what extent should regulatory requirements accommodate public~ perceptToii Y The draft paper states that the staff has inenrporated only technical considerations in the developmer,t of its proposed positions. In particular, they have not attempted to accommodate external factors, . e applaud this restraint. And we counsel such as public perception. W the Commission to keep safety regulations unambiguously related to protection of the public health and sinfety. l 15

i The Honorable Lardo W. Zech, Jr. July 20, 1988 Extra capacity in decay heat removal and scram systems The three DOE designs provide much more capacity in decay heat removal and scram systems than are provided in present LWRs. While these important systems in LWRs must be tolerant of single failures, the advanced reactors go well beyond that. The reason for this is the intent to build more robustness into the first two layers of defense in depth and thus permit less in the last two layers, containment and emergency plannino. Two independent scram systems are provided in two of the three pro-poseo designs. Each system is somewhat diverse in design and toler-ant, within itself, of single failure. All three design proposals have multiple systems for decay heat removal. In addition to being diverse and resistant to single failure, the extra systems have inherent passive attributes. They apparently will function effec-tively without motive power or operator intervention. However, a caution is necessary. Experience in operation and analysis has indicated that redundancy, i.e., extra systems or components, is not as powerful in improving reliability as might be expected. Too often the nature of initiating challenges, or of the complex sequence of events in accidents, seems to cause the extra parts of a system to be faulted along with the main system. The diverse and passive nature of the three designs being considered might ameliorate such unwanted interdependency, but further study is warranted. In addition, while the three proposed designs have these positive features, it is not clear that the NRC's proposed requirements would provide assurance that these desirable diverse and passive attributes would be guaran-teed. Need for prototyping The staff proposes only modest requirements for prototype testing of the advanced reactor designs. Although, they have recently added a proposed requirement that any designs not incorporating a containment must be tested in prototype at a remote site, we question whether this is enough to carry the process to a point at which the NRC would be willing to license an unlimited number of new power plants. For example, the metallic LMR cores are claimed to have very favorable, inherently stable characteristics in responding to possible tran-sients. These characteristics were not well understood a decade ago. 1 t An excellent experimental and analytical program by ANL with the EBR-II reactor at INEL has effectively demonstrated that the EBR-II system does exhibit such inherently stable and predictable behavior. However, it is not yet clear that such characteristics can be assured j 16

The Honorable Lando W. Zech, Jr. July 20, 1988 for the larger and different & Rs to be used in commercial electric power production. We believe that a more and extensive series of prototype tests will be necessary before design certification could be granted. Use of cost-benefit analysis The staff paper proposes that prospective licensees should be required to demonstrate through cost-benefit ' analysis that design features alternative to those being proposed are not warranted. Presumably, the NRC staff would review such analyses and perhaps suggest alterna-tives. We believe this is an unworkable and unnecessary strategy. The NRC should concentrate its efforts on specifying design require-ments that will result in plants that are in conformance with the safety goal. Consideration of alternatives and costs is properly a function of the designer and owner of a plant. The NRC should have enough confidence in its safety goal that it does not feel the need for the proposed approach. Design for resistance to sabotage It is often stated that significant protection against sabotage can be inexpensively incorporated into a plant if it is done early in the design process. Unfortunately, this has not been done consistently because the NRC has developed no guidance or requirements specific for plant design features, and there seems to have been no systematic attempt by the industry to fill the resulting vacuum. We believe the NRC can and shculd develop some guidance for designers of advanced reactors. It is probably unwise and counterproductive to specify highly detailed requirements, as those for present physical security systems, but an attempt should be made tc develop some general guidance. Operation and staffing Little is said in the staff paper about requirements for operation and l staffing of advanced reactors. We find this to be a serious over-l sight. Experience with LWRs has shown that issues of operation and I staffing are probably more important in protecting public health and safety than are issues of design and construction. The designers of the three reactor proposBis seem to be claiming that the designs are so inherently stable and error-resistant that the questions of opera-tion and staffing, so inportant for LWps, are unimportant for the j advanced reactors. And that, in feet, the advenced plants can be L operated with only a very sma*1 staff. We believe these claims are j. unproven Lnd that more evidence is required before they can be ac. l cepted. 17

The Honorable Lando W. Zech, Jr. July 20, 1988 The two major accidents that have been experienced in nuclear power, those at TMI-2 and Chernobyl 4, were caused, in large measure, by human error. These were not simple " operator errors" but instead were caused by deliberate, but wrong, actions. There are some indications that the advanced reactor designs being considered have certain characteristics tending to make them less vulnerable to such mal-operation. But, this has not been demonstrated in any systematic way. The traditional methods of PRA are not capable of such analyses; but, we believe a systematic evaluation should be made. There seems little merit in making claims for the improved safety of new reactor designs if they have not been evaluated against the actual causes of the most important reactor accidents in our experience. Will regulatory criteria evolve? The Staff proposal provides for a future milestone in the ongoing design-review-licensing process at which the NRC will step back and make sure that the agreements reached early in the process are still valid, given possible new information and understandings. We believe this is wise and necessary, although it does place a potential licen-see at some risk. It should be recognized that this milestone activ-ity might have to include the possibility of changes in the actual requirements, as well as interpretations of requirements. Focus on the most important residual uncertainties Although the staff paper discusses uncertainties relative to the i development of requirements and designs, it should provide a clearer i statement of what the staff believes to be the most important of these. This would assist policymakers in making judgments about the designs and requirements and, perhaps, about whether certain avenues of research should be further pursued before or in parallel with licensing. Additional comments by ACRS Member Carlyle Michelson are presented below. Sincerely, h i Willliam Kerr Chairman l Additional Coments by ACHS Member Cerlyle Michelson It is not clear to me that the safety goal in its present form was intended to apply to advanced reactors which do not have conventional 18

The Honorable Lando W. Zech, Jr. July 20, 1988 containment systems. The guidelines for regulatory implementation might have been different if the Comission had considered that the defense-in-depth approach might not include a containment system on future plants. It would be unfortunate if the frequency of large release criterion suggested in the present guidelines is used as a basis for justifying the omission of a containment system for an advanced reactor plant at a time when advanced LWRs which might be 6ble to meet the same crite-rion are required to have containments.

References:

1. Draf t Commission Paper from Victor Stello, Jr., for the Commis-sioners,

Subject:

Key licensing issues associated with DOE sponsored advanced reactor designs, dated February 9, 1988 2. U.S. Nuclear Reaulatory Commission, NUREG-1226 " Development and Utilization of the NRC Policy Statement on the Regulation of Advanced Nuclear Power Plants," published June 1988 L l l 19

s / 'o UNITED STATES g 8 NUCLEAR REGULATORY COMMISSION o 3 I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 o, October 13, 1988 The Honorable Lando H. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

PREAPPLICATION SAFETY EVALUATION REPORT FOR THE MODULAR HIGH TEMPERATURE GAS COOLED REACTOR Introduction During the 342nd meeting' of the Advisory Committee on Reactor Safe-guards, October 6-7, 1988, and in previous meetings of the Committee and our Subcommittee on Advanced Reactor Designs, we reviewed a draft of the subject Safety Evaluation Report (SER). During these meetings, we had the benefit of discussions with representatives of the NRC staff and its consultants, with representatives of the Department of Energy (00E), and representatives of General Atomics, the chief design contractor for the We also had the Modular High Temperature Gas Cooled Reactor (MHTGR). benefit of the. documents referenced. The MHTGR concept is a product of a joint DOE / industry program to develop a design for a nuclear power plant using HTGR technology and having important inherently safe characteristics. The NRC staff is reviewing the concept under the advanced reactor policy to help assure that the final design will develop along lines acceptable to the NRC. The draft SER indicates that the staff believes the conceptual design is generally satisfactory and that work directed toward eventual certifica-tion should continue. The staff-has provided a number of conditions along with this endorsement and also believes that a continuing program of research and development will be necessary to support final design and eventual licensing. I We tre in general agreement that design and development thould continue along the lines outlined by the NRC staff. We can agree to moving forwar6, however, only because we undtrstand that an NRC endorsement 6t this time does not imply a final commitment either to the general design I or to its details. We believe that ongoing research and development can resolve important safety issues before licensing. We bave a number of ~ comments discussed below about the design. 21

The Honorable Lando W. Zech, Jr. October 13, 1988 Key Features of the MHTGR 1 The MHTGR differs in important ways from existing light water reactor (LWR) plants and from previous gas cooled reactor plants, including i several new safety characteristics. The goal of the designers is that L the improved safety features will more than make up for the absence of others (e.g., containment). They believe the MHTGR design will provide a plant that is safer than LWRs. Safety of the MHTGR is keyed to properties of its unique fuel particles. Millions of these microspheres of enriched uranium oxycarbide, each the size of a grain of sand, are in the reactor core. Each fuel particle is coated with four successive protective shells that includes a buffer layer of a porous carbon and then bonded with others into a fuel rod which is, in turn, sealed in vertical holes in graphite blocks. These graphite blocks provide neutron moderation and are the chief structural material in the core. l The ma'imum fuel particle temperature in normal operation will be about 1150*C. An expected very small fraction of defective particles will cause a r.easurable, but acceptably low, level of chronic fission-product activity in the coolant and reactor systems. l So long as the particles are maintained below 1600*C, fuel, transur-anics, and fission products will be retained by the particle coatings, with very high efficiency. At temperatures above about 2000*C, failures of particle coating will become significant, and above about 2300 C the coatings will fail completely. All other safety features of the reactor systems are designed to assure that particles will remain below 1600*C over a wide range of challenges and circumstances. It is expected that temperatures can be maintained below 1600*C, in any conceivable reactor transient, because of two favorable characteristics of the reactor ccre: (1) Strong negative reactivity changes with increased temperatures in fuel or moderator and (2) Large thermal inertia of the core and fuel structure. It is also expected that temperatures will be maintained below 1600'C even with loss of normal decay heat removal because of the following important features: (1) The same strong temperature-reactivity effects will assure a very 4 low equilibrium power eren with failure of reactivity control and 1 shatdown systems. (2) At these low or decay power levels, if normal heat transfer systems - fail, all heat can be removed from the reactor by a pessive heat i transfer system that permits atmospheric air to flew by natural ] 22

The Honorable Lando W. Zech, Jr. October 13, 1988 convection through a cavity surrounding the reactor vessel. Under these conditions, the reactor core and the vessel will attain temperatures only slightly above their normal operating values. (3) If this passive heat removal system should become unavailable (e.g., by blockage of air flow), heat at low power or at decay heat levels would be transferred from the reactor cavity by conduction directly to the earth surrounding the reactor building. Under these conditions, fuel would remain below 1600 C, but the reactor vessel would eventually heat to well beyond its normal operating temperature. Whether the reactor could be returned to normal operation after exposure of the vessel to such overtemperature is problematic at the present time. But, the vessel would remain sufficiently intact for the safe removal of decay heat. The passive heat transfer functions in items (2) and (3) above require that the reactor core and vessel be small enough so that heat transfer can be accomplished without core temperatures becoming excessive. This dictates the reactor size and leads to the modular design and the long, small-diameter core. The reactor core is normally cooled by inert helium gas circulated through the core at high pressure. Certain improbable failures of the reactor vessel could permit air to enter the core. However, air flow through the core by natural convection would be at a very low rate. With this restricted supply of oxygen, oxidation of graphite would be so slow that after many hours only a small fraction of the graphite would be consumed and the core would remain structurally intact. Even if the graphite should burn, through some undetermined mechanism, the indica-tions are that the graphite temperature would be well below the 1600*C critical temperature for the fuel particles. The combination of nuclear decay and conbustion heat would not be expected to increase core tem-perature to greater than 1600*C. The Safety Issues The challenge in assuring that the key safety characteristics claimed for the HHTGR design are realized in an actual plant is, in simplest terms, in assuring that the following issues are adequately addressed: l l (1) Fuel particles must have the retention capabilities attributed to them and this must be assured with recognition of inevitable l l variability and imperfection in the fuel particles and their compaction process. This will require a higher level of quality in manufacture than has been achieved and must be experimentally verified. 23

The Honorable Lando W. Zech, Jr. October 13, 1988 i l 1 (2) The reactivity and temperature-reactivity characteristics used in safety analyses are based on limited data. Further verification of these characteristics as a function of fuel burnup, core shuffling, and a variety of operational transients is needed. (3) Inadvertent ingress of water or steam into the core must be pre- .cluded with high reliability. Water or steam could cause corrosion and mechanical damage to the graphite and would also add a positive reactivity contribution. This seems to be a possible complication of, for example, steam generator tube failures that is not present in LWRs. Internal flooding of the underground reactor cavity could lead to similar problems. (4) There must be assurance that decay and low-power heat transfer can be accomplished without causing excessively high core temperatures. Performance of the passive atmospheric cooling system and the ability te conduct heat to the surrounding earth must be demon-strated. (5) The structural properties of the graphite must be demonstrated and assured. (6) Some of the important safety benefits of the design (e.g., passive decay heat removal and resistance to graphite burning) depend upon the core geometry remaining unperturbed. Questions of seismic resistance, effects of aging, and the possible cascading effects of certain reactor accidents remain to be fully answered. A major issue is whether a conventional containment structure or some other mitigation system or process should be required. Neither the designers, the NRC staff, nor the members of the ACRS have been able to postulete accident scenarios of reasonable credibility, for which an additional physical barrier to release of fission products is required in order to provide adequate protection to the public. This does not mean that a conventional containment should not be provided or required as further defense in depth against unforeseen and unforeseeable events. J However, it does meen the.t.the design basis for a containment would have to be arbitrary, not altogether unlike what was done in the early days for LWRs. We believe that the decision to require a containment will 1 have to.be made on the basis of technical judgment, with appropriate { consideration of the effects on other technically based safety features I now a part of the design. In addition, there may be safety and economic tradeoffs ' between provision for containment and provision for passive decay heat removal, l 24

i f I The Honorable Lando W. Zech, Jr. October 13, 1988 Recommendations A substantial program of research and development must be continued to support the final design for the MHTGR. This program should concentrate on providing assurances relative to the safety issues we have discussed above. General Atomics has generated extensive data on fuel performance, but a comprehensive program on the reference fuel appears to be needed. This would include testing of irradiated fuel, fuel from large-scale man-ufacturing, and fuel exposed to a variety of environmental conditions and temperatures such as might be encountered in possible accidents. A hot critical experiment may be necessary. The core is of an unusual geometry and has nuclear characteristics different from those in previ-ous HTGRs. Assuring that the safety response of the plant is as pre-dicted will require comprehensive information on the reactivity charac-teristics of the core over a broad range of normal and accident con-ditions. More extensive analysis is needed of the response of the plant to accidents that might change the core geometry. Certain accident scenar-ios can be hypothesized that would affect core geometry and influence coolant distribution and reactivity characteristics. A prototype should be built and appropriately tested before design certification. Concepts for a containment or another sort of physical mitigation system require further study. Finally, there are two issues identified in our letter to you dated July 20, 1988, " Report on Key Licensing Issues Associated With DOE Sponsored Reactor Designs, that we believe should be given early consideration as the design of this plant progresses. These issues are related to design for (1) resistance to sabotage and (2) operation and staffing. The appropriate excerpts from that letter are attached. Additional coments by ACRS Hembers Forrest J. Remick and Charles J. Wylie, and William Kerr are presented below. Sincerely, hhV William Xerr Chairman 25

The Honorable Lando W. Zech, Jr. October 13, 1988 Additional Comments by ACRS Members Forrest J. Remick and Charles J. Wylie In general, we agree with our colleagues in the above letter.

However, we cannot in good conscience recommend a design of a nuclear power plant for design certification which does not have a conventional containment or other mitigation system which would serve as a more robust external barrier than is currently proposed to protect the public from radio-logical releases.

The designers of the MHTGR deserve much credit for their effort to incorporate inherent and passive safety features in the design concept. However, even though we believe that the proposed design has a good potential for providing enhanced safety, experience has shown that new reactor designs have technical unknowns. Because of the possible technical unknowns, the known uncertainties associated with the pos-tulated inherent and passive safety features and the lack of experience with operation of a reacter of this new design, we do not recommend these reactors for design certification without a more extensive ex- -ternal barrier consisting either of a conventional containment structure or other appropriate mitigation system. We think it important that the ACRS and the Commission make this techni-cal judgment at this time in order that the designers of this promising reactor concept have ample opportunity to thoroughly consider alternate designs. Additional Comments by ACRS Member William Kerr I remind the Commission of the comments on containment included in the Committee's letter of July 20, 1988, namely: "We are not prepared at the present time to accept these approaches to defense in depth as being completely adequate. Further, we are not prepared at this time to accept the orgaments that increased prevention of core melt or increased retention capacity of the fuel provide adequate defense in depth to justify the elimination of the need for conventional l containment structures. This is not to say that we could not I decide otherwise in the future, in response to an unusually persuasive argument." That is still my position on the containment issue. I would add only that I have not yet heard the persuasive argument." l 26 1 L

I' l The Honorable Lando W. Zech, Jr.- October 13, 1988 l-

References:

1. Office of Nuclear Regulatory Research, " Pre-Application Safety Evaluation Report for the Modular High Temperature Gas Cooled Reactor," dated August 1988 (Predecisional Draft) (DOE Contract), l Stone. Webster-Engineering Corporation 2. HTGR-86-024, "HTGR Preliminary Safety Information Document for the Standard MHTGR," Volumes 1-5, 1986 3. GA Technologies, Inc. (DOE Contract), DOE-HTGR-86-011, "HTGR Probabilistic Risk Assessment for the Standard Modular High Temperature Gas-Cooled Reactor," Volumes 1-2, January 1987

Attachment:

Excerpts from' July 20, 1988 ACRS Letter, " Report on Key Licensing Issues Associated With DOE Sponsored Reactor Designs" 27 __u

ATTACHMENT TO ACRS LETTER ON MODULAR HIGH TEMPERATURE GAS COOLED REACTOR Excerpt from July 20, 1988 ACRS Letter, " Report on Key Licensing Issues Associated With DOE Sponsored Reactor Designs" Design for resistance to sabotage It is often stated that significant protection against sabotage can be inexpensively incorporated into a plant if it is done early in the design process. Unfortunately, this has not been done consistently because the NRC has developed no guidance or requirements specific for plant design features, and there seems to have been no systematic attempt by the industry to fill the resulting vacuum. We believe the NRC can and should develop some guidance for designers of advanced reactors. It is probably unwise and counterproductive to specify highly detailed requirements, as those for present physical security systems, but an attempt should be made to develop some general guidance. Operation and staffing Little is said in the staff paper about requirements for operation and staffing of advanced reactors. We find this to be a serious over-sight. Experience with LWRs has shown that issues of operation and staffing are probably more important in protecting public health and safety than are issues of design and construction. The designers of the three reactor proposals seem to be claiming that the designs are so inherently stable and error-resistant that the questions of opera-tion and staffing, so important for LWRs, are unimportant for the advanced reactors. And that, in fact, the advanced plants can be operated with only a very small staff. We believe these claims are unproven and that more evidence is required before they can be ac-cepted. The two major accidents that have been experienced in nuclear power, those at TMI-2 and Chernobyl 4, were caused, in large measure, by human error. These were not simple " operator errors" but instead were caused by deliberate, but wrong, actions. There are some indic6tions that the advanced reactor designs being censidered have certain characteristics tending to make them less vulnereble to such mal-operation. But, this has not been demonstrated in any systematic way. The traditional methods of PRA are not capable of such analyses; but, we believe a systematic evaluation should be made. There seems little merit in making claims for the improved safety of new reactor designs if they have not been evaluated against the actual causes of the most important reactor accidents in our experience. 28

i UNITED STATES / \\ NUCLEAR REGULATORY COMMISSION 1 / g 3 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS j WASHINGTON, D. C. 20555 o November 22, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D ' 20555 Dear Chairman to

SUBJECT:

SAFETY EVALUATION REPORT FOR THE " POWER REACTOR INHERENTLY SAFE MODULE" (PRISM) DESIGN During the 343rd meeting of the Advisory Comittee on Reactor Safe-we reviewed a draft of the subject safety 17-18,).1988, guards, November (SER The ACRS and its Subcommittee on Advanced evaluation report Reactor Designs have reviewed these matters in previous meetings. During these meetings we had the benefit of discussions with representa-tives of the NRC staff and its consultants, and with representatives of the Department of Energy (DOE) and its contractors, including represen-tatives of the General Electric Company, the lead design contractor. We also had the benefit of the documents referenced. The PRISM conceptual design is a product of a DOE program to develop designs for possible future power reactor systems that would have enhanced safety characteristics. Other design projects in the program are the Modular High Temperature Gas-Cooled Reactor (hHTGR) and the Sodium Advanced Fast Reactor (SAFR). The NRC staff is reviewing these designs in accordance with the Comission policy on Advanced Nuclear l Power Plants. These preapplication reviews are intended to provide NRC l guidance on licensing issues at a relatively early stage of design development. The ACRS has previously commented to you on NUREG-1226, l " Development and Utilization of the NRC Policy Statement on the Regula-tion of Advanced Nuclear Power Plants," in June 1987, on key licensing issues assuciated with the entire program in July 1988, and on the SER for the MHTGR in October 19P8. We understand that issuance of the SER will not constitute approval of the PRISM design. Further engineering development and documentation will be required to support a future application for design certifica-tion. The PRISM design incorporates several small, modular reactors cooled by liquid sodium. Trie standard PRISM plant would consist of nine reactor modules, each generating 425 MWt, providing a total plant output of 1245 29 i

The Honorable Lando W. Zech, Jr. November 22, 1988

MWe, Each reactor, along with its intermediate heat exchangers and pumps, is imersed in a pool of sodium.

A steel vessel containing this pool is located within a secondary steel container. The steel con-tainers share a common head. Each such unit is installed within an underground concrete silo. Secondary sodium coolant flows to steam generators which are also located below grade, but are outside the silo along with the remainder of the " balance of plant" (B0P) equipment. The PRISM design provides several features for enhancing safety of a nuclear power plant.

• a passive system for emergency removal of decay heat
• inherent mechanisms for negative feedback of reactivity

' large thermal inertia in the pool of sodium coolant ' metal fuel, offering greater opportunity for on-site fuel reprocessing

• small component sizes, providing opportunities for factory fabrication
• opportunity for prototype testing of a single module

' separation of safety-related functions from 80P systems On the basis of its review, the NRC staff has concluded that the PRISM design has the potential for a level of safety at least equivalent to current light water reactor (LWR) plants, provided that a number of specific issues are resolved. Our general recommendation is that, from the perspective of safety and licensing, design development of PRISM should continue, taking into account the points made by the staff. i A number of safety issues remain to be completely addressed, a program l of continuing research and development is necessary to support further design, and plans for extensive prctotype testing should be developed. In the following paragraphs we coment on a number of specific safety issues which we believe should be considered by the staff in its final SER, and by DOE in its continuing development and design activities. Containment i Although a secondary vessel is provided to contain leakage of sodium coolant, the PRISM design does not include a conventional containment capable of resisting high temperatures and pressures. It is contended that the potential for core disruptive accidents, for which such a 30 i

1 i The Honorable Lando W. Zech, Jr. November 22, 1988 l containment might provide mitigation, is so low that a conventional containment is not needed. Both deterministic and probabilistic argu-ments are made in support of this contention. Although these arguments have technical merit, we are not yet convinced. Our position is as stated in our report to you of July 20, 1988 on the key licensing issues associated with DOE sponsored reactor designs and our report to you of October 13, 1988 on the preapplication safety evaluation report for the modular high temperature gas-cooled reactor. However, there is a problem. One reason for providing a strong physical containment is to protect the public against unforeseen accidents.

But, precisely because they are not foreseen, the design requirements for a containment are not obvious.

Therefore, engineering and policy judg-ments must be made about the need for, and nature of, containment that might be used with PRISM. We believe that further study is appropriate before final judgments are made. Absence of a Backup Shutdown System The PRISM design provides a control rod system consisting of six control rods, a safety grade means of scraming these rods by gravity, and a safety grade electrical system to drive the ro(s into the core. How-ever, the design provides no backup to this cot trol rod system other than the inherent characteristics of the core. We question whether these inherent characteristics are adequate as a backup system, for two reasons. First, they may not act fast enough to compensate for certain fast transients without scram. Second, they are not capable of making the reactor subcritical and taking it to cold shutdown conditions. Therefore, we believe the need for a backup system or suitable demon-stration of scram reliability deserves further study. Need for Local Flow and Temperature Monitoring The PRISM safety analysis indicates that blockage of flow through one fuel assembly may possibly damage that assembly, but will not damage adjacent assemblies. Early work with oxide fuel has demonstrated that propagation is unlikely, but experiments and analysis with metal fuel have not been as exttnsive. Especially because the design does not provide for monitoring flow and effluent temperature from individual assemblies, we believe this requires further study. Individual Rod Worth Each of the six control rods is sufficient, individually, to shut down the reactor and maintain it in cold shutdown. Therefore each rod has a very large reactivity worth, about two dollars. There is thus potential 31

The Honorable Lando W. Zech, Jr. November 22, 1988 for serious consequences from a rod ejection accident. This potential is ameliorated in two ways. First, for startup, rod operations are interlocked so that the rods can be withdrawn only in a carefully orchestrated sequence. This rod sequencing system will have to be very carefully designed, operated, and maintained. Second, for power opera-tion, the expected reactivity change of a core through its lifetime is expected to be so flat that only very small rod insertion will be necessary at the beginning of core life, thus reducing the effect of a rod ejection accident. These features will be effective only with accompanying administrative controls on core design and rod operation over the lifetime of PRISH plant operations. This should be acknowl-edged in the SER. Role of the Operator We believe that insufficient attention has been given to the role of the operator. Claims that a PRISM plant would have such inherently stable and safe characteristics that the operator will have essentially no safety function are unproven. Operation of nine reactors, possibly in several different operational states at any given time, may be a daunt-ing challenge for the small operations crew envisioned. Opportunities for cognitive error, which might defeat favorable safety characteristics of the reactor, might be more abundant than is now recognized. Further study appears to be desirable. We believe insufficient attention has been given to the physical securi-ty of the plant's operating and technical support staff. It is claimed that the control room, with all of its contents, including operating personnel, can be destroyed and that the plant can be safely shut down from remote contrn1 stations that are within the physical security controlled areas of the plant. Therefore, the control room and techni-cal support areas are now proposed to be located outside the physical security boundary. We believe, given an external threat, such as an attack by terrorists, that it is essential to preserve the operating and technical expertise on-site, and recommend that the control room and appropriate technical support personnel be located within the physical securf ty boundary. Other Operational Considerations In addition, certain features that have been found to be desirable in LWR plants are not provided in the PRISM design. No technical support center is provided. Although remote shutdown capability is provided, it appears to lack some of the attributes of such systems in current LWR plants. Also, the design does not include Class IE AC electric power systems, but relies entirely on IE DC power from batteries. It is not clear that adequate consideration has been given to the potentially 32

i The Honorable Lando W. Zech, Jr. November 22, 1988 1arge power needs of essential auxiliary functions such as space cooling and emergency lighting. Protection Against Sabotage With regard to the need for designing protection against sabotage, the following statement from our report of July 20, 1988 should be given early consideration as the design of this plant progresses: "It is often stated that significant protection against sabotage can be inexpensively incorporated into a plant if it is done early in the design process. Unfortunately, this has not been done consistently because the NRC has developed no guidance or requirements specific for plant design features, and there seems to have been no systematic attempt by the industry to fill the resulting vacuum. We believe the NRC can and should develop some guidance for designers of advanced reactors. It is probably unwise and coun-terproductive to specify highly detailed requirements, as those for present physical security systems, but an attempt should be made to develop some general guidance." Sodium Fires Further study of the potential for and suppression of sodium fires and consideration of their possible consequences is needed. Such studies should include the possibility of fires resulting from earthqua ke effects. Sincerely, Forrest J. Remick Acting Chairman

References:

