Text

Requests That Team Members Be Informed to Include ASLB Hearing,Records,Depositions & Ofc of Investigations Repts in Ref Section of Sser Writeups When Info Germane to Issue Being Evaluated.W/Seven Oversize Organization Charts
	ML20210T883
Person / Time
Site:	Comanche Peak
Issue date:	09/10/1984
From:	Ippolito T; NRC - COMANCHE PEAK PROJECT (TECHNICAL REVIEW TEAM)
To:	; NRC - COMANCHE PEAK PROJECT (TECHNICAL REVIEW TEAM)
Shared Package
ML20209B909	List: ML20209B904; ML20210T713; ML20210T743; ML20210T758; ML20210T767; ML20210T782; ML20210T866; ML20210T883; ML20210T896; ML20210T908; ML20210T925; ML20210T940; ML20210T951; ML20210T959; ML20210T973; ML20210T989; ML20210U006; ML20210U027; ML20210U066; ML20210U082;
References
	FOIA-85-59 NUDOCS 8606020114
	Download: ML20210T883 (141)
	v • d • e

{{#Wiki_filter:_ ';)hW ..,b* ..,s%, / UNITED STATES l' NUCLEAR REGULATORY COMMISSION o gg C WASHINGTON, D. C. 20555 %,*****/ September 10, 1984 MEMORANDUM FOR: Technical Review Team Leaders ;FROM: Thomas A. Ippolito, Project Director Comanche Peak Technical Review Team

SUBJECT:

REFERENCES IN SSER WRITEUPS Please infom yo'ur team members to include ASLB Hearing Records, Depositions, and OI reports in the qcference section of their SSER writeups when this infomation is gemane to the issue being evaluated. These references have not been appearing in writeups reviewed so far and they are essential for background inforination. See attachment for reference formats. n t Director Comanche Peak Technical Review Team ec: Walter Oliu Cathy Brown hh &a., s, ' ; - l~ l l 8606020114 860528 PDR FOIA GARDE 85-59 PDR O e

e Attachment i C i Depositions ej 1. Deposition of W. J. Doe (Volume 2) before the ASLB, August 3,1984, pages 77,364 - 77,366. c Hearing Records i / i ,/ 2. TUGCO, et al. Hearing before the ASLB, December 3,1981, page 602. OI Investigetive Reports 3. NRC OI Repor.t A4-83-005, May 23,1983. ~. Op ge i il. e e o e+e O""~

/ }ll [ fj-TUQ.2206 TEXAS UTILITIES GENERATING COMBLNY i 1 OFFICE MEMOR ANDUM To - S ' " E - '- - ~ _ Glen Rose, Texas July 9,1984 IE Bulletin 79-14 subject Confirming a previous conversation. we have apprised the NRC Resident Inspector of our decision to cease direct QA involvement in the IE Bulletin 79-14 Program. The NRC Resident Inspector had no adverse comments on this. If you have any questions, please contact the undersigned. ~ hd /*~' A. Vega TUGCO Site QA Manager AV/bil ,_2 NORM

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COMANCHE PEAK COMPLETION OF OUTSTANDING REGULATORY ACTIONS CHRONOLOGY SPECIAL MANAGEMENT EFFORT DIRECTED BY ED0 MARCH 12, 1984 SPECIAL REVIEW TEAM ONSITE INSPECTION APRIL 3-13, 1984 COMANCHE PEAK PLAN PUBLISHED JUNE 6, 1984 TECHNICAL REVIEW TEM STARTS ONSITE REVIEW JULY 9, 1984 SESstm I JULY 9-20, 1984 SESSICN II JULY 30-AUG. 10,1984 SESSION III AUG. 20-31, 1984 SESSION IV SEPT. 10-21, 1984 SESSION V OCT 1-12, 1984 PUBLISH SSEo, LATE OCTOBER APPLICANT READY FOR FUEL LOAD (EGC0 EST.) LATE OCTOBER l m O +j O*'. k_) ', W l ~ } "*4, s 'J,d 't '.ym

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_CO_MANCHE PEAK - FAJOR OFEN REGULATORY ACTIONS LICENSING ACTIONS o APPLICANTS' PROPOSED DELETICN OF PIPE BREAK JET SHIELDS o TRANSAERICA DELAVAL BERGENCY DIESEL GENERATOR o REVIEW OF CYGNA INDEPENDBU ASSESSENT PROGRAM o ENVIROIMNTAL QUALIFICATION OF ECHANICAL AND ELECTRICAL EQUIPMBfT o CONFCWANCE OF FIRE PROTECTION TO APPENDIX R EARING ACTIONS o DESIGN ADEQUACY AND QUALITY ASSURANE o CONSTRUCTION ADEQUACY o CONSTRUCTION OVALITY ASSURANCE o INTIMIDATION / HARASSMENT OF QC INSPECTORS AND CRAFTSPERSONS INSPECTION ACTIONS o COMPLETION OF ROUTINE INSPECTION PROGRAM o RESOLUTION OF OPEN INSPECTION ITEMS ALLEGATION ACTIONS o RESOLUTION OF APPROXIMATELY 500 ALLEGATIONS o COMPLETION OF 01 INVES.TIGATIONS G L

COMANCHE PEAK PLAN FOR THE COMPLETION OF OUTSTANDING REGULATORY ACTIONS INTRODUCTION NUMBER OF REGULATORY ACTIONS TO COMPLETE: 821 i LICENSING ACTIONS: 37 HEARING ACTIORS: 3 INSPECTION ACTIONS: 377 ALLEGATIONS ACTIONS: 004 PLAN AND SCHEDULES ESTABLISHED TO ASSURE OVERALL C0 ORDINATION OF OUTSTANDING ACTIONS BY NRR, IE, 01, OELD, AND REGION IV ~ ASSUMES: APPLICANT MEETS SCHEDULE LEADING TO OCTOBER 1, 198u, FUEL LOAD NO MAJOR NEW ISSUES STAFF REVIEWS REACH ACCEPTABLE CONCLUSIONS RESOURCES NEEDS ESTIMATED AT APPROXIMATELY 100 PAN-MONTHS ADDITI0flAL EFFORT. l N i

t. TECHNICAL REVIEW TEAM (TRT) Project Director T, A. Ippolito, AE00 Assistant J. Gagliardo, IE Staff A. VTetti R.C. Tang R. Wessman Electrical / Civil / Mechanical QA/QC Leader Coatings Leader Test Programs Instrumentation Leader Leader Leader J. Calvo..NRR L. Shao, RES H. Livermore, R-III P. Matthews, NRR R. Kemig, R-I i I /N I 1 8- _ _L - - - __/ N_ _L_ .__ _J - Technical Issues Technical Issues Programmatic and Technical Issues - Technical' Issues Generic technical - Allegations - Allegations issues and Allegations - Allegations allegations - QA/QC Related - QA/QC Related QA/qC Related - DA/QC Related to E, I&C to Civil /Hech. - Integrate QA/qC to Coatings to Test Programs from other groups - Interface with other groups on QA/QC l

TRT DETAILED GUIDANCE

SUMMARY

METHOD AND APPROACH FOR IDENTIFICATION AND DISPOSITION OF ALLEGATIONS TRACKING SYSTEM PREPARATION OF DOCUMENTATION AND RECORDS PROTECTION OF INDIVIDUALS INITIATION OF SPECIAL NRC ACTIONS, SUCH AS CONFIRMATION OF ACTION LETTERS OR 50.54(F) LETTERS MANPOWER ACCOUNTING e O e 4 e 4 0 J

TRr ICTIhus AT CCtWCE PEAK SESSICN IV - tityMER 9-21, 1984 m APPRCCGRT11Y 50 PEOPLE CN 'IET+1 (SCME PARP THE). APPPOXDE"ELY 500 ALIIGATIONS EEEiG WRKID CN. (!WIY ARE GENERAL AND VAGUE. 'HIIS IS MI DEREASE OF ADCUr 100 ALLH3ATICNS E101 THE "PIAT'). DNESTIGATICN OF ELECTRICAL, CIVIL /.vuu.wt<AL, MECHANICAL / PIPING, COATINGS AND 'IEST PRCGRAM ALIB3ATICNS WILL BE EN CQ@IETE; DRAET SSERS IN F3NAL STN3ES OF REVISICN. QA/QC AREA WILL PROBABLY RDQUIRE A FIITH SESSICN. TURNNER OF CONTRACIOR PERS0 thel HAS D@ACTID JCIIVITY AND CIMPLETICH OF REVIINS. REGICN IV COMPIETED 91 LINE ITDiS AND 30 INSPECTICN PFC.m]RES AS OF AUGUST 31, 1984. ArfmATICNS BREAIGrWN (AS OF AUGUST 31, 1984) NO. OF NO. T AREA LEADER CATIrORIES* ATIFATICNS DPAPI SSERS a EI38ICAL CALVO 9 53 9 CIVItt SIRUCIURAL SHAD /JENG 17 51 17 ^ MEDIANICAL/ SHAO/HCU 47 ~ 147 45 PIPING QA/QC LIVEBE RE 62** 125 5 CDATINGS MATIHEh3 7 62 0 TEST PROGRAMS KEIMIG 7 19' 7 l l MISCELIANECUS BANGARP 22 M J 'IDIALS 171 481 87 I FJOi CATB3ORY WIII BE'A SEPARATE SSLR HIPUr. i

- QA/QC MAY HAVE MORE 'IEAN ONE SSER INPUT FOR SCNE CATBGO.M.

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DEALING WITH FUTURE ALLEGATIONS 1. REPORT TO PROGRAM MANAGEMENT AND TRACK ALLEGATION 2. MAKE BOARD NOTIFICATION, IF APPROPRIATE 3. PROVIDE TECHNICAL ASSISTANCE TO 01 IN TAKING STATEMENTS FROM ALLEGERS 4. REVIEW ALLEGER'S STATEMENTS AND IDENTIFY ADDITIONAL TECHNICAL ISSUES. 5. ENTER TECHNICAL ISSUES INTO TRACKING SYSTEM. ASSIGN ALL TECHNICAL ISSUES TO APPROPRIATE GROUP IN TRT. 6. DESIGNATED TECHNICAL REVIEWER PREPARE QUESTIONS TO APPLICANT, BASED ON TECHNICAL ISSUES. 7. OBTAIN PROGRAM MANAGER APPROVAL OF PROPOSED CORRESPONDEPCE TO APPLICANT 8. DIRECTOR, DL, DISPATCH CUESTIONS TO APPLICANT 9. TRT REVIEW APPLICANT RESPONSE AND INCLUDE IN EVALUATION 10. DOCUMENT RESULTS. e e 9 g... TASK AND DISCIPLINE CODES I. Task Codes A - Allegations B - IE Bulletins, Circulars, Information Notices C - CAT inspection findings 0 - Deficiency Reports 50.55(e), and Part 21 Reports E - Inspection findings, (violations, unresolved items, and open items) G - Generic letter issues H - Hearing open issues I - Investigation Reports issues L - SALP open issues M - Congressional or Connission concerns 0 - Operator Licen;ing open items P - Inspection Program Status 0 - Final OA Reinspection issues R - Room / System Turnover status (those done, those to do) S - SFR open items T - Task force special inspection items U - Unfinished punchlist items V - IDVP open issues Z - Vendor open items II. Discioline Codes 0 - QA/QC (Genersi) B - Bolts C - Concrete /Rebar E - Electrical W - Welding P - Pipe (Construction) H - Hangers I - Intimidation 0 - Coatings T - Test Program O - Design of pipe / pipe supports M - Miscellaneous A - Independant Assessment V - Vendor / Generic S - Instrumentation R - Records d N - NDE K - Cables /Teminations F - Fire Protection ^ U - Procedures / Instructions Review p

e COMANCHE PEAK OPEN ISSUE ACTION PLAN Task: Improper weld examination and testing Ref. No.: AW-48, AQW-20, AQW-21, AQW-8 Characterization: Various allegations covering plug welds, scrap material used in a welded support, deficiencies due to pressure to finish a job, radiographic irregularities. Initial Assessment of Significance: Source: Mechanical & Piping Cat. No. 7 Approach to Resolutions: 1. Locate and receive applicable NDE procedure (s) for liquid penetrant and radiographic. 2. Discuss " improper" use of liquid penetrant with QC welding personnel or velding supervision to determine if records / recall reveal that such a complaint was made in July 1982; determine if any followup action was taken. 3. Check the "T" holes in penetrameters to see if they have been reamed to a large size. 4. Review in detail 79-12 to d ne if the items were closed out i h correctly and any required follo up has been implemented. 5. Attempt to identify and locate QC trainee involved to determine if he recalls concern. 6. Discuss liquid penetrant and radiographic procedures with personnel who should be knowledgeable in this area to determine if they are aware of procedure (s), received adequate procedure training, and were/are adhering to procedures. 7. Refer any examples of wrongdoing or significant deficiencies to TRT l manager. l 8. Evaluate allegations for generic / safety implications. 9. Report on results of review / evaluation of allegations. [0 Ev4( af e generic / safe.f3 i = dibia'en a a nd p o4* ate'a l vela fen s. hbbbO-Od T/le3 L

t Related Open Issue Identification: 1. Using system codes, pull open items, previous inspection findings, etc., from the tracking system open item list. (Region IV identify and add to this work package.) 2. Review activities necessary to close or partially close related items, either based on inspection conducted above or reasonable additional inspection while the inspector is familiar with the areas. 3. While performing physical inspections above, examine surrounding systems, components, and structures for any apparent defect or indicator of faulty workmanship. 4 If workmen are still in the area of a physical inspection, interview them for any related knowledge of other potential deficiencies. 5. Complete portion of IE Module on welding if it relates to effort made on allegations. Status: l Review Lead: Meet..ical : d Piping V. Ferrarini Support: OI Estimated Resources: 4 man days Estimated Completions: 8/27/84 CLOSURE: Reviewed by: TRT Leader i l = G l t l

s., COMANCHE PEAK OPEN ISSUE ACTION PLMI Task: Improper veld preparations Ref. No.: AW-45, AW-46, AW-61, AW-66 Characterization: Various improper veld preparation concerns involving purging, grounding, inerting and concrete dust contamination. Initial Assessment of. Significance: The type of allegations are of a nature and specific enough to warrant further follow up. Source: Mechanical and piping allegations (Cat. No. 6) Aeprorch to Resolutions: 1. Review weld preparation portion of welding procedures related to allegations. 2. Review QC records associated with hardware involved with allegations to determine if weld preparation portion of procedures were followed, appropriate QC inspections were conducted in this area, and corrective action taken in those areas found to be deficient. 3. Discuss adherence to weld preparation procedures with available personnel knowledgeable with welding practices; e.g., welders, weld foreman, QC welding inspectors. 4. Refer any examples of wrongdoing or significant deficiencies to TRT leader. 5. Evaluate allegations for generic / safety implications. 6. Peport on results of review / evaluation of allegations. Related Open Issue Ide/sa4=b ir rb'c,aNons anci pokoh'ai r. E valua4e generic vivia tien s. ntificatigp 1. Using system codes, rM! ois items, previous inspection findings, etc., from the trac!in : >y: ::t open item list. (Region IV identify and add to this work gackas.) 2. Review activities necessary to close or partially close related items, either based on inspection conducted above or reasonable additional inspection while the inspector is familiar with the areas. 3. While performing physical inspections above, examine surrounding systems, components, and structures for any apparent defect or indicator or faulty workmanship. r n.t G t- .w i 1 'w

