ML20205D150

Summary of ACRS 314th Meeting on 860605-07 in Washington,Dc. Meeting Agenda & Apps Encl

ML20205D150
Person / Time
Issue date:	03/25/1987
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2428, NUDOCS 8703300343
Download: ML20205D150 (255)

Text

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[db'~k pDxs/xp1 TABLE OF CONTENTS MINUTES OF THE ]f < ;LgU["f'hTj WM 84

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t t 94- i HASHINGTON, D.C. b'lG s $ b)i d C I. Chairman's Report (0 pen)............................ 1 II. South Texas Nuclear Plant, Unit 1, Operating License Review (0 pen).................... 1 III.

a Recent Operating) Experiences at NuclearFacilities(0 17pen........................l A. Possible ATWS Event at La Salle Unit 2......... 17 B. Reactor Scram at Palisades..................... 19 C. Repeated Snubber Failures at' Troian. . . . . . . . . . . . 22 D. Single Failure at Miniflow Logic at Pilgrim.... 24 9

IV. Source Term for Nuclear Power Plant Accidents (0 pen). 25

- ' V. ! Briefing Regarding IAEA Meeting on Chernobyl (0 pen).. 27 VI. Meeting wi th NRC Comi ssioners (0 pen) . . . . . . . . . . . . . . . '29 VII. ExecutiveSessions(0 pen)........................... 34 A. Subcomittee Assignments

1. ACRS Review of the Hanford N-Reactor /

Chernobyl Reactor......................... 34

2. Membership on the ACRS Managerrent C om i t te e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34

3. Subcomtttee Assignments.................. 34 i

( 4. Report of the ACRS Management Comittee... 35 r y;

, B. Reports, Letters, and Memoranda................ 37 x 1. ACRS Comments on the NRC Safety Research Program and Budget for Fiscal Year 1988... 37

2. ACRS Report on South Texas Project, Units 1 and 2............:................ 37 l

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I 314TH ACRS I1EETING ii

3. ACRS Comments on NUREG-0956, " Reassessment of the Technical Bases for Estimating Source Term" -- Review Copy / Final.......... 37

4. ACRS Recommendations on Hope Creek........ 38

5. Proposed NRC Policy Statement on Standardized Nuclear Power Plants......... 38

6. ' ACRS Status............................... 38 C. Future Agenda.................................. 38

1. Future Agenda............................. 38

2. Future Subcommittee Meetings.............. 38 D. B&W Program on Trip Reduction.................. 38 E. Memorandum of Understanding with the EDO....... 39 F. Proposed Amendment to ACRS Bylaws. . . . . . . . . . . . . . 39 G. -Conduct of Members............................. 39 H. Testimony by H. W. Lewis....................... 40 I. Implications of Chernoby1...................... 40 J. Report of the Nominating Panel................. 40 K. Reappointment of ACRS Member................... 40 L. Agenda for the Wingspread International Meeting......................................... 40 l

M. Member Complaint Regarding Lack of Documentation in Safety Evaluation Reports...... 41 l Topics for the 315th ACRS Meeting...............

N. 41 South Texas Project Safeguards Information Supplement l

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iii TABLE OF CONTENTS APPENDICES TO MINUTES OF THE 314TH ACRS MEETING JUNE 5-7, 1986 Appendix I Attendees Appendix II Future Agenda Appendix III ACRS Subcomittee Meetings Appendix IV Applicant's (HL&P) Introduction Appendix V Construction Organization, Philosophy and Status Appendix VI Plant Layout Appendix VII Three Train Design Appendix VIII Nuclear Assurance Appendix IX Operational Phase Organization Appendix X Operations Department Appendix XI NRC Regional Inspection Program Status Appendix XII NRR Presentation on South Texas Appendix XIII Possible ATWS Event at Lasalle County Station, Unit 2 Appendix XIV Reactor Scram at Palisades Plant Appendix XV Snubber Failure at Trojan Appendix XVI Recent Significant Events l

Appendix XVII NUREG-0956 - Final Report Staff Briefing i

Appendix XVIII Staff P,esponse to ACRS Letter of December 12, 1985 Appendix XIX Additional Documents Provided for ACRS' Use l

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51, No.104 / Frid y M';y 30, 1986 / ]ces 19638 .

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willreportbriefly Media) Arts AdvisoryPanet; Meeting currentinterest to e Copunittee.

1100 Pennsylvania Av'enue.NW., &d5AJL-12:30RM.: South Texos W;shington.DC 20506, Pursuant to section10(a)(2} of the, Pmject, Units 2 ond 2 (Open/ Closed}-

This meeting willbe open to the ederal Advisory Committes Act(Pu .

L 92-463), as amended, notice is he a Y %e members willhear and discuss the public on a space available basis. The. reports ofits subcommittee, the NRC t: pics willinclude guidelines and poli given that a meeting of the Media in Staff, and the Applicant regarding the Advisory Panel (Radio ProgrammI request for an operating license for this If y:u need accommodations due to the Arts Section) to the NationalCquncil disability, please contact the Office fo facility.

Special Constituencies, National on the Arts that was tobe held on MaYPortions of this session willbe closed 29,1986, from 9:00 am430 p.m. id roons as required to discuss Proprietary Endowmentforthe Arts.110U - 716of theNancyHanksCenter.1100 Information applicable to this facility Pennsylvania Avenue NW., Pennsylvania Avenue.NW., / and detailed security arrangements for

- W:shington, DC 20506, 202/682-5532; .

Washington, DC has been changed.This this project.

"ITY 202/882-5496 atleast seven(7) .

meeting will not be held on Junej12, f ISPJf-1:45 RM.i Topicsfor 1986, from 900 a.m.430 p.m. inl room days prior to the meeting.Furtherinformationwithrefere/

ncetor 716 of the Nancy Hanks Centeritim Meeting with NRCCommissioners (Open/ Closed)-The members will -

l this meeting can be obtained from Mr. Pensylvania Avenue,NW., Washington discuss the contents ofits report of J hn H. Clark, Advisory Committee.a DC 20506, / January 14,1986 to the NRC regarding Management Officer, National This meetingis for the purpose of the FinalDesign Approvalof the End;wment for the Arts Washington. Panel review, discussiom evaluatio&

'DC 20508, or call 202/882-5431 and recommendation on ap lications for CESSAR11BWR/6 NuclearIsland'

/ sess n i be l'osed gg un a no e Arts a d t rtions i as necessary to discuss Proprietary fo tie Arts. Humanities Act ofl965, as amended, rot $n orfon o,e including discussion of information Information and detailed anangements May 23.19ee.  !. for plant security for this class of givenin confidence to thd Agency bY nuclear plants.

(TR Doc.S1210s 'Filed 5-2 Samgrant tes aml applicants. In acenrdance with the gmpp 3:30PM.: Meeting with satsia caos rsar-ows I determinationof the Chairman NRCCommissioners (Open/ Closed}--

publishedin the FederalRegisterof February 13,1980, these sessions willbeThe members of the committee willmeet Expansion Arta Advisory Panet; closed to the public pursuant to with the NRC Commissioners to .'discuss the Committee's report of January 14 Meeting hnd(9)(B) of subsections (c)(4). (6)d, United States1986 regarding the CESSAR 11 Final section 552b of Title ~

Design Approval as noted above.

Pursuant to section 10(a)(2) of theCode. /

F d:ral Advisory Committee Act(Pub. Portions of this session will be closed L 92-483). notice is hereby given that a Furtherinformati withreference to as necessary to discuss Proprietary this meeting can b btained from Mr. Information and detailed arrangements meetingpFthe Expansion Arts Advisory, JohnH. Clark Ad soryCommittee for plant security for this class of Panel [ Overview Meeting) to the Management Offigler National nuclear plants.

Nati:nal Council on the Arts will beEndowment . for the Arts,Washingtom ads AM.-5:45 AMiNRCSofety held on June 16-17,1986, from 900 a.mr DC 20506, or call (202) 682-5433. Research Pmgmm (Open}-The 5:30 pm in room 714 of the' Nancy Dated:May23. see. members will discuss portions of the Hanks Center,1100 Pennsylvania Av:nue.NW., Washington.DC 20506. John E C1.,h, proposed ACRS report to the NRC Dinctor. Coun andPoneloperarlons, regarding the proposed safety research This meeting will be open to the j budget for FY 1988-89 NationalEnd rnentfor the Arts. f public on a space available basis.The 82 Filed 5-2444 &45 t31 5:45 Pht.-8:45 PM.: Future ACRS topics will include guidelines, policy and [FR Doc.E.,m Activities (Open/ Closed)-The th3 Five-Year Plan.

,,u.,,acoo,

' ,j members will discuss anticipated ACRS If you need accommodations due to a -

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dis;bility, please contact the Office for I "8 # ""*g gti I t e ull Special constituencies. National UC1. EAR REGULATORY Committee, COMMISSION Endowment for the Arts,1106 j , Portions of this session willbe closed Pennsylvania Avenue,NW., Advisory Comrnittee on Reactor as required to discuss National Security Weshington.DC 20506,202/682-5532,, Safeguards Nuclear Regulatory Infonnation.

TTY 202/682-5496 atleast seven (7)Commission; Meeting Agenda sy, June 6. M daya prior to the meeting. in accordance with the purposes of S:30 AJf.-10t30 A3f.: Recent Further information with reference to sections 29and182b.of the Atomic Opemting Experiences at Nuclear this meeting canbe obtained from Mr. Facilities (Open/ Closed)-the members

, Energy Act (42 U.S.C. 2039,2232b). the

- J:hnH. Clark. AdvisoryCommittee Advisory Committee onReactor will hear and discuss the reports of its Management Officer. National Safeguards willhold a meeting on June subcommittee, and representatives of Endowment for the Arts,Washingto& 5-7.1986,in Room 1046,1717 H Street, the NRC staff. Representatives of the

,f DC 20506, or call 202/682-6433 . NW, Washington.DC. Notice of this nuclear industry will participate as i Deted May 23.1dem meeting was pubtfshed in the Federal d ay 19, 86. 8p{ 0n f this session willbe closed I p,g g er to discuss Proprietary Information and Direcid Officeo/CouncitandPonel 'Itursday, June 5,198I detailed security arrangements for the h Operoyons.Notiano/Ebdowmentforthe 8:30 Arts.A3f.445

- EM.: Report of ACRS

facilitiee being discussed.