1. Office of Nuclear Regulatory Research, " Safety Evaluation Report for the Power Reactor Inherently Safe Module (PRISM)/ Liquid Metal Reactor Conceptual Design," dated September 10,1988(Predecisional Draft) 1 2. General Electric / Nuclear Systems Technology Operation (DOE Con-tract), GEFR-00793, " PRISM Preliminary Safety Information Docu-ment," Volumes I through V, IS86 33

>AKtCy l '/ 'o UNITED STATES 8[ J ^,% NUCLEAR REGULATORY COMMISSION .,E ADVISORY COMM!TTEE ON REACTOR SAFEGUARDS o g WASHINGTON, D. C. 20555 I %..*l \\ 5 July 20, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

REPORT ON THE INTEGRATION PLAN FOR CLOSURE OF SEVERE ACCIDENT ISSUES (SECY-88-147) During the 339th meeting of the Advisory Committee on Reactor Safe-guards, July 14-16, 1988, we discussed with members of the NRC staff a plan for the integration of the various severe-accident-related programs as described in SECY-88-147, " Integration Plan for Closure of Severe Accident Issues." This plan was also considered by our Severe Accidents Subcommittee during a meeting held on July 13, 1988. We also had the benefit of the documents listed as references to this letter. We commend the NRC staff for its efforts to develop an integrated approach for dealing with the various severe accident issues and to centralize responsibility for resolving them. SECY-88-147 describes the first step toward developing such a plan, namely, identifying the relevant issues. However, it gives little information on how the various issues are to be integrated. Rather, it discusses the severe-accident-related issues and programs that should be integrated, but 4 does not describe the process to be used. l The need for additional integration is illustrated in the discussion of external initiators. In several recent PRAs, externally initiated 4 sequences are major contributors to risk. This fact appears not to have been considered in SECY-88-147. Considering only internal initiators may well provide a distorted picture of the " major vulner-abilities" for a particular plant. This may result in an inappro-priate allocation of resources for plant-specific fixes, unless all system changes are delayed until external events are treated. This does not seem to be the procedure to be used. Further, the statement is made, in support of delaying a consideration of external initi-ators, that no new sequences are likely to be initiated by seismic events. This seems to contradict the conclusions of a Brookhaven 35

The Honorable Lando W. Zech, Jr. July 20,1988 study of the GESSAR PRA which concluded that relay chatter, produced by a seismic event, could be a major risk contributor. Furthermore, it ignores the fact that a large seismic event has the capability (much less likely for other initiators) of simultaneously initiating a large number of risk-significant sequences. The comments on severe accident management provide no indication of how the licensee is to proceed. Although for this issue, immediate action is not required in connection with the Individual Plant Exami-nations (IPEs), the implication is that enough information now exists to permit a licensee to formulate an appropriate program. We note that on March 13, 1985, the ACRS sent a memorandum to the then-EDO, l Mr. William J. Dircks, in which we asked if enough information existed l to provide guidance to plant operators in a situation in which core melting had proceeded without a source of cooling. Our question was l whether a situation could develop in which, if coolant became avail-able after core melt had begun, adding coolant to the in-vessel melt would exacerbate the accident. We have yet to receive a response to our memorandum. This, we think, is a rather fundamental question. If the staff does not have the information to answer this question, how is a licensee to reach a decision? Does existing instrumentation provide the information needed? Does the instrumentation suggested in Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," suffice for the task? For accident manage-1 ment, answers to such questions are required. l We observe that, in the evaluation of containment performance, the licensee is to " consider" direct containment heating (DCH), a pos-tulated event 6 bout which there are major uncertainties. However, the proposed gercric letter on IPEs (Reference 3) states that no major changes in the containment are to be made until the NRC research program has produced information required to decide what, if any, system changes should be made. Is anything to be done in the mean-time? What is the " consideration" by licensees to produce? We note that a Panel report on source term uncertainties (Reference 4) con-cluded that information needed to determine the effects of DCH is unlikely to be available for a long time. The Panel recommended that, rather than wait for the results of the needed research, the probabil-ity of DCH should be made negligibly low by hardware changes or procedural measures. Furthermore, in describing the resolution of some of these important issues, the process to be used is left so vague as to be uninterpret-able. For example, from the discussion of the way in which it is proposed to deal with severe accidents for advanced light water reactors (LWRs), one gets the impression that if some as yet undefined 36

The Honorable Lando W. Zech, Jr. July 20,1988 process, possibly rulemaking, is put into place, the problem will somehow become resolved. In the area of containment performance criteria for advanced LWRs, it is especially important that some early decisions be made. The review process currently being considered appears to endorse the use of design criteria based on " design-basis accidents" formulated before the Reactor Safety Study (WASH-1400), which indicated a need to consider severe core damage accidents. This seems, at best, imprudent in light of all that has been learned since these criteria were first formulated. Designs using these " obsolete" criteria are now being cerisidered in the licensing process. In our discussions with the staff, we explored how the Reactor Risk Reference Document (NUREG-1150) will be used in the rasolution of the severe accident issues. Although we were told that the information in this document will play a key role, we were unable to get a clear picture of just how. If NUREG-1350 is to play a key role, it is important that its accuracy and credibility be established. We believe that subjecting the final version of NUREG-1150 to a thorough peer review is required as part of the process of establishing credi-bility. We believe a glossary of terms used in SECY-88-147 would be helpful. We suggest that SEVERE ACCIDENT, DAMAGED CORE, CORE DAMAGE, CORE MELT, VULNERABILITIES, RADI0 ACTIVE RELEASE, LARGE RADI0 ACTIVE RELEASE, CON-TAINMENT PERFORMANCE, CONTAINMENT FAILURE, and CONTAINMENT BYPASS be defined. In addition, definitions for FRONT END, BACK END, LEVEL I PRA, PEEVENTION, and MITIGATION as used in this paper might be l helpful. Finally, we encourage the staff to continue its efforts toward inte-i gration of the various programs being developed for resolution of the severe accident issues. We believe that the most recent draft generic letter describing the IPE program (Reference 3) represents a move in the direction we have recommended in our letter to you of May 10, 1988. We are convinced that further integration can conserve re- ) sources of both the staff and the licensees and can contribute to a i more effective process for risk reduction in operating plants. Sincerely. W. Kerr i Chairman l l 1 37 l

u l-The Honorable Lando W. Zech, Jr. July 20, 1988

References:

1. SECV-88-147, Memorandum dated May 25, 1986, for the Commissioners from V. Stello, Executive Director for Operations,

Subject:

Integration Plan for Closure of Severe Accident Issues 2. - Brookhaven National Laboratory Draft Report, "A Review of the GESSAR II BWR/6 Standard Plant Seismic Probabilistic Risk Assess-ment," September 1984 (Unpublished-Predecisional) 3. Memorandum dated April 1,1988, from T. Speis - (NRC) to W. Kerr (ACRS), " Documentation Necessary for the Initiation of the Severe Accident Policy Implementation" (Draft Predecisional Attachments - Portions Updated as of June 28,1988) 4. Brookhaven National Laboratory Report, NUREG/CR-4883, " Review of Research on Uncertainties in Estimates of Source Terms from Severe Accidents in Nuclear Power Plants," H, Kouts, April 1987 5. U.S. Nuclear Regulatory Commission, WASH-1400 (NUREG-75/014) " Reactor Safety Study," October 1975 i 6. U.S. Nuclear Regulatory Commission, NUREG-1150, " Reactor Risk Reference Document," Draft for Comment, February 1987 1 ) 38 l

o ctroq 4't, UNITED STATES NUCLEAR REGULATORY COMMISSION M E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Y WASHINGTON, D. C. 20555 %...../ June 7, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

NRC RESEARCH RELATED TO HEAT TRANSFER AND FLUID TRANSPORT IN NUCLEAR POWER PLANTS During the 338th meeting of the Advisory Committee on Reactor Safe-guards, June 2-4, 1988, we considered a report from our Subcommittee on Thermal Hydraulic Phenoraena pertaining to its review of research activities sponsored by the NRC on reactor thermal-hydraulic phenomena. This subject had been considered during the 337th ACRS meeting, May 5-7, 1988 and a number of previous meetings of the ACRS and the Subcom-mittee. We also had the benefit of the documents referenced.

Background

The technical subjects of heat transfer and fluid transport are of cardinal importance when considering the safety of nuclear power plants. They are of chief concern in relation to LOCA and reactor transients, and the performance of ECCS, steam generators, secondary systems, and containment phenomena. These issues have been studied extensively in experimental and analytical programs sponsored by NRC and the industry. The earliest reactor safety research dealt principally with reactivity addition accidents. Later research, sponsored initially by the Atomic Energy Commission and then by the NRC, was devoted almost exclusively to providing)a better understanding of the hypothetical large-break LOCA (LBLOCA and the related performance of ECCS and containment systems. It was soon perceived that the complexities of two-phase flow l and the time sequences involved in a LBLOCA were such that straight-forward experimental representation was difficult. It was further perceived that traditional methods for application of empirical data to plants were subject to challenge. In response to this, the NRC spon-sored development of complex computer codes at the Department of Energy's (DOE's) national laboratories. j These codes were intended to provide consistent treatment of the relationships among various plant systems during rapid transients and 39

The Honorable Lando W. Zech, Jr. June 7, 1988 to bridge gaps in data from the various test programs. In time, as the physical representations of two-phase flow and heat-transfer phenomena and the plant systems were made more detailed, broad interpolation and extrapolation from experiments were attempted and came to be relied upon. This general strategy, that is, primary dependence on detailed math-ematical models of physical phenomena coded for rapid analysis by computers, has been adopted by the NRC for studying other technical areas involving complex phenomena and interactions. Another product of this era, in addition to the extensive base of I experimental information and the codes, has been a skilled cadre of experts. These experts can be found in the NRC, the national labora-tories, the universities, and the industry. These people understand thermal-hydraulic phenomena associated with LBLOCA probably as exten-sively. as almost any other similar subject in modern technology is understood. These experts are also well-schooled in the general strategy described ' above. It is important that a cohesive group of experts be maintained. The next period of thermal-hydraulic research followed the accident at TMI-2 and the gradual assimilation of the perspectives provided by probabilistic risk assessments. Interest shifted toward small-break LOCA and plant transients. De-emphasis of LBLOCA began. The large system codes, developed in the previous era, were available and, with modification, came to be the primary means by which these less dramatic reactor events were analyzed. It was recognized - that these codes, I originally written to incorporate certain conservatism in'an attempt to envelop uncertainties in LBLOCA analyses, would serve their new purposes only if they could more realistically track the evolution of i transients. The conversion to realistic or "best estimate" codes is i now complete. J In each of these periods of thermal-hydraulic research, scientific interest was confined to phenomena and time sequences associated with normal plant conditions and with faulted conditions extending -to, but not beyond, the point at which a coolable core geometry is lost; research activity also included consideration of single-and two-phase flow, heat transfer, and nonequilibrium conditions. General Recommendations The following comments include recommendations for future research in the traditional " thermal-hydraulic" area, including specific recom-mendations for the code development program. 40 l

The Honorable Lando W. Zech, Jr. June 7, 1988 For the sake of discussion we have posed our comments in this _ section as a series of questions, with our recommendations following' as an-swers:

• Is there a need to continue a program of experimental research in the traditional thermal-hydraulic area?

Yes, but not indefinitely nor without specific purpose. There are several matters-that currently need attention and will require several more years of experimental work at a moderate rate. Specific recommendations will be given below. NRC has played a key role in the fundamental development of thermal sciences related to nuclear power plants. It should continue to furnish leadership, perhaps by more clearly defining basic research needs or directions for the DOE and industry. l l ' Is the strategy of dependence on large system codes as primary tools for analysis valid? This strategy has both strength and weakness. As strength, the system codes have the ability to model, in a consistent and reasonably accurate way, the dynamic relationships among the various elements in a plant heat transport system. They are weaker in the accuracy with which they model the complex physical behavior of system subelements, especially in extreme off-normal conditions. This weakness becomes an important problem because analysts and decisionmakers tend to overlook the inaccuracies and to behave as if the codes were revealing physically correct and validated information about the plants. These codes are also very expensive to use and require specialists to use them properly. l These codes can be useful if they are regarded as simply one input, albeit 'often an important one, to the understanding of plant transients. The codes can be dangerously misleading if they are used without engineering judgment and to the exclusion of simpler but less comprehensive analyses. We are concerned that those conducting research in severe accident phenomena have fallen into this trap.

• Should traditional (i.e., LOCA) code development be continued?

The codes are now adequate for the purposes for which they are needed and further development is unjustified. First, they satisfy the regulatory need related to the ECCS rule. For this the Code Scaling, Applicability, and Uncertainty (CSAU) Program is 1 41 __-___ - _ a

The Honorable'Lando W. Zech, Jr. June 7, 1988 p helpful. Second, they are adequate as general-purpose tools 'for exploring and gaining understanding of-other plant transients, from a safety rather than a regulatory perspective. In this use, analysts should be guided by the consnents above. In' making this recommendation, we recognize that the codes are not without flaws. However, we believe that not all of the imper-fections in the codes can possibly be corrected by any reasonable program of research and code development. Marginal improvements that 'could be made over the next -few years by extrapolating the i recent levels of development work will not.be sufficient to attain a significantly higher plateau. of code accuracy and validation. The. code development' effort has been a substantial technical achievement and the endes have made an important contribution.to nuclear power plant safety. Further refinement is unnecessary. The CSAU Program will provide a reasoned perspective on the accuracy of the existing codes. With that perspective available, we endorse the general. strategy proposed by the RES staff toward maintenance of existing codes. This would provide for completion of RELAP-5 and. TRAC-PWR development through the International Code Assessment Program-Consortium in 1989. A - modest level of re-sources would be provided to maintain the codes overall.(including TRAC-BWR, COBRA-NC,andRAMONA-3B), based'onregulatoryneeds. Nuclear power plants' are complex machines, even in normal operat-ing modes; they have many interrelated systems and processes. We believe that computer codes can model normal operating behavior accurately and usefully, if extreme physical-phenomena are not involved and if the codes can be validated by comparing their results to measurements of plant operating parameters.. There'is a significant resource in code development expertise at the national laboratories. Consideration _should be given to using this re-source with ' an approach to code development.that takes advantage of inherent strengths in the present codes. Efforts should be t concentrated on including all of the plant systems, providing code versions ' validated for specific plants and providing modeling and j-interfacing that is transparent and understandable for use by those expert in plant operation rather than just those expert in analysis by computers, i

• Is it essential that a cadre of experienced people be maintained?

It is essential to maintain such a cadre, because-questions of fluid and heat transport will always be central to reactor safety. The NRC should maintain a center of expertise in experimental and 42

l i The Honorable Lando W. Zech, Jr. June 7, 1988 analytical research in thermal-hydraulic phenomena. The Technical l Support Center at the Idaho National Engineering Laboratory serves this purpose. However, the NRC should limit the program to: (1) confirming selected information supplied by industry and (2) exploring important issues that the industry is not addressing. Involvement of universities and other nongovernment research organizations should be encouraged. There should be free exchange of information with industry and international experts. - Specific Recontrendations (1) The CSAU method, or something similar, can be used in other areas of safety analysis, that is, beyond the currently conceived purpose of assessing uncertainty associated with calculations by thermal-hydraulic codes. In particular, its application to severe accident studies and risk assessments could serve, not only to provide an improved perspective on uncertainty, but also as a guide to allocation of research resources. This should be inves-tigated. (2) The current programs of research on B&W reactor systems and once-through steam generators should be continued only to the point that the technical understanding of B&W systems is compara-ble to that of other nuclear steam supply systems. In particular, it should be demonstrated that adequate capability for predicting B&W system performance is in hand. (3) Analysis of industry experience with water hammer events suggests that water hammer is not a significant initiator of nuclear power plant accidents. However, insufficient consideration has been given to whether water hammer, occurring as a consequence of other initiators, might contribute to unexpected failures that could compromise core cooling. This issue should be investigated. (4) The recent steam generator tube rupture (SGTR) at North Anna has been explained as the result of a series of mechanisms which indicate that multiple SGTRs are no more probable than has been believed. The licensee's technically complex explanation was based on poorly understood phenomena. The NRC should explore this issue sufficiently to confinn the licensee's conclusions. (5) Although the feed-and-bleed cooling process is not directly required by regulations, it is given credit in assessing the overall safety of individual plants and of the population of plants in the United States. The contribution made by feed-and-bleed cooling to the safety of plants needs to be better 43 l

The Honorable Lando W. Zech, Jr. June 7, 1988 established. It is regarded as a "last ditch" cooling mode that can be effective in some plants. Risk assessments are ambiguous about its importance. There is significant uncertainty about the reliability with which this process can actually be carried out in many plants, perhaps in most. In particular, there are questions about the flow capacity and reliability of the valves (usually power operated relief valves) essential to provide bleed flow and blowdown quenching capacity. In addition, the complex flow path and the effects of uncovering the core do not seem to be well understood for all plants. Research should be directed toward resolving the key uncertainties related to providing assured feed and bleed at plants that depend on the process for a margin of improved safety. (6) The LBLOCA, the design-basis accident for certain plant systems, should be reconsidered in view of the results of research on leak-before-break and the revision to General Design Criterion 4. Thermal-Hydraulic research will be necessary in support of this etfort. (7) The designs for so-called evolutionary LWRs and especially the " passive" LWR being developed by the Electric Power Research Institute and DOE, will require research by the NRC to confirm certain favorable characteristics being claimed. The DOE Advanced Reactor Severe Accident Program is not sufficient for this purpose. The NRC should use existing codes to review these designs so there is sufficient lead time to conduct more experimental or code development work, if necessary. (8) There is some uncertainty about applicability of the RELAP-5 code to BWRs and to LBLOCAs. This should be resolved. (9) Full documentation should be completed for the NRC codes that are maintained for active use. This should include not only user manuals but developmental assessment reports and "models and correlations" documents. Ideally, these would be published as NRC documents in the NUREG series to ensure widespread availability. (10) Thorough analyses have generally been made only for the initial period of reactor accidents such as LOCAs. Analyses of the follow-on transition to stable long-term cooling have been less comprehensive. We recommend that NRC determine whether a more systematic and complete study of the reliability of such tran-sitions should be undertaken. 44

The Honorable Lando W. Zech, Jr. ) i ACRS Members William Kerr, Harold Lewis, and Forrest Remick did not participate in the review of this matter. i Sincerely, I 00.aQ David A. Ward Acting Chairman I

References:

1. U.S. Nuclear Regulatory Commission, NUREG-1080, Volume 2: "Long-Range Research Plan FY 1986 - FY 1990," Office of Nuclear Regula- ) tory Research, August 1985. ) 2. U.S. Nuclear Regulatory Commission NUREG-1266, Volume 2: "NRC 1987," Office of Safety Research in Support of Regulation Nuclear Regulatory Research, May 1988. i 3. U.S. Nuclear Regulatory Commission, Draft NUREG-1252: " Thermal Hydraulic Research Program Plan," Office of Nuclear Regulatory Research, November 9, 1987. 4. U.S. Nuclear Regulatory Commission, NUREG-1236: "NRC Thermal Hydraulic Research Plan for B&W Plants," February 1988. 5. R. A. Dimenna, et al, Idaho National Engineering Laboratory: "RELAP-5/M002 Models and Correlations" (Draft Report), December 31, 1987. 6. D. R. Liles, et al, Los Alamos National Laboratory: " TRAC-PF1/ MODICorrelationsandModels"(DraftReport),providedtotheACRS in December 1987. l 45 i ______________N

WE:q'o UNITED STATES 8 NUCLEAR REGULATORY COMMISSION o a,E ADVISORY COMMITTEE ON REACTOR SAFEGUAf1DS -e g WASHINGTON, D. C. 20556 July 20,1988 Mr. Victor Stello, Jr. Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

COMMENTS ON THE NRC STAFF'S DRAFT SAFETY EVALUATION OF THE WESTINGHOUSE TOPICAL REPORT, WCAP-10924, " WESTINGHOUSE LARGE-BREAK LOCA BEST-ESTIMATE METHODOLOGY" During the 339th meeting of the Advisory Comittee on Reactor Safe-guards, July 14-16, 1988, we met with members of the NRC Staff and their consultants as well as representatives of the Westinghouse Electric Corporation and licensees of the Prairie Island and Point Beach nuclear l power. plants and reviewed the subject safety evaluation. Our Subcom-mittee on Thermal Hydraulic Phenomena met on May 27 and June 21, 1988 to discuss this matter. We also had the benefit of the documents listed as references for this letter. This review concerns an improved method for predicting the performance of the emergency core cooling system (ECCS) during a large-break loss-of-coolant accident (LOCA) in two-loop pressurized water reactor l (PWR) plants of Westinghouse design. The ECCS in these plants is different from those in the majority of PWR plants in that low-pressure injection flow is provided to the reactor vessel through special nozzles entering the upper plenum rather than through the cold leg piping. Several years ago, the NRC Staff recommended that certain thermal-hydraulic conditions unique to the upper plenum injection (UPI) config-uration were not adequately modeled in the ECCS codes being used. Accordingly, licensees at UPI plants were instructed to develop improved models, to demonstrate that these models provided compliance with the ECCS rule, and to use these models in future licensing submittals. Westinghouse has developed such an improved model for the utilities operating the Prairie Island and Point Beach nuclear power plants. That is the subject of the present review. Models for two other UPI plants are being developed elsewhere and are to be given a separate review. The improved analysis is provided by a code called WCOBRA/ TRAC that has been developed from existing codes TRAC-PD2 and COBRA-TF. COBRA-TF is the reactor vessel part of the analysis while TRAC-PD2 models the 47

Mr. Victor Stello, Jr. July 20,1988 overall system. COBRA /TF describes three flow fields in three dimen-sions. It thus has the capability of modeling thermal-hydraulic phenom-ena important. in the UPI configuration, including details of counter-current flows of vapor and liquid. In addition to the improved phenomenological modeling, the overall methodology presented in WCAP-10924 incorporates a so-called "best-estimate" approach to calculation of limiting plant parameters. The general approach described in the referenced NRC document, SECY-83-472, is used. The staff's review has concluded that the WCOBRA/ TRAC code provides adequate rrodeling to represent the UPI plant configuration for large-break LOCA analysis. They have also concluded that the best-estimate methodology, including allowances for uncertainty, adequately conform to the provisions outlined in SECY-83-472. However, they have not accepted the best-estimate methodology, as. presented, for other plants or for use with the proposed new ECCS rule when that becomes available. We find no reason to disagree with these conclusions of the staff. A cautionary word about the so-called "best-estimate" approach: we have previously expressed our approval of, and, in fact, applauded, this approach to analysis of reactor transients. This applies to both the 1 SECY-83-472 approach and the proposed new rule for large-break LOCA analysis. Best-estimate analysis, in this sense, has two parts: (1)a realistic analysis, with no purposeful biases, to provide a central estimate of the parameters of interest and (2) a conscious and explicit estimate of the margin that should be provided from this central esti-mate to achieve a desired level of confidence in conclusions to be drawn from the analysis. 1 In most practical engineering situations, including LOCA analysis, the I relationships are so complex and the data so sparse that mathematical i rigor in defining the desired confidence level and necessary allowance is impossible. However, the method is still of value even though it may involve what are largely engineering judgments about confidence level and the magnitude of allowances. The problem is that, too often, practitioners of the best-estimate analysis or users of the results describe their analysis and the results in terms that imply mathematical rigor and give an impression that statistical relationships have been developed with great precision, whereas, the actual data and methods of l analysis are approximate. Terms such as "95% confidence interval" are used when only a term such as "a reasonably high confidence level" is justified. In addition, distinctions among variability, uncertainty, ? l and confidence level are not observed, and statistical relationships are often used carelessly and inaccurately. We recommend that, in the future, the NRC Staff should involve professional statisticians in the review of these matters. 48

J l Mr. Victor Stello, Jr. July 20,1988 1 Also, there will be a greater technical challenge for the staff in reviewing best-estimate analyses compared with evaluation-model reviews carried out in the past. We believe that agency management will have to make a special effort to provide appropriate resources. We hope that our comrrents will prove useful. Sincerely, 3 i W. Kerr i Chairman i

References:

1. Memorandum dated July 12, 1988, f rom M. W. Hodges, NRC, to P.

Boehnert, NRC, transmitting draft " Safety Evaluation of the Westinghouse Electric Corporation Topical Report, WCAP-10924,

' Westinghouse Large-Break LOCA Best-Estimate Methodology'" (Proprietary) 2. U.S. Nuclear Regulatory Commission Staff Document, " Emergency Core Cooling System Analysis Methods," SECY-83-472, dated November 17, 1983 3. Westinghouse Electric Corporation, WCAP-10924-P: " Westinghouse large Break LOCA Best Estimate Methodology - Volume 1: Model Description and Validation," June 1986 (Proprietary); and WCAP-10924-P, Volume 2, Revision 1: " Application To Two-Loop PWRs Equipped With Upper Plenum Injection," April 1988 (Proprietary) 4. Westinghouse Electric Corporation: " Responses to NRC Questions on Westinghouse Large Break LOCA Best Estimate Methodology, WCAP-l 10924-P, Volume 1," October 1987 (Proprietary) l 5. Westinghouse Electric Corporation: WCAP-10924-P, Volume 2, Adden-i dum 1: " Responses to NRC Questions on WCAP-10924-P, Volume 2 (Addendum to Westinghouse Large Break LOCA Best-Estimate Method-ology, Volume 2: Application to Two-Loop PWRs Equipped With Upper { Plenum Injection)," April 1988 (Proprietary) 49 D

s p 28C / 'o,, UNITED STATES 8 NUCLEAR REGULATORY COMMISSION n ,I ADVlsORY COMMITTEE ON REACTOR SAFEGUARDS 7 o wAsmNGTON, D. C. 20665 j June 7, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

RULEMAKING ON EMERGENCY PLANNING AND PREPAREDNESS 2 During the 338th meeting of the Advisory Committee on Reactor Safeguards, June 2-4, 1988, we met with the NRC Staff to discuss the proposed rulemaking actions related to: (1) 10 CFR Parts 30, 40, and 70, " Emergency Preparedness l for Fuel Cycle and Other Material Licensees," and (2) 10 CFR Part 50, "Emer-gency Planning and Preparedness Requirements for Nuclear Power Plant Fuel Loading and Initial Low-Power Operations." The Comittee previously com-mented on this matter in a report dated July 17, 1985. Regarding these two activities, we offer the following coments: 1. We concur with the proposed rulemaking for fuel cycle and other material l licensees. This will codify via rulemaking what is now being addressed l through orders. The rule is limited in application and will require emergency planning for only those facilities which have a potential for significant accidental impact on public health. The total. number of such facilities is about 30, and we understand that all of them already Nye emergency plans that fulfill the requirements of the proposed rule. 2. We also concur with the proposed rulemaking that would require the develoriment of limited emergency plans for nuclear power plants during fuel loading and initial low-power operation. During these periods of time, the radioactive material source term is minimal and the proposed I degree of emergency preparedness appears commensurate with the risk. Before a plant would be licensed to operate at higher power levels, a full-scale emergency plan would be required. Sincerely, I W. Kerr Chainnan 51 ___-_______a

i l I UNITED STATES o,, / NUCLEAR REGULATORY COMMISSION n { 7,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS O WASHINGTON, D. C. 20555 May 10, 1988 Mr. Victor Stello, Jr. Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dear Mr. Stello:

SUBJECT:

FIRE RISK SCOPING STUDY In our July 16, 1986 letter to the Commission concerning fire protection issues, we urged reconsideration of the budget and manpower allocations to the fire-related portions of the NRC safety research program. In response, in a memorandum dated July 24, 1986, Chairman Zech recommended that the Staff work closely with the ACRS to assess further research needs and to consider what priority should be given to fire protection research. The Staff acted in January 1987 by initiating the Fire Risk Scoping Study at the Sandia National Laboratories (SNL), and we provided our views on the scope and direction of this Study in a report to the Commission dated August 10, 1987. During our 337th meeting, May 5-7, 1988, we met with representatives from the Office of Nuclear Regulatory Research and SNL to discuss the results and conclusions of the Fire Risk Scoping Study. This matter was considered by our Subcommittee on Auxiliary Systems during a meeting on fiarch 9, 1988. We also had the benefit of the document referenced. We were informed that the Staff is now considering what actions should be taken regarding the disposition of the recommendations resulting from the Study, and a decision is expected by the end of FY 1988. If some of l the asserted results survive deeper scrutiny, they could be important. Therefore, we recommend that the Staff evaluate the results and conclu-sions of the Study and decide on a course of action on a schedule which permits any high-priority research to be initiated in FY 1989. We wish to be kept informed of further developments, and we expect to provide coments after the Staff has identified its proposed plans. Sincerely, i W. Kerr Chairman j 53 l i l

h 1 L. i 1 1 Mr. Victor Stello. Jr.