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- 4 If workmen are still in the area of a physical inspection, interview them for any related knowledge,of other potential deficiencies. 5. Complete portion of IE Module on welding if it relates to effort made on allegations. Status: . 4f Review Lead: %echanical ip n V. Ferrarini Support: 0.T. Estimated Resources: 10-man days Estimated Completions: 8/10/84 CLOSURE: Reviewed by: TRT Leader A -, - + - -

f -f COMANCHE PEAK OPEN ISSUE ACTION FLXI Task: Anchor bolt installation problems - l Ref. No.: AB-4, AB-5, AB-6, AB-8, AQB-1, AQB-2, AB-9, AB-10, AB-ll, AB-7 Characterization: Various concerns regarding anchor bolt installation modification, torquaing and related procedures assignments, and workmanship. Initial Assessment of Significance: AB-9 initially dispositioned in IR-80-16 AB-10 may have been documented, but referenced documentation is not applicable (unauthorized modifications). AB-4 initially dispositioned in IR-79-26. A follow up on all 3 will be undertaken. AB-4 AB-5, AB-6, AB-8, AQB-1, AQB-2, AB-7 need further investigation. The allegations raise questions on potential failure of supports which may have safety implication. /7 Mechanical & Piping allegations (Cat No. p - _= ^^ Source:

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1 Approach to Resolutions: 1) Review all installation, rework and testing /QA verification procedures for Hilti and or anyother types of anchor bolts used on project. 2) Review and research the allegations described in the task's above. 3) Review and inspect were appropriate the actual installation where the allegation occurred. 4) Review samples of similar type installadons and procedures for adequacy. 5) Compare the procedures reviewed with indust'ry accepted methods looking for deviations. 6) Refer any examples of wrongoding or significant deficiencies to TRT t l leader. 7) Evaluate allegations for generic safety implications. 8) Report on results of review and evaluation of allegations. 3) 6.valuale generic./Jafely i*aplicaheeir Aacl polenFal v ioleLo s, [mf) p. ". a g~ Lc w-l j l T ltn g

I' \\ ' Related Open Issue Identification: 1. Using system codes, pull open items, previous inspection findings, etc., from the tracking system open item list. (Region IV identify and add to this work package.) j 2. Review activities necessary to close or partially close related items, either based on inspection conducted above or reasonable additional inspection while the inspector is familiar with the areas. 3. While performing physical inspections above, examine surrounding systems, components, and structures for related apparent defects or l indicators cf faulty workmanship. (u6 4. If workmen are still in the area of,44 physical inspection, interview them for related knowledge of other potential deficiencies. 1 Status: i Review Lead: Nechanica Pipin V.Ferrarini Support: OI Estimated Resources: 15 man days Estimated Completions: 8/10/84 \\ CLOSURE: Reviewed by: TRT Leader i

COMANCHE PEAK OPEN ISSUE ACTION PLAN S/G lafariA( supperf belfs insproperlj Sborfened Task: AB-12 Ref.No.:((({[DF Characterization: Allegation stated that the bolts on the Steam Generator upper lateral supports were out and are too short. Initial Assessment of Significance: If the bolts are too short they may not be adequate for the intended function. Source: Mechanical & Piping Cat. No. 18. Approach to Resolutions: Review 1) G&H Drawings 2323-17, Rev. 1 2) AFC0 Steel Drawing 303 3) P. O. 35-1195-14915 C07 4) Meterial Received Records (MRR) 060860, 61000, 61150 i"' P '**I* " 8 ** *I P***"f / i vi,(. tan s w;t t be evalu aiad. (sneric./safefd li Related Open Issue Identification: 1. Using system codes, pull open items, previous inspection findings, etc., from the tracking system open item list. (Region IV identify and add to this work package.) 2. Review activities necessary to close or partially close relate'd items, either based on inspection conducted above or reasonable additional inspection while the inspector is familiar with the areas. 3. While performing physical inspections above, examine surrounding systems, componente, and structures for any apparent defect or indicator of faulty. workmanship. 4. If workmen are still in the area of a physical inspection, interview them for any related knowledge of other potential deficiencies. 5. Complete portion of IE Module on welding if it relates to effort made on allegations. 7 C, e.; k p?C-m d pidia l s

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Status: Review Lead: ' Mechanical ; rielug, V. Ferrarini Support: OI Estimated Resources: 2 days Estimated Completions: 8/27 CLOSURE: Reviewed by: TRT Leader a

fh 2 COMANCHE PEAK OPEN ISSUE ACTION PLAN Task: W-68 Under$Ued Wolds ^ Raf. No.:(GAP witness C, I# A d- (,8 Characterization: Supprts on tanks to RH heat exchangers has undersized welds (Westinghouse). Initial Assessment of Significance: The undersizing of support welds on this equipment could have safety implications. Source: Mechanical / Civil (Cat. No. 19) Approach to Resolutions: 1. Try to get more detailed information on the support 1.e. (Support #, elevation, etc), from the alleger. 2. Review the RH heat exchanger drawings and do a field inspection of the velds. 3. Discuss this allegation with site QA/QC personnel knowlegeable in the area. Related Open Isgener;c/ safely tbj:licaU+ns and 4. Evaluele p.fenMI viol.Wns ue Identificatio 1. Using system codes, pull open items, previous inspections findings, etc. 2. While performing physical inspections above, examine surrounding systems, components, and structures. Status: Review Lead: Mechanical Civil V. Ferrarini Support: OI Estimated Resources: 4 man days Estimated Completions: 8/31/84 I CLOSURE: l Reviewed by: TRT Leader ~s s. p j T Is7

2b .I COMANCHE PEAK OPEN ISSUE ACTION PLAN l MM3 P pia 3 ' bolb material 4raceaMfy Task: t Q4-84-001 1/9/84[ A d f. 2,, A A G - 3 Ref. No. statement, Characterization: Material traceability - - WeVGd W "'" Avat o A,o - cwfoMwc.4. got:rs 4 crw c t cs7-T'D u pt Initial Assessment of Significance: Use of improper material or inability 4,(- to_ properly document material Meck.a Ia L 4 P'P'aj C4. z o fi Source: -2+ $p, A ', t F' y Approach to Resolutions:

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Review A-3 statement. g 2. Review procedures, nonconformance report and NRC requirements. 3. Discuss adequacy of procedures with field personnel. 4. Examine items in testimony (if possible) to determine effect or traceability problems. 5. Refer any examples of wrongdoing or significant deficiencies to TRT manager. 6. Evaluate allegations for generic / safety implications. 7. Report on results or review / evaluation or allegations. P. Evaluate ene ric / s44*+3 '&pUenW. nr a.I ,4e.f,. ( v,, l. +/, n e Related Open I ue Identification: Status: Not started Review Lead: Pleckoni..l ( f r-keIpert W. H4barJ Support: 8 Estimated Resources: S man days Estimated Completions: 8999994 E's-3/ CLOSURE: Reviewed by: TRT Leader 1 .4 4. ) ' QQi~jN 1 l ^ T 108

2b -k j COMANCHE PEAK OPEN ISSUE ACTION PLAN I reEA A"b, Task: 60P-3. A0%'-23 "#h" O "d* O*" N#" '"3'^#f*' '^' Ref. No. : (84-006, IR-82-1@ Characterization: Pipe whip restraints improperly altered plus faulty receiving inspection of welds. Initial Assessment of Significaace: Unknown. Mechanic. ( ( pip.'a} (Ai. Source: % LL Approach to Resolutions: 1. Review 84-006 and IR-82-10. 2. Review procedures, codes / standards, design requirements for adequacy at time work was performed. 3. Discuss adequacy of procedures with field and engineering personnel. 4. Review sample of similar-type design / installation procedures for adequacy. 5. Refer any examples of wrongdoing on significant deficiencies to TRT manager. 6. Evaluate allegations for generic / safety implications. 7. Report on results of review / evaluation of allegations. Evalu.fe generic./ =4ety i rlieefa' ens #nd pdenfial violab' ens F. Related Open Issue Identification: 5 Status: Not started. Meckaa'e I 4 rir 7, Robeef w. Hubba<ci 'I Review Lead: Support: Estimated Resources: 1 man days Estimated Completions: Fri?4/54 8 31-OY CLOSURE: Reviewed by: TRT Leader p f..," p,+. f.g )[ \\qM O ? 't .h % Y Y ~%l l A l

S 2 COMANCHE PEAK OPEN ISSUE ACTION PLAN f FalsificaiE*n of WtAdi") reco*CIS S M $^ Task: Q' C. ' G' f Ref. No.: (TR-79-12, 84-006 3/7/8g A aW-1, AG ul-D, Ad H - f 6 Characterization: ?_' _ _... '. _#; = nd,;. to&< or svG WS

4% G') t MM Coevetou mvm wtWe Initial Assessment of Significance: htent _ r1 b-: hl; Y. LA/qc g;5 g --

Qo r15w rt M. T % s<.a6es t7tt C. " ^;. ; e f- 00CuvM psw6 t. Source: es, Hec harica( ( P rlag Ca4. 2 3 l Approach to Resolutions: 1. Review IR-79-12, 84-006 3/7/84 2. Review QC and document control practices and procedures, if available, for adequacy at time work was performed. 3. Discuss adequacy of procedures W/QC and document control and field personnel. 4. Refer any examples or wrongdoing or significant deficiencies to TRT manager. 5. Evaluate allegations for generic / safety implications. 6. Report on results of review / evaluation of allegations. 3ener/c./ sam belicafi nr aaJ p. fen &( violaH.ns. 7 Evalua(* 3 Related Open Issue Identification: Status: Not started Review Lead: ecka n'ca l 4 g. 7lng, Ro bert W. NuNatcI Support: to Estimated Resources: 3 man days Estimated Completions: Spia$iCET 8 - LY ' CLOSURE: Reviewed by: TRT Leader <- ~ ,',y ij i s,J..i' V U ~ TlIlo

1 COMANCHE PEAK OPEN ISSUE ACTION PLAN 1mproper Weldin3 pracket Task: -1,AQW-12,AQW-14,AQW-2D Ref. No.: t04-82-000s. HA nn d Characterization: Improper weld practices. Initial Assessment of Significance: Allegations are non-specific but weld QA program may have broken down. Source: 20, 25, 23 MechanicM f pipin3 Co f.14 Approach to Resolutions: 1. Review Q4-82-0005 8/2/82, 84-006 3/7/84. Tv6 CD 2. Review Brown & Root and M welding QA document control practices and procedures. 3. Discuss adequacy of proced'ures with QC and document control and field personnel. 4. Refer any examples of wrongdoing or significant deficiencies to TRT manager. 5. Evaluate allegations for generic / safety implications. 6. Report on results or review / evaluation of allegations. Related Open Issue Iden{tification:salaig impuestion s and polenital v;otah' ens 7. Esalusta gener <- Status: Not started Mecbanical bf r "3,h koherf W. Nulabard I Review Lead: ikkhk Support: lo Estimated Resources: $ man days Estimated Completions: ftidikst24 6-10 -dk CLOSURE: Reviewed by: TRT Leader ,,+ VM W. j'1A J 4 J 4J L VV,, U 'w ' TlHI

S k$ 8 + COMANCHE PEAK OPEN ISSUE ACTION PLAN Incorreek wetd:n3 docu~entat.'on Task: (AQW-23, AQW-25, AQW-29, A d V-6h Ref. No.: 15-83-0 T, 84-UU6, Jie/os, a-J stateme g LOST kIMLo M Characterization: In n_:- 5 weld documentation. 4 Initial Assessment of Significance: Difficulty in tracing weld data. Fywa4 1 t;;;'&m in Q A /qc ;y;;;;. Source: E -.2.3 Mechanical ( f elaj Ca+. 2r i Approach to Resolutions: 1. Review IR-83-07, 84-006, A-3 statement. 2. Review weld documentation control practices and procedures. 3. Review drawings and specifications of welding involved in allegation. 4. Discuss adequacy of documentation control with field personnel. 5. Refer any examples of wrongdoing or significant deficiencies to TRT manager. 6. Evaluate allegations for generic / safety implications. 7. Report on results of review / evaluation of allegations. F. Evaluaf e 9eners'<- Related Open Inue Ide[ntification:saf ef.impl catJens and p.+e 4,.i violab'.n 2 3 Status: Not started Mechan' cal 4 gl h, Reber+ w. Hunarcl Review Lead: Support: 3 Estimated Resources: I man days Estimated Completions: 8185E4 8-M-d CLOSURE: Reviewed by: TRT Leader e.umme g Il2. 4 a A s

COMANCHE PEAK OPEN ISSUE ACTION PLAN unqualifiacl welders Task: (AQW-Z, AQW-3, AQW-D Raf. No. : (54 -006 3/7/84, IR-79-11, IR-79-20, IR-79 ~ Characterization: Use of unqualified welders. Initial Assessment of Significance: Allegations are non-specific but the welder qualification program should be reviewed. Source: 467-pdechanical ( Pipia3 Cat. 26 Approach to Resolutions: 1. Review IR 79-15 IR 79-22, IR 79-20 to determine if documentation adequacy supports findings. 2. Review Brown & Root & TUEC procedures and training programs. 3. Review welder qualification program. 4. Refer any examples of wrongdoing or significant deficiencies to TRT manager. 5. Evaluate allegations for generic / safety implications. 6. Report on results of review / evaluation of allegations. 7. F_ valuate generic. Saleh implicab'en s and p 4enfial vi.(afi.ns. Related Open Inue Ide tificKeion: Status: Not started M/ -RWTP/ h4ecban* cal ( Fl[ j, Rol,er f W. Hu%aret Review Lead: Support: 4 j Estimated Resources: & man days Estimated Completions: 6 7-U-CLOSURE: Reviewed by: TRT Leader . ~ .h h.h AS 5 F ll3

l4 k 66 COMANCHE PEAK OPEN ISSUE ACTION PLAN LloquaGf t'ed We.lcl inspecfer Task: (f0W-5, AQW-ll, AQW-16, AQW-17, AQW-18, AQW-19, AQPD Fef. No.: dI4-006 J///84, IR-79-15, A-18 3/8 7 Characterization: Weld inspectors not properly qualifiedsel W " M *

eS3 t d L6-Initial Assessment of Significance: Ph breakdown of weld inspector training and qualification program.