4 [IR Dec. so-121st Filed MS-et 6 Chairman ' 45 am)"(Open)-The

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. . . . ... .. . .. t Fed:r:1 Rw st:r / Vcl. 51, N:.104 / Frid:y, May 30, 4986 / N tices' . 19837 to discuss ACRS recommend:tions on Fraley (telephone 202/634-3265), '

ItN5AAf.-12:00 Noon: Reactivation ofDeferredand CancelledNuclear severe accidents.The Committee will . - between 8:15 A.M. and 5:00 P.M.

Plants (Open)-The members will hear also complete discuss!on of matters Dated:May 27,1906. '

I o briefing regarding major issues in considered during this meeting. N g 'I ,, i * ' ' ' '

reactivation of nuclear power plant Portions of this session will be closed Adm.ory CommitteeManagement Ofp.cer.

construction prolects. as necessary to discuss controlled and (FR Doc. 86-12154 Filed 5.as-as; 8:45 am]

1:00PJf.-2:30PJt ACRS classified information as noted above.

Subcomittee Activities (Open)-The Procedures for the conduct of and sa mscoo m e m 4*

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members will hear and discuss a report participation in ACRS meetings were " 7 by its subcommittee on thermal published in the Federal Register on -

Hydraulic Phenomena regarding October 2,1985 (50 FR 191). In i et Nos. Emet and 50-3701 .. ,

i proposed NRC activities in this area, accordance with these procedures, oral "

2:30PJf-3:00PJL NRCSofety or written statements may be presented em he n u fon of the 1$1 $teo th

%n nt M mhn and Committee's proposed report to NRC portions of the meeting when a * "-

• regarding the proposed NRC safety transcript is being kept, and questions research program for FY 1988-89. may be asked onlyb members of the TheU. .Nuclea'rRegulatory 3:15 PJL-5:15 PJL: Source Term for Committee, its consuftants, and Staff. Commisa n (the Commission)is' NuclearPowerPlantAccidents Persons desiring to make oral conside issuance of amendme'nts to

. (Open)-The members will hear and statements should notify the ACRS Facility rating License Nos.NPF-e discuss proposed revisions to the Executive Director as far in advance as and NPF- saued to Duke Power accident source term used in evaluation riate Company o erstionof the McGuire of nuclear power plants. practicable arrangementsso canthat approfe to allow the Nuclear Sta be ma Units'1 and 2, located necessary time during the meeting for in Mecklenburg County, North Carolina.

Su om t t v[ s( pen)-The such statements. Use of still, motion ne amendinents would authorize on Ih ddi th picture and television cameras during an emergency basis a one. time release

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, rt o fi s Manag ent Subcommittee this meeting may be limited to selected of the existing contents of the \

regarding procedural topics considered rtions of the meeting as determined Conventional (non. radioactive) i d its subcommittee meeting on June [y the Chairman. Information regarding WastewaterBasin,containingtrace \

• the time to be set aside for this purpose amounts of tritium,into the Catawba \

5 45 PJL-&30PJf: A oin may be obtained by a prepaid telephone River. Technical Specifications (TS)

Activities of ACRS t. fem 7ers (tment and Open/ call to the ACRS Executive Director, 3.11.1.1 and its referenced Figure 5.1-4, \-

Closed)-The members will discuss the R.F. Fraley, prior to the meeting. In view ** Site Boundary forI.lquid Effluents" define the authorized discharge point f repo*t ofits Nominating Panel regarding of the possibility that the schedule for g candidates nominated for appointment ACRS meetings may be adjusted by the radioactive material released in 11guld x to the ACRS.ne members will also Chairman as necessary to facilitate the effluents to unrestricted areas as being discuss the proposed reappointment of a conductof themeeting. persons only to Lake Norman.The proposed member of the Committee and the non- planning to attend should check with the authorization would be accomplished by ACRS activities of ACRS members. ACRS Executive Directorif such the addition of a footnote to TS Figure Portions of this session will be closed rescheduling would resultin major 5.1-4 at the discharge point for the as necessary to discuss information the Conventional Wastewater Basin into the release of which would represent a inconvenience.

Ihave determinedin accordance with Cat ba River, stating that this clearly unwarranted invasion of disch point is authorized for a one-personal privacy. section 10(d) Pub. L 92-463 that it is necessary to close portions of this time arge of water which contains Saturday, June 7,1986 meeting as noted above to discuss trace unts of tritium in addition to Proprietary Information (5 U.S.C. the no lly processed effluents of the 8:30 AJf.-22:30 PAL: Preparation of Waste Wpter Collection Basin, effective ACRS Reports (Open/ Closed)--The 552b(c)(4)) applicable to the facilitier being discussed, detailed information the date o(Commission approval.The members will discuss proposed reports regarding matters considered during this related to the security arrangementslimits at a change woWd not affect any existing or procedures regarding the meeting. nuclear power plant (5 U.S.C.

552b(c)(3)). information the release of processing of conventional (i.e.,non.

Portions of this session will be closed as necessary to discuss Proprietary which would represent a clearly radioactive) contaminants.

Information. detailed security unwarrantedinvasion of personal These revisions to the technical 5 U.S.C. 552b(c)(6)). classified specificstions would be madein arrangements. National Security privacy (d data (5 U.S.C. 552b(c)(3)), and response to the licensee's application for Information, and information concerning restructe amendments dated May 20.1986.

initiation, conduct. or disposition of a information concerning initiation, conduct, or disposition of a formal Beforeissuance of the proposed formal agency adjudication applicable to the matters being discussed. agency adjudication (5 U.S.C. license amendments lthe Commission

. will have made findings required by the Jz30PJf-2.MPJf. ACRSProcedures 552b(c)(10)).

Further information regarding topics Atomic Energy Act of1954. as amended (Open}--The members will discuss s proposed changes to ACRS Bylaws and to be discussed, whether the meeting (the Act). and the Commission's has been cancelled or rescheduled. the regulations.  ;

procedures for the conduct of ACRS An unexpected release of tritium lnto activities. Chairman's ruling on requests for the 2W PJL-3;00 PJf.! Afiscellaneous opportunity to present oral statements the Conventional Wastewater Basin has (Open/ Closed)-The members will hear and the time allotted can be obtained by created the need for prompt action as a report by a member of the Committee a prepaid telephone call to the ACRS proposed above for two reasons, both -

regarding participation on an ANS Panel Executive Director, Mr. Raymond F. stemming from the fact that the Basin is s

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UNITED STATES )

E' n NUCLEAR REGULATORY COMMISSION M :E ADVlsORY COMMITTEE ON REACTOR SAFEGUARDS

%  ! WASHINGTON, D. C. 20555 4., * * *** /

Revised: June 3, 1986 SCHEDULE AND OUTLINE FOR DISCUSSION 314TH ACRS MEETING June 5-7, 1986 WASHINGTON, D. C.

Thursday, June 5, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.

1) 8:30 - Bi45 A.M. ReportofACRSChairman(0 pen) 1.1) OpeningStatement(DAW) 1.2) Itemsofcurrentinterest(DAW /RFF)

2) 8:45 - 12:30 P.M. South Texas Nuclear Plant, Unit 1 (0 pen)

(BREAK-10:30-10:45) 2.1) . Report of ACR5 Subcommittee regarding an OL for this unit (JCM/MME)

TAB 2 --------- 2.2) Meeting with NRC Staff and Applicant (Note: Portions of this session may be closed to discuss Proprietary Information and security arrangements for this facility.)

12:30 - 1:15 P.M. LUNCH

3) 1:15 - 1:45 P.M. Preparation for Meeting with NRC Comissioners (0 pen / Closed)

(Note: Portions of this session will be closed as necessary to discuss detailed security provisions and Proprietary Information applicable to GESSAR II.)

4) 2:00 - 3:30 P.M. Meeting with NRC Commissioners (0 pen / Closed)

TAB 4 -------- 4.1) Discuss ACRS report on GESSAR II dated January 14, 1986 (Note: Portions of this session will be closed as necessary to discuss detailed security provisions and Proprietary Information applicable to this matter.

3:30 - 3:45 P.M. BREAK

17) 3:45 - 4:15 P.M. Future Activities (0 pen) 17.1) Briefing by H. R. Denton, NRR, regarding IAEA meeting on the Chernobyl reactor accident

5) 4:15 - 6:00 P.M. ReactorSafetyResearchProgram(0 pen) 5.1) Discuss proposed ACRS report to NRC regarding the Safety Research Program for FY 1988-89 (CPS,etal./SDetal.)

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. 314th ACRS Meeting Agenda '

6) 6:00 - 6:45 P.M. Future ACRS Activities (0 pen / Closed) 6.1) Anticipated Subcommittee activity (MWL) (0 pen)

TAB ----------

-TAB ---------- 6.2) Proposed items for ACRS consideration (DAW /RFF)

(0 pen)

TAB ---------- 6.3) Consideration of N-Reactor review (0 pen / Closed)(SJSP/RKM)

(Note: Portions of this session will be closed as required to discuss National Security Information.)

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. . - I 314th ACRS Meeting Agenda 13 REVISED: June 6, 1986 Friday, June 6, 1986, Room 1046, 1717 H Street NW, Washington, D.C.