Reference:

' Draft Report dated March 1988, Sandia National Laboratories, NUREG/CR-5088, SAND 88-0177, " Fire Risk Scoping Study: Investigation of' Nuclear Power Plant Fire Risk, Including Previously Unaddressed Issues" I 1 I i l l 54

I p C E: o UNITED STATES g NUCLEAR REGULATORY COMMISSION g j ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wAsmNGTON, D. C. 20666 April 12, 1988 ) The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

PROPOSED RESOLUTION OF USI A-47, " SAFETY IMPLICATIONS OF CONTROL SYSTEMS" -- ACRS COMMENTS During the 336th meeting of the Advisory Committee on Reactor Safe-guards, April 7-9, 1988, we discussed the NRC Staff's proposed resolu-tion of USI A-47, " Safety Implications of Control Systems." We prev-iously met with members of the NRC Staff and reviewed the proposed resolution during our 331st meeting on November 5-7, 1987. Our Sub-committee on Instrumentation and Control Systems met on October 29, 1987 and March 24, 1988 to consider this matter. We also had the benefit of the documents referenced. The proposed resolution of USI A-47 requires certain design modifica-tions or upgrades for many plants, technical specification modifications for all plants, and issuance of an information letter to all applicants and licensees. These matters are currently scheduled to be prblished for public comment in April 1988, and the final resolution is scheduled to be issued in April 1989. Although we agree with the proposed resolution of USI A-47 as it relates to steam generator and reactor vessel overfill, we believe that the scope of the issue has been unduly truncated. The problem description of USI A-47 in the last revision of the Aqua Book (NUREG-0606, "Unre-solved Safety Issues Sumary," now discontinued) gave a much broader issue for evaluation, with steam generator / reactor vessel overfill and overcooling transients identified only as a subtask. Other tasks included evaluation of all control system failures that have safety implications and evaluation of the effects of loss of power on control systems. It is not clear that these tasks were performed. We believe that they should be. When reconsidering the scope, we recommend including an evaluation of the safety implications of failures in nonsafety-grade control systems that result from common-cause external events such as earthquakes, fires, and other potentially far-reaching events such as high or mod-erate energy pipe breaks. Such events were not evaluated in USI A-47, and we do not believe that they are adequately treated elsewhere in the context of this USI. 55

i The Honorable Lando W. Zech, Jr. April 12, 1988 We recommend also that other events be included, such as the degradation or loss of control power or control air and the improper functioning of heating, ventilating, and air conditioning systems (particularly when j temperature-sensitive devices are in the affected environment). Our i questions to the Staff concerning these events met with less than satis- ) factory assurances that they have been considered in the program. ( In conjunction with these efforts, the Staff should develop its position on how similar events will be included when resolving USI A-17, " Systems Interactions in Nuclear Power Plants," which considers safety-grade protection systems and protective actions. Since such events are likely to affect more than one system or component, whether safety grade or not safety grade, it is necessary to show how the evaluations performed for A-17 and A-47 collectively cover the situation. We cannot agree that the Staff's recommendation.s constitute resolution of USI A-47 as originally defined. We recommend that the proposed resolution be issued as the resolution of an appropriately redefined generic issue and that the remaining concerns, as identified above, be included in a new generic issue, which need not necessarily be accorded USI status. Sincerely, W. Kerr Chairman

References:

1. U.S. Nuclear Regulatory Commission, Draft NUREG-1217, " Evaluation of Safety Implications of Control Systems in LWR Nuclear Power Plants," Technical Findings Related to Unresolved Safety Issue A-47, April 1987. 2. U.S. Nuclear Regulatory Commission, Draft NUREG-1218, " Regulatory Analysis for Proposed Resolution of USI A-47, Safety Implications of Control Systems," April 1987. 3. U.S. Nuclear Regulatory Commission, NUREG/CR-4265, "An Assessment of the Safety Implications of Control at the Calvert Cliffs 1 Nuclear Power Plant," Volumes 1 and 2 April 1986 and July 1986, respectively. 4. U.S. Nuclear Regulatory Commission, NUREG/CR-3958, " Effects of Control System Failures on Transients, Accidents and Core-Melt Frequencies at a Combustion Engineering Pressurized Water Reactor," ) March 1986. j I 56 I

l ) I The Honorable Lando W. Zech, Jr. April 12, 1988 5. U.S. Nuclear Regulatory Commission, NUREG/CR-4047, "An Assessment of the Safety Implications of Control at the Oconee 1 Nuclear Plant," March 1986. 6. U.S. Nuclear Regulatory Commission, NUREG/CR-4386, " Effects of Control System Failures on Traasients, Accidents, and Core-Melt Frequencies at a Babcock and Wilcox Pressurized Water Reactor," December 1985. 7. U.S. Nuclear Regulatory Commission, NUREG/CR-4387, " Effects of Control System Failures on Transients, Accidents, and Core-Melt Frequencies at a General Electric Boiling Water Reactor," December

1985, 8.

U.S. Nuclear Regulatory Commission, NUREG/CR-4385, " Effects of Control System Failures on Transients, Accidents, and Core-Melt Frequencies at a Westinghouse PWR," November 1985. 9. U.S. Nuclear Regulatory Commission, NUREG/CR-4326, " Effects of Control System Failures on Transients and Accidents at a 3-Loop, Westinghouse Pressurized Water Reactor," Volumes 1 and 2, August 1985 and October 1985, respectively. 10. U.S. Nuclear Regulatory Commission, NUREG/CR-4262, " Effects of Control System Failures on Transients and Accidents at a General Electric Poiling Water Reactor," Volumes 1 and 2, May 1985. l l l l 1 l 57 l

-p ue +8 'o UNITED STATES g 8 NUCLEAR REGULATORY COMMISSION o U E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS i r WASHINGTON, D. C. 20555 l Y %... / April 12, 1988 l The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, DC 20555

SUBJECT:

EFFECTIVENESS OF PROGRAMS RELATING TO GENERIC AND UNRESOLVED l SAFETY ISSUES - ACRS COMMENTS l In your memorandum of September 18, 1986, to then ACRS Chairman Ward, you requested that the ACRS advise the Commission on the effectiveness of programs that address unresolved safety issues (USIs) and generic issues (GIs). During a meeting of the ACRS with the Commission on December 11, 1986, we reviewed some of our activities in this area and i asked whether your concern was with the effectiveness of the overall process for dealing with these issues or with the extent to which the implementation of the resolution of GIs and USIs has increased safety. In response, you said that advice on both aspects was desired. The ACRS Subcommittee on Generic Items was assigned the task of developing information on this subject. The Subcommittee has held several meetings with representatives of the Office of Nuclear Regulatory Research (RES) and the Office of Nuclear Reactor Regulation (NRR) to review this matter. Items discussed by the Subcommittee during its review included: Description of the entire process dealing with GIs and USIs, includ-ing identification, prioritization, resolution, imposition, implemen-tation, and verification. Ranges of time involved for completion of each step in the process. Scopes of various issues and the way in which the scopes of individ-ual issues were defined in relation to other issues and to the broader issue of prevention of core damage or mitigation of its l consequences. 1 59 _ _______-__ w

The Honorable Lando W. Zech, Jr. April 12, 1988 i l Role of the NRR project managers in the imposition, implementation, and verification stages. Resolution process for four specific case histories. These included two USIs, A-46, Seismic Qualification of Equipment in Operating Plants, and A-49, Pressurized Thermal Shock, which required five to six years for resolution and resulted in requirements for significant hardware backfits to the affected plants, and two GIs, No. 20, Effects of Electromagnetic Pulse on Nuclear Power Plants, and No. 61 SRV Line Break Inside the BWR Wetwell Airspace of Mark I and II Containments, whose resolution required 13 to 33 months and resulted in no new requirements on operating plants. Description of the Safety Issues Management System (SIMS). In addition, the Subcommittee met with representatives from the Duke Power Company to learn and understand the process of implementation from the licensee's viewpoint, the role of the NRC Staff in this process, and the reasons. for delays in implementation. Specific case histories discussed were: USI A-24, Qualification of Class IE Safety-Related Equipment, USI A-44, Station Blackout, and USI A-46, Seismic Qualifica-tion of Equipment in Operating Plants. During our 336th meeting, April 7-9, 1988, we discussed the information l gathered by the Subcommittee. Our specific comments are presented below for each phase of the process. Identification The identification of GIs is completely open to anyone: NRC staff, industry, ACRS, or members of the public. This is as it should be. However, since every issue identified must proceed at least through the prioritization step, identification must be subject to some discipline. The procedures for identification provided in Attachment I to RES Office Letter No. I dated December 3,1987, are appropriate for use by the NRC Staff and the ACRS. We believe it especially important that these guidelines be observed by task forces investigating operational i incidents. We note, for example, that the feedwater failure event at i the Davis-Besse Nuclear Power Station engendered 34 new GIs. The fact that only 7 of these eventually were judged of High or Medium priority i suggests a lack of suitable discipline in evaluating the significant j issues relating to this event. j 60

) The Honorable Lando W. Zech, Jr. April 12, 1988 Prioritization Each identified GI is assessed and assigned a priority by the NRC Staff. This is an essential step. Since fewer than half of the issues iden-tified are subsequently deemed worthy of resolution (assigned priorities of High or Medium), it is important that issues of low or negligible safety significance be weeded out in order that the limited resources can be applied to the more important issues. It is worth noting that priorities are based primarily on safety significance; the cost-benefit ratio is a clearly secondary consideration at this stage. We review and comment periodically on the priorities assigned to the various issues. Although occasionally we disagree with the Staff's ranking, we believe that overall the priorities have been appropriate. The average time required to assign a priority to a GI is about six months. We do not consider this unreasonable in view of the importance i of this step and in view of the fact that a significant portion of this time is required for peer review. If you wish to reduce the backlog of issues waiting to have priorities assigned, it is necessary only to provide additional resources. Resolution We have reviewed the resolution of essentially all of the USIs and some, but by no means all, of the GIs. We have provided you with letters l indicating our agreement or disagreement with the resolution of each issue we have reviewed. Where we have disagreed, the Staff has, ir, most cases, attempted and sometimes succeeded in resolving our concern. In general, we believe that the resolution of GIs and USIs has been carried out by the Staff in a professional and effective manner. The Staff in its briefing of the Conrnission on October 21, 1987 proposed various means to reduce the time required for resolution of an issue. Some of these were managerial or procedural but others were technical. The Staff proposed to explore whether some issues could be combined and resolved as a package or via the Individual Plant Examination (IPE) portion of the Severe Accident Policy. Whether this will accelerate resolution we do not know, but it addresses some of our concerns about the definition of scope for issues that will be elaborated on below. We are not sure that the technical resolution of complex safety issues can or should be accomplished much more rapidly than is now the case. Better management may provide greater continuity to the effort and may reduce the time for review and concurrence. As in the case of pri-orities, the backlog can be reduced by assigning more resources. 61 _J

The Honorable Lando W. Zech, Jr. April 12, 1988 Imposition, Implementation, Verification The decision to impose the resolution of an issue on a particular plant is made by NRR, the licensee is responsible for its implementation, and either NRR or Regional Staff verify its implementation. The Staff in its briefing of the Commission has recognized that there I frequently has been poor definition of what the licensees are expected to do and on what schedule, and has proposed improvements in the resolu-tion and imposition packages to remedy this situation. We recomend that every effort be made to make such improvements. Moreover, we believe that greater involvement of the affected licensees, through Owners Groups or other means, at the resolution stage, would contribute greatly to improved understanding by the licensees of what is expected or required. The Staff said also that they would explore whether it would be benefi-cial to combine the implementation of various GI resolution packages. We endorse this concept as one that might contribute not only to more timely implementation but also to improved safety. Each issue is now prioritized and resolved as if it were the only one outstanding. As a result, the improvement in safety provided by implementation of issue C after issues A and B have been implemented may be less than would have been the case if issue C had been implemented alone. Considering these three issues together might lead to an entirely different, faster, and more effective fix. In other words, application of the principles and procedures proposed for the Integrated Safety Assessment Program (ISAP) II becomes more and more desirable. The representatives from Duke Power Company offered several suggestions to improve the implementation of resolved generic issues. Some of these are mentioned below. It seems that a major reason for delays in implementation is ineffective communication between the Staff and the licensees. The resolution of an issue is so worded that the licensee frequently does not know what the Staff wants or will accept. In some cases, because of change of re-viewers or simply the passage of time, the Staff's requirements change. Attempts by licensees to act quickly, or even to anticipate the require-4 ments, most frequently have been unsuccessful, requiring redesign or { rework to satisfy the. Staff's ultimate requirements. Some experience i suggests that implementation has been achieved more easily and more j rapidly when the industry has been involved with the resolution, and 62 -____-_____-_-__-_-_-_-_-_A

The Honorable Lando W. Zech, Jr. April 12, 1988 l l thus understands more clearly what is required for implementation. It is equally important to note that implementation will never be accom-plished " overnight." Even if only a Technical Specification or proce-dure change is needed, time is required to write and review the proce-dure, to obtain approval internally and from the NRC Staff (for a Technical Specification change), to incorporate it into the operator training programs, and to train the operators in its use. A substantial portion of the time required to implement a change, either hardware or procedural, is that required for the NRC Staff to review and respond to the licensee's proposal. In the examples looked at by the Subcommittee, it was not unusual to find a turnaround time of as much as one year. Scope of Issues On several occasions, we have raised questions about the scope of individual GIs and USIs; some too narrow, some too broad. The Staff has been concerned that if an issue is defined broadly, such as systems interactions, it may not be resolved within a reasonable time. Although this position is understandable, we are concerned that it represents a reaction by the Staff to pressures from the Commission and the Congress for timely resolution of issues rather than the ultimate objer'ive of improving safety. We do not believe that broadly defined is_ are inherently undesirable. Nor do we believe that such issues are unman-ageable. We see no reason why broad issues cannot be subdivided for resolution without losing sight of the ultimate objective. For example, USI A-47, Safety Implications of Control Systems, is clearly a subset of USI A-17. Systems Interactions, although it was not identified as such. In an even broader view, USI A-17 can be considered a subset of the issues involved in the Severe Accident Policy. We note, however, that the USIs and unresolved GIs have been singled out in the Severe Accident Policy Statement as separate requirements. We believe that achievement of improved safety, defined as reducing the probability of severe core damage and release of radioactive materials to the environment, is not well served by the compartmentalization that we now see. The identification and definition of a collection of USIs and GIs, all subject to outside pressures for speed of resolution, may be a way to show progress, but there is little assurance that cost-effective and needed improvements will be achieved. We believe that the Staff understands this, but is so committed to refining the existing process that it has no incentive to undertake the larger and more formidable task of developing and implementing a more comprehensive and more holistic approach. 63

The Honorable Lando W. Zech, Jr. April 12, 1988 1 Increased Safety Has the implementation of the resolution of USIs and GIs reduced risk? We cannot answer this question on the basis of facts. We did not determine the extent to which all of the resolved issues have been implemented for all of the operating plants. We have seen no proba-bilistic safety assessments that compare risks before and after the implementation of resolved issues. Although the assignment of pri-orities and the regulatory analyses accompanying the resolutions include estimates of risk reduction, these are calculated for each issue sepa-rately and do not consider the combined or cumulative effects of all issues. Nor do these analyses necessarily consider possible effects adverse to safety. Lacking hard data, the answer to this question is a matter of judgment. The Staff believes that their efforts have reduced risk. We must agree, since in most cases we have endorsed the resolution reached by the Staff. The representatives of Duke Power Company were asked to comment on this question. They thought that the following had contributed to safety: standby shutdown facilities resulting from Appendix R re-quirements, provisions to avoid failure of low-pressure systems con-nected to high-pressure systems (Event V), improvements in reactor trip switchgear reliability, improved training, and improved emergency operating procedures. On the other hand, they thought that only ques-tionable improvement to safety had resulted from: ATWS requirements, reactor vessel water level instrumentation, portions of Regulatory Guide 1.97, and portions of the requirements for equipment qualification. Overall, we believe that most of the USIs have improved safety where they have been implemented end that many of the GIs have also. We believe that a more comprehensive approach to resolution and a more integrated approach to implementation, both based on probabilistic safety assessments, would make it much easier to determine that risk had been reduced in a cost-beneficial Wianner. This suggests the desirabil-ity of implementing the Severe Accident Policy in a more integrr.ted manner, rather than dealing separately with USIs, GIs, and IPE results. It also emphasizes the benefits that might be obtained from implementing plant fixes in accordance with the principles of ISAP II. Additional remarks by ACRS Member Harold W. Lewis are presented below. Sincerely, W. Kerr Chairman 64

The Honorable Lando W. Zech, Jr. April' 12, 1988 1 i ) Additional Remarks By ACRS Member Harold W. Lewis The Committee is comfortable with the six months necessary to set priorities (there is no such word as prioritize), and accepts the position that the only solution to the backlog question is the assign-ment of additional resources. This glosses over the fact that most of the time is spent in interoffice coordination and peer review and that it is not uncommon for additional resources (see the works of L. Peter or F. Brooks) to slow down a process. I do not prepose a global solu-tion, but note only that this appears to be one of the many NRC activ-ities that suffer in the effort to forge an agency out of semi-autonomous units. 65

I

1. s%q'o UNITED STATES g

8 NUCLEAR REGULATORY COMMISSION o ,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Q 4 WASHINGTON, D, C. 20555 June 7, 1988 MEMORANDUM FOR: Victor Stello, Jr. Executive Director rations FROM: R o . Fraley Executive Director, ACRS

SUBJECT:

PROPOSED PRIORITY RANKINGS OF GENERIC ISSUES: FIFTH GROUP During the 338th meeting of the Advisory Comittee on Reactor Safeguards, June 2-4, 1988, the members reviewed the adequacy of the proposed priority rankings for a group of Generic Issues identified in the attached Table A, and their comments are contained in the following attachments. Attachment I lists those issues for which the ACRS agrees with the priority rankings proposed by the NRC Staff. includes those issues for which the ACRS agrees with the priority rankings proposed by the NRC Staff, but has coments. ' Attachment 3 identifies the Generic Issue for which the ACRS disagrees with the NRC Staff's proposed priority ranking along with the reasons therefor. The members have requested that the NRC Staff provide written responses to the comments identified in Attachments 2 and 3. The ACRS will continue its review of the adequacy of the proposed priority rankings for additional Generic Issues when they become avail-able. Attachments: As Stated 67

TABLE A GENERIC ISSUES REVIEWED BY THE ACRS DURING THE 338TH, JUNE 2-4, 1988, MEETING GENERIC PRIORITY RANKING ISSUE TITLE PROPOSED BY THE REFERENCE NUMBER NRC STAFF DOCUMENT 88 Earthquakes and Emergency RESOLVED Memorandum from Plar.ning E.S. Beckjord, dated October 6, 1987 106 Piping and the Use of MEDIUM Memorandum from Highly Combustible Gases E.S. Beckjord, dated in Vital Areas November 3, 1987 113 Dynamic Qualification HIGH Memorandum from Testing of Large Bore E.S. Beckjord, dated Hydraulic Snubbers July 2, 1987 125 Davis-Besse Loss of All Feedwater Event of June 9, 1985 - Long-Term Actions 125.1.1 Availability of the Shift DROP Memorandum from Technical Advisor (Thesafetyconcern E.S. Beckjord, dated of this Issue has July 2, 1987 been addressed by licensees in response to TMI Action Plan Item I.A.1.1, " Shift Technical Advisor") 125.I.3 SPDS Availability MEDIUM Memorandum from E.S. Beckjord, dated May 6, 1988 68

Table A (Cont'd) GENERIC PRIORITY RANKING ISSUE TITLE PROPOSED BY THE REFERENCE NUMBER NRC STAFF DOCUMENT 125.I.4 Plant Specific Simulator DROP Memorandum from (The safety concern H.R. Denton, dated of this Issue has February 20, 1987 been addressed in the resolution of 1NI Action Plan ItemI.A.4.2(4), " Review Simulators for Conformance to Criteria") 125.I.6 Valve Torque Limit and DROP Memorandum from Bypass Switch Settings (The safety concern E.S. Beckjord, dated of this Issue has December 31, 1987 been addressed by licensees in response to IE Bulletin 85-03 and by the Staff in the resolution of TMI Action Plan Item II.E.6.1, " Test Adequacy Study") 125.I.7.a Recover Failed Equipment DROP Memorandum from (The safety concern E.S. Beckjord, dated of this Issue has December 31, 1987 been addressed in the resolution of TNI Action Plan Items I.A.2.2, " Training and Quali-fication of Opera-tions Personnel," andI.A.2.6(5), " Develop Inspection Procedures for Train-ing Program," and Human Factors Issue HF 2.1, " Evaluate IndustryTraining") 69 l

Table ~A(Cont'd) GENERIC PRIORITY RANKING ISSUE TITLE ' PROPOSED BY-THE REFERENCE NUMBER NRC STAFF -DOCUMENT 125.I.7.b Realistic Hands-on DROP-Memorandum from Training (The safety concern H.R. Denton, dated of this Issue is March 10, 1987 being addressed by the Policy Statement i on Training.and Qualifications,and by the work related to the resolution of Human Factors Issue j HF 3.1, " Develop Job Knowledge Catalog") l 125.I.8 Procedures and Staffing DROP Memorandum from l for Reporting to NRC (Thesafetyconcern E.S. Beckjord, dated j Emergency Response Center of this issue has June 3, 1987 been addressed in the resolution of .TMI Action Plan Item III.A.3.4, " Nuclear 2 Data Link") l 125.11.1 Need for Additional Actions on AFW Systems j l' 125.II.1.a Two-Train AFW Unavail-DROP Memorandum from ability (The safety concern H.R. Denton, dated of this issue has October 8, 1986 been addressed in the resolution of I Generic Issue 124, { " Auxiliary Feed-water System Reli-ability") 125.II.1.b Review Existing AFW HIGH Memorandum from Systems for Single H.R. Denton, dated Failures October 8, 1986 70 I i i i l

I 1 l h i TableA(Cont'd) L I i I. GENERIC PRIORITY RANKING ISSUE TITLE PROPOSED BY THE REFERENCE NUMBER-NRC STAFF DOCUMENT 125. II.1.'c .NUREG-0737 Reliability DROP Memorandum from Improvements H.R. Denton, dated October 8, 1986 125.II.1.d AFW Steam and Feedwater DROP Memorandum from Rupture Control System / (Thesafetyconcern H.R. Denton, dated j ICS Interactions in B&W of this Issue has October 8, 1986 Plants been addressed in the resolution of Generic Issue 124) 125.11.2 Assess the Adequacy of DROP Memorandum from Existing Maintenance (The safety concern E.S. Beckjord, dated Requirements and Their of this issue has June 11, 1987 Impact on the Resulting been addressed in-Reliability of Safety-the resolution of Related Systems Human Factors issue HF 8. " Maintenance and Surveillance Program") l 125.II.5 Thermal-Hydraulic Effects DROP Memorandum from of Loss and Restoration (This safety concern E.S. Beckjord, dated of Feedwater on Primary of this Issue has June 17, 1987 l System Components been addressed in the resolution of USI A-49, " Pres-surized Thermal Shock") 125.11.6 Reexamine PRA-based DROP Memorandum from Estimates of the Likeli-(Thesafetyconcern H.R. Denton, dated hood of a Severe Core of this-Issue has March 26, 1987 Damage Accident Based been addresed in the on Loss of All Feedwater resolution of USI A-45, " Shutdown Decay Heat Removal Requirements," and Generic Issue 124) 71

Table A (Cont'd) $ GENERIC PRIORITY RANKING ISSUE TITLE PROPOSED BY THE REFERENCE NUMBER NRC STAFF DOCUMENT 125,II.8 Reassess Criteria for DROP Memorandum from Feed-and-Bleed Initiation (The safety concern H.R. Denton, dated of this Issue has March 26, 1987 been addressed in the resolution of Generic Issue 122.2, "Initi-ating Feed-and-Bleed") 125.11.10 Hierarchy of Impromptu DROP Memorandum from Operator Actions (The safety concern H.R. Denton, dated of this Issue has February 21, 1987 been addressed in the resolution of Human Factors Issue HF 4.4, " Guidelines for Upgrading Other Procedures") 125.11.12 Adequacy of Training DROP Memorandum from Regarding PORV Operation (Ths safety concern H.R. Denton, dated of this Issue has March 10, 1987 been addressed in the resolution of Human Factors Issue HF 3.1, " Develop Job Knowledge Catalog") 125.11.13 Operator Job Aids DROP Memorandum from (Thesafetyconcern E.S. Beckjord, dated of this Issue has March 1, 1988 been addressed ty the INP0 Training 4 Accreditation Pro-gram which was endorsed in March i 1 1985 by the Com-mission Policy Statement on Train-ing and Qualifica-1 tion of Nuclear Power Plant Person-nel) I 72

l Table'_A(Cont'd) ' GENERIC-PRIORITY RANKING i 1SSUE. TITLE PROPOSED BY THE REFERENCE NUMBER NRC STAFF DOCUMENT .I -126 Reliability of PWR LICENSING ISSUE Memorandum from. Main Steam Safety (Resolved) E.S. Beckjord, dated Valves March 10, 1988 127 Maintenance and Testing LOW Memorandum from of. Manual Valves in E.S. Beckford, dated Safety-Related Systems June 1, 1987 128 Electrical Power HIGH Memorandum from Reliability T.P. Speis, dated November 28,1986, and Supplement 6 to NUREG-0933, "A Prioritiza-tion of Generic Safety Issues," dated March 1987 130-Essential Service Water HIGH Memorandum from Pump Failures at Multi-H.R. Denton, dated plant Sites March 10, 1987 133 Update Policy Statement LICENSING ISSUE Memorandum from on Nuclear Power Plant H.R. Denton, dated Staff Working Hours February 25, 1987 134 Rule on Degree and HIGH Memorandum from Experience Requirement. H.R. Denton, dated February 20, 1987 135 Steam Generator and MEDIUM Supplement 7 to NUREG-Steam Line Overfill 0933, "A Prioritiza-tion of Generic Safety Issues," dated March 1988 i - 136 Storage and Use of LICENSING ISSUE Memorandum from Lar9e Quantities of 'E.S. Beckjord, dated 1 Cryogenic Combust-March 8, 1988 ibles on Site 73 1

1 i ATTACHMENT 1 LIST OF GENERIC ISSUES FOR WHICH THE ACRS AGREES WITH THE l -PRIORITY RANKINGS PROPOSED BY THE NRC STAFF l l F GENERIC ISSUE NO. TITLE 88 Earthquakes and Emergency Planning 106 Piping and the Use of Highly Combustible Gases in Vital Areas 113 Dynamic Qualification Testing of Large Bore Hydraulic Snubbers 125.I.1 Availability of the Shift Technical Advisor l F 125.I.3 SPDS Availability 125.I.4 Plant Specific Simulator 125.I.6 Valve Torque Limit and Bypass Switch Settings 125.I.7.a Recover Failed Equipment 125.I.7.b Realistic Hands-on Training 125.I.8' Procedures and Staffing for Reporting to NRC Energency Response Center 125.II.1.a Two-Train AFW Unavailability 125.II.1.b Review Existing AFW Systems for Single Failures 125.II.1.c NUREG-0737 Reliability Improvements 125.II.1.d AFW Steam and Feedwater Rupture Control System /ICS Interactions in B&W Plants 125.11.2 Assess the Adequacy of Existing Maintenance Require-ments and Their Impact on the Resulting Reliability l of Safety-Related Systems 125.11.5 Thermal-Hydraulic Effects of Loss and Restoration ci Feedwater on Primary System Components 74