Mechanie.[f piping Cay, 2 7 Source: Approach to Resolutions: i 1. Review 84-006, IR-79-15, A-18 104C* 2. Review Brown & Root TUSC weld inspector training and certification program. TOGC0 3. Discuss adequacy of program requirements with Brown & Root & *,JEC' QA/QC departments, and other personnel involved. 4. Refer any examples of wrong doing or significant deficiencies to TRT manager. 5. Evaluate allegations for generic / safety implications. 6. Report ou results of review / evaluation of allegations. 1. Evaluate generiefsafet impliceUens and potential vietations. Related Open Inue Identifica ion: Status: In progress W/ 4H#r-/ Mechanicq( (@g, Rolperf w/. Nubbard Review Lead: Support: 7 Estimated Resources: 9 man days Estimated Completions: 8Hagl82 7-2.7 OY CLOSURE: Reviewed by: TRT Leader F01A-85-59 }\\\\4

COMANCHE PEAK OPEN ISSUE ACTION PLAN lingualified G44[dlc personne.( for welcling clecueni s Task: QQW-6, kW-7, Act P - z d Ref. No.: 64-006 3/7/84 IR-79-l_[ hsO Cp0+LA 9 PrLO fn W sM W. T' (2 0 Characterization: _ _ ; _ Q-h;'Q-:,, _...;l fr: Endling - ciding l MS T~rbt LA Yt 0W O r' C( ***-f f T I W GCHfh ^~* Initial Assessment of Significance: P" d i ',-ok - .4;.iQ" p :;rr-- m (t art $~ I hsMe enti woM f36- / US T4U l'C o rVCeW 46 Mechan;e I /r pip n3 Ca4. L 2 Source: Approach to Resolutions: 1. Review IR-79-12, 84-006, 3/7/84. 2. Review training program and methods used to determine qualification of key QA/QC personnel. 3. Review results of involved personnel work and evaluate concerns against requirements. 4. Refer examples of wrongdoing or significant deficiencies to TRT manager. 5. Evaluate allegations for generic / safety implications. 6. Report on results of review / evaluation of allegations. Related Open Issue Iden[tification:plicab'en.s wl pofen%( vielaf/en h-7. Evaluata 4eneric. sa f e4S Status: Not stated M/ 'b) -RWA / Mechane* cal I ;-yEmg, Ra 'w 'l - l Review Lead: '1 Support: M m d-I /Y1 A 4 * */ #*

5
l Estimated Resources: $ man days Estimated Completions: M4 & -l b8Y l

CLOSURE: Reviewed by: TRT Leader i j .,.r-l )II6 ~ ,-..-._-s e

8 COMANCHE PEAK OPEN ISSUE ACTION PLAN Iwfrof erly Cef UNE bhE0 (An*Yf**5 "S*"i Task: AYM Ref. No.:(Case letter 3/18/83 to Characterization: Use of improperly certified liquid penetrant material. Initial Assessment of Significance: Liquid penetrant examination with f aulty material could be rejected resulting in questionable welds. No.cb4 decal 4'f'h[n]bi. Source: F Approach to Resolutions: 1. Review case lecter 3/18/83 to ASLB. 2. Review purchase specifications,' receiving inspection procedures and related documents. 3. Review related nonconformance reports and evaluate closure data. 4. Discuss adequacy of NCR closures with field personnel. 5. Refer any examples of wrongdoing or significant deficiencies to TRT manager. 6. Evaluate allegations for generic / safety implications. 7. Report on results of review / evaluation of allegations, li RelatedOpennsueI/antifIcation:i-P c4 Ue n 2 Ancl fd* anal velat '.n s t. EvaNafe 9eae*!* Jaftf3 Status: Not started Review Lead: Mechani eal }, Eb->I W. Ndbard Support: y p e-s en A :. o u foM 7 Estimated Resources: $ man days Estimated Completions: e/M O -8 f'-8 Y CLOSURE: Reviewed by: TRT Leader ( L t. ,4 >.e : -m 'Q $ J N, Y 4 e l iib 1

s I h0 COMANCHE PEAK OPEN ISSUE ACTION PLAN Improper receivlog nspec.f.'en of We(d3 i Task: / Ref. No.: 64-006, 3/7/84) A d w - 'I' Characteri::ation: Irpreper re:21cin luncuivu m: ~_2 m bndenser. "60 o $ M TJ (b.cc.sv#L4 rZ,45W # 4:564 TT 48 M

Initial Assessment of Significance: Pe;.al.1 L t..ii m in ^A/^' ; ;;c.

j Ca v00u7Sc W., m W GG MW e Source: et Mechanical f' pipeh Ced. 3 o i Approach to Resolutions: 1. Review 84-006 for completion of allegation. 2. Review Brown & Root and TUEC receiving inspection procedures and practices. 3. Discuss adequacy of procedures with QC and field personnel. 4. Review drawings and specifications of material received. 5. Refer any examples of wrongdoing or significant deficiencies to TRT manager. 6. Evaluate allegations for generic / safety implications. 7. Report on results of review / evaluation of allegations. RelatedOpenIssueIde{ntification: 9. Gvalua4e generis ta-(el3 i v' p Usatio n s a nd pole., b'.\\ vio laWen s Status: Not stated Mechanical h[q R.ber+ w. uha<cl Review Lead: Support: 3 Estimated Resources: $ man days Estimated Completions: M 8 -I O " N CLOSURE: Reviewed cy: TRT Leader 0~?h ,y -h

7r r J" LK v-a-i g i \\ ca r c$ q L ~ -M = W~d H R , ~Ra F1, v' t- - h f. R J Pt

34-4 33' dow d2 4 / (MW-4[)

- M. RWH e ~ N /rG W ' 7 / RWH A

A m.-. LO ~~ t4 u a ( S W 1 ~1 ' r 1 s - N .t n. m

9 4

9' b k 'T d>o ~ ee;" es ] d i e I i .e Y 2% .~. _.a. k ~ N V k h e 4 n l . r __ 4 k M,

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I 4 J T' c +ar i Q i d w ~Z-UE ) c. 8 %T-Y N r+5 mm D ' r> L, -m 1 3 -1a s g 17 d 4 5 .v a O l X d h ' 1. El LL-I o e s cw 3, 1 Ot-3 10 M U 9 8 9

.os .., c a==... =L....:.; x: i e QA/QC Allegation Review Categories i 4 i Category Est. Mandays Allegation Package Assigned S:hedule No. Subject to Complete Nos. Prepared to Open Close Remarks 1 NCR Activities 5 AQ-30. AQ-31, AQ-32, AQ-34, AQ-36 AQ-37 1 2 QC Inspection AQ-33, AQ-35 l Reports { r. 3 Adherence to AQ-52, AQ-61 AQ-78, l Procedures AQ-79 l. j 4 Incorrect QA/QC l (Items trans-Procedures

ferred to 1

other cate-l f gories) I 5 Hanagement AQ-1 AQ-2, AQ-6, AQ-25, AQ-62, AQ-66, AQ-67, AQ-69, AQ-81 AQ-80 AQ-56 i 6 General Document AQ-3, AQ-21, AQ-45 Control AQ-49, AQ-57, AQ-71, AQ-72 AQ-74, AQ-84, AQ-8 AQll-1

l.,

{ h, ,{'5 yr "+ ? { .b i- '+ lIto i ~

c. *,..,,

~, ~ .i ~ j QA/QC Allegations Review Categories j ~ ; '; Category Est. Mandays Allegation ~ Package Assigned Schedule No. Subject to Complete Nos. Prepared to Open Close Reserks l 's

lt 7

Drawing Control AQ-17 AQ-18, AQ-22, AQ-42, AQ-58 .i 9 ') 8 Document Review AQ-75 AQ-76 l { l Program 1 = 1 9 Design Change AQ-16. AQ-59, l I. 8 Control AQ-60, AQ-70 h i j 10 False / Wrong AQ-9, AQ-7, AQ-10, e I Documents AQ-11 AQ-55, j l AQ-83. AQ-15, AQ-44 i j 11 Document, Clerk AQ-73 Training \\ 12 QC Inspector AQ-23, AQ-24, AQ-26, (or Supervisor) AQ-27, AQ-28 AQ-29 j Qualification / AQ-63 Training i [ 13 Craft Qualification / (Items transferred 'i Training to other categories) 1 (Items transferred 14 Improper Testing l tp other categories) 15 Inadequate Inspection AQ-38 AQ-39, AQ-50 and Certificat;1on, ~ 5 c,..,', i-

~ 2 ..c,.. v3. a ua. s : 'l J DA/QC Allegation Review Categories 8 I i ~ l Category Est. Mandays Allegation Package Assigned Schedule No. Subject to Cnselete Nos. Prepared to Open Close Remarks 16 Traceability of Haterial AQ-5. AQ-40. AQ-41 AQ-53. AQ-77 t 17 Improper Upgrade / Downgrade AQ-12 AQ-13, AQ-14 l of Material e I 18 Undocumented Activity / AQ-43. AQ-47, AQ-51 i fI Rework f. 19 Cleanliness Control AQ-65 AQ-54 k 20 Inadequate Staff AQ-4 f f.7 Resources ? 21 Defective Material AQ-20 l 22 Control Over Stamps AQ-19. AQ-48 j and Devices i I i 23 Disruptive Activities AQ-46 24 Evaluation AQ-64 25 Maintenance AQ-68, AQ-82 . I. } [ i k. 4 5 t t 1 ) ~ i i 1: ,s /s '.

.1......__..... ...._4. .Le u!.e Ng 7/ tole + ( e Mechanic'al si Piping Allegation Review Categories f ) / t I \\ p ~ j Cctegory Est. Handays Allegation Pacirage Assigned Sched'ule ? [ ~ No. Subject to Complete Nos. Prepared to Open Close Remarks j .. sl ? i I Uelding - incorrect or AW-31, AW-34, AW-35, g,47 no proccedure AW-37, AW-38 s i y 2 Welding - procedure AW-32, AW-36, AW-54, 5 I.IT c-

j adherence AQW-28, AQW-30, AQW-10, AQW 33 j

. f g) + I 4 3 Improper or defective AW-39, AW-40, AW-41, g welds AW-42, AW-43, AW-47, AW-49, AW-50, AW-52, t ,i ' AW-33, AW-57, AW-58,

p AW-59, AW-60, AW-62,
)

? AW-64, AW-65 .[ .t 4 Plug welds AW-51, AW-55 ET 5 Weld designs AW-44 j-i fit k 6 Improper. weld AW-45, AW-46, AW-61, g. }. e-preparations AW-66 i 7 Weld examination and AW-48, AQW-8, AQW-20.- b I testing AQW-21 l' ~ spT l 8 Weld repairs AW-63 l, 6 g.. _ e 9 Weld rod control AW-56f AdN-2-$ f [ l f i 10 Dar. aged pipe AP-5, AP-8, AP-10 C 11 Pipe installation

AP-4. AP-9, AP-13 W P:C j.

l 12 Reactor Vessel installation AP-ll [ 13 Repairs & modifications AP-7, AP-12, AP-16 c, t to pipe ? .j l; f I y c .- :e '.

t-... __....m,..__. 1 i g< g Mechanical & Piping Allegation Review Categories l' l g l-I i I-i 3 ] Category Est. Mandays Allegation Package Assigned Schedule No. Subject to Complete Nos. Prepared to Open Close Remarks !t i 14 Hiscel. piping AP-6. AP-14, I AP-15, AP-17 W 'O t-S 15 Defective hangers AH-3 AH-4, 4}i-6, );y 4.- g AH-13 y ,1 16 Ilanger design problems AH-5, AH-8, AH-l4 8d (d

1 f {

17 Ilanger damage AH-9, AH-10 All-!! R

f. ? --

AH-7, AH-12, A H -(J~ g,1)1 - ~ ..{ 18 Itani,erconstruction

t 19 Illit i b olc a AB-4 AB-5, AB-6, V mfp p

AB-4, AQB-1, AQB-2 k i 20 Concrete anchor bolt AB-9 AB-10, AB-11 'I 21 Nuts-torquing AB-7 y{q {.

22. g NCR Activities AQP-1, AQH-2 y&+tt I

23,g g Welding - Undocumented AQW-14. AQW-25, AQW-26, ggg g i I activity rework AQW-29, AduJ-s./ j l i 24 R2 d Design change control AQW-22, AQP-3 % W- [ t i ~ 25 q False / Wrong documents AQW-9, AQW-12,, AQ H-IS IlY b l j l AQW-13. AQW-23 I l i k 26 AtAld QC Inspector training AQW-1. AQW-2, AQW-3, 'g . AQW-4. AQW-5, AQW-6,'. ( AQW-7, AQW-11, Aflf-23 i 4 j , I:. 1 + { f !I 'h f. a.

e ' e.,,

+

... - a u. 7... .s. L- ~u.-. ._s.-+. ...a l

2.... 2 -'

kr s . L L. -. a ~:..h 5 ' ' ' ' * '* - L' *~19-E -q s s, r ~, Hechanical & Piping Allegation Review Categories { 3 i.- i i .i Category Est. Mandays Allegation Package Assigned Schedule d No. Subject to Cosiplete 14 0 s. Prepared to Open Close Remarks [ r i 27 Q Inadequate inspection & AQW-15. AQW-16, AQW-17. guj H certification AQW-18,AQW-19,AQW-27 j 28.Qgj Piping & Bolts:.i-ITrage-AQP-2 k ability.of' material .t .[

3.,

.j .i. s' w llab i M k .'I tj. 1] i '/.,17 4 59A ta k, t. kn IAd-'t,f j hut ueL~.y 3 n-g. ,) 3o d efl~ - L jr-It, k[it, AP-u jg b ,. H g p p w rr d ' i g.q,<9_w y y g (,, j 3\\ sf4 lA.A ut,A L& J as a2 v& W r$.Ja%4. i i i i i t e 5 r ? E { t ' j Ed L;. l .rt IE 4 1 b L s. ..,o s.1 ~.

8 e

e.. y e w w-e e -t v +- e

...._.u........ .L ._..-.f... 1 3 f s y Civil / Structural Allegation Review Categories i i I r, 9 } ? Category Est. Mandays Allegation Package Assigned Schedule -) i No. Subject to Complete Nos. Prepared to Open Close Remarks ' k i-I 1 ; t Inadequate materials AC-16, AC-19, AC-20,

i used in concrete AC-21. AC-27' i

t 2 Concrete placements AC-22 AC-23 I ? ~ 3 Poor weather conditions AC-24, AC-35 4 ,y 4 placement of concrete 5 k. w .),; .] Concreteyoids/ cracked / 4 AC-25, AC-32,.AC-34, -l crumbled - AC-41, AC-28, AC-33 g ~.. kl i 5 Miscel. concrete AC-17, AC-18, AC-26,

g AC-29, AC-31, AC-36, L..