7) 8:30 - 11:15.A.M. NRC Safety Research Program (0 pen)

(10:00 - 10:15 - BREAK) 7.1) Discuss proposed ACR5 report (CPS, et al/SD, et-al) .

6) 11:15 - 12:00 Noon FutureACRSActivities(0 pen / Closed)

. TAB ~---------- 6.1)- Anticipated Subcommittee activity (MWL) (0 pen)

TAB ---------- 6.2) Proposed items for ACRS consideration (DAW /RFF)

(0 pen)

TAB ---------- 6.3) Consideration of N-Reactor review (0 pen / Closed)(SJSP/RKM)

(Note: Portions of this session will be closed as required to discuss National Security Information.)

12:00 - 1:00 P.M. LUNCH

10) 1:00 - 3:00 P.M. Recent Operating Experiences at Nuclear Facilities (0 pen / Closed) 10.1) ACRS Subcommittee Report (JCE/HA) 10.2) Briefing by representatives of NRC Staff (Note: Note: Portions of this session will be closed as required to discuss Proprietary Information and Safeguards Infonnation applicable to these facilities.)

3:00 - 3:15 P.M. BREAK

11) 3:15 - 5:15 P.M. Source Term for Nuclear Power Plant Accidents (0 pen)

TAB 11-------- 11.1) Report of ACR5 Subcomittee (WK/MDH) 11.2) Meeting with representatives of NRC Staff and contractors as appropriate

12) 5:15 - 5:45 P.M. ACRSActivities(0 pen) 12.1) Report of Management Committee regarding June 4,1986meetingitems(DAW /RFF)

13) 5:45 - 6:45 P.M. Appointment / Activities of ACRS Members (0 pen / Closed)

SEE HANDOUT---- 13.1) Report of ACR5 panel regarding nomination of candidates for appointment to the ACRS (HWL/ALN)(Closed) 13.2) Reappointment of ACRS member whose term is expiring (WK/ALN)(Closed)

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• 314th ACRS Meeting Agenda ' REVISED: June 6, 1986 13.3) Non-ACRS activities of ACRS members (0 pen / Closed) 13.3-1) H.W. Lewis testimony regarding nuclear ~

future (0 pen)

(Note: Portions of this session will be closed as required to discuss information the release of which would represent an unwarranted inva-sion of personal privacy.)

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• 314th ACRS Meeting Agenda. REVISED: June 6, 1986 Saturday, June 7, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.

14) 8:30 - 12:30 P.M. ACRS Reports to NRC (0 pen / Closed) 14.1) Discuss proposed ACRS reports to NRC regarding:,

14.1-1 NRC Safety Research Program (SD,et al) 14.1-2 SouthTexas, Unit 1(JCM/PetE) 14.1-3 Reassessment of Source Tem (WK/MDH) 14.1-4 Recent operating experience at nuclear facilities (JCE/HA) (tentative)

TAB ---------------- 14.1-5) Examples of systems interactions (D0/RPS)(Tentative)

(Note: Portions of this session may be closed to discuss Proprietary Information, detailed security arrangements for the plants being-considered, and information that will be involved in an adjudicatory proceeding.)

12:30 - 1:30 P.M. LUNCH

15) 1:30 - 2:00 P.M. ACRSProcedures(0 pen) 15.1)

TAB ---------- ProposedchangeinACRSBylawsregarding)

Preparation of Minority Reports (DAW /RFF

8) 2:00 - 2:45 P.M. ACRSSubcommitteeActivity(0 pen) 8.1) Report of Themal Hydraulic Phenomena Subcom-mittee regarding activities in this area (DAW /PAB)

16) 2:45 - 3:00 P.M. Miscellaneous (0 pen) 16.1) Complete discussion of items considered during this meeting (Note: Portions of this session may be closed to discuss Proprietary Infomation, Detailed Security Arrangements for plants being discussed, infomation the release of which would represent an unwarranted invasion of personal privacy, National Security Infomation, and infomation concerning' initiation, conduct or disposition of a fomal agency adjudi-cation.

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MINUTES OF THE 314TH ACRS MEETING JUNE 5-7, 1986 gf*.;.gf'[a Q~ hL L MG2b

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The 314th meeting of the Advisory Committee on Reactor Safeguards, held at 1717 H Street, N.W., Washington, D.C., was convened by Chairman D. A.

Ward at 8:30 a.m., Thursday, June 5, 1986.

[ Note: For a list of attendees, see Appendix I. D. Okrent, F. J.

Remick, and C. P. Siess did not attend the meeting on Saturday, June 7'.]

Chairman D. A. Ward noted the existence of the published agenda for the meeting, and identified the items to be discussed. He noted that the meeting was being held in conformance with the Federal Advisory Consnit-tee Act and the Government in the Sunshine Act, Public Laws92-463 and 94-409, respectively. He also noted that a transcript of some of the public portions of the meeting was being taken, and would be available in the NRC's Public Document Room at 1717 H Street, N.W., Washington, D.C.

[ Note: Copies of the Transcript taken at this meeting are also avail-able for purchase from ACE-Federal Reporters, Inc., 444 North Capital Street, Washington,D.C.20001.]

I. Chairman'sReport(0 pen)

[ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.]

Chairman Ward indicated that he and T. G. McCreless, ACRS Assistant Executive Director, visited the Wingspread site and found it very satisfactory. He noted that an invitation has been extended for participation by representatives of the Soviet Union. The new NRC Chairman, L. W. Zech, Jr., will formally open the meeting.

Chairman Ward noted the retirement of R. B. Minogue, the Director of the Office of Nuclear Regulatory Research, and the retirement from Government of D. Eisenhut as of June 13, for a position with NUS Corporation.

II. South Texas Nuclear Plant, Unit 1, Operating License Review (0 pen)

[hote: M. M. El-Zeftawy was the Designated Federal Official for thisportionofthemeeting.]

J. C. Mark described the site of the South Texas project indicating that it is a Westinghouse 4-loop PWR of 3,800 MWt. He suggested that the three-train system for cooling might be of interest to some members as is the fact that the RHR pumps, which are in the containment, can withstand the enntainment environment under all conditions. D. Okrent indicated that he has two general questions that he would like to have addressed during the presentations. The first of these involved quality control and quality assurance

Minutes of 314th ACRS Meeting activities at the South Texas plant and a discussion of the de-tailed results of a preliminary scoping probabilistic risk assessment study.

J. H. Goldberg, HL&P, presented a brief history of the South Texas Project (seeAppendixIV). He dirscribed the site and the ownership of the project by four partnert Carolina Power & Light, HL&P, and the cities of Austin and San Antonio, Texas. He indicated that in November 1979 a special NRC inspection team investigated a number of quality concerns that focused on harassment of quality control inspectors and difficulties with concrete and nuclear welding. In December of that year a stop-work order was issued on complex concrete placement. In the spring of 1980, the NRC stopped work on nuclear welding and shortly issued a show-cause order and assessed a civil penalty of $100,000. With the help of numerous consul-tants, by October 1980 the project was able to demonstrate that key problems that impeded the quality of the job were under control and welding was restarted. The pouring of complex concrete was re-started in January 1981. F. J. Remick asked the reason for the April 1980 show-cause order. J. H. Goldberg indicated that it was to show cause why the construction permit should not be suspended.

It was directly related to welding and concrete placement problems and the harassment of quality control inspectors.

J. H. Goldberg indicated that by September 1981, after years of frustratingly slow progress on the project, the project owners reluctantly agreed that the project's interest would be better served with a more experienced architect-engineer. Bechtel Corpo-ration was hired in the fall of 1981 and Brown & Root elected to withdraw totally from the project. In February 1982 Ebasco Ser-vices was hired to take over the duties as constructor. The current project structure is one with Houston Light & Power Company functioning as project manager, Bechtel functioning as architect-engineer aad construction manager, and Ebasco Services functioning as constructor. J. C. Ebersole noted again that the plant is a Westinghouse design. He asked if the balance of plant is basically a Brown & Root design, an Ebasco design, or a Houston Light & Pcwer design. J. H. Goldberg indicated that the basic structural config-uration of the station is a Brown & Root design with a considerable amount of the nuclear analysis done in 1975 by NUS Corporation.

Nevertheless, Brown & Root did not do much of the design of the safety-related cable trays and raceways and virtually none of the nuclear piping design. Most of the mechanical and electrical auxiliary building design was done by Bechtel Corporation. The containment design was a collaboration of Eochtel and Westinghouse.

G. A. Reed complimented HL&P on the turbine orientation, noting that the layout of the turbine building is such that this plant is one of the first to have an arrangement where turbine missiles are not a factor in the penetration of key areas such as the diesel rooms, control rooms, or the containment. He suggested that Brown

Minutes of 314th ACRS Meeting & Root should be given credit for recognizing early on the best turbine building arrangement. M. R. Wisenberg, HL&P, pointed out that the orientation of the turbine building was an issue of concern in the early design stages of the project and modifications were made as a result of NRC staff questions.

J. H. Goldberg discussed the construction organization, its philos-ophy and status. He briefly sumarized HL&P's management philoso-phy. These include a commitment to build and operate the South Texas station in full compliance with applicable regulations, to '

reouire that people who do work take full responsibility for its, quality, and to require quality assurance to independently confirm the quality of activities being performed. There is extensive management oversight of the entire program to ensure compliance with applicable program requirements. HL&P also reports in a timely and forthright manner all matters requiring attention by a regulatory authority (see Appendix V). HL&P upper management organization was discussed.