Generic Issues - Attachment 1 125.11.6 Reexamine PRA-based Estimates of the Likelihood of a Severe Core Damage Accident Based on Loss of All Feedwater 125.II.8 Reassess Criteria for Feed-and-Bleed Initiation 125.11.10 Hierarchy of Impromptu Operator Actions 125.II.12 Adequacy of Training Regarding PORY Operation 125.11.13 Operator Job Aids 126 Reliability of PWR Main Steam Safety Valves 128 Electrical Power Reliability 130 Essential Service Water Pump Failures at Multi-Plant Sites 133 Update Policy Statement on Nuclear Power Plant Staff Working Hours 136 Storage and Use of Large Quantities of Cryogenic Combustibles on Site 75

J ATTACHMENT 2 GENERIC ISSUES FOR WHICH THE ACRS AGREES I WITH THE PROPOSED PRIORITY RANKING BUT WITH COMMENTS j 3 Generic Issue No: 127

Title:

Maintenance and Testing of Manual Valves in Safety-Related Systems Priority Ranking PFo~ posed By The NRC Staff: LOW ACRS Comments: The ACRS agrees with the proposed priority ranking for this issue, but offers the comment given below. Certain manual valves which may or may not be located in safety-related systems are identified in safety analyses or emergency procedures for manipu-lation during low-probability situations such as transients or accidents involving multiple component or system failures. There should be a requirement for surveillance and testing of such manual valves. This could be covered under a new generic issue or included as an extension of the severe accident program. Generic Issue No: 135

Title:

Steam Generator and Steam Line Overfill Priority Ranking Proposed By The NRC Staff: MEDIUM ACRS Coments: The ACRS agrees with the proposed priority ranking for this issue, but offers the connent given below. With the sudden, complete, and unexpected steam generator tube rupture event that occurred at the North Anna Nuclear Plant on July 15, 1987, it would appear that another mechanism for initiating an overfill scenario has been discovered. This would indicate that the mechanisms for initiating steam generator and steam line overfill are not yet completely understood. The Staff should consider such mechanisms under this Generic Issue, and should recognize that new operating experience could change current expectations of the frequency of steam generator and steam line overfill events. 76

ATTACHMENT 3 GENERIC ISSUE FOR WHICH THE ACRS tISAGREES WITH THE PRIORITY RANKING PROPOSED BY THE NRC STAFF Generic Issue No: 134

Title:

Rule on Degree and Experience Requirement Priority Ranking Proposed By The NRC Staff: HIGH ACRS Recommendation: DROP Reasons: In its report of August 12, 1987 to the Commis-sion regarding degree requirements for senior operators, the ACRS commented: A number of Job and Task Analyses (JTAs) have been performed by licensees as part of the conversion to performance based training; analysis of these JTAs has not shown that a college degree is necessary for Senior Reactor Operators (SP.0s) to perform the tasks of their jobs to ensure safety of plant operations. A Peer Advisory Panel appointed by the Commission came to the same conclusion in 1982 and recommended against a degree requirement for SR0s. We have not been informed of any technical rationale for requiring a degree for SR0s at nuclear power plants; we conclude, therefore, that a degree requirement for all SR0s is primarily a policy issue. Based on the above, the ACRS believes that this issue should be dropped from further consideration. l 77 I -_--_a

• Ec%(o UNITED STATES o8'-

'g-NUCLEAR REGULATORY COMMISSION o 5 . E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o WASHINGTON, D. C. 20555 s...../ August 16, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

PROPOSED RESOLUTION OF USI A-17, " SYSTEMS INTERACTIONS IN NUCLEAR POWER PLANTS" During the 340th meeting -of the Advisory Committee on Reactor Safe-guards, August 11-13, 1988, we discussed with members of the NRC Staff the proposed resolution of USI A-17, " Systems Interactions in Nuclear Power Plants." This matter was considered also during our 338th meet-ing, June 2-4, 1988. Our Subcommittee on Auxiliary Systems discussed this matter with representatives from the NRC Staff and Oak Ridge l National Laboratory during its meeting,on August 10, 1988. During the review, we had the benefit of the documents referenced. Our most recent comments on this subject were provided to you in' a report dated May 13, 1986. The systems interaction issue was first identified as a safety concern by the ACRS in 1974, and designated as a USI in 1978. Since then, there has been considerable difficulty in defining the issue and establishing its scope. As now defined by the Staff, USI A-17 is limited to: " Actions or inactions (not necessarily failures) of various systems (subsystems, divisions, trains), components, or structures resulting from a single credible failure within one system, component, or struc-ture and propagation to other systems, components, or structures by inconspicuous or unanticipated interdependencies...." If such a systems interaction is found to be undesirable from the safety viewpoint, it is called an adverse systems interaction. We interpret this definition to mean that an initiating event such as an equipment failure is required before a systems interaction can take place. The initiating event may cause many things to happen including the failure of other equipment to function. Most of these resulting failures will not be considered systems interactions because expected failure of equipment to function or to malfunction in a particular way j is an anticipated occurrence. Only if an occurrence involves incon-spicuous or unanticipated interdependencies or failure modes does it become a systems interaction, and it is an adverse systems interaction only if it is found to be undesirable (i.e., adverse to plant safety). 79

The Honorable Lando W. Zech, Jr. August 16, 1988 t 1 I -The proposed resolution does not encompass the full spectrum of poten-tial systems interactions that have been identified. Some of the omissions are the result of assumptions and limitations imposed by the Staff in order to achieve resolution. Our remaining concerns are expected to be addressed in the Multiple System Response Progrra (MSRP). j Although the proposed resolution of USI A-17 does not represent a comprehensive, and probably not a final, resolution of our concerns about systems interactions, we believe that the potential for a con-tinuing effort through the MSRP is a step in the right direction. In addition, and equally important, systems interactions, some of which may be adverse to safety, will continue to be revealed by operating expe-rience in existing plants. These should be evaluated by the Staff as they occur, and the lessons learned incorporated into the requirements I and practices of.the agency. In summary, we believe that the proposed resolution has a real potential to reduce risk. Since the systems interactions issue is so comprehen-sive, it is unlikely that it will ever be " resolved" in'the sense that all. adverse systems interactions will be found and corrected. We are willing to accept the proposed resolution of USI A-17. This acceptance, however, is based on the expectation that the efforts in the MSRP and the search for systems interactions in operating experience will be continued. We point out, however, that the proposed resolution does not address systems interactions for future plants, especially those plants in-volving design certification based on conformance with the EPRI re-quirements for advanced light water reactors. These requirements commit only to conformance with the resolution of USI A-17, which, as we have pointed out above, does not constitute a truly comprehensive resolution of the full spectrum of our concerns regarding systems interactions. The proposed resolution of USI A-17 represents a useful step in the direction of reducing plant risk due to adverse systems interactions for the present generation of plants. We recommend that it be issued for public comment. Af ter the public comment period, we expect to review the proposed final resolution together with the public comments. Sincerely, W. Kerr Chairman 80

i The. Honorable Lando W. Zech, Jr. August 16, 1988 )

References:

1. U.S. Nuclear Regulatory Commission, Draft NUREG-1174, " Evaluation of Systems Interactions in Nuclear Power Plants," Technical Find-ings Related to Unresolved Safety Issue A-17, April 1987 2. U.S. Nuclear Regulatory Commission, Draft NUREG-1229, " Regulatory ) Analysis for Proposed Resolution of USI A-17, Systems Interactions in Nuclear Power Plants," May 1987 3. Draft Generic Letter, with Attachments, from NRR to All Holders of Construction Permits or Operating Licenses, regarding Resolution of Unresolved Safety Issue A-17 " Systems Interactions in Nuclear Power Plants" (undated) i 4. Draft Letter Report,." Concerns Regarding Resolution of a Number of Specific Regulatory Issues," prepared by Nuclear Operations Analy-L sis Center, Oak Ridge National Laboratory, dated May 29, 1988 (internal use only) 81 _-_________m

p nog .f o UNITED STATES 8j,s(b/,j NUCLEAR REGULATORY COMMISSION IT d-A-r ADVISORY COMMITTEE ON REACTOR SAFEGUARDS k i'l .[ WASHINGTON, D. C. 20555 s~+, e... August 16, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

PROPOSED RESOLUTION OF USI A-47, " SAFETY IMPLICATIONS OF CONTROL SYSTEMS" During the 340th meeting of the Advisory Committee on Reactor Safe-guards, August 11-13, 1988, we discussed the NRC staff's proposed resolution of USI A-47, " Safety Implications of Control Systems." This was also the subject of a meeting of our Subcommittee on Regulatory Policies and Practices on August 10, 1988. We also had the benefit of the documents referenced. We had previously discussed the subject during our 336th meeting, April 7-9, 1988, and wrote you a letter dated April 12,1988, " Proposed Resolution of USI A-47, ' Safety Implications of Control Systems - ACRS Comments.'" In our letter of April 12 we observed that the scope of the staff work had been " unduly truncated," that there were many relevant subjects omitted, and that the relationship to other USIs was unclear. We therefore concluded that the staff recommendations did not constitute resolution of USI A-47, as originally defined, and recommended that the problem be dealt with by redefining the USI, issuing the proposed resolution as relevant to the newly defined objective, and including the unresolved issues in "a new generic issue, which need not necessarily be accorded USI status." We commend the staff for its effort to expand its list of the kinds of threats that can be posed by control system failures, though the work remains necessarily and substantially incomplete. Some of the specifics of the omissions noted in our April letter have been addressed, and others have been assigned to the Multiple System Response Program, for which priorities are yet to be set. We continue to recommend, as we did in April, that this " resolution" be issued, but without any pretense that it is a resolution of the total original concern. We do not wish here to provide a detailed list of 83 l M __._.__m__

The Honorable Lando W.-Zech, Jr. August 16, 1988 . things that ought to be fixed, since that would involve us too heavily in an iterative process with the staff, inappropriate for an advisory committee to the Commission, but will supply some examples of the sorts of issues that require further work. The procedure used by the staff was, in simplified form, to postulate the failure of an element of the control system, and to perform a Failure Modes and Effects Analysis-(FMEA), until one' reaches a point.in the sequence at which further development is arrested _by a safety-grade protective system. At that point, the sequence is presumed terminated. We believe that this takes the regulatory term " safety-grade" too literally. The ' failure probability of a safety-grade system is not zero. For the most part, as is common in FMEA studies, only failures are con-L sidered, with the probability of degradation or misbehavior barely treated. Such questions as the extent to which degradation of a control system can provide incorrect or misleading information to an operator were left untouched, though they were primary factors in our conclusions that control systems represent an important issue. There are different sorts of failures possible in these areas, and the complexity of possible responses to small disturbances of an electronic system, especially a computer system, is far greater than other sorts of failures. This area of potentially great significance remains largely unexplored. We note parenthetically that the NRC has not made any substantial effort to strengthen its staff capabilities in this in-creasingly important area. One cannot, therefore, criticize the present. staff. During.1985 the staff issued Regulatory Guide 1.152, " Criteria for 4 Programmable Digital Computer Software in Safety-Related Systems of Nuclear Power Plants," which endorses an ANSI /IEEE-ANS standard. In the Discussion Section in that guide, the staff encouraged the application of. this technology "if such advanced technology serves to enhance safety." Thus, the acceptability would appear to require some measure of'the new threats that are the inevitable accompaniment of new systems 'to match against the benefits. The staff recognized in the Regulatory Guide. "the unique nature of programmable digital computer systems." In particular, the Regulatory Guide observed that computers are "more vulnerable to subtle failure modes and unauthorized manipulation." This unique feature is not part of the USI A-47 effort. We think it is important. These concerns do not impel us to change the recommendations in our letter to you of April 12, 1988, as stated in paragraph 2, above. However, we feel it important that the NRC undertake to promptly and 84

l l J The Honorable Lando W. Zech, Jr. August 16, 1988 systematically broaden its area of expertise to encompass new and increasingly _ important technological trends. Sincerely, W. Kerr Chairman

References:

1. U.S. Nuclear Regulatory Commission, Draf t NUREG-1217, " Evaluation of Safety Implications of Control Systems in LWR Nuclear Power Plants," Technical Findings Related to Unresolved Safety Issue A-47, April 1987. 2. U.S. Nuclear Regulatory Commission, Draft NUREG-1218, " Regulatory Analysis for Proposed Resolution of USI A-47, Safety Implications of Control Systems," April 1987. 3. U.S. Nuclear Regulatory Commission, NUREG/CR-4265, "An Assessment of the Safety Implications of Control at the Calvert Cliffs 1 Nuclear Power Plant," Volumes 1 and 2 April 1986 and July 1986, respectively. 4. U.S. Nuclear Regulatory Commission, NUREG/CR-3958, " Effects of Control System Failures on Transients, Accidents and Core-Melt Frequencies at a Combustion Engineering Pressurized Water Reactor," March 1986. 5. U.S. Nuclear Regulatory Commission, NUREG/CR-4047, "An Assessment of the Safety implications of Control at the Oconee 1 Nuclear Plant," March 1986. i 6. U.S. Nuclear Regulatory Commission, NUREG/CR-4386, " Effects of Control System Failures on Transients, Accidents,- and Core-Melt Frequencies at a Babcock and Wilcox Pressurized Water Reactor," December 1985. 7. U.S. Nuclear Regulatory Commission, NUREG/CR-4387, " Effects of Control System Failures on Transients, Accidents, and Core-Melt Frequencies at a General Electric Boiling Water Reactor," December 1985. 8. U.S. Nuclear Regulatory Commission, NUREG/CR-4385, " Effects of Control System Failures on Transients, Accidents, and Core-Melt Frequencies at a Westinghouse PWR," November 1985. 9. U.S. Nuclear Regulatory Commission, NUREG/CR-4326, " Effects of Control System Failures on Transients and Accidents at a 3-Loop, Westinghouse Pressurized Water Reactor," Volumes I and 2, August 1985 and October 1985, respectively. 85 a

1 The Honorable Lando W. Zech, Jr. August 16, 1988 10. U.S. Nuclear Regulatory Commission, NUREG/CR-4262, " Effects of Control System Failures on Transients 'and Accidents at a General Electric Boiling Water Reactor," Volumes ~1 and 2, May 1985.

11. Letter from Victor Stello', Jr., Executive Director for Operations

.to William Kerr, Chairman, ACRS,

Subject:

"ACRS Coninents on Proposed Resolution of USI A-47, ' Safety. Implications of Control Systems,'" dated May 20 1988 1 l L 86

o UNITED STATES '/ g 8 NUCLEAR REGULATORY COMMISSION o B ,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS f W ASWNGTON, D. C. 20656 September 14, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

PROPOSED RESOLUTION OF GENERIC ISSUE 99, " IMPROVED RELI-ABILITY OF RHR CAPABILITY IN PWRs" During the 341st meeting of the Advisory Committee on Reactor Safe-guards, September 8-10, 1988, we considered a proposal by the NRC staff that would serve as part of the resolution of Generic Issue 99, " Improved Reliability of RHR Capability in PWRs." Our subcommittee on Decay Heat Removal Systems previously discussed this issue during a meeting on July 27, 1988. We also had the benefit of presentations by the NRC staff and of the documents referenced. Generic Issue 99, as originally posed, addressed concerns about the possible failure of core cooling that could result during shutdown operations of PWR plants from the inadvertent closing of suction valves for residual heat removal (RHR) pumps. Later, the issue was broadened to consider the possible loss of core heat removal during periods of so-called "mid-loop" operation. In the latter status, inventory of coolant in the reactor primary system is purposely reduced, for example to permit steam generator maintenance. In such l circumstances there have been incidents in a number of PWR plants in which lapses in control of water level have caused loss of suction head and simultaneous failure of pumping from all RHR pumps. In some of these incidents reactor coolant has heated to boiling. Boiling for. relatively brief periods, tens of minutes to hours, could cause enough loss of coolant inventory to uncover part of the core and overheat the fuel. Risk analyses indicate the probability of core damage from loss of RHR during shutdown to be of about the same magnitude as the prob-ability of core damage from all causes during power operation. However, the risk to public health would seem to be substantially smaller during shutdown. Core decay power would be quite low because shutdown operations of such nature as to put the core at risk are normally conducted no sooner than about two days after shutdown. Also, the temperature and pressure of reactor coolant would be low. Guidance to PWR licensees from the NRC staff in the form of a generic letter issued in 1987 seems to have accomplished little in tenns of reducing the rate of occurrence of such events. As a result of this 87

The Honorable Lando W. Zech, Jr. September 14, 1988 unsatisfactory experience, the staff has proposed new guidance in a draft of another Generic Letter.. This letter makes a number of rec-commendations and requests that PWR licensees inform the staff of their action in response to each recommendation. -Most of the recommendations concern improvements in procedures and instrumentation designed to make it less likely that RHR flow will be interrupted during mid-loop operations. We are in agreement with these recommendations. Another proposes procedures to insure that the containment will be closed in the event there is a threat of core damage caused by loss of RHR cooling during mid-loop operation. We have some concerns about this recommendation. Most occasions for mid-loop operation will occur during periods of reactor shutdown when the containment building is open for mainte-nance activities. The staff proposal calls for licensees to maintain equipment in readiness and have procedures and personnel available so that hatch covers and seals can be quickly installed if there is a threatened loss of RHR cooling. The. proposal would not require containnient to be sealed as in normal operation; the argument is that with the depressurized condition of the reactor coolant system in any such circumstance, there is little source of energy to pressurize the containment. The NRC staff has judged that even partially bolted i containment closures would successfully contain fission products released from a damaged core and mitigate consequences of such an accident. We recognize that, in the circumstances envisioned, decay power and the quantities of short-lived and volatile fission products in the core will be small. We also recognize that the other steps recom-mended by the staff can be expected to reduce the likelihood of a q core damage incident to a very small value. Therefore, the proposed method of containment closure may be a sufficient measure of protec-tion. However, we question whether the proposal has been suffi-ciently analyzed in light of the many varieties of containments and containment closures in service. We believe further study of the practicality and efficacy of the closure proposal is needed. If it cannot be shown that the proposed method is adequate, then con-sideration should be given to a requirement that would prohibit mid-loop operation except when the containment is fully closed. Sincerely, William Kerr Chainnan t 88

l 1 Th'e Honorable Lando W. Zech, Jr. September 14, 1988

References:

1. Memorandum dated August 10, 1988 from J. Sniezek, Office of Nuclear Reactor Regulation, NRC, to E. Jordan, Office for Analysis and Evaluation of Operational

Data, transmitting proposed Generic Letter, " Loss of Decay. Heat Removal" 2.

U.S. ' Nuclear Regulatory Commission, NUREG/CR-5015 " Improved 1 Reliability of Residual Heat Removal Capability in PWRs as Related to Resolution of Generic Issue 99," dated May 1988 3. U.S. Nuclear. Regulatory Commission Generic Letter dated July 9, 1987, " Loss of Residual Heat Removal (RHR) While the Reactor Coolant System (RCS) is Partially Filled (Generic Letter 87-12)" I 89

- O

CORRECTED COPY: 2/28/89 omf ~ UNITED STATES - (see Page 2)' k* 8- . NUCLEAR REGULATORY COMMISSION 'j ADVISORY COMMITTEE ON REACTOR SAFEGUARDS l g WASHINGTON,D.C 20686 September 14, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C.. 20555

SUBJECT:

PROPOSED RESOLUTION OF UNRESOLVED SAFETY ISSUE A-45, " SHUT-D0WN DECAY HEAT REMOVAL REQUIREMENTS" During' the 341st meeting of.the Advisory Committee on Reactor Safe-guards, September 8-10, 1988, we discussed the proposed resolution of USI A-45, " Shutdown Decay Heat Removal Requirements." We had previ-i= ously met with members of the NRC staff and representatives of NUMARC during our 340th meeting, August 11-13, 1988. Our subconunittee on Decay. Heat Removal Systems also recently met with the NRC staff and with industry representatives to discuss this subject. In addition, there have been a series of reviews of this subject by the subcom-mittee and the full committee dating back to 1981. We also had the benefit of the documents referenced.

Background

Task Action Plan A-45 was. intended to answer the question of whether I decay heat removal systems in existing U.S. nuclear power plants can l be counted on to reliably transfer core decay heat to En ultimate l heat sink under a variety of situations and challenges. Experience L and analyses indicated failures in decay heat removal systems to be a leading contributor to core melt risk. In addition, several European nations initiated programs to add dedicated decay heat removal systems to some of their existing plants. Through USI A-45, the staff undertook to determine whether significant vulnerabilities existed in operating U.S. plants, whether regulations related to decay heat removal were adequate, and, in particular, whether add-on systems might be warranted. The heart of' the USI A-45 program was a limited PRA study of six plants selected as a sample of the general population. The study was limited in that it concentrated on the decay heat removal systems and their. contributions to core melt risk, it was conducted only as a Level. 1 PRA, and it used approximate analytical methods in some 91 i

The Honorable Lando W. Zech, Jr. September 14, 1988 instances. The study did, however, assess the consequences of both external as well as internal initiators. It was, in fact. when compared with other less-than-full-scope PRAs, a thorough and compre-hensive risk analysis. In addition to the risk analyses, the A-45 program included prelim-inary designs of concepts for-possible decay heat removal system improvements, including dedicated add-on systems. Cost estimates were made for these improvements, as well as estimates of expected benefits in terms of reduction in the risk of core melt. Resolution Proposed by the NRC Staff The NRC staff has concluded that add-on, dedicated decay heat removal systems should not be required as backfits for existing plants. Fur-ther, they have proposed that no specific resolution be provided for USI A-45.- While important vulnerabilities in decay heat removal systems were identified in the A-45 study, it appears that they are highly dependent on specific details of each plant design. There-fore, the staff has proposed that the issues important to safe decay heat removal be subsumed by the program for individual plant examina-tions (IPEs) being developed under the Severe Accident Policy. ACRS Recommendation We agree that a general new requirement for add-on decay heat removal systems is not warranted, and we have no quarrel with the general strategy for incorporating the A-45 issue into the IPEs. We have,_in fact, recently proposed to you that the IPEs should be made more comprehensive in covering a wide range of outstanding safety issues. We do, however, have suggestions for additional steps that should be considered as a part of the A-45 resolution. We believe enough has been learned in the A-45 program to suggest that certain actions related to plant decay heat removal systems may be prudent, even if they are not justified by specific risk reduction arguments. Given this perspective, we suggest consideration be given to three additions to the A-45 resolution: 1. Certification of Feed-and-Bleed Capability At Each Plant. Studies in A-45 and other programs have shown that feed-and-bleed cooling can be an important contributor to reducing the risk of core melt in operating nuclear power plants. There is evidence, however, that the efficacy and reliability of this cooling method may be very different at some plants. The process has not been analyzed for all plants. Whether feed-and-bleed cooling can be accomplished depends upon several 92 CORRECTED PAGE 1

i The Honorable Lando W. Zech, Jr. September 14, 1988 characteristics of a plant, including the arrangement of major j reactor system components, the capability and reliability of key pumps and valves, the accuracy and completeness of procedures, and the training of operators. We note that most of the equipment called on for the feed-and-bleed operation is not safety grade. This has led to ambiguity about the extent to which NRC should regulate or influence operation of systems used for feed-and-bleed. For this reason, we believe that each licensee claiming specific benefit from feed-and-bleed cooling should be required to ] certify that its plant has the capability, in both hardware and in procedures and training to successfully cool by this means. The certification would include a description of the entire path from the core to an ultimate heat sink. This would incorporate l the ability both to supply and remove sufficient flow from the reactor coolant system and also to remove heat from the contain-ment for the period and in the circumstances necessary. Alternatives to feed-and-bleed cooling could be certified for plants where this is possible and necessary. For some PWR plants, it might be shown that secondary side blowdown and condenser cooling accompanied by primary depressurization using f high-capacity pressurizer spray systems is adequate for emer-gency decay heat removal. l What we mean by " certification" is documentation that sufficient analysis has been done to demonstr .to the satisfaction of the NRC staff that feed-and-bleed c asonably be relied upon as an emergency means for cool he core. The certification process should be designed. _ ae staff to avoid unnecessary and burdensome requirements, m. We realize that tgf4ecomended certification effort could be a part of the Jfg* hd accident management programs, but believe t that it i ydnmentioningspecificallyatthistime. t 2. Dir<4 'Use of Feedback From Experience. The Institute of f T[Jdear Power Operations (INPO) has studied operating experience f with decay heat removal systems and concluded that a number of improvements in plant equipment and procedures would increase j the reliability of these systems. These recommendations have been published in a number of Significant Operating Event Reports (SOERs). These SOER recommendations have not, as we understand, been justified by risk arguments but were derived more directly from experience and judgment. We believe that 93

The Honorable Lando W. Zech, Jr. September 14, 1988 licensees could make use of these recommendations in the course of their IPEs. 3. Protection Against Sabotage. Finally, the original intent of A-45 was to include consideration of the need for protection of the decay heat removal function against. sabotage. Dedicated add-on systems might have been configured to add such pro-tection, but this possibility has been lost with the decision against requiring such special systems. It is probably more sensible to consider the need for sabotage. protection from a broader perspective than just decay heat removal. However, the IPE program will not consider sabotage because it has been found generally not useful to treat sabotage in a risk perspective. We have recommended elsewhere that design for sabotage resis-tance should be considered in the advanced reactors program. Consideration should be given to a continued study of how ( sabotage resistance might be improved in existing plants. We trust that these recommendations will be helpful. Sincerely, William Kerr l Chainnan l

References:

1. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Draft NUREG-1289: " Regulatory and Backfit Analysis: Unresolved Safety Issue A-45, Shutdown Decay Heat Removal l Requirements," April 1988 l 2. U.S. Nuclear Regulatory Commission, Draft NUREG-1292: " Shutdown Decay Heat Removal Analysis - Plant Case Studies and Special l Issues: Summary Report," October 1987 3.. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, NUREG/CR-4710: " Shutdown Decay Heat Removal Analysis of a Combustion Engineering 2-Loop Pressurized Water Reactor-Case Study," August 1987 4. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, NUREG/CR-4767: " Shutdown Decay Heat Removal Analysis of a General Electric BWR-4/ Mark I-Case Study," July 1987 5. U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation, NUREG/CR-4458: " Shutdown Decay Heat Removal Analy-sis of a Westinghouse 2-Loop Pressurized Water Reactor-Case Study," March 1987 94

l j 'i The Honorable Lando W. Zech, Jr. September 14, 1988 6. U.S. Nuclear Regulatory Commission, Office of. Nuclear Reactor Regulation,NUREG/CR-4762: " Shutdown Decay Heat Removal Analysis of a Westinghouse 3-Loop Pressurized Water Reactor-Case Study," March 1987 7. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, ' NUREG/CR-4713: " Shutdown Decay Heat Removal Analy-sis of a Babcock and Wilcox Pressurized Water Reactor-Case Study," March 1987 8. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,NUREG/CR-4448: " Shutdown Decay Heat Removal Analysis of a General Electric BWR-3/ Mark I-Case Study," March 1987 9. Nuclear Safety Analysis Center, "EPRI/WOG Analysis of Decay Heat Removal Risk.at Point Beach," NSAC-113, dated March 1988 I 95 f ___________________D

REVISED: 5/9/88 'o UNITED STATES. g NUCLEAR REGULATORY COMMISSION E o y ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 0 WASHINGTON, D. C. 20655 c *** o April 12, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