AC-42, Ace 43 ' ). [ i 6 Rebar improperly installed / AC-30, AC-37, AC-38, 1 ]. drilled or omitted AC-39, AC-40 ,i i L .. j 7 Concret/Rebar - Undocu-AQC-16, AQC-12, AQC-13, mented activity / rework AQC-14. AQC-15 l, lf l. 8 False / Wrong dac n ats AQC-1, AQC-2, AQC-3, u.' l AQC-7 j t ,) i 9 QC Inspector training AQC-9 if l and qualifications ll i e ( A 10 Improper testing AQC-4, AQC-5, AQC-6, j; AQC-8 AQC-11 ? f n. a e i. i } i

)

l. _

t I. E L e

i I Coatings Allegation Review Categories l 4

j j'

l l c, i-P Category Est. Mandays Allegation Package Assigned Schedule No. Subject to Complete Nos. Prepared to Open Close Remarks q 5 l 8 f I Coatings applied to AQ-32, AQ-33, l improperly prepared A0-34, A0-38, th a surfaces AD-39 [ j i j ,I. 1 ] 2 Incorrect coating AO-37 ] , 'M ' ~ li J :

j

.J used p. ,' + >:'

j p

a 3 Coating must be reworked AD-4Q l

f

+ ,o

5 h.

t 4 Coatings in pipe hangers $0-39 ,f { could reach sump lg .s-1 u 5 Miscel. coatings AD-36, AD-45 AG-42 e 5 t i. 9 l (sensitive) {

f.

p. 6 Coatings - NCR activities AQO-8, AQO-9 AQO-10 ( ,i 1 i AQO-11, A40-12e AQO'-13 .4 7 Coatings - Procedure AQO-2. AQO-3, AQO-16, !i i adherence AQO-17, AQO-18, AQO-19,

D j

I AQO-20, AQO-23, Ago 29c }* c 8 Incorrect QA/QC procedures AQO-15 .f i g 9 Hanagement AQO-1, AQO-28, AQO-294 i 446 30, A90 51 t l l l 10 Coatings - Undocumented AQO-5, AQO-26 e activity / rework [ t t. i False / Wrong documents AQO-21 ii. { t y -{ 12 QC Inspector qualifications AQO-27 t I l s ,e t: ? I i t i <L

h

. $.~ _. ..._ _...._. m _ __.. u .._.._ ; w..e

_a..... _....;. e., m aft _. m i. l L

.L Coatings Allegation Review Categories f [ .\\ e-e s' Ii 1

Category Est. Mandays Allegation Package Assigned Schefule

] No. Subject to Complete Nos. Prepared to Open Close llenar'ks e h i 1 13 Craft qualifications AQO-29b ,c and trainin8 i l r t . s. .I I { 14 Improper testing AQO-4, Ago-22 Ago-24,Aq0-25 '{ 15 Inadequate inspection / Ago-7,Aqo-b { s.l. certification 16 Coatir.gs - Traceability " AQO-29e, Ago-6 ,f ' p ~ { .1 of material + I N ,1. +-

d' g.

e t , ' ~ l ' j' ~ .l. i 1, i s 't .h ( g 9 L 4 .i 1 t I l 4, l ~=. l j i. 3 ~ a I-i l l 1 e t a 4 s '.. 0, es. e e 't e

a ......u. m 3 1 Test Program Allegation Review Categories 1 i I ,I } L: i t l Cat: gory Est, Mandays Allegation ' Package Assigned Schedulo No. Subject to Complete Nos. Prepared to Open Close: Remarks i [ -) i 1 Ilot functional testing 4T-1, AT-2, AT-3, i1 [ p ,f AT-4, AT-5, AT-6, ' [' l l' 4T-7. AT-8 'd. /.f."- . ' i. n. M 'A.,. .u t i 2 NRC inspection of test $: AT-9, AT-10, AT-11, , R. S it. h% f ,.r, ; 7,, program / reporting ..r AT-13 g' p:: i. 'q;H:i..

g!

3 Unit 2 testing r AT-12 ,3 lt \\ Y-c - l.e - ( V- 't r e ': 'O }. t. 5- ,{ I t t t. i

d 1

'5 t o f.i l l ?. l '. / t 1 h f. <1 i 't i e i q 1, ~ 1 'f li i .t i i e 2 5 t 4

t.. _

)

Lj 1 1. 7 i

s =.....

2.._. .......a._ ...._.2.. i. I. N l . t. >l Electrical Allegation Review Categories I P n , l. .i 1

t.. " :

i-Category Est. Handays Alleg4 tion Package Assig'ned Schedule 3...~:e I. ~ l No. Subject n to Complete Nos. Prepared to Open Close.0.! Remarks 2 ..n 4, 3 1 Improperly installed AE-13 1 ,'Q[ , ! Jg ' t 'J. 2I electrical compor.ents . yg - ......;}

s '.

3 's tray hangera (elev. 610'). AE-14 M. ' I N. [' $ ' ! dk C"' Ij) 2 Safaguards building csble. e ;. !>( 'c.:[ '.I,T Y. ; e.. - l [ ' ' d, )..' i. j.4 - * { not constructed properly. ' r, ', )' 'S i r . i. - [.* 5, '.' "

y

.t. '.,(,~.. AE-15.f 5 '.' /. ? 3 Separation crir.sta ,,M '., t \\', ..{ ,c, e, '. i; ',;.'..... '- ( 1, . violated 1 1 eL .c'- J C T[ s ..e j..f d: q 4 Safeguards 1 panel on 79Q8 ... ' AE-16 - .,j..; h ' ' ',E ;. j f .}4:'

h. >,.

level hac loose bus bars P and ground wire connectior.s- ^ '.. ?. f r. ' J ' F-d i f T- ' p,... [l.,. * 'j 5 Control room area deficiencies AE-17.'[.l i] . w /,. 1 (field run conduit, drywall, 0 and lights) [ ,./;, Ji,.' ..r. n 7 .y 6 NCR Activities AQE-1,'4QE-4,A9E-5 'f t.{.

.t-7 Electrical Inspectors directed AQE-6 j

not to follow procedures ..I '/, ' O f .I - 8 Hanagement AQE-{1 { t 1 -[ 9 Electrical - Undocumented AqE-2,'AQE-3* [ i [ ' d, activity / rework ti i 4, g J.j : 10 QC Inspector training / AQE-8 1 q

y. :C A

qualifications l }

.g f: \\

[ 11 Craft qualification & AQE-10 .{ .; I( training .j ,s \\ I 12 Inadequate inspection / AQE-7, A'}E-12 is 4 certification s \\.' i, e ( i

a t

3

1..... g.

... N. 2,.. A u

w....

u. l

u........a .... h ..._._..o .u__. ,,_.,..r... J [ I. {*- I 'd MiscellaneousAllegationReviewCategor1I I 1 (tag.w.sr) 8 3 . i' ~ j Catigory Est. Mandays Allegation Package Assigned Schedule i t No. Subject to Cosolete Nos. Prepared

to Open Close Remarks f,

I Hearing - 19 issues AM-1 I t 1 1

.i 6

2 Reactor fuel AM-2 ,1 e e '5 {e Reactorpressurevessel) AM-3, AM-23 l. I j 3 ,j s 1:' ] 4 FSAR error (10,2-11 4 1$) AM-4 Ig i

. si g

I. 5 Radioactive material release AM-5 n 6 IIP turbine AM-6 I d-:I [ 4l- .,-i P +

I, 7

Pressurizer AM-7. j; j l, ' '

o IN 8

Condenser AM-8 AM-9, AM-10 'I I 1 p j ) .s i. 'h,d .m te j - f 9 Reactor / fuel bldg. liners AM-Il j f, y 10 CCW system AM-12 [ l (.;

b s

11 Pumps - Ilayward Tyler AM-13 i s. n 12 Diesel generator AM-14 .I [ j 13 Polar crane - shimming AM-15 AM-16 i I d' C 3 l- '-'j 14 Containment doors AM-17 .) [ v.~ ! 3 l .i 15 Safeguards tunnel AM-18 r ;l g. .,1 16 WRC form posting AM-19 { i l iy 17 Material false statement (plant AM-20 l' management to ASLB) }' -:j s . b 18 Drug Abuse AH-21 ij tl 8 .?l (' ] 19 IIVAC system AM-22 l1 1 1

i

.-. e ' s....

7.1.1 t*f o "'. c.., :. r .2 QA/QC A1)cgation Revi'cw Categories

a tegory Est. Mandays Allegation Package Assigned Schedule No.

Subject to Complete Nos. Prepared to Open Close Remarks 1 HCR Activities 5 AQ-30, AQ-31, AQ-32, 54 434 -AQ-34,-AQ-36,4Q-37 / y v A00-8, A00-9, AQ0-10, ' AQ0-11,~AQ0-12, 'AQ0-13 / ./ / AQE-1, AQE-4, AQE-5 /AQP-1 l / AQH-2 /n . t i,,. l s v 2 QC Inspection AQ-33,AQ-35 l Reports s s 3 Adherence to AQW-28, AQW-30 Procedures / 'AQ0-2 00-3 AQ0-16 -17," -18,100-19,

3. -..,

'AQ0-20,4Q0-23,1Q0-29c .,j - 1 l /AQE-6 /AQB-1 m E. 4 5.tc. ( w a G A /t W ) 5 -AQp--le / / )f.Q-79 Q-52, AQ-61, AQ-78, j i .\\ )

l QA/QC A11cgatien, Review Catigories Category Est. Handays Allegation Package Assigned Schedule flo. Subject to Complete Nos. Prepared to Open Close Remarks 1 4 Incorrect QA/QC

/AQW-10 Procedures

'AQ0-15 +AQk/-33 a v 5 Management AQ-1,AQ-2,AQ-6, 'AQ-25,4Q-62, - AQ-66. -AQ-67, -AQ-69,'AQ-81,. < AQ-80,'AQ l go- 'AQ001 AQO-28,

~ 3 0

/ AQ0-3

AQE-11

/ s s 6 General Document AQ 1, AQ-21, AQ-45, f'AQ-71,vAQ-72, Q-49,4Q-57, Control VAQ-74,/AQ-64, / AQ-8 AQH-1 l I

m ,g, ..,7.., 1 { I QA/QC Allegation Review Cattgsries Category Est. Handays Allega tion Package Assigned Schedule fio. Subject to Complete Nos. Prepared to' Open Close Remarks / 7 Drawing Control 'AQ-17, AQ-18, ' AQ-22,#AQ-42, /AQ-58 /AQE-9 ,/ 8 Document Review 'AQ-75,AQ-76 Program J 9 Design Change AQ-16,AQ-59, Control

AQ-60,#AQ-70

/ AQW-22, a. I AQP-3 t 4 / / 10 False / Wrong AQ-9, AQ-7, AQ-10, Documents /AQ-11,4Q-55, /AQ-83,4Q-15, AQ-44 / s AQW-9, AQW-12, /AQW-13,4QW-23 AQC-1,1(QC-2, /AQC-3,'iQC-7

AQ0-21 i

3

v ,e 5 I QA/QC A11egaticn Review Categ: ries Ca tegory Est. Mandays Allega tion Package Assigned Schedule flo. Subject to Complete Nos. Prepared to Open Close Remarks / 11 Document Clerk AQ-73 Training i / 12 QC inspector AQW-1, AQW-2, (orSupervisor) ' AQW-3,40W-4, Qualification / /AQW-5,'AQW-6, Training "AQW-7,4QW-11

AQC-9 4Q0-27 AQE-8
AQ-23,AQ-24,

'AQ-26."AQ-27,#AQ-28, /AQ-29,'AQ-63 S 13 Cra f t Qualification /

AQ0-29 h Training 14 Improper Testing 100-4,AQCI,AQC-6
AQC-8,"AQC-?1 s

s 0-Q0-22, AQ0-24 AQ -25 / AQW ./ AQW-8, AW-20, AQW-21 i I s } m--

~.s ,m 4 .**/-, \\ QA/QC Allegation Review Categ ries l 3 Category Es t. Handays Allegation . Package Assigned ' Schedul e tio. Subjec t to Complete Nos. Prepared to Open Close Remarks s s 15 Inadequate inspection AQW-15',3QW-16, 'AQW-17, 'AQW-18, and Certification /AQW-19,-AQW-27

AQE-7, AQE-12

/ AQB-2 s J ,AQ-38,AQ-39,AQ-50, J AQO-7, AQ0-14 v / v 16 Traceability of AQ-5,AQ-40,AQ-41, Materia 1 'AQ-53. AQ-77 i L lhM 'AQO-QQO-p AQP-2

AQB-3 17 Improper Upgrade /

AQ-12,AQ-13,AQ-14 Downgrade of Haterial h MC } I l

^ QA/QC Allegatien Review Catcgsries Ca tegory Es t. Handays Allegation Package Assigned Schedule flo. Subject to Complete Nos. Prepared to Open Close Remarks / / 18 Undocumented AQW-14,AQW-25, Activi ty/ Rework /AQW-26,4QW-29 /AQC-10,IQC-12, VAQC-13,4 QC-14 VAQC-15 AQ 26 A QE-2, AQE-3 ( AQ-43, AQ-47, /AQ-51 ~ ~ / 19 Cleanliness Control AQ-65,JAQ-54 20 Inadequate Staff /AQ-4 S Resources Wy#>J - 21 Defective Material AQ-20 22 Contril Over Stamps, (p -f. / / AQ-19, AQ-48 Weld Rod, and Devices j 1 i }

.e,,- a m f--, I e .j, QA/QC A11egaticn Review Categ: ries Ca tegory Est. Handays Allegation Package Assigned Schedule No. Subject to Complete Nos. Prepared to Open Close Remarks 23 Disruptive Activities AQ-46 24 Evaluation JQ-64 J ./ 25 Maintenance AQ-68,AQ-82 l O 4 9 i

p HECHANICAL & PIPING ALLtunIION REVIEW CATEGORIES , Category Est. Handays Allegation Package Assigned Schedule No. Subject & Complete Nos. Prepared to Open Close* Remarks ..,1. g Welding - incorrect AW-31, AW-34, AW-35, or no procedure AW-37, AW-38 ,. ~ 2. Welding procedure AW-32, AW-36, AW-54 adherence A \\ i 3. Improper or defective AW-39, AW-40, AW-41, welds AW-42, AW-43, AW-47, AW-49, AW-50, AW-52 AW-53, AW-57, AW-58, AW-59, AW-60, AW-62, AW-64, AW-65 t b 4. Plug welds AW-51, AW-55 Id5. Wel,d designs AW-44 6 Improper weld AW-45, AW-46, AW-61 N. preparations AW-66 ' 7. p Weld examination and AW-48 testing g 8. Weld repairs AW-63

9. g g Weld rod control AW-56 wpc 10.

Damaged pipe AP-5, AP-8, AP-10 g f C, 11. Pipe installation AP-4, AP-9, AP-13 wp c, 12. Reactor vessel AP-11 installation 13. Repairs & modifica-AP-7, AP-12, AP-16 q,c tions to pipe i ~ y p 14. Miscel. piping AP-6, AP-14, AP-15, j I AP-17 \\ f

~' 's 'HECilANICAL & PIPING ALLEGATION REVIEW CATEGORIES ( Category Est. Handays Allegation Package Assigned Schedule No. Subject to Complete Nos. Prepared to Open Close Remarks sf ph g

15. y y Defective hangers AH-3, AH-4, AH-6, AH-13 16.