D. Okrent asked how many of the managers have reasonable technical insight into the potential causes of severe accidents with respect to potential scenarios that severe accidents can follow given the different sets of constraints. J. H. Goldberg indicated that HL&P has two basic engineering organizations. An engineering group on the project and an off-project engineering team called Nuclear Engineering which handles most of the analytical work (core phys-ics, thermal hydraulics analysis, and probabilistic risk assessment). He noted that within the off-project engineering team at least a half dozen engineers might well qualify regarding knowledge of severe accidents. P. Dodson, HL&P manager of engineering, indicated that a half dozen individuals on the project engineering team would also quality. The nuclear engineering group on the project had participated in actual running of some of the Westinghouse codes. South Texas operations people have been very close to the Westinghouse Owners' Group Emergency Response Guidelines and emergency operating procedures. These individuals would also be knowledgeable of severe accident scenarios. D.

Okrent agreed that very few in operating engineering groups may have knowledge of severe accidents but he noted that no mention was made of upper management. J. H. Goldberg mentioned the capabilities of J. G. Dewease, Vice President of Operations; W.

Kinsey, Plant Manager; and K. K. Chitkara, Manager of the Off-Project Muclear Engineering Group, as well as E. Dodson, Manager of Project Engineering, who bring extensive experience to the South Texas project.

J. C. Ebersole pointed out the fact that HL&P has a number of problems with the Westinghouse turbine. He spoke of the extensive turbine inspections perforned by Westinghouse to protect the plant against physical damage potentially caused by turbine explosion.

., m Minutes of 314th ACRS Heeting J. H. Goldberg agreed with J. C. Ebersole's assessment of the situation.

E. Dodson discussed the plant layout and identified some of the

-project's unique features (see Appendix VI). The 46 acre essential cooling pond was described. He noted that 'it functions as the ultimate heat sink for the plant to provide cooling water for safety-related systems. W. Kerr asked the maximum temperature of the pond during a worst-case accident. M. Wisenberg indicated that the temperature used in the design calculations inat'were performed for a thirty-day capable pond was 10?.* Fahrenheit. E. Dodson pointed out that the site consists of two plants that are slide-along duplicates. Each plant is physically separated from the other and has its own safety-related and nonsafety-related systems. Only a few nonsafety-related support systems are connon to both plants. C. Michelson asked if the heating, ventilating and air conditioning system is just for the control room. E. Dodson indicated that there is an entirely separate system for the control room besides the rest of the electrical auxiliary building. J. C.

Ebersole pointed out that the distribution of air occurs in coseon duct work which is fed by redundant chillers in air moving systems.

D. W. Moeller noted that all have a connon outside air intake. E.

Dodson agreed but indicated that it is a concrete duct. D. W.

Moeller wondered if isolation of that concrete duct to the control room would turn off the air supply to the electrical auxiliary building. E. Dodson agreed to discuss it later in the session. He mentioned that the diesel generator building is compartmentalized into three compartments for the three identical Cooper diesel generators which service Class 1E sources for on-site AC power. C.

Michelson asked if the diesel is self-supporting in terms of its cwn battery supplies. E. Dodson indicated that it does not have to have external power. C. Michelson noted that water spray deluge systems are used for fire protection. E. Dodson indicated that the water spray deluge is in the diesel room itself actuated by a pre-actions signal. It is seismically cualified. C. Michelson thought it significant that South Texas can operate the diesel generators with the deluge system on. He wondered if all the equipment in the rooms was qualified for deluge including the switchgear, batteries and other electrical equipment. In answer to a question by F. J. Remick, E. Dodson indicated that all three diesel generators are independent from the standpoint of fuel and air supply.

E. Dodson explained that the South Texas project uses a three-train design instead of a two-train design which is physically segreqated and electrically independent. There are no shared components for heat removal from the core or containment atmosphere or heat rejection to the ultimate beat sink. He pointed out that the South Texas project has the capability to shut the plant down with one of three trains rather than one of two. Three trains also provide

Minutes of 314th ACRS Meeting greater margins since, for the majority of the analyzed possible accidents, one of three trains can successfully mitigate the accident. There is single train shutdown capability for fire protection for small break loss of coolant accidents, small breaks in general, and normal shutdown. It was revealed during a brief Committee discussion that the South Texas project can be credited with three 100 percent trains with the exception of the design basis accident, the non-mechanistic double-ended pipe break which requires two trains. E. Dodson added that if the double-ended pipe break is postulated to occur it is assumed that one train would ,

spill and one train would experience active failure. One train would be expected to inject. Therefore, three 100 percent trains are required for this scenario.

The electrical auxiliary building air distribution was discussed extensively including fire protection features , including fire dampers and chilled water systems in places such as the switchcear roomwherethecirculatingaircannotremoveenoughheatinafIre.

D. W. Moeller voiced some concern regarding the fact that water i deluge systems are placed on the charcoal filters on the recir-culating system to comply with Regulatory Guide 1.52 even though the ACRS has complained that water deluge systems should not be required in this application. There have been cases where they have activated and inadvertently shorted electrical cables. C.

Michelson asked if this deluge system is seismically qualified as was the case with the diesel generator room. E. Dodson indicated that the deluge systems are seismically qualified not to operate.

D. W. Moeller asked if there is a plan to test the emergency ventilation system for the control room regarding measurement of the rate at which temperature increases in the control room. W.

Kinsey, HL&P, indicated that HL&P plans to do a preoperational test of the control room ventilation system and will check temperature rise, recirculation and leakage in accordance with technical specifications at 18-month intervals. D. W. Moeller asked if HL&P was familiar with control room habitability studies that the EC staff has had underway for the past few years. E. Dodson indicated that they were.

E. Dodson explained that the three-train design coupled with the l plant layout provides considerable advantages for fire protection l including two ways to shut the plant down in the event of a fire in any area. He pointed out that the three-train capability extends

! to auxiliary shutdown casabilities including control of all three

, trains and the capability to maintain cold shutdown from the l

auxiliary panel. He noted that HL&P has compartmentalized the l plant to limit the vertical propagation of a fire by creating l separate fire areas et each of the elevations. He explained that

! the auxiliary feedwater system is a four-train system including a steam-driven feed pump with segregation of the trains with cne train for each of the four steam generators. Three independent i

Minutes of 314th ACRS Meeting Class IA power sources feed the three trains. G. A. Reed pointed out a potential vulnerability of the auxiliary feedwater system en loss of all AC power (station blackout) since there is only one turbine-driven pump. E. Dodson acknowledged the weakness. C.

Michelson pointed out that the NRC Staff's SER mentions a safety-related cooling water system with three 50 percent capacity trains. He asked the applicant to explain how HL&P characterizes them as three 100 percent capacity trains. E. Dodson explained that the trains are characterited by HL&P as 100 percent trains depending upon the amount of load shedding that is necessary. HL&P -

considers them 100 percent trains for the purposes of accidents since all that is needed is one out of those three trains. J.

Bailey, ilL&P, indicated the situation can properly be characterized as three 50 percent trains regarding the destgn basis accident but 100 percent trains for lesser events. G. A. Reed asked if the South Texas Project has feed-and-bleed capability. E. Dodson indicated that the plant does have feed-and-bleed capability but not with a loss of AC power because there are no other feed pumps except the one turbine-driven pump that can operate with ro AC at all.

E. Dodson explained that the control room of the South Texas Project fully comslics with NUREG-0737, Supplement 1. The control room design review integrated the human factors design post-accident monitoring instrumentation, safety parameter displays, emergency operating procedures, safety-grade cold shutdown capabil-ity, bypass and inoperable status monitoring for engineered safety features equipment and enunciator alarm prioritization (see Appen-dix VII). F. J. Remick asked where the SPDS CRTs are read. E.

Dodson indicated that they are on the main panel and also on various boards. R. L. Balcom, HL&P, explained that the dedicated SPDS CRT is on the operator's console and there are also CRTs in the Technical Support Center and the Emergency Operatiens Center.

C. Michelson asked about the fire protection system for the cable spreading rooms and the switchgear. S. West, NRC, indicated that the switchgear rooms are considered heavily cabled a-eas and the Applicant has agreed to put in a fixed deluge supprtision system manually actuated.

E. Dodson explained that another major aspect of the control room integration effort was the Qualified Display Processi1g System (QDPS). The objectives of this system was to optimize the in-strumentation design to include evolving regulatory requirements such as Three Mile Island, Appendix R and Safety-Grade Shutdown Criteria, at well as to provide optimized cable routing using the latest digital technology. J. C. Ebersolo asked about the # gree of redundancy of the 00PS. E. Dodson indicated that it is a fully separated three-train system. E. Dodson explained that the QOPS is a digital monitoring system which offers graphic displays which support the operating proceduros, while using fewer panel

Minutes of 314th ACRS Meeting '

indicators and simpler control panels. It relieves the operator of l the burden of cross checkin checking of input signals,g redundant simplifiesindicators, performs instrumentation quality of signal distribution using the data links while monitoring itself through on-line diagnostics and self-calibration.- D. Okrent requested further information on quality checking on input signals. He was l

concerned regarding the possibilities and under what circumstances the computed averages and deviations of redundant input sensors

, might lead to wrong infomation for the operator. T. H. Crawford, HL&P, indicated that the chances are very small and would require multiple channel failure of that signal. E. Dodson described the extensive verification and validation program to ensure the proper functioning of sof tware and hardware. He noted that the actual software and hardware is being tested in the Unit 2 system. R. L.

Balcum explained that the data are displayed either as questionable data or bad data based upon the redundant sensor calculation. The operator has backup indications and is not taught to believe the indicators blindly but to use the other plant parameters and the knowledge of events to further analyze the situation.

D. Okrent reintroduced his concern regarding the adequacy of the design and construction from a quality and quality assurance point of v'ew. He raised the specter of finding significant gaps in the overall plant quality after the ACRS issues its operating licensee report. C. J. Wylie and C. P. Siess assured D. Okrent that the Applicant would talk later in the session regarding the unique l features of his nuclear assurance program.