PROPOSED RULE ON FITNESS FOR DUTY PROGRAM -- ACRS COMMENTS During the 336th meeting of the Advisory Comittee on Reactor Safe-guards, April 7-9, 1988, we reviewed the proposed rule on Fitness for Duty Program. Our Subcommittee on Human Factors met on March 28, 1988 to discuss this matter. We also had the benefit of discussions with the NRC Staff and of the documents referenced. This subject was also considered-during the 280th meeting on August 4-6, 1983, and we pre-viously commented on this matter in a report to the Commission dated August 9, 1983. The Commission issued a policy statement on Fitness for Duty of Nuclear Plant Personnel on August 4,1986. The policy statement was issued in place of rulemaking in recognition of industry efforts to voluntarily establish a fitness for duty program. The Committee considered the policy statement during our 316th meeting on August 7-9, 1986, and provided comments in a report dated August 12, 1986. In that policy statement the Commission indicated that it would reassess the possible need for further' action by considering the success of those programs during an 18 month trial period. All nuclear utilities have developed and implemented fitness for duty programs using the "EEI Guide to Effective Drug and Alcohol / Fitness for Duty Policy Development." Although these programs have helped provide reasonable assurance that nuclear power plant personnel are fit for duty, they do not utilize uniform standards. For example, only about a third of the licensees are conducting random chemical testing of body fluids, testing or cutoff levels vary from program to program, and management actions taken in response to positive chemical test results vary. The proposed rule on fitness for duty programs is intended to correct the above noted nonuniformities in existing programs. We indicated in our report to Chairman Zech of August 12, 1986 that we endorse the random chemical testing of body fluids as an element in effective fitness for duty programs and now recomend that the proposed { 97

The Honorable Lando W. Zech, Jr. April 12, 1988 rule be issued for public comment. However, we make the following detailed comments about the proposed rule and related action: 1) The draft of the Federal Register Notice of the proposed rule which we reviewed is, in general, a well-written and well-organized document which contains a wealth of information on the effects of certain drug uses, an extensive bibliography and a thorough ration-ale for the proposed rule. It also identifies a number of topics for which public comments are solicited. However, in the defini-tion of " impairment" ( 26.3), in the descrip'. son of " program elements and procedures" (5 26.20), and at severd other locations in the draft Federal Register Notice, one receives only the faint impression that licensce's fitness for duty programs should address a broad range of possible impairments to the ability of personnel to perform their duties; whereas the document addresses, almost exclusively, the subject of drug and alcohol abuse as an impair-ment. No prescription, guidance, or examples of the other types of impairments to be addressed in such programs are provided. This will result inevitably in confusing those wanting to provide comments on the proposed rule and will result in nonuniform and inconsistent fitness for duty programs. We are advised orally by the NRC Staff that this weakness has been corrected; however, we have not received a copy of the revision of the proposed rule. 2) On March 9,1988 the Commission published in the Federal Register (53FR7534) a proposed policy statement on the Nuclear Power Plant Access Authorization Program which defines the policy of the NRC regarding unescorted access to protected areas and vital areas at nuclear power plants. The purpose of the proposed policy statement is to establish access authorization programs to ensure that individuals who require unescorted access to protected areas or vital areas of nuclear power plants are trustworthy, reliable, emotionally stable, and do not pose a threat to commit radiological sabotage. The access authorization programs are to have several elements similar to those that are to be required in the fitness for duty programs (e.g., supervisor training and observation for detection of alcohol and drug abuse). However, neither the rela-tionship and similarity to, nor the differences from, fitness for duty programs are mentioned. This proposed policy statement was developed for the Commission by the Office of Nuclear Regulatory Research (RES). We were not prcivided copies to review. The proposed fitness for duty rule is to provide for the public health and safety by eliminating access to protected areas (pre-sumably this includes vital areas, although not so stated) at nuclear power plants by personnel who are judged unfit for duty. In the proposed rule, reference is made to the proposed policy statement on access authorization; however, overlap and incon-sistencies exist between the two documents. This proposed rule was developed for the Commission by the Office of Nuclear Reactor Regulation (NRR) and was brought to us for review. We will not be 98

l The Honorable Lando W. Zech, Jr. April 12,1988 ) the only ones to be confused by the fact that the Commission almost 1 simultaneously is publishing for comment a proposed policy state- { ment developed by RES and a proposed rule developed by NRR, both of j which address unescorted access to protected areas and which contain commonalities, differences, and inconsistencies. Surely this will contribute to confusion and will adversely affect respect for the regulatory process. 3) In our previous reports to the Commission on fitness for duty programs, we stressed the importance of NRC employees who have unescorted access to protected areas at nuclear power plants being subject to a fitness for duty program comparable to that being imposed upon licensees. We note that such a program has not yet been implemented by the NRC and continue to stress its importance. Additional comments by ACRS Members William Kerr and Harold W. Lewis are presented below. Sincerely, F t J. Remick c<,ing Chairman Additional Comments by ACRS Members William Kerr and Harold W. Lewis We cannot endorse this proposed rule. It is ambiguous and lacks focus. If, as we were told by the NRC Staff, the principal purpose is to make legal the random testing for drugs of those having access to protected ( areas of a plant, a much shorter and more focused rule should be formu-lated. The proposed rule is subject to misinterpretation, misapplica-tion, and probably to successful legal challenge.

References:

1. Draft SECY rsport for the Commission from Victor Stello, Jr., Executive Director for Operations, NRC,

Subject:

Proposed Rule-making - Fitness for Duty Program (Predecisional), transmitted to ACRS by memorandum dated March 17, 1988. 2. Proposed Policy Statement on Nuclear Power Plant Access Authori-zation Program published in the Rderal Register March 9, 1988 (53FR7534). 99 Revised Page

/po M0 9'o UNITED STATES 0 ~g ! ~ ^b NUCLEAR REGULATORY COMMISSION o E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5*k WASHINGTON, D. C. 20555 o,v / May 10, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

PROPOSED COMMISSION POLICY STATEMENT ON THE PROFESSIONAL CONDUCT OF NUCLEAR POWER PLANT OPERATORS (SECY-88-57) During the 337th meeting of the Advisory Committee on Reactor Safeguards, May 5-7, 1988, we discussed the Proposed Commission Policy Statement on the Professional Conduct of Nuclear Power Plant Operators. This matter was reviewed by the Human Factors Subcommittee on April 27, 1988 with the NRC Staff, and at this same meeting related industry initiatives were discussed with representatives of the Nuclear Management and Resources Council (NUMARC) and the Institute of Nuclear Power Operations (INPO). We also had the benefit of the documents referenced. Based on our discussions and review, we recomend that the Comission not issue the proposed policy statement for public coment. We make this recom-mendation for two reasons: (1) the proposed policy statement and associated supplementary information have not been adequately developed; and (2) issu-ance at this time could be counterproductive to a more comprehensive effort under way by industry. Instead, we recomend that, at least for the present time, the NRC monitor the broader and more comprehensive industry effort and defer the decision on the need for such a policy statement until a later time. We provide the following elaboration of these two reasons for our recomenda-tion: Inadequate Development: The proposed policy statement has not been developed to the stage that it is ready for issuance. In contrast to being a broad statement of policy suitable for Comission issuance, it is a list of limited prescriptive do's and don'ts suitable, at most, for inclusion in a lower-level document. Further, one can conclude that complying with the list of proposed do's and don'ts would constitute an adequate standard of profes-sional conduct for operators. We do not consider this to be the case. Confusion and inconsistencies exist among the policy statement, the associ-ated SECY document, and the enclosed supplementary information on such matters as to which operating personnel (i.e., licensed? unlicensed? both?) the policy applies. The document is replete with all-inclusive statements, such as: "... should have knowledge of all aspects of plant status"; "... should be alert to prevent and mitigate any operational problems"; "All 101

The Honorable Lando W. Zech, Jr. May 10, 1988 on-duty operators at all times must be alert;" "The operator's attention must be given to the condition of the plant at all times"; "All of the operator's senses must be focused on carrying out..." Lemphasis added]. Such expecta-tions are unrealistic when dealing with humans. The proposed policy statement also addresses the matter of unauthorized individuals being allowed to manipulate controls. This matter is quite clearly covered in the Comission's regulations [e.g.,10 CFR 50.54(h) and (1) and 10 CFR 55.13]; therefore, it should not be included in a proposed policy statement on professional conduct. Further, the proposed policy statement indicates that licensees should discourage the use of electronic entertainment devices such as radios and tape players. The NRC Staff has not been able to show that the use of such devices necessarily is disruptive of professional conduct and poses a public health and safety problem. The accompanying SECY paper indicates that licensees either must provide assurance that the use of electronic entertain; ment equipment in the control room maintains or enhances operator perform-ance, or must prohibit the use of such equipment. In contrast to expecting licensees to prove the unprovable, we think that licensees should be asked to ensure that, if such devices are permitted in control rooms, they do not interfere with normal control room operations. Counterproductive to Industry Effort: P. representatives of NUMARC and INP0 briefed the Human Factors Subcommittee on two related industry efforts. These are the top-down effort to develop management Principles for Enhancing Professionalism of Nuclear Personnel and the bottom-up effort to develop a Professional Code for Operators. The first effort is to establish principles by which management can provide an environment in nuclear power plants that is conducive to excellence and professionalism. This effort includes princi-ples for corporate management and will include not only principles for managing operations personnel but also principles for managing maintenance, technical, and engineering personnel. The bottom-up effort encourages every ) i nuclear utility to assist its operators in developing a professional code for operators (both licensed and unlicensed). Documents entitled, " Key Elements of Professional Code for Operators" and " Suggestions for Developing and Implementing a Code" have been distributed to all utilities. These were developed by a select group of Senior Reactor Operators from each utility. INPO has asked each member utility to have its professional code for opera-tors in place and in use by July 1988. The proposed policy statement is much less comprehensive than the industry effort to establish a professional code and will lack the pride of authorship of those who must utilize it. Issuance of the policy statement at this time would be counterproductive because some utilities may await issuance of the impending policy statement in contrast to participating fully in the industry-wide effort. We think this would be unfortunate. For these reasons, we recommend that the Comission not issue the proposed policy statement for public comment. Instead, we recomend that the 102

The Honorable Lando W. Zech, Jr. May 10, 1988 Commission encourage the industry effort and monitor its effectiveness following implementation. Sincerely, W. Kerr Chairman

References:

1. SECY-88-57 dated February 29, 1988, for the Commissioners from V. Stello, Executive Director for Operations, NRC,

Subject:

Proposed Commission Policy Statement on the Professional Conduct of Nuclear Power Plant Operators, with enclosures (ACRS Internal Use Only). 2. Letter dated March 14, 1988 from Zack T. Pate President, INPO, to R. Patrick Mcdonald, Senior Vice President, Alabama Power Company, re-garding development of professional codes for operators. 3. Letter dated March 30, 1988 from Zack T. Pate, President, INPO, to Joseph M. Farley, President, Alabama Power Company, regarding the devel-opment and adoption of a set of management principles to enhance profes-sionalism. 1 103

eMac o UNITED STATES / 'k ' NUCLEAR REGULATORY COMMISSION { r ADVlsORY COMMITTEE ON REACTOR SAFEGUARDS (N WASHf NGTON, D. C. 20555 ' o, p ',g s July 20, 1908 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

REPORT ON PROPOSED REVISED POLICY STATEMENT ON NUCLEAR POWER PLANT STAFF WORKING HOURS During the 330th meeting of the Advisory Committee on Reactor Safe-guards, July 14-16, 1988, we discussed the proposed revised policy statement on nuclear power plant staff working hours. During our meeting, we had the benefit of discussions with the NRC staff and of the documents listed as references for this letter. Although we believe that revised guidance on nuclear plant working hours should be issued for public comment, we do not recomend that the version of the proposed revised policy statement that we reviewed should be issued, for the reasons which follow. The staff confuses what constitutes a policy statement and what consti tutes a rule or regulation. On the one hand, the staff, in addressin CRGR questions (Enclosure 4 to the referenced draft Commission Paper makes statements such as (1) the policy statement revises present guidanceonshiftschedulingandovertimebyaddingguidanceon...;(2) the revised policy statement is not a rule; (3) the revised policy statement will apply to nuclear power plent licensees on a voluntary basis; and (4) nothing is required. On the other hand, the draft Con, mission Paper, the summary of and background to the policy statement, and the policy statement itself are replete with such statements as (1) i administrative procedures shall be develeped; (2) the procedures shall, apply to the plant operating staff; (3 the utility must specify limits; (4)' the plant manager must approve;

5) the documentation required by this policy statement; and (6) limits that shall not be exceeded. Such statements of mandatory action are not appropriate for a policy state-ment that is to provide guidance to licensees to use on a voluntary basis.

Further, there are major inconsistencies among statements made in this draft Commission Paper and statements made in the summary of, the back-ground to, and the policy statement itse'lf. To mention a few of these inconsistencies, the summary states that (1) to exceed the first set of limits, the plant manager's approval must be documented and retrievable, whereas the policy statement indicates that the plant manager's approval 105 1

The Honorable Lando W. Zech, Jr. July 20, 1988 should be separately dccumented and that these documents should be retrievable and (2) the plant manager must explain the circumstances that require this deviation, whereas the p6Tlcy statement indicates that the circumstances that required deviation from the guidelines should be explained. Such confusion and inconsistency in a document proposed to be issued for public comment will result in responses not focused on the real issues of the revised guidance. We believe for these reasons, the version of the policy statement provided to us for review is not ready to be published for public conrtent. The staff indicates that the mandatory action statements will be uniformly and consistently removed. We do have a concern that for some personnel not engaged in routine shif t work, the recommendations in the policy statement may be over-restrictive. If so, we trust that during the public comment period knowledgeable and experienced experts will provide more specific com-ments for consideration in the final formulation of the policy state-

ment, j

In the staff's presentation, they gave numerous references that provide the bases for the revised limits and recommendations indicated in the proposed policy statement. We believe these references should be listed in the Federal Register notice in order that the bases for the revision can be carefully examined. Additional comments by ACRS Member Harold W. Lewis are presented below. Sincerely, l William Kerr Chairman Additional Comments by ACRS Member Harold W. Lewis I find the assertion that "nothing is required" somewhat empty. A Commission Policy Statement carries weight, and experience suggests that a licensee who ignores it ii. curs a certain risk.

Reference:

Draft Proposed Commission Policy Statement on Nuclear Power Plant Staff Working Hours with enclosures, received with memorandum dated June 17, 1988, from J. W. Roe, NRR, to R. F. Fraley, ACRS. 4 a 106 I

UNITED STATES o,'n NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS r, O WASHINGTON, D. C 20555 March 15, 1988 The Honorable Lerdo W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON EMBRITTLEMENT OF STRUCTURAL STEEL There has been a recent exchange of letters between the ACRS and the NRC concerning the possibility of threats to the safety of nuclear power plants due to the more rapid than expected embrittlement of pressure vessel supports from radiation at low temperatures. During the 335th meeting of the ACRS, March 10-12, 1988, we heard presentations from the NRC Staff on its further evaluation of this matter. Our Subcommittee on Metal Components met on February 18, 1988 to discuss this matter. We'believe that the program the NRC Staff is undertaking to resolve this matter is appropriate. There are clearly important uncertainties remaining but the effort being implemented should generate the informa-tion needed in a timely manner. We would like to be kept informed on the progress and resolution of this matter. Sincerely, l William Kerr Chairman

References:

1. Letter dated July 15, 1987 from William Kerr, ACRS Chairman, to Victor Stello, Jr. Executive Director for Operations, U.S.

NRC,

Subject:

ACRS Coninents on the Embrittlement of Structural Steel 2. Memorandum dated October 7, 1987 from Victor Stello, Jr., Executive Director for Operations, U.S. NRC, to William Kerr, ACRS Chairman, Regarding the Embrittlement of Structural Steel 3. Letter dated December 8,1987 from William Kerr, ACRS Chairman, to Victor Stello, Jr., Executive Director for Operations, U.S. NRC,

Subject:

ACRS Comments on Memorandum from Victor Stello, Jr., EDO, dated October 7, 1987 Regarding the Embrittlement of Structural Steel 107 _-- _ _J

/ 'o UNITED STATES g 8 ' NUCLEAR REGULATORY COMMISSION n ,E. ADVISORY COMMITTEE ON REACTOR SAF EGUARDS WASHWGTON, D. C. 20666 June 7, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

INSERVICE INSPECTION OF BCILING WATER REACTOR PRESSURE VESSELS Durit.9 the 338th meeting of the AdJisory Committee on Reactor Safe-guards, June 2-4, 1988, we met with the NRC Staff to review the status of the inservice inspection program of boiling water reactor (BWR) pressure vessels. This matter was also considered during a meeting of the ACRS Subcommittee on Metal Components on May 26, 1988. We were told that the NRC Staff is proposing a plan that would require the inspection of all accessible shell welds in BWR pressure vessels. -(Very few of these are now being inspected in the older plants due to lack of appropriate equipment and difficulty of access.) The plan would also recommend a performance demonstration to show that the inspector and equipment are capable of detecting and sizing significant flaws in the reactor pressure vessel welds. Although we believe that catas-trophic failure of a BWR pressure vessel should continue to remain outside the design basis, recent experience demonstrates that flaws can grow from the coolant side of the pressure vessel into the steel pres-sure boundary. The proposed inspection program is desirable to maintain the appropriate defense-in-depth, and we therefore encourage its imple-mentation. Sincerely, W. Kerr Chairman 109 l

1. pren UNITED STATES' I

fi NUCLEAR REGULATORY COMMISSION f, j ADVISORY COMMITTEE ON REACTOR SAFEGUARDS e., wasumavow, p. c. noess February 19, 1988 I l The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS REPORT ON THE TENNESSEE VALLEY AUTHORITY'S MANAGEMENT REORGANIZATION During 'the ' 328th, 331st, and 334th meetings of the Advisory Committee on Reactor Safeguards on August 6-8 and November 5-7, 1987 and February 11-13,. f 1988, respectively, we met with members of the NRC Staff and the Tennessee i Valley Authority (TVA) Staff and reviewed the status of the resolution'of the issues pertaining to reorganization of TVA's management structure as related to the operation of TVA's nuclear power plants. In this review, we had the benefit of meetings of our -Subcommittee on TVA Organizational Issues on November 5, 1987 and February 2-3, 1988. In addition, we had the benefit of advice from two Comittee consultants and the benefit of the documents - referenced. We previously comented on this matter in a report dated August 12, 1986. In the August' 12 report several issues were identified that needed to be dealt with to correct TVA's problems and prevent their recurrence. The following measures have been taken by TVA to deal with these issues: 1. .We recomended that TVA establish a forward-looking, longer range, structured management development program to ensure a. continued supply of competent nuclear power managers. TVA has taken positive steps to - address this need. Most of the critical positions have been filled with experienced TVA employees or experienced managers from outside the organization. TVA has initiated a formal program for the selection and development of individuals from within TVA who evidence management capa-bilities. This program is. described in the Corporate Nuclear Per-formance Plan and, if properly implemented, should develop an adequate in-house source of management talent. Implementation will take con-siderable effort and time, and progress needs to be monitored by the NRC Staff. 2. We observed organizational ambiguity as to where the focal point in nuclear safety resided within the TVA corporate management structure, and recommended that this nuclear safety cognizance be reestablished and focused at the top management levnl. TVA has taken steps to reestablish a focus for nuclear safety by increasing upper management awareness of nuclear. activities and clearly defining lines of responsibility and accountability. These measures-include: 111

The Honorable Lando W. Zech, Jr. Februa ry 19, 1988 a. The Office of Inspector General has been established and reports directly to the Board of Directors. b. The Manager of Nuclear Power reports directly to the General Manager and Board of Directors, thereby focusing top management attention on nuclear plant activities, c. A separate Nuclear Safety Review Board to provide recommendations and advice independently to the Manager of the Office of Nuclear Power has been established for each plant. d. Communication channels have been established to keep the Manager of the Office of Nuclear Power aware of the daily activities in each of the plants. 3. One of the cor.cerns expressed in our August report related to the effectiveness of TVA procedures for the selection of personnel. TVA has taken steps to address this concern. Although the present organizational chart of the Office of Nuclear Power still shows a large number of line and staff managers reporting directly to the Manager, this organization appears to have perfomed effectively to date. However, we still are concerned that an organization with as many people reporting to a single manager may cause serious problems if continued indefi-nitely. This issue should be revisited during the transition to a permanent organization. The statutory limit on salaries at TVA remains in effect and the ability to retain top level nuclear managers over the long term will continue to be a problem. TVA has been successful in employing new top level managers by establishing home relocation services and relocation incentive plans for i newly hired managers. However, nothing has been done to provide additional compensation to retain key personnel who have been with the TVA nuclear program through the difficult years. In our report of August 12, 1986, we noted that the.NPC Staff would be facing a difficult task in reviewing the reorganization efforts of TVA and reaching judgments about whether those efforts were adequate. This problem is not confined to the evaluation of management problems only at TVA, but extends to a number of other utilities. We intend to explore the NRC Staff's progress in developing the needed resources and reserve our judgment until the completion of that review. An area that we explored in detail with both TVA and the NRC Staff was the effectiveness of TVA's internal safety reviews. Our concern was whether the process that TVA uses to address unresolved safety issues allows considera-tion of issues other than those relating directly to nonconformance with regulations or commitments. We believe that significant improvements have been made by TVA, but further improvements can and should be made. 112 l l 1 l

'The Honorable Lando W. Zech, Jr. February 19, 1988 Management systems and procedures for controlling the design and modifica-tions have also been improved significantly. Particular attention needs to be given to the mechanisms which assure that systems integration is properly addressed. In sumary, we believe that significant improvements have been made in TVA's organization and management capability and that most of our previous concerns have been addressed. Sincerely, William Kerr Chaiman

References:

1. U.S. Nuclear Regulatory Comission, Office of Special

Projects, NUREG-1232. Volume 1, " Safety Evaluation Report on TVA Revised Corporate Nuclear Performance Plan," July 1987 2.

U.S. Nuclear Regulatory Comission, Office of Special

Projects, NUREG-123?, Volume 2

" Preliminary Safety Evaluation Report on TVA: .Sequoyah Nuclear Performance Plan, Part 1: Programmatic Evaluation, January 1988 3. Letter dated November 6, 1987, from Stewart D. Ebneter, Office of Special Projects, to S. A. White. TVA,

Subject:

Sequoyah Unit 2 Integrated Design Inspection (IDI) 4. Letter dated February 3, 1988, from Lynne Bernabei, Attorney for Andrew Bartlik, to NRC Commissioners, regarding TVA Nuclear Power Program 5. Letter dated December 24, 1987, from Kurt N. Larson, American Nuclear Insurers, to David H. Marks, TVA,

Subject:

ANI/MAELU Nuclear Liability Insurance Inspection 113

/ UNITED STATES NUCLEAR REGULATORY COMMISSION o ,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS O WASHINGTON, D. C. 20555 l April 12, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

Dear Chairman Zech:

) i

SUBJECT:

UPDATED POLICY STATEMENT ON TRAINING AND QUALIFICATION OF NUCLEAR POWER PLANT PERSONNEL -- ACRS COMMENTS During the 336th meeting of the Advisory Committee on Reactor Safe-guards, April 7-9, 1988, we considered a Subcommittee report and dis-cussed a draft of the updated Comission Policy Statement on Training and Qualification of Nuclear Power Plant Personnel. Our Subcommittee on Human Factors met with representatives of the NRC Staff on March 28, 1988 to discuss this matter. We also had the benefit of the referenced documents. We endorse the Staff proposal to issue the proposed policy statement for public comment. We are of the opinion that the training and accredita-tion program of the industry and the Institute of Nuclear Power Opera-tions has been highly successful in establishing effective training programs for the personnel responsible for the operation of nuclear power plants. We encourage continuation of the INPO-managed accredita-tion program and believe it is unnecessary for the NRC to consider rulemaking related to this issue. Dr. Forrest J. Remick did not participate in the Comittee's delibera-tions regarding this matter. Sincerely, W. Kerr j Chairman

Reference:

Draf t SECY report for the Commissioners from Victor Stello, Jr., Execu-tive Director for Operations, NRC,

Subject:

Updated Commission Policy Statement on Training and Qualification of Nuclear Power Plant Personnel (Predecisional), transmitted to ACRS members by memorandum dated March 21, 1988. 115 1 i

1. ' p M Ezog UNITED STATES og

./ NUCLEAR REGULATORY COMMISSION o .f ,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 6.. September 13, 1988 l The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

NRC STAFF'S TVA LESSONS LEARNED EFFORT During the 341st meeting of the Advisory Comittee on Reactor Safe-guards, September 8-10, 1988, we met with members of the NRC staff and reviewed the staff's TVA lessons learned effort. Our subcommittee on TVA Organizational Issues also discussed this matter with the staff during its meeting on July 22, 1988. During our review, we had the benefit of the documents referenced. We believe that the final lessons learned recommendations appear to be well thought-out and responsive to the problems which were experienced at TVA. An important task ahead is to complete the implementation of the philosopny embodied in these recommendations.. However, in doing so, the NRC should not dictate the management organization and style of a utility and should avoid prescriptive and detailed regulations or evaluation of utility management, methods, and structures. The NRC should observe and monitor the performance of licensees and assess their capability for continued safe operation of their nuclear. power plants, and should communicate its concerns to the appropriate level of utility i i management on a timely basis. Sincerely, William Kerr Chairman

References:

1. SECY-86-334, Memorandum dated November 12, 1986 for the Comis-sioners from V. Stello Executive Director for Operations, NRC,

Subject:

TVA Preliminary Lessons Learned SECY-87-211, Memorandum dated August 20, 1987 for the Commissioners i 2. from V. Stello, Executive Director for Operations, NRC,

Subject:

Final Report on TVA Lessons Learned (Includes comments from former TVA and NRC managers on SECY-86-334) 117

h L l JMitoq'q p UNITED STATES E NUCLEAR REGULATORY COMMISSION n ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g WASHINGTON, D. C. 20655 March 15, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON THE NEED FOR GREATER COHERENCE AMONG NEW REGULATORY POLICIES During the 335th meeting of the Advisory Comittee on Reactor Safeguards (ACRS), March 10-12, 1988, we discussed the problem which we see as an increasing lack of coherence and integration among several separate areas of policy making within the NRC. We understand that the NRC Staff is attempting to develop an integrated approach to implementing the Commission's Severe Accident Policy. In the usual course of events, we would expect to review the Staff's proposals at a later date.

However, we believe it might be helpful to forward a few coments to you at this time.