Ifanger design AH-5, AH-8 d UN problems n 17 74 Hanger damage AH-9, AH-10, AH-11, g yf f 18. Ilanger constructiote AH-7, AH-12 f pff 19. Illiti bolts AB-4, AB-5, AB-6, AB-8 pff 20. Concrete anchor bolt AB-9, AB-10, AB-11 ~ 3 p; 21.' Huts-torquing AB-7 t t t l

,p COATINGS ALLEGATIb.,HEVIEW CATEGORIES Category Est. Handays Allegation Package Assigned Schedule No. Subject to Complete Nos. Prepared to Open Close Remarks 1. Coatings applied to A0-32, A0-33, A0-34, improperly prepared A0-38, A0-39 surfaces 2. Incorrect coating A0-37 used 3. Coating must be A0-40 reworked 4. Coatings in pipe A0-35 hangers could reach sump 5. Miscel. coatings A0-36, A0-41, A0-42 (sensitive) O$ I l l i

_.n ?, a < CIVIL / STRUCTURAL ALLEs.. ION REVIEW CATEGORIES Category Est. Handays Allegation Package Assigned Schedule No. Subject to Complete Nos. Prepared to Open Close Remarks 1. Inadequate materials AC-16, AC-19, AC-20, used in concrete AC-21, AC-27 2. Concrete placements AC-22, AC-23 3. Poor weather conditions AC-24, AC-35 placement of concrete 4. Concrete voids / cracked / AC-25, AC-32, AC-34, crumbled AC-41, AC-28, AC-33 5. Miscel. concrete AC-17, AC-18, AC-26, AC-29, AC-31, AC-36, 1 AC-42, AC-43 6. Rebar improperly AC-30, AC-37, AC-38, installed / drilled or AC-39, AC-40 ,omitted f, k e f e ss I I e e \\ f I

TEST PROGRAM ALLEGA,.sN REVIEW CATEGORIES Category Est. Mandays Allegation Package Assigned Schedule No. Subject to Complete Nos. Prepared to Open C1'ose Remarks 1. Ilot functional AT-1, AT-2, AT-3, testing AT-4, AT-5, AT-6, AT-7, AT-8 2. NRC inspection of AT-9, AT-10, AT-11 test program / AT-13 reporting. 3. Unit 2 testing AT-12 t O 9 l o e, a e

s,r. f., MISCELLANE0US ALLEGA1 2a REVIEW CATEGORIES Category Est. Mandays Allegation Package Assigned Schedule flo. Subject to Complete Nos. Prepared to Open Close Remarks 1. Hearing - 19 issues AM-1 2. Reactor fuel AM-2 3. Reactor pressure AM-3, AM-23 vessel 4. FSAR error AM-4 (10.2-11 & 12) l 5. Radioactive materia'l AM-5 release 6. HP turbine AM-6 7. Pressurizer AM-7 l 8. Condenser AM-8, AM-9, AM-10 ? 9. Reactor / fuel bldg. AM-11 liners 10. CCW system AM-12 11. Pumps - Hayward Tyler AM-13 12. Diesel generator AM-14 13. Polar crane - AM-15, AM-16 shiming 14. Containment doors AM-17 15. Safeguards tunnel AM-18 16. NRC form posting AM-19 )

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5 p MISCELLANE0US ALLEGATION REVIEW CATEGORIES Category Est. Mandays Allegation Package Assigned Schedule No. Subject to Complete Nos. Prepared to Open Close Remarks 17. Material false AM-20 statement (plant management to A5LB) 18. Drug Abuse AM-21 19. HVAC system AM-22 i 4 9 4 4 0 9 I I e i n 4 i ,C s p7, r-i..: ' )

, :~. e .m,. ( ELECTRICAL ALLEGATIb., REVIEW CATEGORIES ~- Category Est. Mandays Allegation Package Assigned i Schedule flo. Subject to Complete Nos. Prepared to Open Close Remarks 1. Improperly installed AE-13 electrical components 2. Safeguards building AE-14 cable tray hangers (elev. 810') not i constructed properly 3. Separation criteria AE-15 violated 4. Safrguards I panel on AE-16 790' level has loose bus bars and ground wire connections 5. Control room area AE-17 l deficiencies (field ) run conduit, drywall, i and lights) f~% 1 ,Q ('h

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I MRC T6CHNICAL REVIEW TEAM INTRODUCTION 1.. F. Fikar/B. R. Clements COMANCHE PEAK OVERVIEW J. T. Herritt, Jr. A. Organizational Chart e B. General Facility Locations ENGINEERING AND PROCEDURES 'k. R. McBay DOCUMENT CONTRCL j..Richman/H.Hutchinson QA/QC PROGRAM 'T Vega D h We* 5 6 Eg T/l?A F- ,y-, g

NRC TECHNICAL kEVIEW TEAM COMANCHE PEAK TEAM Project Manager John Merritt (101) CONTACTS 1. QA/QC Tony Vega (321) Bob Scott (859) 2. Electrical and I&C Larry Popplewell (721) Charlie Britt (171) R. B. Kelley (276) 3. Preep, Starting & Testing Dick Camp (711) Tom Miller (255) 4. Protective Coating Ron Tolson (705) 5. Civil-Mechanical Claude Moehlman (271) Billy Ward (726) George Tanley (262)

6. DCC ' p,%T cM Q }

Heyward Hutchinson (257) 7. Engineeri.ng Procedures Richard Baker (573) 8. Operations Dick Jones (897-4856, #3202) FACILITIES, ARRANGEMENTS, MISC. Bob Deatherage (695) John Dittmar (816) Jack Rcdding (451) James Ca11 cutt (430 P. M. Milam (713) Don Anderson (103) Bill Nelson (852) David Wade (252) Joe Johnson (518) 5 ,u

3 z 3 -; a z x a m m s., M*:hanicc1 & Piping Alleg% tion R: view C2tegorisa { si i e i Cat; gory Est. Mandass Allegation Package . Assigned Schedule Ho. Subject to Completa Nos. Prepared to Open Close Remarks 1 Welding - incorrect or AW-31, AW-34, AW-35, no proecedure AW-37, AW-38 l t 2 Welding - procedure AW-32, AW-36, AW-54, I. adherence AQW-23, AQW-30, t. AQW-10, AQW-33 i 3 Improper or defective AW-39, AW-40, AW-41 I [* welds AW-42, AW-43 AW i-I. AW-49, AW-50, AW-52, l. AW-53, AW-57, AW-58, f 1-AW-59, AW-60. AW-62, j AW-64, AW-65 i i 4 Plug welds AW-51, AW-55 l; ~ i 5' Weld designs AW-44 i' 1 68 Improper weld AW-45, AW-46, AW-61, preparations AW-66 j 7 Weld examination and AW-48, AQW-8', AQW-20, i -l l testing AQW-21 f' [b.k jy/Q" p 8 Weld repairs AW-63 - d- 'F 9 Weld rod control AW-56 e i 10 Damaged pipe AP-5, AP-8, AP-10 11 Pipe installation AP-4 AP-9, AP-13 i 12 Reactor Vessel installation AP-11 t ,~ 13 Repairs & modifications AP-7,'AP-12, AP-16 to pipe b,., I r i

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3 r

Hechanical & Piping Allagation Review Categories 'i ,s i i ~ i. - Cat: gary Est. Mandays Allegation Package ^ Assigned , Schedule No. Subject to Complete Nos. - Prepared to J Open Close Remarks j. 14 Miscel, piping AP-6, AP-14 .i AP-15, AP-17 i, 1 .? l 15 Defective hangers AH-3. AH-4 AH-6, 's I AH-13 f;t j i s r 16 Hanger design problems AH-5. AH-8 'I { I' 17 llanger da:nage AH-9, AH-10, AH-11 i-b l .i. j 19 ililti bolts ~ f:' j 18 Ilanger construction AH-7, AH-12 s ? AB-4, AB-5, AB-6, l; l-j AB-8, AQB-1. AQB-2 ,y 3 j 20 Concrete anchor bolt AB-9, AB-10, AB-11 1 ~ 5 )i i i 21 -Nuts-torquing AE-7 >} l i ](;I i 2'/ NCR Activitien AQP-1, AQH-2 j j i 23 Welding - Undocumented AQW-14. AQW-25,'AQW-26, ll l } activity rework AQW-29: {; t 24 Design change control, AQW-22, AQP-3 25 False / Wrong documents AQW-9, AQW-12,, AQW-13. AQW-23 I 26 QC inspector training AQW-1 AQW-2 AQW-3, [ i j . AQW-4 AQW-5 AQW-6. ', I 8' i, AQW-7, AQW-11 l { I <s l*. i 1 I. .r. e t 2 p.,_ ,s LL r....

m_ mm. _.e.__. ea ...._.~ s 'i Mechanical & Piping Allegation Review Categories 4 i I f' s: Cat; gory Est. Mandays Allegation Package Assigned Schedule No. Subject to Complete-- Nos. Prepared to-Open Close Remarks t i 27 Inadequate inspection & AQW-15. AQW-16, AQW-17, certification AQW-18. AQW-19, AQW-27 I: i lI l I h I i 28 Piping & Bolts: -.;Trage-AQP-2 ability.of'_ material j{ t' 3' I. 1 f f l-t O i ,T t-0 l t* t s' s h' 3 I' i l~ i. i I I 4 i i ei f-l i j t. l i' {!.. li .I, t i'

sl: l t i ,i SH Shou-nien flou, N u.c. 3 RW14 Robe r 4-W. Hubbard, sartpsrorte f, 1 v1P C W. Paul Chen, ETEC E rnie C1. Th oe,,yson, ETec EGT Psoberf J. Macierson, EAS,Inc. RJ M i Vic F. Ferrarini, E AS, Inc. VPF Charles P. Richards, AI I cp8 i l F0lA-85-SS /lz7

s I SH Shou-nien Hou, NRC RW H Rober 4-w. Hubbard, sartprrorlE WPC W. Paul Chen, ETec E rnie G. Th o,,pson, e rec EGT Roberf J. Moderson, E A S,.In c. RJ M VPF .Vic F. Fe r ra rini, EAS,Inc. Charles P. Richards, AI c p Ps James H. Malonson, TetaPYNE JHM I 4 i e am

L i SH Shou-nien Hou, N ac... RW H Rober 4-W. Hubbard, sar19:Torla l l vlP C W. Paul Chen, ETec Il E rnie G. Th o e,, yso n, eTec EGT RJ M Roberf J. Nacierso n, E A S,.In c. j .Vic F. Fe r ra r in i, EAS,Inc. { i l VPF Charles P. Richards ' AI ( l CP8 f James H. Malonson, TELEDYNE e ~J H M 1 i j 1 I I c 1 i i l 1, 1

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-~. .A Y,'?h$$,'***f L}. - , -, ~, o .] 2 L m ;..:= :: ~. - 0 IUJ'?.ia?f'.C ] Pl.'") - I C1 loading pits to ascertain if the commitments stated in the PSAR [ M rd 2323-55-18, Rev. 2 were N.k and Gibbs & Hill (G&H) specificationThe inspector reviewed Grewn & Root (81 t a yU being implemented. construction procedure 35-il95-CCP-38, " Stainless Steel Liner W~ .5 hf Erections," and B&R QA procedures CP-QCP-2.ll, "Inspc.ction of j4,- Stainless Steel Posi Liner Systems," and CP-QCI-2.ll-1, " Weld W g-Inspection and Fit-Up of Stainless Steel Liners," to ascertain Addi tional P{ 2 if the above stated requirements had been implemented. b QA and work procedures in the areas of weld expend . Ni surveillance were reviewed to assess control of these activities. No items of noncompliance or deviations were identified. [fp j 3 Observation of Work Activities j I5 b. w.s (1) Stainless Steel Liners Is F-The welding of fillet joints for the attachment of leak chase

L channels and of tacks for the attachment of backing bars for the butt weld seams for stainless steel liners was inspected.

[ Weld procedures and welders were found qualified in accordance The-with the requirements of the ASME B&PV Code, Section IX. welding was performed in accordance with WPSs 99020 and 86023 Work and and placed as specified by B&R drawing WRB-10559. i-inspection activities were performed as prescribed by the procedures discussed in the pr'tvious section. i 9f No items of noncompliance or deviations were identified. g p (2) Reactor Coolant System Component Succorts i. A limited inspection of the Vertical Columns - Cl as shown r, f and described on Westinghouse drawings 1457F29 and 1457F27 r C The inspector l was perfonred in the site storage yard. reviewed the PSAR a E; Rev.1, " Fabrication Requirements For the Reactor Coolant System Component Supports," and determined the vertical fabrication requirements were ASME B&PV Code, column The Section III, Div.1 NF,1974 edition as a minimum. inspector was unable to find any documentation in the l preliminary data package and certificates of conforman i accordance with ASME III, NF and that volumetric inspection l of the full penetration welds had been perforced as prescribed The licensee is obtaining by ASf'E III, MF, paragraph NF-5212.the complete data packa l items were fabricated and inspected as prescribed. This item is considered unresolved. T** 6 ; r. a -e l ^ l,,. [.y g.-am e v U