E. Dodson discussed station blackout and the fact that the South

, Texas Project falls into the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> plant station blackout general i design criteria. He noted that a plant-specific procedure for station blackout has been developed in conjunction with the West-inghouse Owners' Group Emergency Response Guidelines to ensure that proper operator action would be taken. The maximum seal leakage l has been calculated to be approximately ?5 gpm per pur:p or 100 gpm total. Ecch Class lE battery can supply station blackout loads for approximately 8 hcurs. E. R. Dodson noted that seal cooling can be maintained by operator action through a positive displacement pump powered from a balance of plant d< esel which is in addition to three standby diesels on-site. In addition to this there are five balance of plant diesels that can be hooked up if necessary and five balance of plant batteries that can be made available. He l noted that the reactor coolant pump seal leakage was based uson I

results of tests that were run in France on a 7-inch seal assem 1y which shewed 16 gpm. D. Okrent pointed out that the French have chosen to provide a direct source of reactor coolant pump seal cooling water as part of a beckfit to all of their plants. He l noted that the British Sirewell B plant will have a similar modi-l fication. He asked if HLAP had specifically considered that

Minutes of 314th ACRS Meeting technique and discarded it. E. Dodson indicated that HL&P has not considered that actual technique but had looked at the situation primarily from the reliability of the electrical grid. HL&P has not looked at the cost benefit or option of backfitting the steam-driven charging pump as being done in France.

E. Dodson briefly discussed prevention of an explosion involving the diesel generator oil storage tank. He indicated that the room in which the tank is located is continuously ventilated from ceiling to floor to remove potential flames. The fan is spark-proof on the IE bus and the rooms are provided with a foam-water fire sup 3ression system. Doors to the rooms are water-tight and locked closed and the tank level is monitored in the control room.

As a result, an explosion is not deemed to be a credible event. C.

Michelson and il. C. Ebersole expressed particular interest in the fire protection aspects regarding overfillirg of the diesel fuel tanks. it . C. Ebersole suggested that the problem is the level indication system on the fuel tank and its singular capacity to fail. H. P. Kadambi, NRC, agreed that it is a hazardous situation but indicated that the Staff had not actually looked at the scenar-10 of overfilling the diesel fuel tank. He suggested that this question could be addressed in an SER Supplement. M. W. Carbon asked if there might be some sort of coninon mode failure of filling all the oil tanks with a supply of bad oil. M. L. Balcum indicated that there are technical specification limits on the fuel oil put into those tanks and HL&P has a rigid sampling program prior to filling the tanks. Only one tank is filled at a time and a source would be sampled before filling. J. C. Ebersole asked if HL&P has sought to avoid crash cold starts on these diesels. M. L. Balcum indicated that HL&P has a surveillance program for a once-a-month start of the diesels from an emergency start signal. The diesels .

are unique, however, in the fact that they are not cold started since there are support systems that maintain them hot. C. J.

Wylie asked if a vibration analysis is done using extended runs on the diesels to ensure that piping disconnections do not take place.

He asked if such in situ tests will be run to pick out vulnerable spots for vibratio~ii and fatigue of pipe connections, it may take 750 hours0.00868 days <br />0.208 hours <br />0.00124 weeks <br />2.85375e-4 months <br /> of continuous running to do such a vibration analysis.

E. Dodson indicated that HL&P is doing vibration analyses through-out the plant but not such a test on the diesels.

E. Dodson discussed additional design features of the South Texas Project, including the qualified residual heat removal system inside the containment, the backup power for the chemical and volume control positive displacement pump, and steam generator sludge supports at the preheater. Similar design aspects such as the three-train systems that are being used in both France and Belgiun were mentioned. He noted that the South Texas Project has made several modifications to the Model E Westinghcuse steam generators at the secondary side of the plant to protect their

l'inutes of 314th ACRS Meeting

• investment. J. C. Eberstle remarked about the comonality of all the steam generators regarding the header arrangement from main feedwater and recent check valve failures at other plants. He postulated a burst in the main feedwater header system with improp-er function of check valves in common which could drain all the steam generators backwards through the feedwater system. He asked if HL&P had considered the dynamic reverse flow problem and the rapid closure of valves. E. Dodson indicated that the block valve

, is a gate valve and it closes equally well in either direction as does the flow control valve which is upstream of that valve. G. A.

Reed pointed out recent removal of tilting disk check valves in that same location and the insulation of swing checks because of problems with valves. E. Dodson indicated that HL&P removed the swing check valves.because of the problems that that type of valve has been experiencing.

C. Michelson brought up the issue of the detailed acceptance criteria developed by the NRC staff for the selection of pipe systems under CDC 4 to which revisions to the pipe break hangers will be applied. He noted that in the case of South Texas Unit 1 this provision is to be applied new. He asked the Staff how they intend to apply their criteria. V. ficonan, NRC, indicated that the appropriate Staff members were not in the meeting room to address that question. He indicated that the Staff would nake a submittal to the ACRS on this subject. C. Michelson thought that since the Staff intends to apply these criteria to South Texas now he did not believe it unfair to ask the Staff for the acceptance criteria at this time. C. Michelson asked the Applicant what they have re-quested of the Staff regarding GDC 4. M. Wisenberg indicated that HL&P has asked the Staff for pernission to take advantage of tbc existing rule on GDC 4 for main loop hreaks. There is a discussion pending regarding a submittal relative to balance of plant breaks.

All of those breaks will be inside containment. C. Michelson expressed concern since the Staff appears to have decided on application of GDC 4 for breaks outside of containment. He noted that the South Texas SER suggests this fact. He did explain that his concern would go away if the piping were only inside contain-nent. D. Okrent expressed his interest in also seeing criteria for pipes outside of containment.

J. E. Geiger, HLap, indicated that the South Texas f;uclear Assur-ance Program during the operational phase will consist of Op-erations QA, an independent safety engineering group (ISEG), a safe team program (employ)ee (see Appendix VIII . concerns),

He defined the and the Fitness forofDuty responsibilities the Program Quality Engineerinp Group and the Quality Control Inspection Group.

He trentioned a Technical Services Division which will perform other necessary and important tasks to implement a comprehensive QA program. Technical Services Division assignrtents are not in the nature of day-to-day support as is the operations QA division

o Mirutes of 314th ACRS Meeting previously described. J. C. Ebersole brought up the issue of welding problems that the South Texas Project had had back in the early days with Brown & Root. He asked how the issue of weld inspection (metallurgical review) was handled from a quality control and quality assurance standpoint.

J. Geiger discussed the formation of the Independent Safety Engi-neering Group whose responsibilities include providing continuing systematic and independent assessments of plant activities includ-ing maintenance and modifications. D. Okrent asked what, if anything, this group would be doing about systems interactions. J.

Geiger indicated that they will perform reviews and some analyses of selected problems that occur at South Texas and will do, on a selected basis, some root cause analysis. He stressed that he was not singling out systems interactions as an independent activity but said that the group would undertake activities with one of the important features of those activities being systems interactions.

He described the Safe Team Program as an administrative program for the purpose of providing a forum for South Texas Project employees to identify concerns in the area of nuclear safety quality. D. W.

Moeller wondered about the number of responses from employees regarding concerns and deficiencies. J. Geiger indicated that since September 1984 the company has contacted almost 18,000 individuals and received 580 concerns related to nuclear safety or quality. D. A. Ward asked if any of those 580 concerns have resulted in some significant change to systems or the program. J.

Geiger indicated substantiation of roughly 110 concerns which did result in some modifications to the course of business. The Committee discussed verification of the qualifications of welders which arose as a result of an employee allegation.

J. Geiger discussed a fitness for duty program based on the Edison Electric Institute Guide. He indicated that the progran at the South Texas Project has ten key elements which include top manace ment support, written policy, programs training, liaison with Iaw-enforcement as well as chemical testing. He mentioned a urinalysis test for illegal drugs and a breathalizer test used for drunken-ness. He mentioned strong emphasis on behavioral observation of employees by supervisors. A supervisor can reouest that a subordi-nate be given a chemical test at random. The Comittee discussed the chemical testing program and its implications. G. A. Reed raised the issue of the qualifications of QC personnel with regards to the issue of indesendence of QC versus technical qualifications.

J. Geiger indicatec that all inspectors on-site are certified current to relevant ANSI standards.

J. Geiger discussed the construction t ransition prog ram, the transfer of responsibility fron Brown & Root to Bechtel. In answer to D. Okrent's question regarding the possibility of significant quality or quality assurance issues arising during the remainder of

~

O Minutes of 314th ACRS Meeting the project, he spoke of 230 work packages in the transition program which covered items such as current status of the engineer-ing, including design verification, licensing items that were pertinent, such as I&E bulletins, circulars, and necessary SER changes, recomendations for any significant corrective actions, a summary of work in process, and assumptions of special conditions.

All factors were cross referenced. Open non-conformances in this transition from Brown & Root to Bechtel/Ebasco had been made the responsibility of Bechtel/Ebasco. Houston Light & Power personnel performed QA audits and surveillances and the NRC conducted in-spections and reviews. J. H. Goldberg added that Bechtel accepted technical responsibility for the work previously performed by Brown

& Root as a contractual condition. D. Okrent expressed concern regarding the depth of sophistication and the incentive Bechtel had to find Brown & Root errors. J. H. Goldberg indicated that Bechtel had a strong incentive to do a thorough review of Brown & Root's work in part because their professional reputation was at stake for any mistake that might be comitted however unintentional. There was also no financial penalty to conduct extremely detailed re-views. J. Geiger indicated that Bechtel reported several major findings as a result of their reviews in the following areas:

penetrations through the main cooling reservoir and erosion of the soil around pipe penetrations; transformer-size insufficient to handle loads on safety-related buses; defective weld joints discovered during inspection of emergen-cy cooling reservoir; capacity of safety-related HVAC insufficient to handle safety-related beat loads chiller capacity increased after Bechtel took over as architect-engineer and construction manager.