The Severe Accident Policy is only one of a number of new Comission policies and programs concerning nuclear power plants that have been advanced over the past two or three years. Others relate to the safety goal, standardized plant designs, ISAP, and advanced reactors. In addition, the resolution of USIs and GIs has led, or might lead, to important new requirements and guidance for licensees in several areas. Although the NRC Staff, the ACRS, and the Commission have provided some overall guidance toward integration of these policies and new require-ments, we believe this has been insufficient. As a result, licensees can be confused or burdened with contradictory new requirements from different parts of the NRC Staff. For example, a part of the resolution of USI A-45, " Shutdown Decay Heat Removal Requirements," proposed some months ago, incorporated the use of safety objectives similar to, but not the same as, objectives being developed in implementation of the Safety Goal Policy. 119

1 The Honorable Lando W. Zech, Jr. ! March 15, 1988 l We offer two suggestions for your consideration: I (1) The attempt to integrate evolving policies and issues should not be } limited to those embodied in the Severe Accident Policy, but should I incorporate the entire range recently addressed in policy state-ments or currently before the Comission. 1 (2) The Safety Goal Policy should not be viewed as just one of the several policies and issues on today's table. Instead, it should be seen as an umbrella policy which should be the principal tool for integrating and providing coherence to the entire set. We expect to review the integration program being developed by the NRC Staff when it is available. Sincerely, William Kerr Chairman l l l 120

'o,, UNITED STATES NUCLEAR REGULATORY COMMISSION g r ADVISORY COMMITTEE ON REACTOR SAFEGUARDS [ WASHINGTON, D. C. 20655 O April - 12,1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D. C. 20555

SUBJECT:

ACRS REVIEW 0F SHUTDOWN NUCLEAR POWER PLANTS The ACRS plans to review several nuclear power plants that have either-been shut down for an extended period and are now being considered for restart or hav'e been delayed in initial startup for a longer period than usual. We will not as a matter of course issue reports prior to a restart decision unless we have concerns that we believe should be brought to the Commission's immediate attention or you have specifically requested such a report. In the latter case, we would appreciate early notification since this could influence our priorities in scheduling these reviews. Sincerely, 1 W. Kerr Chairman 121

UNITED STATES J [ o,, 8 NUCLEAR REGULATORY COMMISSION { o ,N ADVicORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20655 i February 16, 1988 4 \\ i The Honorable Lando W. Zech, Jr. Chairman U. S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON SECY-87-314 INTERIM POLICY STATEMENT ON MAINTENANCE OF NUCLEAR POWER PLANTS, DATED DECEMBER 30, 1987 During the 334th meeting of the Advisory Committee on Reactor Safe-guards, February 11-13, 1988, we were briefed by the NRC Staff on the proposed Interim Policy Statement on Maintenance of Nuclear Power Plants (SECY-87-314), dated December 30, 1987. Although we are unable to provide definitive advice on the basis of this briefing, we have learned enough about the proposed Policy Statement to have serious concerns about its being made imediately effective without benefit of coments from the public and the industry. If promulgation as an imediately effective Policy Statement can be delayed appropri-ately, we will be able to review its intent and implications in suitable depth and provide you with our advice or recommendations. Sincerely, William Kerr Chairman 123

[

o UNITED STATES - _i g e: s' NUCLEAR REGULATORY COMMISSION 1 n ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g WASHINGTON, D, C. 20665 September 13, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission l Washington, D.C. 20555 l

SUBJECT:

PROPOSED RULEMAKING RELATED TO MAINTENANCE OF NUCLEAR POWER PLANTS During the 341st meeting of the Advisory Committee on Reactor Safe-guards, September 8-10, 1988, we discussed the proposed amendment to 10 CFR Part 50 related to maintenance of nuclear power plants. We had previously discussed this topic with the NRC staff during our 340th meeting, August 11-13, 1988, and during meetings of our Maintenance Practices and Procedures Subcommittee on June 15 and September 7, 1988. During these meetings, we had the benefit of comments by a representative of the Nuclear Management and Resources Council (NUMARC). We also had the benefit of the document referenced. We cannot endorse the proposal to establish a maintenance rule. The proposal appears to be based upon the following hypotheses: 1. Maintenance of nuclear power plants, as now performed, poses a significant risk to public health and safety. 2. The existence of a maintenance rule would reduce this risk. There is some evidence to support the first of these hypotheses, although' there is no direct quantitative indicator. There is anec-dotal evidence that poor maintenance -has contributed to plant un-availability and has even led to the existence of plant states that are interpreted as accident precursors. Common sense suggests that a well-maintained plant poses less risk than one poorly maintained. Some operating plants have better maintenance programs than others, and some have programs that should be improved. We have seen no evidence to support the second hypothesis. The regulatory analysis provided by the staff makes the arbitrary assump-tion that a reduction in risk will occur as a result of the rule, and bases its cost-benefit conclusions on a guess about the amount of risk reduction expected. Nor have we seen evidence that the existence of a rule would not make things worse. Indeed there are characteristics of regulations, and especially the way in which they.are typically enforced, that lead us to believe that, under a rule, a move toward uniformity would occur, and this is likely to decrease the effective-ness of some of the better existing programs. 125

The Honorable Lando W. Zech, Jr. September 13, 1988 ) Finally, it appears to us that maintenance practices in the industry are improving and that a rule may be disruptive to the substantial industry initiatives that have been developed to accomplish this improvement. Additional comments by ACPS member Harold W. Lewis are presented below. Sincerely, W. Kerr Chairman Additional comments by ACRS member Harold W. Lewis I want to take this opportunity to observe that it is not necessarily true that mcre maintenance is better maintenance -- a substantial number of events are initiated by testing and maintenance operations. There exists a well-developed theory of. reliability which deals with such matters as the optimum level of maintenance -- there are books on the subject -- and it would be useful to bring some expertise into the analysis of this question.

Reference:

Letter cated August 29, 1988 from Bill M. Morris, Director, Division of Regulatory Applications, Office of Nuclear Regulatory Research, to Raymond F. Fraley, Executive Director, Advisory Committee on Reactor Safeguards, enclosing Draft Commission Paper for Notice of Proposed Rulemaking for Maintenance of Nuclear Power Plants I 126 l

1 'o UNITED STATES g 8 NUCLEAR REGULATORY COMMISSION n y

,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o.,

WASH.NGTON, D. C. 20666 June 7, 1988-1 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

NRC PROPOSED RULE ON EARLY SITE PERMITS, STANDARD DESIGN CERTIFICATION, AND COMBINED LICENSES FOR NUCLEAR POWER REACTORS During the 338th meeting of the Advisory Committee on Reactor Safe-guards, June 2-4, 1988, we reviewed a proposed rule,10 CFR Part 52, which would provide for issuance of early site permits, standard design certifications, and combined construction permits and conditional operating licenses for nuclear power reactors. We had the benefit of briefings by the NRC Staff during a subcommittee meeting on May 31, 1988 and during the full Committee meeting. We also had the benefit of the documents referenced. The ACRS provided comments on this subject in its letters of August 12, 1986 and October 15, 1986. We have three concerns, as articulated below. In addition, we suggested changes to the requirements for ACRS review, which the NRC Staff agreed to, and which presumably will be made in the draft submitted to you. We recommend that, in 10 CFR Part 52, Subpart B, the scope and level of detail of information required by the Staff for design certification be defined more fully by incorporating the information identified for this purpose in the NRC Policy Statement on Standardization of Nuclear Power Plants. Although we encourage the development of a clear enunciation of Comis-sion regulations for early site permits, standard design certifications, and combined licenses, we question whether all three should be addressed in the same Part of Title 10 of the CFR. The Comission's regulations concerning standardization of manufactured and duplicate plants and the Staff review thereof are contained in Appendices M, N, and 0 of Part 50. The portion of proposed Part 52 relating to standard design certifica-tion is an elaboration of Section 7 of Appendix 0 of Part 50. To make this elaboration a significant portion of a new Part of the regulations, which also includes two other complex matters, will add to the com-plexity and inscrutability of the Comission's regulations. Part 50 is already confusing because it is a multipurpose regulation that includes power reactors, nonpower reactors, and fuel cycle facilities. We ] l 127 ) i

The Honorable Lando W. Zech, Jr. June 7, 1988 recommend against promulgating another multipurpose Part of the regu-lations. The Staff agrees in principle with these views but indicates that it does not have the resources to develop the new regulations in a more orderly fashion and thus offers the proposed patchwork. We can think of no better time in the agency's existence for improving the scrutability of the regulations. We see a need to distinguish between the amount of design detail re-quired for the NRC Staff review of a request for certification and the design detail that is included in the certifying rule. It is highly desirable that nuclear power plant designs submitted for certification be essentially complete in both scope and detail. However, if the certifying rule includes the same amount of detail, rulemaking will be required in order to correct errors in the documentation or to make minor but desirable changes in the design. It is therefore essential that great care be taken in defining what is to be included in the design certification. In this respect, we believe that alternatives to certification by rulemaking have not been adequately explcred. These are the only major comments we have to offer at this time. We l will continue our review and offer comments as appropriate as the process develops. j Sincerely, W. Kerr Chairman

References:

1. U.S. Nuclear Regulatory Comission, Proposed Rule,10 CFR Part 52, "Early Site Permits; Standard Design Certifications; and Combined j Licenses for Nuclear Power Reactors," received May 24, 1988. 2. U.S. Nuclear Regulatory Commission, Policy Statement,10 CFR Part i 50, " Nuclear Power Plant Standardization," 52 FR 34884, dated September 15, 1987, 128

n aroog#o a UNITED STATES 8 NUCLEAR REGULATORY COMMISSION n 3 I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS O WASHINGTON, D C. 20555

• • .+

March 15, 1988 1 Mr. Victor Stello, Jr. Executive Directcr for Operations U. S. Nuclear Regulatory Commission Washington, D. C. 20555

SUBJECT:

ACRS COMMENTS ON REGULATORY GUIDE 1.99, REVISION 2, " RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS" DATED MARCH 1988 During the 335th meeting of the Advisory Committee on Reactor Safe-guards, March 10-12, 1988, we discussed the subject Regulatory Guide. Our Subcommittee on Metal Components met on September 4, 1985; July 24, 1987; and February 18, 1988 to consider this Guide. We also had the benefit of discussions with the NRC Staff and its consultants. We pre-viously commented on this matter in a report dated September 17, 1985. We concur in the Regulatory Position of this Guide. Sincerely, ( William Kerr Chairman cc: S. J. Chilk, SECY T. Rehm, EDO E. S. Beckjord, RES G. A. Arlotto, RES P. N. Randall, PES C. Bartlett, RES A. J. Cappucci, NRR 129

i [ 'o,, UNITED STATES 8 NUCLEAR REGULATORY COMMISSION e [ I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o WASHINGTON, D. C. 20665 May 10, 1988 Mr. Victor Stello, Jr. Executive Director for Operations U.S. Nuclear Regulatory Commission ( Washington, D.C. 20555

SUBJECT:

REVISION 2 TO REGULATORY GUIDE 1.100, " SEISMIC QUALIFICATION OF ELECTRIC AND MECHANICAL EQUIPMENT FOR NUCLEAR POWER PLANTS" During the 337th meeting of the Advisory Committee on Reactor Safe-guards, May 5-7, 1988, we concurred in the regulatory position proposed in Revision 2 to Regulatory Guide 1.100, " Seismic Qualification of Electric and Mechanical Equipment For Nuclear Power Plants." Sincerely, William Kerr Chairman

References:

Memorandum dated March ?l, 1988 from Eric S. Beckjord, Office of Nuclear Regulatory Research, to Raymond F. Fraley, ACRS, transmitting: a. Revision 2 to Regulatory Guide 1.100, " Seismic Qualification of Electric and Mechanical Equipment For Nuclear Power Plants," March 1988. b. IEEE Std. 344-1987, " Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations." c. Public Comment Letters d. Resolution of Public Comments cc: S. J. Chilk, SECY T. Rehm, ED0 E. Beckjord, RES G. L. Arlotto, RES S. K. Aggarwal, RES C. Bartlett, RES A. Cappucci, NRR 131

/ UNITED STATES 8 NUCLEAR REGULATORY COMMISSION k' f ADVISORY COMMITTEE ON REACTOR SAFEGUARDS t, wAsHawoTon. o. c.rossa o.... November 22, 1988 Mr. Victor Stello, Jr. Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS ACTION ON PROPOSED REGULATORY GUIDE 1.9, REVISION 3 " SELECTION, DESIGN, QUALIFICATION, TESTING, AND RELIABILITY OF DIESEL GENERATOR UNITS AS ONSITE ELECTRIC POWER SYSTEMS AT NUCLEAR POWER PLANTS" During the 343rd meeting of the Advisory Committee on Reactor Safe-guards, November 17-18, 1988, we met with members of the NRC staff and discussed the subject regulatory guide. As a result of this discussion, we concur in the NRC staff's proposal to issue the subject guide for public comment. After the public coment period, we expect to review the proposed final version of this guide together with the public coments and the NRC staff's responses to them. Sincerely. Forrest J. Remick Acting Chairman

Reference:

U.S. Nuclear Regulatory Comission, draft Regulatory Guide 1.9, Revision

3. " Selection, Design, Qualification, Testing, and Reliability of Diesel Generator Units as Onsite Electric Power Systems at Nuclear Power Plants," September 1988 cc:

S. J. Chilk, SECY j 4 J. L. Blaha, EDO E. Beckjord, RES W. Minners, RES A. Serkir, RES C. Bartlett, RES D. Persinko, NRR 133 l

/ o,, UNITED STATES f. NUCLEAR REGULATORY COMMISSION o -M I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20566 o.*** November 22, 1988 Mr. Victor Stello, Jr. Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS ACTION ON PROPOSED REGULATORY GUIDE, TASK NO. EE-006-5, " QUALIFICATION OF SAFETY-RELATED LEAD STORAGE BATTERIES FOR NUCLEAR POWER PLANTS" During the 343rd meeting of the Advisory Comittee on Reactor Safe-guards, November 17-18, 1988, we met with members of the NRC staff and discussed the subject regulatory guide together with the public i coments and the staff's responses to them. The ACRS had previously comented on an earlier version of this guide in a letter to you dated August 11. 1987. As a result of this discussion, we concur in the regulatory position of this guide. Mr. C. J. Wylie did not participate in the Comittee's deliberations regarding this matter. Sincerely, Forrest J. Remick Acting Chairman

References:

Memorandum from Eric S. Beckjord, Office of Nuclear Regulatory Research to Raymond F. Fraley, ACRS, dated October 25, 1988, trans-mitting: 1. Proposed effective Regulatory Guide, Task No. EE 006-5, "Quali-fication of Safety-Related Lead Storage Batteries for Nuclear Power Plants" 2. IEEE Standard 535-1986, "IEEE Standard for Qualification of 4 Class IE Lead Storage Batteries for Nuclear Power Generating Stations," June -25, 1986 135

[ 7, e Mr. Victor Stello, Jr. - November 22, 1988 cc:.S. J. Chilk, SECY- { J. L. Blaha, EDO ' l E._Beckjord,RES-G. A. Arlotto, RES S. K. Aggarwal, RES C. Bartlett, RES D. Persinko, NRR 1, i-136 _..m___-._ m

p LE% e UNITED STATES '8 o,i NUCLEAR REGULATORY COMMISSION 3 ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS O WA$HINGTON, D. C. 20655 \\ * * * * * $g April 12, 1988 The Honorable lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

PROGRAM TO IMPLEMENT THE SAFETY GOAL POLICY -- ACRS COMMENTS During the 336th meeting of the Advisory Committee on Reactor Safe-guards, Apr.11 7-9, 1988, and in previous meetinas, we discussed NRC Staff efforts to devise methods for implementing the Safety Goal Policy enunciated by the Commission in 1986. Following our report to you of May 13,1987, "ACRS Comments on an Implementation Plan for the Safety Goal Policy," the Commission directed the Staff to develop a plan along the lines we had suggested. The ACRS and its Subcommittee on Safety Philosophy, Technology, and-Criteria have. met several times with members of the Office of Nuclear Regulatory Research (RES) to discuss means by which the implementation plan suggested in our May 1987 report might be " fleshed out." We expect that additional meetings will be necessary over the next few months. Our understanding is that the Staff will then document a description of its program. We believe it would be useful to provide you with some interim comments. These follow: Definition of a "Large Release" The Safety Goal Policy includes a general performance guideline that there should be a probability no greater than 1E-6 per reactor-year of a large release from any operating nuclear power plant. Exactly what is meant by a large release was not defined in the Policy, but it has been suggested that a definition is needed in the implementation plan. We believe the definition of a large release should make clear that a release to the biosphere of a substantial fraction of the core inventory is intended. In particular, it is misleading to define it in terms of health effects; those goals are stated elsewhere in the Policy. Of course, in assessing whether the IE-6 goal is met by the body of regu-latory rules and practice, one will have to consider and evaluate specific sequences for individual plants. All of the probabilistic risk assessments (PRAs) provide examples of such sequences, and it is uni-versa 11y agreed that the " bottom line" estimates thereby derived are among the weakest results of a PRA. Nonetheless, a definition of a 137

The Honorable Lendo W. Zech, Jr. April 12, 1988 l l large release in terms of a number of curies released would provide a l durable objective for the calculations, which will improve with time, and which must be done as best one can at any given time. In the end, of course, a particular number will have to be specified, and it must be consistent with the other elements of the program. I Examples of what we have in mind might be helpful. We regard the release to the environment that occurred at THI-2 in 1979 as not a large release. The release that occurred in the accident at Chernobyl Unit 4 in 1986 was a large release. The fact that there were apparently no " prompt" radiological fatalities among the offsite population at Chernobyl is irrelevant under the proposed definition. Definition of " Core Melt" In our report of May 13, 1987, we suggested a performance objective for " prevention" systems as a calculated core melt probability of less than 1E-4 per reactor-year. Exactly what is meant by core melt, seemingly a simple ouestion, presents a problem for analysts and others considering the details of nuclear power plant accidents. The most likely sequence of core overheating, melting, and displacement can be viewed as a hypothetical sequence of events. Each event is less likely than that preceding it because the sequence may be interrupted at any point, for example, by successful performance of emergency procedures. These events might be defined as: (1) loss of adequate core cooling (core overheating beyond design-basis limits), (2) onset of significant damage to the core, (3) melting and displacement of the core within the reactor pressure vessel (as in the TMI-2 event), (4) passage of molten core out of the reactor pressure vessel (e.g., " core on the floor"). We would apply the IF-4 objective to the first event in this sequence, with the expectation that there is a significant but undefined margin in likelihood between it and the remaining events. We note that referring to this as a " Core Melt Performance Objective" is an unfortunate choice; however, it is one that is well established in the nuclear safety community. 138

I The Honorable Lando W. Zech, Jr. April 12, 1988 1 Definition of the Plant Performance Objective Our May 1987 report recommended that a performance objective expressing "how well the plant is operated" should be developed. The RES Staff has indicated that it doesn't know how to do this and plans no further work I regarding this matter. Without a performance objective of this sort, the severe gap in the logic of the Safety Goal Policy, which led to our original recommendation, remains. The problem is as follows: The Safety Goal Policy is intended, as we understand it, to be a declaration of intent about how safe operating nuclear plants are to be. However, PRA, the primary tool by which performance of plants against the safety goal is to be judged, uses few data on operational performance. Most of the analyses in a PRA depend on attributes of the plant design. Very little information about how a plant is actually operated is used. Where actual operational performance is included (e.g., equipment failure rates and pree dictions 'of operator response), most of the data used are generic to industry experience and little reflect attributes of the oper-ation of the particular plant being analyzed. This is really an inherent weakness in the present art of PRA. Although no means are presently apparent for incorporating a more complete definition of operational performance into the Policy implementation, we believe credibility of the Policy suffers without it. Research could help. It might be possible to somehow better incorporate attributes of operational performance into PRA. If this cannot be done, a prominent caveat, e.g., a warning that PRA results do not tell the full story, should be made a part of the Policy or of the implementation plan. We note that at one of our recent meetings, Nuclear Reactor Regulation Staff described plans for further work having as its objective a better description (for use in PRAs, e.g.) of the contribution to risk or to safety made by the plant operating staff. Results of this work could contribute to the formulation of a performance objective. Use of Cost-Benefit Analysis Cost-benefit analysis has a role in regulatory practice under the backfit rule. In this context, the role of the safety goal should be only to help provide a definition of what is meant by " adequate for safety." If it is found that a regulation is permitting plants to be licensed which seem to have safety performance poorer than the guidance provided by the safety goal, then that regulation should be revised, without recourse to cost-benefit arguments. 139

l The Honorable Lando W. Zech, Jr. April 12, 1988 l Need For Review of Regulations From the Perspective of the Safety Goal Policy There is a need to consider what is meant by implementation of the Safety Goal Policy. We have suggested that the Policy not be "used to make narrowly differentiated decisions about specific plants." Instead, it should be used as a primary means for judging the suitability and necessity of specific regulations and regulatory practices. We include practices in this discussion because many of the requirements levied by the NRC on licensees are not part of formal regulations but proceed from a more informal body of practice. We put aside, for now, the question of whether this in itself is or is not a problem but suggest only that the informal, as well as the formal, regulatory practices should be constrained.by the Safety Goal Policy. The next question is whether the Policy should be used only reactively in assessing proposed regulatory changes evolving from other programs (e.g., new requirements coming from resolution of a USI), or should be used more actively in assessing the present body of regulatory practice. We recommend the latter. The existing, body of regulations and reguletory practice has grown enormously over the past 30 years. This growth has been largely a bottom-up process as the regulatory staff and ACRS have reacted to proposals from applicants and vendors and responded to developing technical information and plant experience. We believe it is possible to make a zero-based assessment of this body of regulations to deter-mine: (a) which parts are contributing effectively to assure that plants are appropriately safe, (b) which parts are unnecessary, and (c) which parts need to be strengthened or better focused. It is the responsibility of the NRC to move in the direction of such an assessment. It will not be easy but should begin now for several reasons: There is a hiatus in applications for new plants. There is now an extensive body of experience with operation and regulation that did not exist 30 years ago. There is now much more complete information about severe accidents l than existed previously. PRA has matured and is available as a tool. And finally, the Safety Goal Policy is available as a thoughtful j and agreed-upon measuring criterion. 140

The Honorable Lando W. Zech, Jr. April 12, 1988 We trust the above comments will be useful as the Staff continues with development of the Policy implementation. Additional remarks by ACRS Member Harold W. Lewis are presented below. Sincerely, W. Kerr Chairman Additional Remarks By ACRS Member Harold W. Lewis The Committee has, in its May 1987 report, defined the term " core melt" to mean loss of as.sured core cooling which can result in severe core damage, to match the probability objective of IE-4 per reactor year. In this report the definition is made even more restrictive. While there is ambiguity in the community about the meaning of the term (as noted by the' Committee)., the redefinition has an enormous impact on the effect of the goal. There is a considerable difference of probability between loss of adequate core cooling and melting of the core, the former more probable but not necessarily damaging. Since the assignment of the term " core melt" to an event which need not melt the core unnecessarily biases the interpretation of the safety goal, I believe it is the job of the Commission to clarify what is meant by the tenn, rather than for the Committee to read minds. I take the simplistic view that a core melt requires a molten core. In the law, the established procedure for resolving apparent ambiguities is to start with the plain meaning of the words. I also believe it important that the Commission (not the Staff) clarify its intent in promulgating the Safety Goal Policy. Though the goals were stated by the Commission two years ago, we continue to hear of Staff actions which are justified by one or another version of "if we can see a way to improve safety, we will." Presumably the Commission, by giving an answer to the how-safe-is-safe-enough question, intended precisely to dampen such unbounded enthusiasm. I believe the Commission should reinforce its guidance to its staff. 141

53rrou 'o UNITED STATES ,8 ~,,. NUCLEAR REGULATORY COMMISSION n r,Y ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o,, WASHINGTON, D. C. 20565 August 16, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. -20555 1

SUBJECT:

REPORT ON NUREG-1150, " REACTOR RISK REFERENCE DOCUMENT" During the 340th meeting of the Advisory Committee on Reactor Safe-guards, August 11-13, 1988, we discussed the staff's plan for - the development of the final version of NUREG-1150, " Reactor Risk Reference . Document," with Mr. V. Stello, Jr., Executive Director for Operations, and members of his staff. We also had the benefit of the documents referenced. In our July 20. 1988 letter to you on the Integration Plan for Closure of Severe. Accident Issues, we stated, "We believe that subjecting the final version of NUREG-1150 to a thorough peer review is required as part of the process of establishing credibility."' Reviews by a number of individuals and groups were highly critical of the original draft of NUREG-1150 In view of the extensive modifica-tions that have been made in response to this criticism, the current version must be regarded as a new document. Also, since this document is intended to play a substantial role in the implementation of the Commission's severe accident policy, its quality and credibility are very important. We recommend that before publication in final form, the final version of NUREG-1150 be subjected to a thorough peer review. { Sincerely, un W. Kerr Chairman i 143 1

l The Honorable Lando li. Zech, Jr. August 16, 1988

References:

1. Memorandum dated August 10, 1988 from V. Stello (ED0) to R. Fraley 1 (ACRS), " Plans for Review of Final NUREG-1150" j 2. SECY-88-147, Memorandum dated May 25, 1988, for the Commissioners from V.

Stello, Executive Director for Operations,

Subject:

Integration Plan for Closure of Severe Accident Issues 3. Brookhaven National Laboratory Report, NUREG/CR-5000, " Methodology for Uncertainty Estimation in NUREG-1150 (Draft): Conclusions of a Review Panel," H. Kouts et al., December 1987 4. Lawrence Livermore National Laboratory Report, NUREG/CR-5113 " Findings of the Peer Review Panel on the Draft Reactor Risk Reference Document, NUREG-1150," W. Kastenberg et al., May 1988 5. American Nuclear Society, " Initial Report of the Special Committee on Reactor Risk Reference Document (NUREG-1150)," L. LeSage et al., Draft Report dated April 1988 I I 144

o UNITED STATES ~g / ) NUCLEAR REGULATORY COMMISSION o f r$ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 'o WASHINGTON, D. C. 20555 December 20, 1988 l The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

EOUIPMENT QUALIFICATION-RISK SCOPING STUDY j During the 344th meeting of the Advisory Committee on Reactor Safe-L guards, December 15-16, 1988, we considered a report from our Subcommit-l tee on Reliability Assurance pertaining to its review of the Equipment Qualif'ication (EQ)-Risk Scoping Study, performed by the Sandia National Laboratories (SNL) for the NRC Office of Nuclear Regulatory Research (RES). This matter was also discussed with representatives of RES and SNL during our 339th meeting, July 14-16, 1988 and during previous meetings of our Reliability Assurance Subcommittee on December 16, 1987; June 14, 1988; and December 12, 1988. We also had the benefit of the documents referenced. The purpose of this study was to assess the risk significance and risk uncertainties associated with current EQ requirements for safety-related electrical eauipment. The approach was to use information from existing PRAs to determine what electrical equipment would be needed to prevent or mitigate the consequences of a severe accident and, at the same time, would be exposed to a harsh environment related to that accident. For the most part, the conclusions and recommendations from this study are plant specific. For this reason, the NRC staff proposed, and we agree, that the insights from this study can be used in two ways: ' As items to be considered further in the Individual Plant Examina-tion and the Accident Management programs.

• As a means to limit or better focus the EQ inspections at existing plants.