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s i'
.= FIRST ORAFT: SEVERE ACCIDENT PROGRAM DESCRIPTON
Contact:
Daniel R. Muller x28026 t I l a. .N 't 6 l - -... - ~...
a, SEVERE ACCIDENT PROGRNi DESCRIPTION Introduction This description of planned NRR activities relating to severe accidents has been prepared in response to the following paragraph from the NRR FY 84 Operating Plan: "0SI, in consultation with other NRR divisions, will develop a program description of planned NRR activities relative to Severe Accident efforts. The description is to include generic methodol-ogies and phenomenological studies and the application of these severe accident techniques in NRR licensing reviews. This program description also will reflect activities associated witn develop-men' of the Commission's Final Policy Statement on Severe t Accidents. The draft program description will be ccepleted by Aaril 1984 (for planning nd budgeting purposes). The final program description should be completed by June 1984." The overall NRC effort on severe accicents began because recent and interrelated events suggest that the regulatory focus relative to severe accident analysis that had been acceptable up to a few years ago need to be modified. Specifically, the following events have resulted in changing insight on the subject: Preparation of the Reactor Safety Study and subsecuent wide use of
PRA, The TMI Accident, Action Plans and Rulemaking responding to THI,
. Industry response to TMI (IOCOR, EPRI, INPO), Better (more extensive) analytic techniques, Accident phenomenology research, Source term Research, and Proposed Safety Goals. The response to these has been a number of changes mandated by the Commission in the design and operation of nuclear power plants. Most of these changes were instituted as a consequence of action items identified in NUREG-0660, The TMI Action Plan, which provided an I integrated plan to improve the safety of power reactors and NUREG-0737, Clarification of TMI Action Plan Requirements, which identifies TMI-related requirements approved by the Commission for implementation. The actions that have been taken to date, collectively, have resulted in imorovement in the safety of nuclear power plants by croviding additional assurance that accidents involving core camage have been to:n reduced in likelihood and 'would be mitigated should they occur. As more information anc insight beccmes available, further changes in the Regulatory requirements may eccur. The pnilosophy behind such cnanges is to be provided in a policy statement on severe accidents. l l One of the items in the TMI Action Plan, (Task II.3.8, "Rulemaking Preceeding on Degraded Core Accidents") specified a generic rulemaking to reach final decisions on severe accidents. This rulemaking would establish policy, goals, and requirements related to accidents involving core damage greater than that of present design basis accidents. The i I I m.m..._ w
a staff proceeded to the point of publ'shing an Advanced Notice of Rulemaking on October 2, 1980 (45FR65474). Events transpiring subsequent to that date have dictated a modification of this planned activity. It now appears that the objectives could be better met if the scope of the rulemaking would be more specifically stated and narrowed. Accordingly, the staff now is directing its efforts toward a policy statement on Commission treatment of severe accidents. The policy issues that will be addressed are: 1. Policy on new custom plants 2. Policy on new standard plants 3. Policy on current OL reviews and operating plants, and 4. Policy on delayed CP's. The Policy Statement is scheduled for review by the Commission in June 1984 The program description called for in the NRR Operating Plan is needec for two purposes. 1. To provide a succinct cescription of NRR related activities in FY 84 for NRR management perspective, and 2. To provide a pr,ojection of NRR related activities for FY 35, 86, and 87 for budget development. .._...--,m._.,.. t
u.. a. n e E e y 4-For the first purpose, this program description wiil identify all related NRR activities relative to severe accidents that have occurred or are to occur in FY 84 The principal source of this information is NRR staff members that have identified activities related to severe accidents in FY 84. For the second purpose, to the degree that specific FY 84 programs continue into FY 85 and beyond, we identify those with some certainty..However, since a policy statement on Commission treatment of severe accidents will not be approved prior to late spring 1984, and since the overall technical and related issues are evolving rapidly at this time (response to IDCOR, source term, safety goal, and hearing board decisions), the projection of efforts beyond FY 84 is less definite. The first point at which there might be a change in direction or emphasis in NRR on severe accident work will coincide with Commission approval of the Policy Statement. At that juncture, this description should ce reevaluated with the cbjective of assuring that it coincices with the dictates of tne Policy Statement. Subsequently, it can be anticipatec that the pregram will be modified based on the outcome of the IDCOR work, the source term work, anc possibly the evaluation of the safety goals. Periodically, as new technical insights evolve, the level and direction of Severe Accident related work will need to be reevaluated based on the relevant facts at that time. The June 1984 date for the final version of this report shoulc coincide with Commission approval of the Policy Statement and thus the program description should be consistent with the cictates of the Policy Statement. en FY 1984 Activities __ N p W M@4- ,y ye
I j ' This program description will draw together the elements of NRC ictivi-ties that relate to severe accidents. The description includes the following program elements in which NRR staff is involved: Review of IDCOR results Coordination with RES Severe Accident Policy Paper Accident Source Term Evaluation of Safety Goals Safety and Environmental Licensing The NRR effort on severe accidents may also be viewed from a technical issue vantage, with the following technical elements: Accident Prevention Source Term Accident Management Core Melt Phenomenology Containment Resoonse Consecuence Evaluation Gas Generation Emergency Preparecness PRA Methodology External. Events Ecuipment Survivability. This descripton will summarize all ongoing and contemplated ac:ivities within.NRR relative to Severe Accidents from both the program issue and =
-6 technical issue vantage. The paper will not be a detailed program plan. Rather, it will be a listing of those activities that the NRR staff is involved in relatisa to severe accidents that are not included in the NRR Operating Plan in FY 84. In some instances some of the work associated with the program or technical element; is identified with programs other than severe accidents. Such work has been identified; and resource estimates have been made; so that a corolete snapshot of all severe accident related work is available. t This program description will thus provide information such that the reader can: focus all related NRR activities in one overall program description, identify major milestones for timely and infor=ea management cecisions, identify areas in which additional severe accident related activi-ties are needed, or areas in which the program shculd be realignec. We will take two cross cuts of the subject catter; by program elements, e.g., IDCOR related work, Safety Goal, etc., and by technical elements, e.g., core melt cheronena, atmosoher-ic dispersion, etc. These cross cuts will provide an aggr'egate overview of the severe acci-
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7 dent issue reccanizing that it is broad in scope, encompasses the tech-nical expertise of many of the staff, and does not necessarily follow traditional organizational responsibilities. The interfaces and interaction between the disciplines represented by the subject matter will provide an opportunity for intensive interdisciplinary communication and cooperation between diverse Branches and Divisions. This diversity is illustrated througnout the sections of this paper in which the staff units that are assigned responsibility for various technical or administrative areas are identified. Finally, we include the major decision milestones such that future closure of the progran elements is demonstrated, provide a list of the various technical assistance FIN plan inputs that will be neeced keyed to the decision milestones, and summarize the NPR manpower dedic4ted to the issue. FY 1985, 1986, 1987 Activities In view of the current status of the !DCOR work, the Folicy Facer, the l l Accident Source Term Program Office effort, and otner relatec work., the situation with regard to FY SS, 86, and 87 is fluid. Never*heless, the final version of this description due in June 1984 will identify programs for FY 85, 86, and 87 that can be identified at that tire. This will be done with full knowledge that the rapidly evolving technology as well as policy will probably result in many changes in the program in late'r fiscal years. l 1 .t...... a _n_
j . 4 II. CURRENT AND PLANNED NRR ACTIVITIES ^ A. Program Elements ) In the following sections we provide a description of the various programs within NRR that relate to severe accident activities, the { principal activity milestones, the total manpower usage anticipated en the program element for FY 84 by NRR Branch, and the Technica! [ Assistance resources used and planned for the same time period. Vnen available, we have also provided a summary of anticipated activities post FY 84 including planned activities, and expected manpower and Technical Assistance usage. O I e l l
9 !!.A.1 REVIEW OF INDUSTRY DEGRADED CORE RULEMAKIt!G (ICCCR) PROGRAM RESULTS OSI/RSB/S. Sheron Introduction The NRC Severe Accident Research Program (SARP) will provide a foundation of experimental data and analytical models to support rulemaking and licensing reviews related to severe accidents. The Program covers a range of topics, including accident progression, core melt phenomena, hydrogen behavior, containment loads and response, and fission product release. The 10COR program is an industry supoorted parallel effort to develcp technically defensible models of the same technical issues considered by SARP. Frecram Description The ICCOR program has been established by industry with the objective o'f developing a ecmprehensive, integrated, and technically sound position to hela support the rulemaking process proposed by the Cenmission en severe accidents. The IDCOR program is supported solely by carticipating industry. The program is managed by Technology for Energy Corporation (TEC). The overall apprcach to meet ne cbjectives involves a series of technical tasks that, collectively, will cover all technical aspects of the severe accident issue. Included willebe an assessment of the capability of existing plants to withstand degraded core accident h j +
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o . phenceena, and evaluations of the operational aspects of accident prevention control, and mitigation will be made. The IDCOR strategy includes development of technical and legal options for the rulemaking, and working with NRC to reach convergence on technical issues. The NRC and IDCOR have committed to participate in a series of technical exchange meetings to define the areas of agreement and disagreement between the two organizations. Focus will be on those areas of disagreement which most significantly affect risk estimates. Both will seek a consensus, where possible. The outcome of this exchange will be embodied in a set of technical issue papers which are to describe the status of knowledge on over fifty specific technical topics. These technical issue papers are being prepared by RES and NRR staff, as assigned. Planned Activities Schedule Technical Meetings 1. (Subject) Date Lead Person 2. 3. 4, Specify Technical Differences Cate Lead Persen a Prioritize Obtain Technical Assistance Close-out Issues
II.A.2 RESEARCH CdORDINATION AND APPLICATION TO ISSUE RESOLUTION OSI/AES/J. Hulman Intreduction The Severe Accident Researen Program (SARP) directed by the Office of Nuclear Regulatory Research consists of aporoximately 60 separate research programs. The 60 programs are subdivided into 13 program elements so that progress of integral parts of the overall research program can be more evectively evaluated. The overall intent of the SARP is to provide a sound technical basis on which evaluations can be made of existing plant design and op,eration of changes in nuclear power plant design and operation. NRR staff participates with Research in direction of these Programs and maintaining general technical cognizance of the results of the research. More recently, and with the initiation of the IOCOR effort, research has been focused on developing better understanding of more than 50 technical issues (currently 55) that encccoass all severe accident related work. The results of the severe accidert research are being or will be usea to aid in the resolution of related il1I issues (hydrogen control, con-tainment venting, and control rocm habitability shielding, environeental cualifications, post accident sacoling, risk reduction at high population censity sites, primary ccolant source control, licuid pathway analysis, and decay heat removal guidance), generic issues, and unresolved safety issues. Program Description NRR organizational elements track the results of the research work in their respective areas of expertise. In some instances a formal g
q 12 - arrangerent has been rade for Ni A staff to participate in Rssearch Review Groups. Yhis activity c:ccliments the activities of the Research and Standards Coordination Branch (RSC3), DST, and the oversight of the Senior Review Group For SARP (chaired by the Deputy Director, RES and - has as members the Directors of DE, OSI, and DST). On request, usually via the RSCB, NRR organization alements tnat have specific needed technical expertise will be requested to conduct peer reviews of anticipated research or of research results. By this and other means potential changes in the direction of research are made. Finally, the NRR staff may identify the na.ed for a new research initiative via a ijser Nged menorandum from the Director, NRR to the Director, RES. The participation of NRP. Civfsion Directors in the SARP Senior Review Group has resulted 19 NRR staff accepting new assignments related to severe accidents. $pecihcally: OST - Preparation of the Savere Accident Policy Statement CE Assestcent of Containment Loading and Performance under Severe Accident Conditiers, and CE - Cevelopment of Containment Performance Guicelines CST - Safety Goal Implementation Plan. L The tJE containment work is supportec by OSI containment Systems Branch.
or examcle, conscnsus opinicns regarding such severe accident chenomena ds Stean spikCs and heating neecs to be factored into PRA reviews anc contairment analysis cr.lculations. Furthermore, findings regarding the expected behavior of the containment under severe accident loading conditions will provide insight necessary to fonnulate Centainment
-a m .. + .44
Performance Guidelines. Planned Activities Schedule ? ..m Pesearch Ccoldination is an ongoing pf% rain with specific goals ano milestenes associated with reyiew, cement, and critique of specific research prograr.s. The initiatives listed above that dre assigfied to DE, CS:, and DST via the $ACIP Senice Review Group fre dit-l cyssed in detail elsewhere in this re; ort (f.ee Skctico 1.A.3 II.B.6, !!.C.5, and II.A.5) in this report. 4 f 9 l I l l l l l I e - ~ a -e e + ** *e. = +*
14 - II.A.3 SEVERE ACCIDENT POLICY PAPER (NUREG-1070) 051/" Spangler Intrcduction The Comission's ongoing Severe Accident Program, cut of which the effort leading to a revised Policy Statement arose, began to take shape almost immediately after the Three Mile Island ac:ident in March of 1979. A number of changes in design and operating procedures of nuclear power plants were mandated by the Commission. These changes were the outcome of numerous investigations of the causes of the Till accident as well as consideration of the vulnerability of all plants to severe acci-dent risk. Following the initial implementation of changes to operating plants and plants under construction, a separate set of requirements was developed for applicants whose construction permit (CP) review had been interrupted. This last set of requtrements, embodied in the Construc-tien Permit / Manufacturing License Rule was published on January 15, 1982 (47 FR 2286)'and oecame effective en February 16, 1982 as 10 CFR Part 50.34(f). As part of the Commission's response to the TMI accicent, an Action Plan (NUREG-0660, May 1980) was issued. Section II.S of that plan deals with the :iting of plants and the requirements for coping with severe acciGents. Ccnsistent with that plan, the Ccmission has imolemented a rule co'ncerning hydrogen control in degraded core cooling accidents [10CFR50.44(c)]. The concept of a generic rulemaking to reach final decisions on severe accidents also took form in the TMI Action Plan, Task 11.B.8, "Rulemaking Proceeding on Degraded Core Accidents". This plan envisioned a long-term rulemaking extended beyond O g n eu. e =ar e- -- ~ ~- ~. y-, ...,,,,,,a,,,,,,e e e ;*=mumme e.,e 5..
l p ~ 1982 to establish policy, goals, an. requirements related to accidents involving core damage greater than the present, design basis for all classes of reactors (those operating, under construction, proposed for construction, or proposed as new standard plant designs). The task also included the interim step of an Advance Notice of Preposed Rulemaking, issued on October 2, 1980 (45 FR 65474). The revised Policy Statement on Severe Accidents (revised Policy Statement) withdraws this Advance Notice of Proposed Rulemaking, and is designed to succinctly state current Commission policy on severe accidents. Program Descriotion On April 13, 1983 The Commission published for public comment a " Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation" (48 FR 16014). Since then events and new technical insights suggest changes in the Proposed Ccemission Policy Statement. The revised Policy Statement takes into ace:unt tne comments received from th? public on the earlier propcsed policy statement and other experience and information developed since that time. The revised Policy Statement deals with three classes of plants: those now existing (operating or under construction), those whose construction has been mothballed, and those in the future. The main purpose of the revised Policy Statement are: To clarify the procedures and recuirements for licensing new nuclear plants; To avoid unnecessary delays of plants now under*construc-tion with potentially serious penalties _to ratepayers; To close out severe accident issues for existing plants r y.,4 ,.;/4 7 J{.;