D. Okrent noted that Bechtel has a penalty clause in its contract.

If an error is found some years from now they have to pay for its repair. He asked if Stone & Webster also had such a clause in its contract. R. A. Frazer, HL&P, indicated that there is a participa-tion agreement with Stone & Webster which obligates them to call to HL&P's attention any matter that they deem to be questionabic or deficient from a technical point of view. D. Okrent asked how large a penalty could be assessed on Bechtcl. J. H. Goldberg indicated that Bechtel could conceivably forfeit its entire fee for the job.

J. Geiger discussed an effectiveness inspection program conducted by HL&P staff as a reinspection of work that had been previously inspected by Bechtel Corporation or Ebasco Services. These inspections were designed to replicate results to reach a

Minutes of 314th ACRS fleeting determination as to the quality of the inspection effort, not the quality of the hardware. In the 1985-86 time frama, these reviews were identified as a limited readiness review audit program. These independent reviews were performed by independent contractors supervised by HL&P management. Previously troublesome topics which were reviewed were seismic interaction, concrete materials control, environmental qualification, structural steel and settlement monitoring. There were no findings of safety-related problems. F.

J. Remick asked if HL&P thought that this readiness review concept was a worthwhile effort. J. H. Goldberg indicated that this review was similar to that undertaken at the Vogtle plant by Georgia Power. It was not particularly useful to the South Texas Project because those issues reviewed were ones with a history of being troublesome and were issues that had been previously solved perma-nontly. Such programs invariably turn up problems that somehow have been missed and the result is a never-ending examination.

J. D. Dewease, HL&P, described a nuclear group organization as con-sisting of groups assigned to plant operations, licensing, nuclear assurance, engineering and construction, special assignments, nuclear safety review board and corporate services (see Appendix IX). He discussed the Nuclear Training Department's program design to apply the systematic approach to training concepts and the major commitment of the organization to performance-based training. G.

A. Reed noted HL&P's use of the Edison Electric Institute POSS and MAST tests for preselection of personnel regarding training and reassignment. J. D. Dewuase indicated that HLAP believes in apti-tude testing.

J. D. Dewease briefly described the Nuclear Security Department (physical protection and safeguards services) and the nuclear construction organization which is a composite of engineering and construction functions. He indicated that the staffing for the operations phase activities continues essentially on schedule for about 1,400 persons for both Units 1 and 2. G. A. Reed thought that the 1,400 person staffing level was ambitious. He wundered tcw all of these individuals could t'e utilized of ficiently. C. J.

Wylic asked where in the organization slant vulnerabilitics and interactions are investigated. J. H. Go'dberg indicated that once the plant is operational, the engineering and construction depart-ment will continue to have a staff of enoincers to conduct basic review of the design from the standpoint of systems interactions.

C. J. Wylie asked who would maintain the PRA reliability analysis.

J. H. Goldberg indicated that that would be done in the Nuclear Engineering Group. C. J. Wylie noted that reliability analysis and systens interactions are interrelated. J. H. Goldberg agreed but noted that the engineers who will be conversant with the physical design and design criteria for the systems themselves will be part of the engineering and construction management group. They will be

?. .

itinutes of 314th ACRS Meeting -

1 supported in PRA analysis capability by the Nuclear Engineering Group as a coordinated effort.

, W. H. Kinsey, HL&P, discussed the nuclear plant operations depart-ment which is responsible for the safe operation, maintenance and testing of the station (see Appendix X). He explained that the Reactor Operations Division will operate six shifts, each with a complement of nine personnel per unit. He indicated that there will be a technical support organization responsible for providing engineering support to the other line organizations that report to the plant superintendent. One section of this organization is  ;

called the Systems Performance Section which is responsible for monitoring plant performance through testing, observation of i operating parameters through plant tours, and review of plant '

l maintenance work requests. The Reactor Performance Section is '

l responsible for routine monitoring of core performance. The engineers in this section will hold an SRO license and serve as shift technical advisors. HL&P believes that the decision to license the shift technical advisors will help to make them an I integral part of the shift's group.

l W .H. Kinsey discussed the Maintenance Division indicating that the ,

maintenance philosophy of HL&P is a strong preventive maintenance program with close supervision of the work. The preventive mainte-nance program will account for approximately 60 percent of expended maintenance man-hours. The maintenance organization is responsible for the station's measuring and test equiment program with the  ;

exception of chemical laboratory equipment and radiation  !

, protection. There is root cause determination for equipment '

i deficiencies. D. A. Ward asked if there is a strategy or plan for j the ratio of resource allocation to preventive maintenance versus corrective maintenance. W. H. Kinsey indicated that about 60 percent of the man-hours projected for maintenance perforrance will go directly into performing preventive maintenance as supporting ,

the preventive maintenance program. C. Michelson asked about color coding of valves and pipes to avoid confusion between Units 1 and  ;

2. W. H. Kinsey indicated that since the two plants are separated by 1,000 yards in distance the confusion between 1 and 2 has been i eliminated.  !

W. H. Kinsey discussed the work of the Radiological Protection $

Section which is responsible for implementing corporate and station i policies regarding radiation protection. He mentioned HLAP's commitment to the concept of ALARA which is reflected throughout i the organization including corporate management. He indicated that t

.1. H. Goldberg has set a maximum limit of 5 rem per year for any l

l individual while working at the South Texas Project and has set an t administrative limit of 4 rem ser year. D. W. Moeller mentioned e the Regulatory Guide 1.97 requ rement for the calthration of the  !

containment high range monitors which should be capable of reading l i

! I 1 t

F'inutes of 314th ACRS Meeting up to 10 0 r/hr. J. Rosenthal, NRR, explained how the Staff deals with calibration of these instruments while avoiding the horrendous potential radiation exposure to technicians. D. W. Moeller noted that the radiation protection manager is really the Health and Safety Services Manager. J. H. Kinsey indicated that HL&P empha-sizes occupational health and safety as well as radiation pro-tection. C. J. Wylie asked which group in the station organization reviews modifications for their safety significance coordination and implementation during the operational phase. J. H. Kinsey indicated that while there is' a coordination process between operations engineering and quality assurance departments the primary responsibility lies with the engineering department. The plant manager's staff is obviously very concerned about review of those rrodifications and the nuclear safety review staff will also be involved, as well as the ISEG. C. Michelson asked how HL&P processes LERs generated by other utilities. The discussion centered on the fact that HL&P as well as other utilities rely on INPO to process, categorize and screen all LERs.

L. Constable NRC Region IV, discussed the overall inspection program at the South Texas Project as well as the status of alle-gations. He indicated that the NRC has spent a great deal more time inspecting the South Texas Project than it would normally spend at a nuclear plant, in part because of the interesting past of this facility, and expects to incur over 30,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of in-spection effort through 1986 (see Appendix XI). He explained that the preoperational inspection program is just getting started and the bulk of the system testing inspection effort so far has been procedural reviews. Generally, the staff has been impressed with most SALP results for this utility. A CAT team inspection in late 1985 uncovered major problems and the Staff is considering appro-priate escalated enforcement actions. An enforcement conference was held with the utility. D. A. Ward asked what the problem areas are. L. Constable indicated that some of the problems involved isolated incidents with individuals regarding QA inspections of welding. G. A. Reed asked if this is influenced by the orga-nizational structure of quality control. L. Constable indicated that it was too early to say if that was the problem. The NRC does seek the benefit of the Applicant's input and has not gotten to the point of issuing violations. The allegations are typically well founded and fairly norrial for a project at this stage in con-struction. The allegations are fairly routine.

N. P. Kadambi indicated that the ACRS Subcommittee members request-ed additional information on five questions af ter the Staff's presentation of license conditions, open items and confirmatory iteris . There were no coninents on the Staff SER. He indicated that the Subcomittee raised concern regarding the degree of protection afforded by the separation between trains in the bunkered system which characterizes the South Texas design, pe11 ability prcblems l

Minutes of 314th ACRS Meeting associated with fire dampers was an issue. He noted that the Staff has issued two information notices and received a report from Ruskin, one of the manufacturers of these fire dampers. C.

Michelson indicated that he was under the impression that there were no cross ties between ventilation trains and no need for fire dampers. H. Dodson, HL&P, indicated that there are darpers from the common supply exhaust intake and outlets into those trains, and there are cases where a comon wall exists between the trains. C.

Michelson asked regarding the powering of these dampers and their failure mode. H. Dodson indicated that they are DC powered. J. C.

Ebersole indicated that it does not matter how they are powered.

They are not redundant and of questionable reliability. N. P.

Kadambi indicated that the Staff has received 50.55(e) reports from the Applicant describing deficiencies in their damper systems and corrective actions. HL&P has concluded that the problem is re-solved. The implementation, of course, is subject to NRC in-spection. S. West, NRR, indicated that the fire dampers in ques-tion are most likely released by fusible links. Any release by a smoke or heat detector would be powered by the diesel generators so they would stay open until there actually is a fire. J. C.

Ebersole wondered whether the fusibic link would function early enough in temperature rise to preclude overheating of electrical apparatus. S. West indicated that fusible links are available with different temperature readings anywhere from 165'F on up. J. C.

Ebersole wondered whether any electrical apparatus could tolerate a 165* ambient temperature. S. West indicated that you would presume the equipment in the fire area actually lost and, even thcugh there night be 165'F in the duct work at the damper, one is not probably to have the same temperature in areas adjacent to the fire area, J. C. Ebersole thought the Staff was counting on an ill-defined temperature gradient.