One conclusion from the study is that the importance of the accident radiation dose in EQ is overemphasized. We believe that this warrants a review of some of the current requirements in Regulatory Guide 1.89, " Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants." Another observation of general significance is that existing PRAs utilize equipment failure rates derived from experience in normal ] 145 \\

The Honorable Lando W. Zech, Jr. December 20, 1988 operating environments. There is no basis for confidence that these failure rates are realistic for equipment required to operate in a harsh environment, whether resulting from a severe accident or a design-basis accident. The significance to risk of this observation is potentially important and deserves further study. The following comments do not relate specifically to this study but do relate to the precess of managing research. We believe that review of the study by a four-person peer review panel contributed significantly to the credibility of the conclusions and the quality of the final report. We were favorably impressed with the expertise and variety of the panel members and the way they interacted with the research team. \\ Finally, we suggest that had a risk-besed scoping study been performed before 'the EQ research program was begun, rether than after it was completed, the nature and scope of the program might have been differ-ent, and presumebly better. While such a study should not be expected to dictate all aspects of a large, complex research program, it should help to distinguish between the clearly important and the clearly unimportant, and perhaps even between what is knowable and what is not. Sincerely, William Kerr Chairman

References:

1. EQ-Risk Scoping Study, Draf t Final Report - L. D. Bustard, Sandia National Laboratories, A. M. Kolaczkowski, G. T. Medford, and J. Clark, Science Applications International Corporation, dated May 1988 2. Letter from George E. Sliter, Electric Power Research Institute, to Mcni Dey, NRC Office of Nuclear Regulatory Research,

Subject:

EQ-Risk Scoping Study Peer Review, dated June 10, 1988 3. Letter from K. S. Canady, Duke Power Company, to Moni Dey, NRC Office of Nuclear Regulatory Research,

Subject:

EQ-Risk Scoping Study Peer Review, dated June 28, 1988 4 Letter from S. P. Carfagno, Franklin Research Center, to L. D. Bustard, Sandia ' National Laboratories,

Subject:

EQ-Risk Scoping Study, dated June 15, 1988 5. Draft Final Comments, Peer Review of the Sandia National Laboratory Equipment Qualification-Risk Scoping Study, A. J. Wolford, H. L. Magleby, EG&G Idaho, Inc., dated June 1988 6. EQ-Risk Scoping Study: Discussion of Peer-Review Comments (un-dated) 146

/ o UNITED STATES g 8 NUCLEAR REGULATORY COMMISSION n ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 0 WASHINGTON, D, C. 20666 May 10, 1988 The Honorable Lando W. Zech, Jr. I Chairman U.S. Nuclear Regulatory Commission d Washington, D.C. 20555

Dear Chairman Zech:

e i

SUBJECT:

PROPOSED REVISION OF THE ECCS RULE CONTAINED IN 10 CFR 50.46 1 AND APPENDIX K During the 337th meeting of the Advisory Committee on Reactor ' Safe-i guards, May 5-7, 1988, we met with members of the NRC Staff and reviewed the final version of the proposed revision to the emergency core cooling system (ECCS) rule contained in 10 CFR 50.46 and Appendix K. Our Subcommittee on. Thermal Hydraulic Phenomena met on April 20,1988 to discuss this matter. We also had the benefit of discussions with the NRC Staff and of the documents referenced. The ACRS previously comment-ed on the proposal to issue this rule for public comment in a letter dated September 16, 1986. The proposed revision to the ECCS rule will eliminate the requirement to use the models specified in Appendix K and allow use of realistic models combined with an uncertainty analysis of the overall calculation. Certain criteria in 10 CFR 50.46, such as 2200 F peak cladding tempera-ture and 17%_ cladding oxidation, would be maintained. The regulatory guide which will accompany _ the revised rule describes features of a realistic evaluation model acceptable to the NRC Staff and contains guidance on performing the necessary associated uncertainty evaluation. No changes have been proposed to the final rule version as a result of the public comments received. The regulatory guide has been modified somewhat to clarify the NRC Staff's intent in certain areas. The ACRS 'has long advocated use of best estimate or realistic evalua-tions for safety analysis. We believe the proposed rule is a major step forward,in this effort, and we support its adoption. We wish to note the following points:

• Work to demonstrate the Code Scaling, Applicability, and Uncer-

. tainty (CSAU) method for the peak cladding temperature calculated to occur in the reflood phase of a large break LOCA has not been completed. This will be needed to establish guidelines for Staff review of future licensee submittals under the new rule. While the CSAU method has been reasonably demonstrated for the so-called 147

The Honorable Lando W. Zech, Jr. May 10, 1988 blowdown peak, application to the reflood demonstration will be more difficult. We do not object to plans to proceed with promulgation of the rule change, but we would like to be kept informed about the development of and allowance for uncertainty in the reflood peak temperature.

• We note that the draft Federal Register notice provided to support the rule change has eliminated reference to any claimed safety advantages for the rule.

We believe the safety advantages are substantial. Additional comments by ACRS Member Harold W. Lewis are presented below. Sincerely, W. Kerr Chairman Additional Comments by ACRS Member Harold W. Lewis I Luve no quarrel with the Committee's letter, but want to seize the opportunity to reinforce a point that has been made before. It is stimulated by unsatisfactory answers to questions at the presentation to 'the Committee. The CSAU " methodology" purports to be a systematic procedure for esti- ) mating the uncertainty in code calculations. That is a laucable objec-tive, and its achievement would be even more laudable. It would be helpful if, in so doing, there were less confusion between the concepts of uncertainty and a probability distribution, and less misuse of the term " confidence limits." These objectives will not be reached unless some professional statisticians become involved. In this case, it is of more than usual importance, since the uncertainty is directly related to the acceptable level of conservatism which must be added to the realis-tic calculations.

References:

1. U.S. Nuclear Regulatory Commission, Draft SECY paper for the Commissioners from V. Stello, EDO, " Revision to the ECCS Rule Contained in Appendix K and Section 50.46 of 10 CFR Part 50," provided to the ACRS, April 20, 1988. 2. U.S. Nuclear Regulatory Commission, Draft NUREG-1230, " Compendium of ECCS Research for Realistic LOCA Analysis," Office of Nuclear Regulatory Research, dated April 1987, 148

'o UNITED STATES g / NUCLEAR REGULATORY COMMISSION o ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20565 O g e.... June 7, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Cemission Washington, D.C. 2055

SUBJECT:

PROPOSED REVISIONS OF 10 CFR 20, " STANDARDS FOR PROTECTION AGAINST RADIATION" During the 338th meeting of the Advisory Comittee on Reactor Safe-guards, June 2-4, 1988, we met with the NRC Staff to discuss the pro-posed revisions to 10 CFR 20. " Standards for Protection Against Radi-ation." This was also the subject of a meeting of our Subcommittee on Occupational and Environmental Protection Systems held on May 31, 1988. i We also had the benefit of the documents referenced. As a general coment, we comend the NRC staff for its efforts in revising these regulations. They have sought to make the regulations compatible with the recommendations of the International Comission on Radiological Protection (ICRP) and the National Council on Radiation Protection and Measurements (NCRP), as well as with the standards of other Federal agencies, such as the Environmental Protection Agency (EPA). In addition, the NRC staff has carefully reviewed and evaluated comments received from several hundred outside organizations and indi-l viduals, and has revised the regulations to include the latest scien-tific data on radionuclides decay schemes and the behavior of specific radionuclides within the human body. Finally, the regulations have been expanded to include dose limits for the embryo / fetus in an occupational setting and for members of the public. In presenting the revised regulations, the NRC staff stated that a number of related issues, such as the designation of radioactive wastes l that are below regulatory concern and the establishment of procedures for estimating doses to the embryo / fetus, would be covered elsewhere. To Several of these issues are to be handled via regulatory guides. ensure that all licensees are aware of the subsequent guidance to be I 4 provided, it would be helpful if a tabulation were prepared for each of these issues, coupled with the mechanism and time schedule on which they are to be resolved. In addition we offer the following specific coments: We agree that application of the comitted effective dose equiva-1. lent is the proper approach to follow in planning for radiation 149

The Honorable Lando W. Zech, Jr. 2 June 7, 1988 protection and in controlling exposures from nuclear activit'.es. However, the committed effective dose equivalent does not consti-tute a sufficient basis in itself for evaluating the potential health effects of radiation exposures in individuals. Such evaluation should be based on estimates of the actual absorbed dose for the period of exposure appropriate to the individual case. For this reason, in the case of radionuclides having long effective half-lives, it is recomended that licensees be provided the option of using the annual effective dose equivalent in the determination of compliance with 10 CFR 20. 2. The proposed regulations exempt " medical research programs" from the given dose limits; in a similar manner, they exempt excreta from medical patients for release to sanitary sewers. We suggest that an analysis be made of the potential health impacts of these exemptions. Also of possible benefit would be a survey of related practices in other countries. 3. Several of the definitions included in the proposed revision appear to be incomplete or to contain errors. These are: " Natural background" - this should emphasize that the exempted a, sources do not include those of natural origin that have been " technologically enhanced." b. "Whole-body" - this definition states that a dose equivalent to the head will be recorded as to the whole body. Consider-ation should be given to the development of weighting factors for converting partial external body exposures into equivalent whole-body doses. 4. The revised regulations do not allow any exemptions from the security requirements that cover access to licensed materials. Quantities of certain radionuclides that represent minimal risk to health should be exempted from these requirements. 5. The proposed regulations require that recipients monitor for radioactive contamination and external dose rates all transpor-tation packages labeled as containing radioactive materials. We believe that monitoring for external radiation levels should be required only for those packages that are required to have a warning label for external radiation. 150

i The Honorable Lando W. Zech, Jr. 3 June 7, 1988 We trust these comments will be helpful. Sincerely, W. Kerr Chaiman

References:

1. Memorandum dated May 20, 1988 from B. M. Morris, NRC Office of Nuclear Regulatory Research, to R. F. Fraley, ACRS, transmitting Proposed Final' Rule, 10 CFR Part 20, " Standards for Protection Against Radiation." 2. National Council on Radiation Protection and Measurements, " Recommendations on Limits for Exposure to Ionizing Radiation," Report No. 91, Bethesda, MD, June 1, 1987 151

o UNITED STATES / y-g NUCLEAR REGULATORY COMMISSION c, a,$ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555

• o#g 4

October 11, 1988 9 o.... l l l l MEMORANDUM FOR: Victor Stello, Jr. j E cutive Director for Operations l FROM: a ond F. Fraley, Director Advisory Committee on Reactor Safeguards

SUBJECT:

ACRS ACTION ON THE PROPOSED AMENDMENT OF 10 CFR 50, APPENDIX J, " PRIMARY CONTAINMENT LEAKAGE TESTING FOR i WATER-COOLED POWER REACTORS" During its 342nd meeting, October 6-7, 1988, the Advisory Committee on Reactor Safeguards was informed of the nature of the proposed Amendment of 10 CFR 50, Appendix J that explicitly permits the use of the Mass Point statistical data analysis method for calcult. ting containment leakage rates. The ACRS has no objection to the issuance of this amendment. Reference Memorandum from Eric S. Beckjord, Director, RES, to Edward L. Jordan, Chairman, CRGR, transmitting (to CRGR and ACRS) the proposed Amendment of 10 CFR 50, Appendix J to permit Mass Point analysis, dated September 15, 1988. cc: S. J. Chilk, SECY J. Hoyle, EDO E. Beckjord, RES C. Bartlett, RES G. Arndt, RES D. Persinko, NRR 153

$anaao

1. o UNITED STATES 8

NUCLEAR REGULATORY COMMISSION o { I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20655 o '+, * * * *

• o*g g

October 13, 1988 The Honorable Lando W. Zech, Jr. Cnairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

LICENSING 0F ALL CHEMICAL IS0 TOPE ENRICHMENT, INC. FACILITIES During the 342nd meeting of the Advisory Committee on Reactor Safe-guards, October 6-7, 1988, we discussed the applications of All Chemical Isotope Enrichment, Inc. (ALChemIE) to modify and operate a facility for the separation of stable isotopes at Oak Ridge, Tennessee. and to similar facility at Oliver Springs, Tennessee, near Oak construct a Ridge. During the meeting, we had discussions with representatives of the NRC staff, the Applicant, and the Department of Energy. We also had the benefit of the documents referenced. The Applicant plans to use centrifugal enrichment machines purchased from the Department of Energy. These machines were designed and built for the enrichment of uranium in a demonstration program, and offered for sale when the project was abandoned. The proposed use is for the commercial enrichment of a wide variety of stable isotopes, in larger quantities than have heretofore been available. Most of these machines, because of their prior use, are slight 1;< contam-inated with uranium, in the centrifuges and in the piping. The amounts total a few tens of kilograms. The average enrichment of the uranium 235 is very slightly over the natural concentration of uranium 235 in natural uranium, again because of the earlier use. There are two principal licensing issues. One involves the potential release of the uranium contaminant, either through leaching into the process stream or through an accident, with consequent health effects to the exposed population. The other involves safeguards, because these machines were designed to separate uranium isotopes, and therefore, regardless of their intended use have that capability. The NRC staff has concluded that the potential for release of uranium is small, both because of the adherence of the contaminant to the surfaces it now covers and because of the small quantities involved. It has estimated that, even in the event of a substantial accident, the public exposure would be negligible. We concur in this finding. 155 'I _______ A

The Honorable Lando W. Zech, Jr. October 13, 1988 The safeguards issues have to do with protecting the machines and inspecting the product. It is proposed to deal with these issues through a conventional program of personnel clearances, area controls, and random unannounced inspections. The NRC staff has expressed satis-faction with the proposed program, and we have no reason to disagree. Subject to implementation of the planned controls, we support issuance of the proposed licenses. Sincerely, William Kerr Chairman

References:

'1. U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Safety Evaluation Report Related to the Construction Modification and Licensing of the ALChemIE Facility-1 CPDF, October 1988 2. U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Safety Evaluation Report Related to the Application for Construction Permit for ALChemIE Facility 2, Oliver Springs, October 1988 3. Security Plan for All Chemical Isotope Enrichment, Inc., ALChemIE Facility 1 - CPDF, Revision 3, August 1988 (Proprietary) 4. Security Plan for All Chemical Isotope Enrichment, Inc., ALChemIE Facility 2, Revision 3, August 1988 (Proprietary) 156

o . UNITED STATES g. NUCLEAR REGULATORY COMMISSION- . [. g .a ' ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g-wasmuoTow.o.c.zosss s February 16 1988 m, The Honorable.Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington', D.C. 20555

SUBJECT:

ACRS COMMENTS ON DEVELOPMENT-OF A METHOD TO ESTABLISH PRIORITIES FOR RESEARCH ACTIVITIES During the 334th meeting of the. Advisory Committee on Reactor. Safe-guards, February 11-13, 1988, we were briefed by our Subcommittee on the Safety Research Program re ding activities' within the Office of - Nuclear Regulatory Research (garRES) to develop and apply a method for the of priorities to research activities. This subject was assignment discussed.with the RES Staff and its contractors' during a Subcommittee meeting in Washington, D.C. on January 29, 1988. The Subcommittee also i had the benefit of the document referenced. The method being developed places major emphasis on: ' translating the NRC strategic goals into. questions that need to be and can be answered through research, and

• evaluatir.g each current or proposed research activity in terms of its usefulness to answer one or more of these questions.

We believe that such an approach is essential if the NRC is to have a research program that is clearly responsive to its needs and has rea-sonable assurance of obtaining results that can be and will be used. We are impressed with the work that has been done so far in developing and testing this method. We believe that the result will be a docu-mented, defensible, and useful tool for the Director of RES and for the Commission. The annual cost of using this method is estimated to be less than one-tenth-of one percent of the RES budget. l 157

l The Honorable Lando W. Zech, Jr. February 16, 1988 For these reasons, we recomend that the development of this method be completed and a program be initiated to implement this method. We expect to follow these activities. Sincerely, 1 i William Kerr Chairman 1 1

Reference:

Scientech, Inc., SCIE-365-87, " Report Office of Nuclear Regulatory Research, Research Assessment Methodology, Task II - Trial of Meth-odology," prepared for U.S. NRC, dated January 1987 158

l \\ c :rgy UNITED STATES o,, 8 NUCLEAR REGULATORY COMMISSION g a ADVISORY COMMITTEE ON REACTOR SAFEQUARDS WASHINGTON, D. C. 20666 February 17, 1988 The Honorable George H. W. Bush President of the Senate Washington, D.C. 20510

Dear Mr. President:

In accordance with the requirements of Section 29 of the Atomic Energy Act of 1954, as amended by Section 5 of Public Law 95-209, the Advisory Comittee on Reactor Safeguards has reported each year to the Congress on the Safety Research Program of the Nuclear Regulatory Comission. In our December 18, 1986 letter to you, we proposed to provide more focused reports devoted to fewer significant research issues that we believe merit Congressional attention rather than one all-inclusive annual report. The Commission has agreed with our suggestion, and NRC Chairman Zech has submitted a legislative proposal in the form of a draft bill on December 2, 1987 to amend Section 29 of the Atomic Energy Act of 1954 to accomplish this. In our December 18th letter we noted also that, in the event that a year passes without the appearance of any significant research issue, we would report that fact to you in writing. Although in the past year we have not encountered any major research issues that we believe merit Congressional attention, we have provided coments and recommendations to the Comission and the NRC Staff on several NRC research efforts. Copies of these reports are attached for your information and use. I We are in the process of reviewing several elements of the NRC research program such as Thermal-Hydraulic _ Phenomena, Severe Accident Management, Plant Aging, Containment Performance, and Seismic Design Margins. Also, we expect to review proposed research associated with Human Factors, Equipment Qualification, and Fire Protection. Upon completion of our reviews, we will provide reports to you on those specific items found to be sufficiently substantive to warrant Congressional attention. Sincerely, William Kerr Chairman 159

I I .The Honorable George H. W. Bush February 17, 1988

• Attachments:

1.. Letter from William Kerr, ACRS Chairman, to Lando W. Zech, U.S. NRC Chainnan,

Subject:

ACRS Report on Proposed Research to Reduce Source Tenn Uncertainty, dated May 13, 1987. 2. Letter from William Kerr, ACRS Chaiman, to Victor Stello Execu-tive Director for Operations, U.S. NRC,

Subject:

ACRS Coments on the Embrittlement of Structural Steel, dated July 15, 1987. 3. Letter from William Kerr, ACRS Chainnan, to Victor Stello, Execu-tive Director for Operations, U.S.-NRC,

Subject:

ACRS Coments on International Cooperation on Research Related to Radiation Protec-tien, dated July 15, 1987. 4. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chairman,

Subject:

ACRS Comments' on Draft NUREG-1150, " Reactor Risk Reference Document " dated July 15, 1987. 5. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chainnan.

Subject:

ACRS Coments on Research into Continuous Containment Leakage Monitoring, dated July 16, 1987. 6. Letter from William Kerr, ACRS Chairman, to Lando W. Zech, U.S. NRC-Chainnan,

Subject:

- Preliminary ACRS Views on Fire Risk Research . Scoping Study, dated August 10, 1987. 7. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chainnan,

Subject:

ACRS Coments on Code Scaling, Applicability and Uncertainty Methodology for Determination of Uncertainty [ Associated With the Use of Realistic ECCS Evaluation Models, dated September 16, 1987. 8. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC i Chainnan,

Subject:

ACRS Comments on Radioactive Waste Management l Research and Other Activities, dated November 10, 1987. 9. Letter from William Kerr, ACRS Chairman, to Victor Stello, Execu-tive Director for Operations, U.S. NRC,

Subject:

ACRS Coments on Memorandum from Victor Stello, Jr., EDO, Dated October 7,1937 Regarding the Embrittlement of Structural Steel, dated December 8, i 1987.

10. ' Letter from William Kerr, ACRS Chairman, to Lando W. Zech, U.S. NRC Chainnan,

Subject:

ACRS Comments on Development of a Method to Establish Priorities. for Research Activities, dated February 16, 1988.

11. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chairman,

Subject:

ACRS Coments on Selected FY 1988 NRC Radio-active Waste Management Research Programs, dated February 17, 1988.

• For Items 1 through 9, see NUREG1125. Vol. 9, 4/88 For Items'10 and 11, see pp. 157 and 167, this volume 160

UNITED STATES f NUCLEAR REGULATORY COMMISSION j ADVISORY COMMITTEE ON REACTOR SAFEGUARD 6 { WASHINGTON, D. C. ;t0886 February 17, 1988 i The Honorable James C. Wright, Jr. Speaker of the United States House of Representatives 1 Washington, D.C. 20515 /

Dear Mr. Speaker:

In accordance with the requirements of Section 29 of the Atomic Energy Act of 1954, as amended by Section 5 of Public Law 95-209, the Advisory Comittee on Reactor Safeguards has reported each year to the Congress on the Safety Research Program of the Nuclear Regulatory Comission. In our December 18, 1986 letter to you, we proposed to provide more focused - reports devoted to -fewer significant research issues that we believe merit Congressional attention rather than one all-inclusive L annual : report. The Comission has agreed with our suggestion, and NRC I Chairman Zech has submitted a legislative proRosal in the form of a draft bill on December 2,1987 to amend Section 29 of the Atomic Energy ' Act of 1954 to accomplish.this. In our. December 18th letter we noted also that, in the event that a year passes without the appearance of any significant research issue, we would report that fact to you in writing.. Although in the past year we have not encountered any major research issues that-we believe merit Congressional attention, we have provided coments and recommendations to the. Commission and the NRC Staff on. several NRC research efforts. Copies of these reports are attached for your information and use. We are in-the process of reviewing several elements of the NRC research program such as Thermal-Hydraulic Phenomena, Severe Accident Management, Plant Aging, Containment Performance, and Seismic Design Margins. Also, we expect to review proposed ' research associated with Human Factors, Equipment Qualification, and Fire Protection. Upon completion of our reviews, we will provide reports to you on those specific items found to be sufficiently substantive to warrant Congres.sional attention. Sincerely, William Kerr Chairman 161

The Honorable George H. W. Bush -2 February 17, 1988

• Attachments:

1. Letter from William Kerr, ACRS Chairman, to Lando W. Zech, U.S. NRC Chainnan,

Subject:

ACRS Report on Proposed Research to Reduce Source Term Uncertainty, dated May 13, 1987, 2. Letter from William Kerr, ACRS Chainnan, to Victor Stello, Execu-tive Director for Operations, U.S. NRC,

Subject:

ACRS Coments on. the Embrittlement of Structural Steel, dated July 15, 1987. 3. Letter from William Kerr, ACRS Chairman, to Victor Stello, Execu-tive Director for Operations, U.S. NRC,

Subject:

ACRS Coments on International Cooperation on Research Related to Radiation Protec-tion, dated July 15, 1987. 4. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC l Chairman,

Subject:

ACRS Comments on Draft NUREG-1150, " Reactor l Risk Reference Document," dated July 15, 1987. 5. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chairman,

Subject:

ACRS Comments on Research into Continuous Containment Leakage Monitoring, dated July 16, 1987. 6. Letter from William Kerr, ACRS Chairman, to Lando W. Zech, U.S. NRC Chairman,.

Subject:

Preliminary ACRS Views on Fire Risk Research Scoping Study, dated August 10, 1987. 7. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chainnan,

Subject:

ACRS Coments on Code Scaling, Applicability and Uncertainty Methodology for Determination of Uncertainty Associated With the Use of Realistic ECCS Evaluation Models, dated September 16, 1987. 8. Letter from William Kerr, ACRS Chairman, to Lando W. Zech, U.S. NRC Chairman,

Subject:

ACRS Comments on Radioactive Waste Management Research and Other Activities, dated November 10, 1987 9. Letter from William Kerr, ACRS Chairman, to Victor Stello, Execu-tive Director for Operations, U.S. NRC,

Subject:

ACRS Coments on Memorandum from Victor Stello, Jr., EDO, Dated October 7, 1987 Regarding the Embrittlement 'of Structural Steel, dated December 8, 1987.

10. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chainnan, -

Subject:

ACRS Coments on Development of a Method to Establish Priorities for Research Activities, dated February 16, 1988.

11. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chairman,

Subject:

ACRS Coments on Selected FY 1988 NRC Radio-active Waste Management Research Programs, dated February 17, 1988.

• For Items 1 through 9, see NUREG1125, Vol. 9, 4/88 For Items 10 and 11, see pp. 157 and 167, this volume 162

WE%'o UNITED STATES g 8 NUCLEAR REGULATORY COMMISSION o f 7 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 9 / WASHINGTON, D. C. 20555 o May 10, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555 )

SUBJECT:

PROPOSED GENERIC LETTER ON INDIVIDUAL PLANT EXAMINATIONS AND THE PROPOSED INTEGRATED SAFETY ASSESSMENT PROGRAM II During the 337th meeting of the Advisory Committee on Reactor Safe-guards, May 5-7, 1988, we discussed a draft generic letter prepared by the NRC Staff as guidance for Individual Plant Examinations (IPEs) for severe accident vulnerabilities. We also discussed the proposed Integrated Safety Assessment Program II (ISAP II) and related informa-tion. Both of these topics have been considered during previous meetings of the ACRS, and we reported our preliminary views on the IPE generic letter in our report of June 9, 1987 and on the ISAP process in our report of July 15, 1987. The ACRS Subcommittee on Severe Accidents met on April 26, 1988 to discuss the latest version of the proposed generic letter on IPEs. The ACRS Subcommittee on Generic Items met on April 27, 1988 to discuss ISAP II. We also had the benefit of discussions on both topics with members of the NRC Staff and industry representatives, as appropriate, and the availability of the documents referenced. These two programs developed by different NRC Staff groups have not been integrated, even though they deal with many of the same issues. It is for this reason that we are providing our comments on both programs in a single letter. The present Staff positions, as we understand them, are that the IPE generic letter should be issued in-its present form and that implementation of the ISAP II should not be pursued at this time. We disagree with both of these positions. Instead, we believe that the IPE program should be expanded to incor-porate all outstanding safety issues, not just those under the severe accident rubric. The generic letter should be revised accordingly. The ISAP 11 approach should then serve as the instrument by which changes in plant equipment or procedures identified by the IPE could be evaluated and assigned priority by the licensees and reviewed by the NRC Staff. i 163

l The Honorable Lando W. Zech, Jr. May 10, 1988 z l 4 \\ 4 We consider the most recent draft of the IPE generic letter an im-l provement over that which we comented on in our report of June 9, 1987. However, in our report of March 15, 1988, we expressed our concern that there was a lack of coherence among the several principal regulatory programs of the Commission. We believe the IPE program offers an opportunity for providing improved coherence. In its present form, the generic letter will, instead, continue the current compartmentalization. 3 We believe that IPE and ISAP II can be recast in a reasonable time and with reasonable expenditure of resources. Radical changes are not necessary. but some modifications and improvements in focus are. We propose a program characterized as follows:

• The purpose of IPEs would be acknowledged as broader than the original intent of " searching for outliers."

Instead, it would call for a general risk recsses'sment of each plant using the body of information available from the TMI-2 accident experience, development of PRA, existing severe accident research, and the general experience of about 1100 reactor-years. All outstanding safety issues. USIs, GIs, etc., would be subsumed by the program. It would be made clear that the intent of the program would be i for this to be the end of new requirements for licensees. This would be changed only by the advent of important new information or experience. We note that the IPE program proposed by the NRC Staff already has been expanded well beyond the " search for outliers" concept. In subsuming USI A-45, " Shutdown Decay Heat Removal Require- ) ments," into the IPE, for example, the Staff has taken a major step in the direction we are suggesting. Our proposal extends this to a more logical conclusion.

• Each licensee would be required to conduct a substantial and systematic risk analysis for their plant. We recommend that such an analysis would be a full scope PRA (at least Level 2) and include both external and internal initiators.

We acknowledge the difficulties inherent in making this an immediate require-ment. However, it should be ppssible to develop a phased ap-proach with the intent that within several years each plant would have been analyzed by state-of-the-art methods.