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u.... 15 - (those in operation and under constructics.) withcut imposing further cackfit recuirements unless these c.an be justified thrcu~ h cur backfit regulations and ;olicies; and, g To acnieve improved stability and predictacility of reactor regulation in a manner that merits pt.blic trust and ccnfi-dence in regulatory decision making. An adjunct to the revised Policy Statement will be a NUREG report that will include a description of other activities related ;c HRC's Severe Accident Program including: tre Severe Accident Research Pregiam, implementation of safety measures resultirg fron lessons learned in the accident at Three Mile Island, safety goal development. resolutien of Unresolvec Safety Issues and otner generic tafety I
issues, rulemaking for source tem revision, deferral of siting policy.
features of a generic escision Strate 4y for resclving Acgulatcry Questions and technical Issues relating to sevefe acciderts, devalopment and regulatory.us2 of new safety ir. formation, treat. Tent of uncertiir.ty in severe 4ccicent 12cisiccmaHng, r.no enccuragement cf Systems Reliability Oregrams by incastry 'or botn existing and future plants to assure Oat th.e realized level of safety is comensurate with the
accident inalys:s used in regulatory decisions.
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. u.. m. Preparation of a revised Policy Statement was reccmmended by the Office e of Researen Senior Research Review Group. Dr. R. Mattson, Director, OSI, NRR, who is a member of the Group, volunteered to prepare the revised Policy Statement. To develop the revised Policy Statement with as broad a constituency as possible, and recognizing that the eventual publication for public comment would be the responsibility of the Office of Research, Mattson assembled a " Committee" of staff cemcers from both NRR and RES that encompass the technical expertise and organizational accumen needed for the endeavor. The following NRC staff members have been actively working on the Policy Statement: R. J. Mattsen, Director 051, NRR R. Bernero, Director, RES ti, Spangler, Special Asst, for Policy Development, 051, NRR J. Henry, "enior Health Physicist, RAB, ORA, RES '. Malero, Chief, Regulatory Analysis Branch, ORA, RES J. Rosenthal, Section Leadar, Reactor Systems Branch, CS!,i:RR
0. f4u11er, Assistant Director for Ractatten Protection, OSI, NRR
!. Rosr*oc:y, Chief, Res, & Stds. Ccordination Branch 057, tiRR Speic, Director, OST NRR In additien, contractor personnel have been iavulved througn Tecrnical Assistacce Centracts-R. ' Denning, Battbile's Coluntus Laboratories - Supported by RES L. Ybarronds, Sciertech, Inc. - Sunported by NRR 4 e e see de ,. esp ...m-g e e a.
.O j 18 - This group of NRC staff members and contractor personnel have been actively engaged in preparing the revised Policy Statement at various - levels of effort since late FY83. Planned Activities Schedule Finalize Policy Statement. April 1984 ACRS Review - May 1984 CRGR Review - June 1984 Comission Review - June 1984 Publish for Comment - Receive Coments - Publish Final Policy Statement - P e 4 h 9 A e 4 .~~my 4.- . g ...g.... t
7 J - 19 II.A.4 ACCIDENT SOURCE TERMS 05 /AES/J. Hulman Introduction During the TMI-2 accident the observed fission product release mix was not as would be predicted based on subsequent knowledge of core damage. Subsequently, there have been reccmmendations from a number of sources to rethink the characteristics of fission product releases from both severe and design basis accidents. This is timely since the previously used source term is based on 10 CFR Part 100 and TID 14844; both about 20 years old. Program Oescription In early 1983, the EDO established an Accident Source Term Program' Office (ASTP0) in RES with a charter aimed at focusing NRC and industry sponsored, research on severe accident radioactive source terms, and sutdecting the conclusions therefrom to a blue ribbon peer review by the Acerican Physical Society. The ccepletion of ASTP0 effort-s are ex:ec,ted in FY85 (presently December 1934). The ASTP0 is staffed by " volunteers" detailed from parts of NRR and RES. At this time the following NRR staff members remain on detail: L. Soffer, SAB*; W. Pasedag, AE3; R. Meyer, RSB. Currently the ASTP0 anticipates that the people on detail will return to normal work assigneents full time late in this calendar year (1984).
R.ecently transferred to RES.
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.e :.mos. e ,G 20 - The work of f1R. centractors on source term (primarily succorted by RES and directed by ASTP0) is being paralleled by similar work undertaken by industry under IDCOR. (See Section II.A.l., Review of IDCOR Results.) Although ASTP0/RES has the lead in this area, NRR is involved through the staff members detailed to AST90, througn its research coordination functicn, and through casework and OR licensing amendment reviews. fWith regard to casework, there is continued pressure primarily during. hearings and in the context of standard plant reviews (G25SAR).to base decisions on the expected outcome of the ASTP0 work. The NRR staff is handling this orfa 'etse-by-case basis with finesse and consistent with the ECO admonition to not base decisions on enticipated outcome. The Severe Accident Research Plan, which includes ASTP0 activitie5. employs c a Senior Research Review Group, for guidance and take', ar.'. active role in review of,, severe accident research plans and results. The applic: tion of ASTP0 sponsored research to regulatory criteria and requirem.ents for flTCL's, CP's, Standard Plants, and OR's is ar. iroortant impending problem confronting the staff in FYSS, 86, and 87, that may have imoact en staff resources and regulatory decisions. When a revised source tem becomes available, affected Standard Review Plans ard Regulatory Guides that are the responsibility of NRR,will recuire evaluation and potent,ial revision. In addition depending on the magnituce of source term, enanges, there will likely be other related work such as changing emergency plans and numerous Tech Spec changes. Planned Activities Schedule Research ---~.-~; n
. Activity Date Lead Receipt of Research Products and Coordination With ICCOR CY84 Review by APS CY85 Receipt of Additional Research Procucts FYS6 Comission Paper on Source Term, Brief ACRS, and Finalize NUREG-0936 on Mathcdology FY86 We anticipate that NRR staff will continue to participate in all of the activities listec acose. Upon cccoletion of the RE5 Source Tem eor. in FY86, there will be considerable effort on the part of HRR to consider the need for and extent of revisions of guidance to applicants and licensees, and to implement the revised Source Term methodology in day-to-day licensing activities extant at that time. This effort will be initiated when preliminary results from the APS review beccme available (late CY84). Those parts of planning for feptementation and preliminary implementation that can be acccmplished efficiently will be dene. It appears that a major effort will extend through FY87. An example of the potential broad impact of changes in Source Term is i l should elements of the Source Tenn change materially from those currently used.,there will be the need to modify those elements of the j l t I l
. emergency procedures of all plant licensei 3 wnich depend upcn source term. While not a direct responsibility cf NRR, it is likely that NRR would be asked for assistance. 0 i l L ) l l 1 l l
W . II.A.5 SAFETY GOAL EVALUATION DST /F. Rowsome Introduction This program constitutes NRR's contribution to the NRC-wide Safety Goal Evaluation Program. The Program's basic objective is during the two-year evaluation period to study the implications of using the safety goals in the nuclear regulatory process. At the conclusion of the evaluation period the Comission will consider what changes, if any, are necessary in the goals and objectives and how they should be applied in regulation. Program Description The overall Safety Goal Evaluation Plan, as published by the Commission, is described in NUREG-0880, Revision 1, " Safety Goals for Nuclear Power Plant Operation," March 1983. The NRR contribution to the work is detailed in "NRR Safety Goal Evaluation Work ' Plan," T. Speis to D. Eisenhut et al., Decemoer 2,1983. Major NRR tasks include the following: 1. Evaluation of 1983 proposed. Safety Goals - This is a retrospective evaluation of whether and hew selected I recent generic safety requirement decisiens--and also some selected existing requirements--would have differed had the safety goals been applied. Special consideration of occupational exposure is included. 2. Evaluation of limited modifications of the 1983 Safety Goals - This is an evaluation of the effects of possible modifications of the 1983 proposed safety goals suggested by problems in applying the goals in trial evaluations. 1 1
. 3. Use of the Safety Goals in the Regulatory Process - This task identifies and discusses the main issues of implementa-tion and outlines a proposed approach. It includes inferences for implementation frcm the trial evaluations and censideration of PRA review, containment performance, and action guidelines. Planned Activities Schedule Interim reports Summer 1984 Complete trial evaluations 11/84 Complete implementation study 01/85 Final report to the Commission 03/85 Note: This is not part of the Severe Accident Program. The ef# ort is budgeted (or apportioned from related work) according to the needs of the Safety Goal Evaluation Program. Coordination between the two programs is planned, to help their deveicoment along ccmaaticle lines. l g-
25 - II.A.6 LICENSING DSI/AE3/J. Hulman Introduction Giver, the volatiltly of severe accident' issue at this time, Safety ano Environmental Licensing are review areas in which the NRR technical review staff must exhibit considerable flexibility in order to use the latest technical input consistent with policy directicn. For instance, there are pressures during the hearing process to use new knowledge, that has evolved, however tenuous. The staff must temper the temptation to use such new knowledge, recognizing that it often is subject to modification as it ages; but also and more important, to recognize the present guidance to the technical staff is to not move ahead of the ongoing methodical research, peer review, and associated decisions yet to be made. The overriding objective of all related work is to maintain credibility in technical issues. Prooram Descriotion Severe accident subjects are handled in the NTOL-licensing process in a variety of ways. Those post THI related issues nat have been resolvec - through changes in staff criteria are incorporated into each staff MTOL, TMI multiplant issue, or related licensing amencment SER via the SRP and staff positions. Other TMI issues which have not been #Ully resolved, and issues related to potential conservatisms in accident source terms used in SER evaluations, including implications for E0 and technical specifications, have been deferred pending the outcome of Severe Accident Rulemaking and ASTP0 efforts, respectively. Note tha't accident evaluations for SER's are based upon the 10CFR100 footnote, TID 14844, Reg. Guide and SRP assumptions of a " substantial meltdown of the core" without loss of containment function; in' essence a severe accident which w.
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. is successfully managed and does not result in loss of containment integrity. These same issumptions and guidelines continue to be used in the reviews of licensing amendments, enforcement actions, standard plant designs (SPD), new construction permit applications (CP), any plant for which a PRA is filed, unresolved safety issues (USI), and generic safety issues (GSI) which do not involve risk assessments and cost / benefit tradeoffs. The other parts of the safety review of SPC's, CP's, PRA plants, USI's, and GSI's relate specifically to risk assessments and cost / benefit tradeoffs of accident prevention and mitigation through the CP/ML rule (the proposed changes to 10 CFR Part 50.34(f) as presently conceived in the Proposed Severe Accident Policy Statement). Reviews of PRAs result in risk assessment reports. The only currently active staff work in the areas in which the staff is now using the evolving ASTP0 sponsored methodolegy in FY 84 is for a GESSAR II Severe Accident SER. A similar apcroacn is articipated for the advanced Westinghouse standard plant review, for the generic evaluation of Mark I and II cortainments for generic studies of accident mitigation, and for PRA reviews in FY 1985 and thereafter. The OSI/RSB is factoring new severe accident research results into the casework reviews. For example, the Branch is currently using the best available thermal-hydraulics methods including fuel-coolant interactions, core-cencrete interactions, and hydrogen phenomena. ~ Severe accident assessments are being undertaken specifically for three NTOL's and four OR's in FY 84 and 85. These indepth, plant-specific ,W 9 b -~. .,,4 n
. probabilistic risk assessments (FRA's) relate to the adequacy of design, construction, maintenance and operation of selected reactors at rela- ' tively high population density sites (Indian Point, Zion, Limerick, Nine Mile Point 2, and Millstone). The staff will also incorporate the results of these PRA's into the accident chapter of the plant Environ-mental Statements. A related severe accident program that is under consideration is the integrated Safety Analysis Program (ISAP) for assessing risks at some operating reactors. This is a continuation of the earlier Systematic Evaluation Program (SEP). Planned Activities Schedule Licensing is an ongoing program with specific goals and milestones associated with licensing actions for specific actions. The severe accident related effort will be incluced as ceterminec to be necessary cy the results of programs such as ASTP0 anc the Severe Accicent Policy Paper. I h 'e l i N e 1 l _,. ~ ..,..-.s,
-. : a. _ _ _ ...--a. Q . II.8 TECHNICAL ELEMENTS rn the following paragraphs we provide in summary fashion, the various technical activities that are being conducted under the aegis of severe accident activities. As will be noted, most of this effort involves m nagement of Technical Assistance Contracts and scee application of this work to regulatory requirements, generic assessments, casework, and aperating reactors. e e =. eM e -w.om.o - m ee w " gm G..
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8 II.B.2 SOURCE TERf1 DSI/AEB/J. Hulman Introduction Three types of severe accident source terms are utilized in staff safety evaluations. For CP and OL SER's the staff generally uses the TID 14844 . based source term for radiological assessments of accidents. For all PRA, DES /FES, GSI and USI evaluations (with two exceptions, GESSAR II and Limerick) the staff utilizes a variation of WASH-1400 source terms. For GESSAR II, the staff is using the evolving ASTP0 sponsored research. For Limerick the staff is using the same WASH-1400 variation, but is also assessing the cost / benefit of additional prevention and mitigation features using the evolving ASTP0 sponsored research. Procram Descriotion The TID-14844 source term is a specified fraction of two fission pro-duct elements (icdine and noble gases) and a general statement on particulates. '4 ASH-1400 includes estimatec fractions of fission products and activation products for seven element groucs. In principal, the ASTP0 source terms will be in terms of a methcdology #cr assessing release time-histories of all radiologically important elements, a leval of detail much greater than WASH-1400 and T!D 1484a. Resources required in applying this new methodology will be both oualitatively and quantitatively diverse. l l 'What is clear is that any modification of the existing review procedures flowing from TID-14844 and WASH-1400 that incorporates the results of l ASTP0 sponsored research is likely to incorporate existing and-newly developed computer codes, and an evaluation methodology for their i n.. -.....n.
. application to specific plants. New severe accident source terms are 1 not likely to result in a simple tabulation of the fraction of the core inventory of a few important fission products that would be released in any severe accident. Rather, a methodology employing several o sophisticated codes and a description of their use is likely. The resource implications of the use of the codes can be significant for most plant specific applications, unless a developmental program can be used to generalize assessments, an undertaking not yet proposed. Additional RES sponsored activities are underway to simplify code use, including consequence analyses, and confirmatory experiments. The results of these activities are being followed by NRR and applied to licensing activities as warranted. There are several important differences between the methodology used to ceveloo WASH-1400 source terms and the evolving ASTP0 sconsorec effort: a. There are significant cifferences in the understanding of relation-ships between temperature and the release of fission products from uncovered fuel between the time WASH-ldOO was written and tocay; b. Significant advances in the understanding and modeling of thermal-hydraulics that drive source term estimates have evolved since WASH-1400; c. WASH-1400 did not evaluate in detail the deposition, plateout and re-evolution of fission products in the RpV, primary system, containment and other compartments; d. The strength of containment and penetrations appear to have been given less credit in WASH-1400 than is presently considered rea-7 bq- ~ - ~ - ~..
e 3 32 - sonable; and e. Post accident radicchemistry appears not to have been fully con-sif.ered in WASH-1400. Planned Activities Schedule e e O .~. _ n - _. - - ---.7-
. II.B.3 ACCIDENT MANAGEMENT OHFS/PSRB/0. Ziemann program Descriotion The DHFS has recently submitted a user-need request to RES to develop generic emergency technical guidelines review criteria for a degraded core. Follcwing is a sumary of this proposed research activity requested by DHFS. GENERIC EMERGENCY TECHNICAL GUIDELINES REVIEW CRITERIA FOR A DEGRADED CORE purpose Recently reviewed generic emergency technical guidelines new being implemented may not adequately address severe accident management. One or more of the published staff SERs covering these guidelines has identified the need for further consideration df ocerator strategies during degradea core accidents. Staff review criteria are necessary to provide a basis for findings on future submittals. Objectives i Investigate and propose criteria to assist in future staff reviews of expected industry submittals on degraded core accident management l strategies. These review criteria would allow a more timely anc sound review process. projects Task 1: By June 1984, provide partial list of ca.ndidate criteria, or as ~ a minimum, outline potential accident management strategies I m... ...~.:..