H. P. Kadambi discussed the ACRS' concern regarding the issue of fire protection in the diesel fuel oil storage areas, especially the proximity of these storage areas to the control room. He indicated that the Staf f has found acceptable the design changes proposed by the Applicant to augment fire protection of these storage areas. He explained that an ACRS question regardinq seaaration of battery rooms from the balance of plant was a case of ambiguous wording in the SER. The staff had in mind not the separation between the battery room and the balance of plant but the balance of the equipment in that train for the particular battery room.

N. P. Kadambi indicated that in the case of fires in cable trays, the Staff has taken into acccunt applicable Sandia National 1. ate-ratory tests in conjunction with the combustion tests on IEEE 383 qualified cables. He noted that the Staff does not believe that some recent tests having to do with fire in cabinets are applicable to the fire potential for cable trays. The spread of fires within

Minutes of 314th ACRS Meeting .

cabinets is characterized by different mechanisms than for cable trays. J. C. Ebersole indicated that the question revolves around whether you have an autocatalytic and progressive fire on the trays caused by burning of a great mass of cables. S. West explained that the ignition resistance and flame spread properties of the qualified cables is much less than for unqualified cables. J. C.

Ebersole suggested that this is just skirting the issue, that the issue is that the cables would burn but not perhaps as briskly as unoualified cables. ,

i N. P. Kadambi indicated that the last of the questions from the Subcommittee meeting involved the relevance of the San Onofre llovember 21,1985, event to the South Texas design. The San Onofre event, a water hammer phenomenon, led to check valve failures. He l indicated that there is a strong defense against a similar event at l South Texas. The defense has to do with the fact that there are l separate lines for the feedwater and the auxiliary feedwater systems. In addition, the feedwater line has an ESF actuated isolation valve in addition to the check valve.

N. P. Kadambi indicated that the current SER does not speak about an txemption from GDC 4. In Section 362 of the SFR the Staff states specifically that South Texas conforms to GDC 4 and pipe rupture postulation and associated effects. The Staff has received exemption requests from South Texas related to the primary coolant loops and pressurizer surge line. The exemption requests for the primary coolant loops has been rendered moot by the limited scope rule recently approved by the Comission in its final form. The surge line exemption is being reviewed at this time. The Staff has I

not developed any criteria by which to accept or reject it. C.

Michelson noted that the broad scope rule out for public comment pertains to all piping inside and outside containment. He asked if the Staff is considering eliminating required breaks including arbitrary intermediate breaks and terminal point breaks outside containment. N. P. Xadambi indicated that the request from the Applicant only applies the leak before break conce)t inside con-tainment. C. Michelson referred the Staff to Append"x G of the SER which implies elimination of arbitrary intermediate breaks both inside and outside containment. N. P. Kadambi indicated that this Appendix is not an exemption from GOC 4 but is viewed by the Staff as a deviation from the Standard Review Plan. M. R. Wisenberg explained that relief regarding arbitrary interrrediate breaks are implied both inside and outside containment. This issue which was handled by the NRC Staff is intended to be outside containment and part of what will be covered ultimately by the broad scope rule.

C. J. Wylie asked if the Staff intends to require testing of diesel oenerators to assure their long term capability of cperation. Such in situ testing would be of vibration, fatigue, or analysis of vibration fatigue. C. Berlinger, NRR, indicated that the Staff

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n does not h9ve a ree,uir ment foi' licensees qualifying a die'sel te 1 run for 10 cycles w was done in the case of the TDI requali- / ,~

fication prograar. Tht program was done primarily because of identified specific roblems with regard to design quality asntt 4

ance, quality contrM, and manufacturing. C. J. Wylie pointed sta that during the TM testir"4 program, it was found that pipih -

designed by the architet/tegineer such as , oil piple to the [

diesels had corroded after uany hours. There were a. number of- ,

reports of broken oil pipes. He suggested that the pipite systems connected to the dierels be vibration tested. C. ~ Berlir.ger in- '

dicated that tiia prot 46.ns experienced with the piping were primar-ily concerned with ' adequate restraints on high )ressure fuel oil piping or actual mechanical or material defects ciscovered in some of the tubing. H4: asserted that if the pipes are adeountely supported, there-should not be a vibration preblem. He also m.ted that the level of vibration on the diesel en '

- and the foundation is monitored with sensgrs.gine support structureAn' ecc ,

do vibration testing is to' utilize walk-do06 examtr.4tions during /

engine operation during preoperational tests. That is industry practice but' is not an MC 'iquirement. t . C. J, general k

contended that a 750 hour0.00868 days <br />0.208 hours <br />0.00124 weeks <br />2.85375e-4 months <br /> test (30 days under full load) ylie was necessary to do a proper test of the piping. V. Noonan, NRO, indicated that the Staff intends to issue two more supplements to the South Texas SER this year. He expected that many of the open ,

items will ba finalized and cloud out in those suprier%nts. , ,

A security briefing on the South Texas project was ' eld in closco session. The discussiot' of this portion of the tres*ing wili be found in a supplement attached to the end of the minutes, III. Recent Operating Experiencea at Nuclear Yapj,j,ljty,s, (0 pen)

(Note: H. Alderman was the Opetsnated Federal Of ficial for tY*

portionoftheT.eeting.]

A. possible AfWS Event at t.a Sa'ieJnit 2 D. Allison, NRC, discussed a possible ATWS event that occurred at the la Salla County Station Unit 2 on June 1,1906 (See Appendix XIf t). Ibe plant was operating at about U percent power and had a feedwatur trent.ient that brought the reactor  !

water level down very close to the trip set point of 12.s inches instruinent level. At that point the operator trok i action and tha level went back un. C. Michelson asked N w;far above the core 12.5 inches is. D. Allison indicated the zero point is about 155 inchus. OurIng the incident, one r ' the four reactor protectian syster. channels indicated a Ic9 la/ci, tripped, and gave a half scram. The other three chi.inels cid not. Apparently at the time the operator thought ha had trade

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it through the transient. During a later review at Comon-

- ' ,. , wealth Edison of the records of the transient, it began to look like the level had gone down to about 6. inches where all four switches should have tripped. It was concluded that the r m tor protection system may have malfunctioned and an alert was declared and the plant shut down slowly. An IE inves-tigation team at the site has so far concluded that the set points on the pressure switches that give the low level trip

,, signal vary by a few inches. It appears that this is what cau,ed the lack of a trip. The switches have displayed 1

se*. point problems at Oyster Creek, but the drift noted there t , , is not nearly enough to cause a real safety problem in this

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particular application. W. Kerr asked the significance of the formation of an inspection team. D. Allison indicated that they are there to assist the region and make sure that the

, problem was only because of the failed switches. E. Jordan, NRC, indicated that the Staff plans to issue an information notice promptly communicating what the Staff knows to this j point. He explained that the plant would remain shut down until the Staff has investigated the problem. There are 130 of these switches used in that particular plant. W. Kerr indicated that he got the impression that these were fairly newly installed switches. D. Allison acknowledged that the switches are newly installed at La Salle and Oyster Creek.

, They are part of an upgraded environmental qualification modification and unlike the previous installation, one cannot

tell that they are drifting except by doing a calibration

! check. The previous installation was a Yarway Level Indicator

, , which 'could be read every shift or every day. These are blind switches in that all that can be done is the application of a to t pressure and a calibration check. C. Michelson asked if Comonwealth Edison is using the Yarway to indicate control rocm reactor level. He surmised that they had replaced the control function of the Yarway with this separate switch. D.

Allison agreed. He noted that the investigation will determine why the switches failed, investigate problems of i feedwater pumps that caused the transient and operator F . reaction and whether the incident was reported to the NRC

)roperly. He mentioned that the company did not realize they ud a problem for some hours and did not report it to the NRC for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. G. A. Reed thought that the Licensee ought not to be criticized. He thought this was good performance on the part of the Licensee in the fact that they reported the incident as soon as it was recognized even thcugh it was many hours af ter it had occurred. E. Jordan noted that once the utility had notified the NRC that their plant had failed to

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trip, they voluntarily took the plant down to try to find out

!, what had happened. D. Allison pointed it out to the comittee that there was a previous trip of the plant on May 9,1986, where the level was going down and kept going down. During

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that incident, the operator notice'd - the level below 12.5 s Linches and . decided to ' scram. An automatic scram occurred before the operator could execute a manual scram.

i B. Reactor Scram at Palisades .

i W. Hehl, NRC Region III, indicated that the Palisades Plan't-experienced a reactor trip from 99 percent power in response '

to a high pressurizer pressure condition on May 19, 1986. The

% high pressurizer pressure condition was the result of the loss i-of control power to the turbine EAC system which allowed the turbine valves to close. Upon reactor trip and during the plant recovery, numerous pieces of equipment failed to per-form. As a result of the reactor trip and associated multiple equipment failures, as well as the potential serious chal-1enges to safety systems posed by these failures and the burden that these failures place on the operating staff, the NRC dispatched a fact-finding team to review the event prior to the unit returning to power. The Augtented Inspection Team (AIT) was tasked with performing an independent review of the May 19 trip to assure that the scope of the equipment failures t was accurately known to evaluate the equipment failures, to gain a perspective regarding the impact of these failures and any existing out-of-service equipment On the Operating staff, and to assess their ability to respond to the plant transients i (see Appendix XIV). W. Hehl presented backgrcund information on the Palisades plant which included the troublesome SALP report covering the perb d' November 1984 to October 31, 1985.

The areas of maintenance surveillance, quality programs, and administrative controls were singled out as problem areas due to a lack of aggressive corrective action by the Licensee and.

poor management controls. Prior events at the facility which began in late 1985 due in part to inadequate maintenance involved safety-related equipment including five separate events related to leaking safety injection tank check valves.