• Conclusions about results of the risk analysis and necessary changes in actual plant systems and procedures would be de-termined by the licensee and reviewed by the NRC Staff through 164

The Honorable Lando W. Zech, Jr. May 10, 1988 the ISAP process. We believe the risk-based approach embodied in ISAP is the most logical means for resolving most safety issues. The risk analysis used in the IPE for each plant will be available for use by the licensee and NRC Staff in their ISAP evaluations. We believe that the approach we have outlined above will provide the opportunity for a more integrated resolution of severe accident issues and other outstanding safety and licensing issues as well. We endorse current efforts on the part of the NRC Staff to formulate an inte-grated' program. Sincerely, W. Kerr Chairman f

References:

1 1. U.S. Nuclear Regulatory Commission, NRR Generic Letter 88-02, dated January 20, 1988, " Integrated Safety Assessment Program II (ISAP II)." 2. Memorandum dated March 1,1988, from T. Speis (NRC) to D. Ross (NRC), et. al., " Commission Paper on Integrated Approach to Implementing the Commission's Policy on Severe Accidents" (Draft). 3. Memorandum dated April 1,1988, from T. Speis (NRC) to W. Kerr (ACRS), " Documentation Necessary for the Initiation of the Severe Accident Policy Implementation" (Draft Predecisional Attach- ] ments). 4. Draft SECY Paper (undated), Integrated Safety Assessment Program II (Predecisional Document), received April 26, 1988, l ) i 165

1. p e ng'o UNITED STATES g

8 NUCLEAR REGULATORY COMMISSION 1 n ) { ,E ADVlsORY COMr ' TEE ON REACTOR SAFEGUARDS o WASHINGTON, D. C. 20555 February 17, 1988 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON SELECTED FY 1988 NRC RADI0 ACTIVE WASTE MANAGEMENT RESEARCH PROGRAMS During the 334th meeting of the Advisory Committee on Reactor Safe-guards, February 11-13, 1988, we met with the NRC Staff to discuss selected radioactive waste management research programs. These dis-cussions included a review of the impacts of the Nuclear Waste Policy Amendments Act of 1987 on these programs. These matters were also discussed with the NRC Staff during a meeting on January 21-22, 1988 of our Subcommittee on Waste Management. Our comments on the research pro-grams follow. Research on High-level Waste (HLW) FY 1988 is the initial year of operation of the Center for Nuclear Waste Regulatory Analyses. We endorse the ongoing efforts by the NRC Staff to ensure that this organization obtains competent personnel to fulfill its assigned responsibilities. In addition, we recommend that procedures be established to provide periodic, critical, technical, external review of the performance of this organization to ensure that its work product is of high ovality. To the extent practical, the ACRS Subcommittee on Waste Management is willing to assist in such reviews. Research on Low-Level Waste (LLW) The available program plans for FY 1988 do not include any research on the effects of organic chelating compounds on the behavior of radio-nuclides in low-level wastes. Since such materials can have significant effects on the mobility of radionuclides migrating from a site, and since there is increasing concern about the management and disposal of mixed wastes that frequently contain such materials, we recomend that plans and resource allocations for FY 1988 be changed to include studies on this subject. We also reviewed the joint NRC/ Canadian efforts to determine the ade-quacy and applicability of models for predicting the movement of radio-nuclides through the groundwater and soil near LLW disposal facilities. Although agreement between the model predictions and the data collected 167

The Honorable Lando W. Zech, Jr. February 17, 1988 at the sites is good, this may be fortuitous. Additional effort should be made to provide data that will make it possible to rationalize the selection and variation of parameters, e.g., retardation factors, in these models. In addition, efforts should be made to develop measure-ment protocols to ensure that data for the key parameters needed for input into the models can be made available. This work should be summarized in a timely manner so that the results can be applied to the needs of the regional compacts currently evaluating and selecting LLW disposal sites. Other topics that were included in our review and need to be further addressed by the NRC Staff are research on: (1) reliable methods for the solidification of LLW, particularly in a concrete matrix, and (2) ' improved environmental monitoring programs for LLW disposal sites. Of the two, we believe that the first, reliable methods for the solidi-fication of LLW, is the more important. Sincerely, William Kerr Chairman 4 168

/ 'o,, UNITED STATES / NUCLEAR REGULATORY COMMISSION o { r,$ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS O WASHINGTON, D. C. 20555 l Apri1 12, 1988 Mr. Victor Stello, Jr. Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS WASTE MANAGEMENT SUBCOMMITTEE REPORT ON Q-LIST TECHNICAL POSITI.0N Enclosed is a report of the ACRS Subcommittee on Waste Management relative' to the Technical Posi. tion 'being developed by the NRC Staff on " Items and Activi-ties in the High-level Waste Geologic Repository Program Subject to Quality Assurance Requirements," which was provided to the ACRS during its 336th meeting, April 7-9, 1988, A copy is forwarded for your information. i Sincerely, 4 Ra ond F. Fraley V Executive Director

Enclosure:

Report of the ACRS Subcommittee on Waste Management dated April 8,1988 cc: Hugh L. Thompson, Jr., NMSS 169 L.__________.--__-

l 1 REPORT OF THE ACRS SUBCOMMITTEE ON WASTE MANAGEMENT l April 8, 1988 l During a meeting on March 17, 1988, the ACRS Subcommittee on Waste Managerient i met with the NRC Staff to discuss the final draft of the Q-List Technical Position that provides the NRC Staff position on QA criteria for licensing a geologic repository for the disposal of high-level waste. On the basis of these discussions, the Subcommittee concluded that the final draft Q-List Technical Position is a sound document and that the NRC Staff has done an excellent job, particularly in subjecting the report to thorough review by outside groups and in being responsive to the comments submitted. The Subcommittee concurs in its publication. Interactions with waste management personnel of the U.S. Department of Energy indicate that, for their purpose, the issuance of Technical Positions in the form of Regulatory Guides would resolve certain questions, would permit the document to be more readily referenced, and would avoid potential problems that could develop at later steps in the repository licensing process. The Subcom-mittee understands that personnel in the NRC Division of High-level Waste Management concur in these comments and are exploring mechanisms to have this and similar documents issued as Regulatory Guides. The Subcommittee concludes that formulation of geologic repository-related Technical Positions in the form of Regulatory Guides is desirable and recommends that steps be taken to accom-plish this. 170

ui muc6tui 8:vutom, ca .o=

i. Ra,on, avu.am,A

.,, m.. ees. Au v., ._me,on. an E*- BIBLIOGRAPHIC DATA SHEET NUREG-ll25, volume 10 Sgt eksimuCTeoNS o.4 tut stytm58 31TLEAmoSv0111Lt 3 Leavestans A Compilation of Reports of the Advisory Committee on Reactor Safeguards. 1988 Annual ) , tan 1 g o. e extaomm January 1989 e oats agront assuto . con T ee vtan April 1989 y pensomuiwG omaam:2Aliom hAuf ANo uatta=G ADoniss,snewer t, C, a PaoJtC1,TA&serons West Nueseth Advisory Committee on Reactor Safeguards '"*"5"^"'"""" U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l t .om..o om A=> Av.o A i *=o a u o Aoomsu,,

t. C,

ii. ivn o, Ateoai Same as above Compilation . nnioo Covinto,, m January thru December 88 33 SWPP tWahT Anv hott$ k

14. ASOTmACT 8380 ewe

ems, This compilation contains 47 ACRS reports submitted to the Commission or to the Executive Director for Operations during calendar year 1988.

It also includes a report to the Congress on the NRC Safety Research Program. All reports have been made available to the public through the NRC Public Document Room and the U.S. Library of Congress. The reports are divided into two groups:ACRS Reports on Generic, ACRS Reports on Project Part 1 Subjects. Part 1 Reviews, and Part 2: contains ACRS reports alphabetized by project name and within project name by chronological order. Part 2 categorizes the reports by the most appropriate generic subject area and within subject area by chronological order. 1 14 ooCvestes? Amatvsas.e stvwomol/otSCRsPfons IS A V A,LAS'bi1 V Nuclear Reactors Safety Engineering Nuclear Reactor Safety Safety Research Unlimited Reactor Operations is SBCualTVCLAssifiCAfloh (Tnn eno., Unclassified . cos=visitas, ort = t=oto vin s <r.. e Unclassified it hvee842 of PAGES is emeC5

• U. $. C0vt RhE. h f DR ikf1NG Orr 1CE i199 9.234-611

c '., f\\p. g ^ * ~ h, f' i ^', h,- ,'.s .;,;+, .u' lj

• " i..

2.. . f ' 4 .l ...v4r ' i' - -&s. - f. s*,'f,,. ,<g ',,,Q,,,,. c y, Y 4 s 3 N .n'.,.?;.gi)'. t o Y i,. ?j:, 't '., - i,'i a ^i.. t.-l ,.n'( ...,'),}. U,c.", g ,..s _-, -y. (o f 'l. ,,L 1-J s I ,..,.~.L,-.i.-,.y [l\\. .'.l. (.,/, .p,1 v l s., a-u <y :an .t i,0., i..;-'., 15f:.-- 0,,,.* : . ';)d5 ,r;l}; a,.b, /, - .,-,. '} q,

?

sg L, .u:;. y < ,.'u.0,:-l,.,, -

,.r

,'.',l .., *,.t f ,.0t .,.-.( g~. '..j' ( !.i

l

'G,,.4 5, s. - s + z 'l".l g,. -,?. ; f,l 4 ,.'.j;..-';:,' i. ; i ^,{,/,. l ,s,. ' '. s [' ',-*;e'.' :. ', o: '.J', 6 i..j:.F.. t _;',l's '+') _ - f,. .y e

. 4p

,'l

'..,o;.;..'

4;'h,::... w, f'.. ",,,.. -: :.gy y. ' w... '.., 4.-,', $;$. :.I'.;%, i. Y, .e- , p,g- ..p,~r.- ) -,d";.v:. f,j '.l. {o ^

g,.

3. 'L ;,. ',.,,. s:- \\'[,:a.g ; c .u',, .i

.e-.lg.

- H8. -:.,,., -^',;..,, - ' =.t '.Vi .l' ,y.- {# 5','.. ', ;,, y y... -' p' ;:*, '. s,G,'Q ', S ' l, -*&. *',,!'),r* -f,' a,',,- s, ;,,. y.,.-. -'. - N <n, -v,, ,L,. '4 7 ,[ -ff s, .-i,+ - c)y. ', ,i .d,, .L;

g.,.,,..*

1 .s; e. ,f[ ) l .,'f.',i .b:.k 1 ;? s. 14-3.s .v.q. " h, ' f.,. ,.) -' -% 'll c'.,,. .A' ? e' h. ( s,.':,i,j. 'n,V' -s -,...,., p p '

• .l J;

>t ^,. "; [t,-,ee,e, -,..'3.

• ,1

'.,? +t,

r..

g _* 1 9' ,/'i,,-.,. ', \\ 'l, l 5, .,.*;.i ,.'c. ',. + 'f y, l;. N'.e

.'l;

.n ^ .;1 :-' i , '..,, -j '.s..,v. 6 4 .

• a.,;.

\\ ,)' {,.?G ,f. .,H' i.;'; -'W ( - N('k, g', t,.. , :,z.,.< 'Q ;'. F .., ^ ; .,.,'6 c.,r' ),t ,. df..- "t

" h' ls. 4,,. -.

.,e,y;~,g $.. c ', p e..' ,.'f. e ,.+'b,.. _ .,e i?.- ?'-W l'.'.'..',,l1', h h',. ',. ,,,-*;-,' Y. 1' ,',5 i' l. l l.- .a. _. 1 o',. '[ 3 g' +l 4,,. 7 [ -. y.,.,jas3< j,, -y ) , i.- ,-c Q.U?' \\,.' .'a _'\\ l R' ,t

• i

): % ?l. .dy

.-l.!,,.,

l, J,., + r 1 n r .;, ' y.i ,9- ' i'3's. .l:,. ,c lQ,. +, ' <;.. f < -,-^/,.- .4,. -,.- *'.i., '..- : Q T '^.o, ' 'L'

u..Y?

l .. +/. s:

• l a

-)lpr s Yv ,Y,: s' e- \\6,; ...4,P a. , -' *)' '.... ~ d,., .',vg". l.:,s ?.: 'l/- 31,.

9

,,.$"-.l.; ,? I',., -., ' ',. ,t ll ^ W'; ,.i, g, -. MT',,- ' i a \\. 'f.'

d. gi.',',.-,

6 .Q, , _. '.;' i. h.:lp N.. +,, f.< ,s .*t ,i -_.,j. ?: ,:. ' aq'_k U:. Yhh;q, ~. s r }.-c t 43h' ..-'j U.,y- ., p

• J^ 'A

,\\'c. 'f., I\\a 5 l' t)y - ..s ':j._, 7, .f.I y 'l, -. )w, :t'. k Vs.

v,-

n ,. g.-

• f,,.

.l:;, }e,!;l b f ,.l e' .g 5.. .-,., e /,'s l L ?' a.

p'[lr

',...q'

u.,. /*..,,.

L.',.o (, . l* c,:_.. sq)n - ; '_'..','.!. 'c ,1?. s, l / . : n..,: .. [q,',:'l s 9

',

g

,j _ -, -.,,N.r t 4.,., - ;j I..(c s'?.,. ', R }.., ' ': ;,Y,,- .l ,.. s .r. . /'. f...,', gh .p> s 3 Y Q ;V. '}l',,

L: ',,... -

4:.; j. ~,. yy 'N'

J.,:

f ,,c i, >g

• 'l

, y' 4... '.. -. f w ;. '.,.1'z l ,g ^

• lI.

I'. +y. As e n:.,',. l s' r' .. $f p ;,p,; ?,..[. l i;, - -. i3,:' 4..-;.. + ', ' 2 ' ~. '44 , - s.,sd,. 7, ^

-j.

.t s. ?,:0

i

." ; I';*.: p;1 .\\ j$.,

h. &,.lf, :, 's ',',. ',

l .-',h, N .->... ' h;;'. ' 4 - 6,, ,tf % j;' :' _,,(, .,-' c,f l ' :' i -. .s - e y' $ ?' l'. n.,.'.-'.,,' c .-+ \\

• • t
• ,}

--l:T.,

.a f.i e.

1 f l.% l, 1 i 4:g'G;. ': ~ ,:,..c,; ... a{l l'-,.yp yfI. .r}.. ; - '. <..

d"'

^ ~ :' t I l .g l,I ,s .,3'>-

f. ',..~ ; : >..,.. [kf'.,:.Nh j.'-

k ~ ?.

Q

f.. : '..[)*,

f , v , ? .s l:

5 Nl' y' i :' c) g ).

-' : ~ _.. -J3..;

?4,s .s., .,*s,8 '.,f q.y; * *. :*,. ',. *

y.,.

n'. e ' ' '..!.^g q .V M .f':_': .i ?'/ *..v 'i., -g %.','.',.:-l ,l .v .L'I "

f sk s;..g,.e-j

'h.'.-. -:,'. ;l ..? .l,4 .,) ,h,_. *.. =,.. = - ',.. =r 3 , ' '..,,;:0,?s., 'V g i. ( c' , ',.si;.,.,.,.,., -1,,,b.f,!,.., .^ ' : *. - - Q ' ' f.ta;,.. f,l - :.':... .[ h({il_. L) .- { I' .,s. s'.;'; y' r * * [' e 1, b; ' -, Y' h',,- v ~. y ,.'.t- ,.e,-( ey,,-.: .l..; ?. .g

l.

i ,?, .:o.s N

i.

g

• ,~.

ll,,.,,.]; . p,.',: - +, 2,;'.. l.", a._. C' ',l ^ } c Lt v._, .-lt; 1 r. s i,'s. .,l','. j ['. l n ' c'/ ,f-, l,s: %. [ t t's:' +, .ml,. ;4 ;1, ;;;.?',... a',* s' e., t f!,*'l/ ..,.s-., f,, - t', t.' o ?' ,. :i '

}

'.el'.I'- ,)

t.

[- s .. i. -,-; p l ay ',l* l.A,,. g,, "' '.,h ) - n 1 .iK,' *3' a. :_ , b. !l c, s 1V6 :l > < -y O.. r j ['_ e .,.".\\. g'a .f, ', )' - ' , ,-'li, N:. ' u;* l,, 136,' l

• ,s,:

y,,,. -." ) b g.'j.,,.l.:'! s k: r ~ ,:t_ g - ..;i' <>,ll.-,l',s l - l4 .'y\\,', Ln ',~ ',... -. ,,'l s' f., t I4., -, -, : [ '_ N, f'l'f'_ r () i,.

, ~ \\

? ' c ,e ;*.4.ll: s 'g I

y, I.

.g

1f.,,\\llq t

. Y: >N.,., ?

,, L

y' 6,, i,i '.- ; ),g

\\.;_

s. '(

• ,a'.

s' . '..a.. ',.. -. l o Y ,3.',- e. ,y;s. f ..- l t.f',.,. ' e ) m.- .{. I

• ','.,c.

h [ l I.,/:. f' l: )> - s<...,.,., ';.-..-.,.;t..',.. -, l. : '... -*;3. .s-

a.. *

,-*\\, .a.Y. ;' f'. Qc,. l' . s ..,.,> y '3

j..-,

.t .'I

\\ i.:, ;

l' ,d. '..

• A,I:*

,., ).r. i.', -. t e' ,M', e c;e.j. G:.- y -d f. }J' '?',

-lf.

',','f N. a,#.d - ;: ; {,,. ~., - 'ps' ,, cm,.,...' - '., - ', a ", Y' '. -'

• l d,

l 3 L',. ),; 'l: -f; [.:' .'.,p : ji

,..' i v i!.

• ..

c,' s 4,' *: **,..,n,. [, ,4.. ;., pj :' g', -. ,';-$N (, nl;l..- 4,1,, 8 s p4, 7 ,. q,,'q l - --,, i f l ~ _.-

• Q

.r

~ u,b ,-.'[ .t [ - ; y;. -( 'o ,; i 'l? v. ,3 c,:,, I..; -l j' l'

9. p *.,

, ^; ,,1.

• l

,v:. J,. ,f,.. ;. -y.:, 'i .i. c, 5' 6' .q' ,_.' i, i.* ' [j 5 i:. -. ;y ; ;. ' ,cf{. ,.. -,.,;..;..$h,;. '.;1. 4*;;: ,,} "[r.- , l 8-f_ '.. a'n' ,s 'g,,*'l.: ,1 b,

• ,'l'.i'

- *e.. t~ y- .-.i *. { s.- ,l- .3 <[;' .,6; ',,1, .'.On,5' ',r,., ',['_k'>l,, . ' [.,' ,1-( vq,.,\\ }; }. ['~ + ^ '4',. ,,,,g' l >,f' ' T :; l' 3;. &(:. l.,%:; 'Q' fl'.- ,o:: l/[' p ', ~ ,.,.- ',' -,. -.,,,-*,'t*..;' < lc j 6 ,.','g.';.. [. ' - I y,-)' i i ;. q- 'v. 1'

U i "

A,' t 'f. .)' -p .,4 r!,

f 8'

.'f,. t '( . *): ;.. '.-,,

.

4,.. ' - ',."

,,I.> e p, \\,: N.~ 1.. 4 ,.,.,4 +', h',. } .; p r >.', . :f, l l,.' '.'f. n, ',fu)5 ;.t\\s,.,- .,.'?, .. ' '.-+ ?/- .A l, -f., s, .a

s.','g y

',, c, O'. o.,.'"),,, ',.,'1, c v...'., ( f-9: ,r (' .c '-l'.

g

,,.s. .. ;,,; F.. 7 Y. G -,. ,v,2 -f. { ,'3. - [ 'E

d. ^ %.l:

t.,[

. F,

.l' f]',.

9,'.',- .<,-4',3 -.sf..

a.-- '...,

i...... ' 4 1 ciy.."'

3,.- 6 .f, '*s-..A..',.a.

• ^

.',f y, L, l.l >., *., l ',,-,'g'. ',,- ' i' ' e 'l. qq j. o V ..tf:,' 'sg., '.3e $.O' 1

e.

\\' [,.. l' ... l ll ,s y. s, 'p,,,. .g, . _. +,,. s,; o.i. ' ' s':.,i.'.

f,6

,. '

',L .'ll ls y e '.;,'.,i ); l;,..., f lf:::,. ' L;. _; s'\\.

,8 s

',o_- , ;e . + 9,. ll'.e s \\ lw 'h'~c

l

/ /^,'^f.h, _i l l..jy;-\\lK,,,. ,?,. }, ' i?.f.. -. ' - r, .M'_.,-: i. e

• 3 & ',.,{!'.

,r

f.. ',*:

. i .g; j e ,c 3, -p..g l' .1, g 4- ?' J

• ,~., '

f

"...'

4 l.; Qs.A,.^..r { ri. ' n - %,,,,, y.(1 [' ;g -g[ j,. >; 'i. ?1 <i,.9,. -),:',,- ii.' ;;. '

t';n1f;',s-;',

's

• ,-,^

.f ,a(4., j...,.., "l. z Y'.'., y: ,; U,;. 4' G.-)jf\\$ .- 3, / t l -l l".- , j . '...:' -l -,g ^ j .'L, ., +.. _ -fa m . '. -.,.4'._,, : : .l, .g eS \\., ,,p;. [ /. :

,>,'e

'Q.,.2 ,;, ;, y[ a,,.. .I

,:'l

.. : -l,yb... G._.. h..,','}',. i l:.. l. (,g.r ' i ."'y1.;,- ;l. '.jy I' ..';.., e -' i. c. ?' l' ^

ls

,ar,glf!. t-c:-%,( N' i,.,- ', ..' '.. ' '.l,,:$. c; l l ,-.:e-. - 'I. \\ .+ ?- ^.l', y,..,; ', : ,1!., e. .-

• J

'-, ' }", ;- Y;A s.,,. ' 'i' n;.- s;f ..-l, ,)

4.,

i 'eI l. 'u,,,f.... l[.,...' -. -,.*f s'.,: '. j.. :, )..,' ,.J ^ c... :...' - :.'{,,f . s3*.':!.,: .,l,.' , f. )p.., {.; g.i -.,l'*;! Y;' ,'l.,; 4.,.

• . V, ^, -

( j -. ',. .g&,',;. ;, 4?.,., ' s' ,e,/,. j - .p 3e y

O ~"

/ .g 'l ,d.*,,.i,. qg?j

• v.

'e.- b'..I ,'[.,. '.. ,h ? J Y,,% r: y V - f f ....i '.? ., [,' - -,, = - :s 7 W.,:. - 'l -' l-'.h:: L.;',- c f ,f'iv. OI ?,i'

au t,,.

i. 1.,';'c..-' i ,,',.-(-

i u

g,- .p c'f";e.i ..s-w r,.-,s.: Y';,,.'.-).'., '". ' .Q}

• .k.',lc[,...f

_..g$' .l \\ L [.ia 'r m. ',.-lf.. .,.y: ,.v., f } t l.-. e,. f. .;.[ f._g r,, g';*zf )0 l' q ij t. .-.,.,.i-,;.

..y

.,y;

\\.

},.,,,-{ l-. } ( >s, h .t;/'r - i -.'.l'<5;j -., - ^.,; , ~ f' y ,-e;. -E .- 3 't .- l:. ,q. -[, d.. t J,.., ',L,;<. v

• _. p: ('_,.s.'

. C>,,,..{';,: !* <:g. /,,lg ]'.,( __.: j.f. Q,-l <.,5

}'

_l.:g,, .c l .t', t .i:' ;I Y,' ',N'.;!. ll-y.s Ul, .r .I,ez.. ;ll.;5.:;'. -a J; f,s 7 7 9. f.* .3, 'll s,.

. 4'-.,-.,

u,i, ,' L f' t f,:'Y' ? .}.*;,.. l,.i ,r, ',s.. r h yQ,.p):::.l' c,- c.-l-..u /...., : ' ! i 's

• ..I

h'

'h. ' \\ % 1'1 G.- . s (.'s.,.a .( Y ., >,j. G,l.* U.,:,'1..w,' p';'y -,, ',..., i.:,' L,..

v. O:

t,l'. '( b,:',9:g,. -;; u, r;

0 Jl i. :; '

.a <y:/ yq[p z t'

.4

,,t r. / 3,, t l. ef.(

h&.Q l

o.\\; +,,., . e.

(

1se' :.*.;} _.-,' ,.e ,?',, k' Y,:.p:.,.,,, t ?', f< )'..",.,,.c\\ gt .s'-'..'\\, Y

nsn '2-s

'b,.. -,.', -., .l l a.; '-,- '. ."./.4. l. >.n ' f. )). f.a:n

c, l

,g} j.px 'f/r,{ -ll'a. q,._,

y.;

, ;? ' N I^ .i. ;,.. n..-, n -- ,t..

.. 9,.

g f,' ?. ' I', .d ya e f.sf,j.- e* -... - -*r .1...- ..4,. 'p ,a.,,g f,..h,,iy "! .s.,. .,- -;)

• . :',.,.,eg4,-],,-s.-*f,,.-

(., .rg' >g,. $<l-3 :,. ' l ' *..,.x.te:a},.[lf -.' [4. ,w p ;. f_ t ) .-'l:h'?P U e,- f_. ;. 3 -..c.., J-c l:', ..s4.s h .:.'L.

• ,:..<a. X

-ca

:7-

.p'.i, 'l i -_. n. ,'s. t..; g' ; & j m. a M. 4,t Q,[,'_, s W

,,4

' Nj.;c:','A t,, i; .j tc t, . [% a .3; .,-f,,, '. j, f g(.' )

$'l. -

'.m ;f..i

' b '

g N: 3 )'. r n '.s .f,' p ..v 14 p l'f., - - s:&,s'.!M 1p5-,,4.e.-.-;,- '& k. ,..+*lq r. t p* ..f 4.j '. f.

%.,.., ','..c,.jT...j..,.,..

i n .,. ' f.. s' 3.',.'^.,.1 r,. [9 :*. :;. ,1 }J.{ .. ~. 4 g G',J -,s. o.p m.:.

4

. '. '_ts' ,, *s< .f

Q. ;,. M l.

I i

>*l' ."[e.js. y.-,

)..,

s..e ' ' 'ct. 'c.' ) 9 't.t.y ,-(.-. 4'r. 7 Y.s.:F s' ,'[a: u'_.:' po; , _. v'

g{

',, -.i ,s:, ,I' n.\\l' '.s 'p, J.;.. $.'^.l~,:',;:.~('.l: .'). f! yt )1,.-. ' w :,,; [:,.1 :.. {- . i.,.}p..: i,U. .)' ?. l?;.:';

).-',

4' ~ yl';s,- 3, I. 'L',; .r t .t,.. u.,:. ,v/ [f'.'.... yi,,ad,. ?, ; lf l y;I]' ;: y r l 9g ., i l

1,,

., i: - l - g' 4:y,h;,.

g

f L - ;,;

1:>, s.s o.. ' [ u.,,-j,,. -Qg - '.*4 -..3, (.'[,. ^ a ;,,*'.!' ,*,, ( !M,r,:M 5 ,.'. -i_, - A ' ;c,..-,gg,- 1 p a

.'.i

.L ; Q,. : '* rl

.?

., 'm j.,. e, ; \\ (.

9'i 1). ,1-/,- s. la_; j, -l E', t;.'s_. ,s E.;.R,,..:., 1s,;- Q(.y.%','.'[', -c G. ,,v<,

,i.,,,

,.-p. -. .t ; ' g. n' j l ,.,y ,3 ,i c,!;i .:%[..'( ^ i IU. '..l, ,t ,.r y l '.slf -D I., -. - - " { c ]' \\,. ,'.b '.v ) e

L'.['

'y:,'

s,

n. /, g u'.

8lq,.',',. 5<. l a' .;c','_;< .';'l' 7)a L

• '*(

, -.-., i ,..,.--.-s.

1 n',.. h.,

f./,1',..: ,.f .:4 .~ s'

c

, f,, '.' ) ._{:9. 'm3.,.. .'\\:). 1,. ;j A;g:'> ?.. e ]', ) . 4, g 3 l

• ...;1,.'b,p..h:

.!' :.O'. ,2: (.. kfi 9,;; L <' s..' ',. j.'.,. $ t' ll >s h f. r. ' s p

l t.,

i .tff5 s { ._.s. aO; :. a ..,'.i.. ( ',..'((..,.-*:iJ,* f., .e .e' :,..f ,n4 .-p:,

f' [,;

.. )\\ i g; a,.,. l ,,p ~

• s

(,. ,3 4 %l'. -,.' ;i,..,,,.,~.J,i,*,A ;,. '. ;:. y.;: l.,.,,y....,Q<,', ,:yl.' 'R, '; l, .y.. t' [ ,*f.,:'. t,. q ?, /.- v.

l.-.

".. g .. ' p., I' p y 4 . i,: ; 4,,'j.- , 6

8 s

'e o kg ? -'.. ',4n,: .u -,.i '.- _ lJl\\. ,..,..:* g,..,!,e l., [ il.,,2, -lf _ -- 1A. : l.-., 'f ,s >L' i \\' l s'_,f- { a i,. ; t 'lt. ,:,. > ' -:.. ?,.k'e... ,..,.i

• f}

~O,g.. l; 3. <,o,, 1 h' e,

i.. H. y,.,,)f,,

yl%.-.

s. f _ r.

l-.i y, i ll; ,} q -- .,gj ,";b; I .- :,es'o ..h.3' i. y s' ,,,-.g+y;.l'<, * : .'l c: ' 4.. ' '. *..., i _i f [ i M. -; ,*e, u.:- '..l, l. 1;: 4,..l',.! J Tn' \\.L'j). "} .l s -' i.,.-",..,,.,.' o ( l.,. ; ,T.1 ,';1 '.!) , >j', <l-' .;'. i, ',{.ll', .-F,'4 ',N;,.<";.. \\' a

f. "'e'. < r

{'< .i\\ .,}pr,, l > t., - .c,.,- .e.,r c t L2 >'.., t::9'.,; r l ,, + - i' -}}