d ...ith a degraded core. Task 2: By June 1985, provide a complete set of criteria for reviewing industry submittals. Office / Division / Branch Resconsible RES/DF0/HFSB l NRC/ Users / Lead Office NRR/0HFS/PSRB The identified tasks are being pursued by RES under the Severe Accident. Research Plan (SARP). SARP Section 5.3 of NUREG-0900 is currently being revised to reflect' these user needs. Administration Awaiting RES (Dr. C. Overpy, HFSB) latest craft of Section 5.3 for concurrence by Directors. CHFS/DSI comments were given to RES in Oc:cber 1983. Alternatives None Evaluation Soundness of technical bases for proposed review criteria. Planned Activities Schedule
. II.S.4 CORE t1ELT PHEh0MEtt0 LOGY tieed input from DSI/RSS / . ~
5 - !S - II.B.5 CCNTAINMENT LOADING OSI/RS./B. Sheron e I ) I l l l l l l l l l l l I l l t f %.-----*+- .--s.a. -. I
. II.B.6 CONTAD!. MENT RESPONSE DE/ECB/V. Noonen Introduction This technical issue involves the response of the containment structure to loads higher than anticipated in the design. Of particular impor-tance is the leakage characteristics of the structure and the potential a failure modes. This project will assist RES in the development of con-tainment failure models for severe accident source term models. By letter dated May 12, 1983, from R. Bernero to R. Vollmer, DE was assigned the lead to furnish a leakage model for containment failure to be used by RES. Prooram Gescription Pressure and temperature response including leakage rates due to imoortant severe accident sequences are estimated incorporating pre-existing leak areas and those leak areas that develop as a result of pressure leading and degradation of seal materials for containment ene-tration closures due to temperature effects. Centainment 'escense 1s. r estimated for levels consistent with low, medium and high levels of confidence. Planned Activity Schedule Publish preliminary draft NUREG-1037 ll/21/83(c) Publish revised draft NUREG-1037 Feb. 1984 Publi'sh third revised NUREG-1037 April 1984 Publish final NUREG-1037 Sept. 1984 l l l { l l l 3.
o . II.B.7 CONSEQUENCE EVALUATIONS OSI/AE3/J. Hulman Procram Descriotion Consequence evaluations of severe accidents usually involve the use of the CRAC code, or its recent modification called CRAC2. RES activities include further updates and simplications of the codes which are sched-uled well into FY85-86. The CRAC code can produce up to 39 accident consequence estimators, ranging from early fatality estimates to land area contaminated. The code has been the subject of considerable controversy since its inception with WASH-1400. Concerning the bases for the more important estimators, the completeness of economic modeling assumptions, and the magnitude of worse case estimates of early fatalities. NRR participates in RES related activities as peer reviewers, and uses the codes for DES /FES and PRA evaluations. Furthermore, NRR centractors use.the CRAC2 code to evaluate USI's and GSI's. The NRR organi:at cnal elements directly involved in application of the code are AEB, GIS, SPES and RSCB and various OL projects branches. Indirectly involvec are RSS. CSB, and RRAB that use the autout from such assessments to evaluate alternative ESF's, to develop containment performance guidelines, and to generate risk profiles, respectively. The estimation of monetary and environmental consequences involves SAB, METB, and EHEB. l Planned Activity Schedule O G e*wwp =gmum - summHElm.BAEEM=, - 9 >*t h ,e.,, e-e'*N 8, '~
II.B.8 C.\\S GENERATION DSI/CSB/W Butler Procram Descriotion m e 4 9 e d
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. II.B.9 EMERGENCY PREP REONESS OSI/AE3/J. Hulman Program Descrintien The primary responsibility for emergency preparedness in NRC rests with IE. NRR's resconsibilities in emergency creparedness are contained in OL project branches, and RAB, METB and AEB; DL project branches because of their licensing management function, and RAB, METB and AEB through DES /FES evaluatio,ns of design basis ano severe accident consequences and as consultants to IE. The consulting role relates primarily to plant specific reviews of Energency Operations Facilities, and Emergency Plan Reviews. In addition, RES initiatives in evaluating alternatives to the existing emergency preparedness regulations (10CFR50, Appendix E) will involve the same branches in review of Commission papers and the bases the refor. Lastly, NRR provides support to the Emergency Operations Center through staff participation in training for and operat on of the i Center. This sucpert is provided through specific individuai assign-nents cutside the normal NRR chain of command. planned Activity Schedule l l .= 6 4
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.. ~...... 1 II.S.10 FRA METHODOLOGY DST /RRAB/A. Thadani Program Oescriotice Methodology for a typical probabilistic assessment of care-damage likelihoods coinorises five major tasks. These are: Plant F.aailiarization Accident Sequence Definition Analysis of External Esents Reliability Data Assessment and Parameter Estimation Accident Sequence Quantification P The Plant Familiarization ta' k invcives identification and acquisition s of plant cesign, operational and administrative information. Accident initiators, systems success criteria, FMEAs anc walk-througn prececures are among the considerations 14 this task The Acciden: Sequence Definition task comprises tne ra'in actlyities required to cbtain cuajitative oefinition of accicent sequences which may lead to core camage and to breach of the containrert. Event trees are develcced to describe how the safety and mitigating functions protect the integrity of the core and contaireent. Systemic event trees e.nd Fault Trees are also develooed to evalua:e the role plcyec 'oy plant l equipment and procedural failures in accident scenarios. i The Analysis of External Events Task considers the plant response to I external accident initiators like seismic events, floods, fires, high wind....etc. Plant structures and plant equipment responses are used to evaluate the likelihoods of induced LOCAs and transients as well as ,f ') r"= f iy
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. their consequences. Further details about the trsatment of externally initiated events are covered elsewhere in this report. The Reliability Data Assessment and Parameter Estimation task is concerned with the quantitative information needs (i.e., data and related models) that will be input to the Accident Sequence Cuantifica-tions task. The generic and plant specific data requirement is defined by the analyses performec in the above tasks. The Accident Sequence Quantification task involves generation of Boolean ~ expressions for accident sequences, quantification and classification of these sequences and the performance of uncertainty, importance and sensitivity analyses. The current NRR Proabilistic Safety Analysis Procedures Guide (NUREG/CR. 2515) provides procedural details for concucting a precabilistic assess-ment of internally initiated events. The August 1984 revision will include treatment of external events as well as refinement of the methodology inasmuch as the state-of-the-art allows. NUREG/CR-2815 emphasizes limitations of the methodology and focuses cc building versa- ~ tile mocels f6r the plant systems. Furthermore, the review of these models and their results will be based on the procedures outlined in tne PRA Review Manual (NUREG/CR-3485) which is currently being supolemented with review procedures for external events. A revised review manual will be available in a draft form in August 1984 a f -**'"N**' -W $4'98,, g_ ,,J = <,,g
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. While significant progress has been made in a number of areas.:f PRA methodology, some areas are still faced with well recogni:ed caficien-cies and maturity problems. Among these areas are: modeling of the human elements in the analysis, treatment of failure dependencies and systems interaction, analysis of external events, treatment of design, construction and installation errors, and evaluation of the uncertain-ties in the results. PRA treatment of human factors should consider aspects of both accident aggravation and mitigation. Models for human reliability, including associated limitations and uncertainties, should be further deve!oped and validated, including operator errors of crission and commission, and for cognitive error. This model is currently being revised to reflect the latest in the state-of-the-art in this rapidly evolving area. More-over, human engineering models still recuire improved data; existing data are marked witn heavy reliance on excert judgment, priorities should be assigned tc developing more acercariate data base. Equipment performance in accident environment require imoroved treatment in PRA methodology. Methods should be developed to predict variations in environmental parameters follcwing an accident, for examole: temoer-ature and pressure buildup, increases in radiation exposure, submer-gence. Fragility data, relating equipment failure probabilities to environmental changes are needed. The ongoing industry sconsorec ecuio-cent qualification test program together with the NRC sponsored prtgrams could provide such information. = 9 f-- .,.p-- 7-. _6
a. 44 - A number of PRA areas af e alf.o suffering frem well recognized difffnul-ties wsth large uncertainties, especially those ariting frem iack of kncwledge of the relative importance of the va8ibus seufcas of 6nce.r-tainties or their ccmbined effect on the risk results generattd. NRR had communicated (memorandum from H. Dento:1, MAR to R. Hinogge, EEE dated November 10,1984) tnese as well as other concerns to tne Off1 e ~ of Nuclear Reactor.lessarch (RES) for consideration in RIS programs (in particular the aaplications oriented RM EP pecgram). Related Documents 1. NUREG/CR-2815 Probahilistic Safety Analysis Preccdures fluidi, Septe.mber 1983. Oujective: To orovide procedural guidance in performing probabilistic safety analysis studics. 4 2. fN2EG/CR-348 PRA Review Panual, Septemoer 1983. Objective: To provice prececural guidatice in reviewinc Orce2-bilistic Safety Analysis (PSA) stucies. 8 Planned Activity Schedule G b a e 4 8 ,4 ,t,-,.,-- e
~ O i !!.B.11 EXTERNAL EVEllT5 ti/R. Jacxson Introdactinn A program pl4n has ceen developed to improv'a the c$pebility of making probabilistic risk asses.lme?.t3 from exfernal events. A sumnry of this plan, whicn is cated Jun 30, 1583, addres:e's the folicwing external events: 591smic Fire Ex.tcrnal floed Wind 2 Man Made tbzar'd Pronram De;criotion In light of resultr, df ecent PPXs indicating the,t external events cculd pr9 vide a significant tent-ibutio1 to co?e damage frequency and risk, there is strong motivaticn for including a fist assessment exarJinati:r. af external events as a sJpol.ecent to the FRA exArrinatien of internal eve.".ts, I$ this supplement, r.o.quideratico cust be given te how the external events pertion cf the ev6re,11 PRA ant. lysis nay have te be
epresen'.ed in a Mnner Qiffefe,9% frCrr, that for it.;4rn31 events,
depending on the maturity of pr.1babilistic methods 'n the sp6cific disciplines 1,9vnived.. Greater inaht.Jsa 'iniolvement and interaction will lead to ircarovec Jser definition of the mar:ner in wnich external events r:"Jst be described and an improved _ app-eciation cf tre li'c.itat ens of d t5ese metnedologies. W(th regard tc the question 'Jf allocationgf c4scurces and prioritization of rssearch efforts in tha external ever.ts drea, there is u,__ e
,m, .s._. 46 - currently no identifiable basis for an importance ranking of the yarious f external events. The relative contribution to plant risk of tnese events is very dependent upon site specific features and the overall plant design. For example, the results of the Indian Point PRA (1932) indicated significant differences between Units 2 and 3. For the Unit 2 plant the fire, seismic and wind contributions to total core damage frequency were 40% to 30% to 10% respectively, with the remaining 20t ~ contribution due to internal events. The corresponding contribution for the Unit 3 plant was about 30% to 2% to 1% with the remaining contribution of about 70% due to internal events. Thus, while some information on the relative importance of different external events may be obtained from previous PRA results, it is important to note that such results are not generic but reflect tne particular plant / site ~ comoination reviewed, and that the results are hignly sensitive to the 4 assumptions employed by the analyst. There is a nr.ed identified for assessing the ' relative centribution te risk from external ' events as ccmpared to internal events. The ef#ect of the external events centribution to risk is compared to risk frcm internal events in a direct manner. The validity of making such a comparison is highly questionable because there has generally been no independent assessment of the similarities and differences in the assumptions, methodologies, and data'used in the two distinctly different areas. In order to provide any'significant progress in the utilization of PRA results, such an evaluation is needed to demonstrate credibility in such ccmparisons.
i I ' In all of the externel event areas discussed there are a number of pecblem areat that have been identified. There is also a great deal of j ongoing work both in hcuse and in RES to address a number of these problems. Thf s effert is quite extensive and should provide for i consicerable prcgress in improving both confidence and capability in PRA. The follawing contrtctual cetivities relate to external events: 1. Satsmic Hacard Characterization (Charleston) Program: A program developed with LLh1 to estimate the probabilities of exceeding
iffferent gret.na motion levels at nuclear power plant sites east of t
the Rocky 'Muhtains. (FIN A0428) 2, SGEB nat a contract with BNL to develop a computer code which will Camputa the prchabilities of exceedence 4f a limit state for con-grete containment and sheat wall structi*es. (FIN A3721) t 3. ,E3 has a contrast with BNL to develop a methodology and data base fer ortbacilistic assesscents of fire events. TMRS of RE3 has a contract with NBS to develop and cemenstrate a fire risk modeling a pp roach. (Firi A3710) 1 REAB and OE are developing a PRA procedure guide for external events. (FIN 43758) I 5. A tecnnit:al assistarce centract has been initiated to develop a technicue anc dats recuirecents to provide a pec:abilistic analysis i of high winds as a function of duration which includes all meteorological pnenomena unich can generate high winds. (FIN 88661) l l 6. PRAs treating axternal events wnich have been or are presently l being reviewed i.9clude Indian Point 2 and 3, Zion. Limerick, i l i
3 4-I Millstone and GESSAR II. e I I 7. A program is being developed in conjunction with RES to compre-hensively address the quantification 6f seismic margins above the design level, 6 Planned Activities Schedule 3 { ^ s d b l 3
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f a f } i E o w l I l j ~.
e o . II.C RELATED AC IVITIES Section B of this paper provides a comprehensive description of the Program Elements and Technical elements that combine to be a description of severe accident related activities. In this section of the paper we will demonstrate the manner in which these elements relate to one another, and to other portions of the operating plan. Within NRR, no one organizational unit can be identified with Severe Accident work that clearly has the lead for technical or administrative direction of the effort. Rather, portions of the work have been assumed by representatives of OSI, DST, DE, and OHFS. The lead organization / person for the Program Elements is as follows: NRR Other IOCOR DST /Speis RES OSI/Rosenthal Sernero Coordination DST, OSI, DE, OHFS (For RES w/RES each Research program, a lead NRR staff member is designated for Technical Evaluation) Severe Acci-OSI/Mattson RES - Bernero, Malero, Henry dent Policy Paper ASTP0 OSI/Hulman RES - Bernero Safety Goal DST /Rowsome
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1 . Licensing (Safety) OST/Thtdani (Environmental) OSI/Hulman The lead NRR organization / program for the Technical Elements is as follows: Accident Prevention OST/Thadani Source Term OSI/Hulman Accident Managemer:t OHFS/Ziemann Core Melt Phenomenology DSI/Rosenthal Containment Loading OSI/Rosenthal . Containment Response OE/Noonan Consequence Evaluation OSI/Hulman Gas Generation DSI/ Butler Emergency Preparedness OSI/Hulman PRA fiethodology DST /Thadani Eiternal Events DE/ Jackson Note that for some of the Technical Elements listed above tne icentity of the lead organi:ation/indivioual does not immediately coincide with that individuals' technical responsibility within NRR. The individuals listed above are those that either have assumed or are generally associated with responsibility within NRR for the technical area. Similarly, for the Program Elements, the lead has evolved to the listed individuals; but in some cases assignment has been more methodical. N ..,wg,,.
1 . The interrelationship between the various elements is comclex, best characterized by informal-ity, and cannot be easily diagrammed or related to traditional organizational resconsibilities. Nevertheless, the attached diagram attempts to show the interrelationships between the various activities. During and prior to FY84, the most central management of Severe Accident related activities was placed in the SARP Senior Review Group. *his group is headed by Deputy Director, Research and has as members thC Directors of DSI, OST, DE. Although not directed by NRR staff, most integration of NRR activities at management level comes from that group. This group is the best available integrated canagement o# Severe Accident work. Out of this group the following specific NRR assign-ments emerged: Severe Accident Policy Paper Work on Containment Loading Response, ano Performance IDCOR coordination. 1 l I i l 1 .-.. n -. -.. - -... -.s,_ y
. ASTP0 levereAccident IDCOR Policy Paper DST R. Mattson, DSI T. Speis in A v j ~ Technical Issues /
1. Accident Prevention OST

2. Source Term OSI
/ Comission SARP Senior
3. Accident Management DST
/ .p 1/ / Review Group
4. Core Melt Pherlomena DSI n
/
5. Containment Leading OSI
/
6. Containment Response OE
/ EDO j
7. Consequence Evalua-DHFS
/ tion / /
8. Gas Generation DSI if

9. Emergency Preparedness DSI L
q ACRS Severe Accident
10. PRA itetnocology OST Research Program

11. External Events CE RES Y
Safety Goal ( Environmental Safety T. Speis Casework Casework
1 a-52 III. SUMNARY-e A. Manpower and Technical Assistance by Program Element and Technical Element (Matrix being developed by J. Hulman. The information may also b'e included in the section.of the report in which each element is discussed.) 8. Integrated Schedule This will be an integration of the principal milestones identified by the lead for each section under Planned Activities Schedule. C. Conclusions The principal conclusion that I can draw at this time is that the direction of the severe accident effort is diffused thrcugh many organizational units in NRR. Please sunnarize any others you think should be made. l t i
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O t i .s
l OVERSIZE
DOCUMENT PAGE PULLED l SEE APERTURE CARDS 1
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