Despite maintenance on these valves during the cycle 5 refuel-ing outage, during cycle 6 two of the 8 valves had to be refilled. Additionally, during the cycle 6 refueling, the Licensee elected not to perform maintenance on the primary coolant pumps despite indicated seal oscillations.

W. Hehl explained that durini a March 1986 startup from a

. refueling / maintenance outage, 'two of four primary coolant

. pumps developed seal problems. Valve leakage problems were also identified in the primary coolant system loop, in check valves, and in two safety injection tank pressure control valves and a manual isolation valve associated with the safety injection tanks in the three way valve in a CVCS system. Also mentioned were on April 10, 1986, exceedance of the technical specificatien limit for unidentified primary coolant systen 1

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Minutes of 314th ACRS Meeting leakage and subsequent shutdown. On April 11, 1986, a derat-ing took place because of a pump packing failure and valve leakage problems in the primary coolant system makeup systen during the period April 23-29, 1986. D. Okrent asked if this pattern of repeated. failures is to be expected and is similar to the average for nuclear plants. W. Hehl indicated that the repetitiveness of equipment failures is far from what would typically be expected. D. Okrent asked if many- of these failures would not have occurred had the Licensee done more maintenance during the cyc1b 5 shutdown for refueling. W.

Hehl indicated that there has been significant concern on the part of Palisades operators with regard to maintenance activ-ities at Palisades and the reliability of equipment. There

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has been experience with leaking check valves since 1983 and vendor assistance has been used in rebuilding these valves.

D. Okrent asked if the Staff knows why this problem has i

occurred. W. Hehl suggested the inadequacies of these valves for the application as a possibility. J. C. Ebersole wondered what the Staff's course of action would be in differentiating between inadequate design and poor maintenance. D. Allison indicated that a new reactor coolant leak is occurring at

, Palisades every two to three days and the Staff does not know whether it is a design or a maintenance problem. J. C.

Ebersole suggested that it is the Licensee's option to put in better equipment. W. Hehl indicated that there are not many nuclear plants like Palisades and Palisades does have some unique problems regarding the location of their safety in-jection tanks. Palisades is somewhat unusual in its experi-ence with this type of repetitive and continuous problem with check valve leakage. G. A. Reed pointed out that the Pali-i sades plant has been in operation since about 1970 and these i- valves were basically sound at that time but possibly in need of careful maintenance in the intervening years. It was his impression that the Palisades maintenance organization is lacking.

W. Hehl discussed the sequence of events during the May 19, 1986 reactor trip at Palisades. He noted that the pressurizer spray valve failed to fully close. G. A. Reed suggested that the Staff identify whether the pressurizer spray valve is a bellow-sealed valve or whether it is just a packed valve because that may have been a factor in its failure to close.

G. A. Reed mentioned a loss of packing on a pressurizer spray valve at the Zion Station suggesting a connection with this incident. D. Okrent noted that the reactor tripped on pressurizer high pressure indicating that the reactor coolant system was heating up and that there was a subsequent cool down. He asked if there was any violation of cool down rate during the aftermath of the transient. W. Hehl indicated that the failure that occurred did not resul t in significant l

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Minutes of 314th ACRS Meeting worsening of the plant transient. The performance of the operators in the other major plant systems performed as expected and well within design criteria.

W. Hehl indicated that the AIT inspection, based on their review of the apparent failure modes, the maintenance history, and discussions with the Licensee's maintenance organization finds that significant weaknesses exist in the areas of diagnostics, troubleshooting repair, and post-maintenance testing. These were contributors to most of the failures that occurred. J. C. Ebersole asked if the Staff has identified a managerial problem at Palisades. W. Hehl indicated that within the last two years significant management changes have occurred at Palisades. With the abandonment of the Midland facility, there was a truncation of the management at Pali-sades incorporating part of the Midland management. Part of the problems cbserved is rooted in the experience level of both tha Palisades plant maintenance operations engineering and the inexperience of the management staff. G. A. Reed asked if Consumers Power uses validated aptitude testing for employment and transfer of their personnel at the Palisades plant. He suggested that the Staff look into this matter since it appears that there have been years of problems with respect to the performance of plant people in operations and maintenance. He noted that while the Big Rock Point Plant is much smaller, it is not noted to be a particular problem. W.

Hehl indi7ted that it is not uncommon to have differing levels of performance within the same utility. G. A. Reed thought this might be attributable to a lack of stan-dardization in the evaluation and processing of new employees.

Perhaps they need more regimentation. W. Hehl indicated that this might be part of the problem. D. Okrent asked if there are any objective indicators that support the Staff's opinions concerning the cuality of maintenance. W. Hehl cited examples of incidences of poor maintenance in the repeated reworking of valves. G. A. Reed suggested that the forced outage rate is probably very high for this plant. This is one of the best indicators of poor performance. W. Hehl cited financial pressures on the utility in part from their harsh treatment by the public service commission of the state of Michigan. Many of their employees at Palisades have had to take significant pay cuts and the attrition rate in their operating department has been of the order of 25 percent. G. A. Reed suggested that possible identification of a parallel case in TVA. He wondered whether the NRC Staff and the regional people will be able to turn this situation around. The committee discussed the fact that because of public utility commission action, Consumers Power may not be able to hire appropriate staff, as with the case with Davis-Besse. E. Jordan noted that the EDO has reviewed the actions of the region taken with respect to

Minutes of 314th ACRS Meeting Palisades and believes that it is appropriate. He also noted that there is a periodic or quarterly look at plants that are most troubled and the Staff wishes to return to the Committee with a status report on the Palisades plant. He noted that it is interesting that the Staff is finding that the problems with Palisades are principally based on balance of plant equipment rather than safety-related equipment. While there is no NRC requirement to cover this situation, there are certainly precursors of serious problems out of the large number of balance of plant failures. D. A. Ward noted that from his review of the SALP ratings for Palisades over the last six years, ratings appeared to improve in the 1980-83 period and have now deteriorated again. He wondered if IE and regional activity has increased in response to the decrease in ratings. E. Jordan indicated that the inspection programs are now adjusted and are based on the poor performers. The better performers get less inspection. The poor performers more.

Palisades is receiving a great deal of inspection attention.

D. W. Moeller observed that is significant that close to 10 percent of the operating nuclear plants are currently shut down and unable to return to power without careful reviews by the NRC Staff.

C. Repeated Snubber Failures at Trojan T. Chinn, NRR, discussed failure of steam generator hydraulic snubbers at the Trojan Nuclear Plant, Unit 1. The Staff's concern was over stressing of the reactor coolant system piping (See Appendix XV). In February 1985, the Trojan Plant was issued surveillance inspection technical specifications for large bore snubbers. In April, Trojan was shut down for refueling and 16 steam generator snubbers were inspected at that time. Two of the snubbers were tested and both failed.

Based on a management decision by Pacific Gas and Electric, all 16 steam generators snubbers were declared inoperable.

, These failures were attributed ever.tually to restrictive i

acceptance criteria for the control valves. All 16 snubbers were disassembled, inspected and reassembled. They were then retested, found to be acceptable, and placed back in service.

Based upon marks that were evident within the snubbers which irdicated that they had exhibited motion, it was not concluded at that time that any of the snubbers had actually locked up during the 1984-85 cycle. During an April 1985 outage, a hot leg to the Steam Generator B pipe whip restraint to lateral support member was found pulled from the wall about 5/8 of an inch. Since 1982, Trojan had observed erratic pressurizer surge line movement and over the past three years had been monitoring this motion in order to determine its cause. A consultant called in by the Licensee to evaluate the pressurizer line surge movement and was made aware of the fact

Minutes of 314th ACRS Meeting j

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l that two steam generator snubbers did not pass. acceptance  ;

tests. The onsultant concluded that the surge line movement was in fact attributable to the locked snubbers. In addition, it was determined that overstressing under the worst case condition assuming the steam generator snubbers were locked from the cold position at the beginning of the. 85-86 cycle, overstressing of the hot leg elbow to B's steam generator could have occurred.

In April 1986, the 16 steam generator snubbers were reinspect-ed and 11 of the 16 were found not acceptable on functional test acceptance criteria. The determination of failure of the snubbers was attributed to inadequacies in the design of the control valve. Because the regional staff were -concerned about possible of overstressing of the RCS piping, the Staff requested several follow-up actions to assure the soundness of the piping. A UT test was performed on the steam generator l elbow to pipe weld and no indications were found. The Licens-ee, NRR, and the Region ~ walked down the portions of the RCS piping and observed some evidence of restrained thermal growth. UT testing was also performed on all four hot leg elbows with no indications found. The snubber control valves were replaced with one of a new design. As a follow-up action prior to restart for the 1986-87 cycle, the Licensee is to monitor the thermal growth of the reactor coolant system during heat up and during operations to assure that the predictive thermal growth and clearances are all acceptable.

The Licensee will also verify the assumption of the locked snubbers causing the erratic pressurizer surge line movement and the damage which was observed on the pipe whip restraint.

The NRR is also reviewing the Licensee stress and fatigue reports to assure that the integrity of the RCS piping is intact.

D. Okrent cbserved that the reason that these snubbers were tested was because there were snubber technical specifications for the first time. He wondered why this was the first time that this type of failure of hydraulic snubbers was observed in a plant. R. J. Kiessel, NRC, indicated that during the 1970's when the initial technical specifications were issued on hydraulic snubbers, a visual examination of all snubbers was expected, as well as a functional testing of a sample of the snubbers. They also included an exemption for any snubbers that were in difficult or unusual locations. There was also a size limitation such that any snubber over 50 KIPS was not required to be tested. Part of the rationale at that time was that there were not sufficient facilities to perform adequate testing on the large snubbers. In November 1980, NRR issued a generic letter which revised the technical specifica-tions by removing the 50 KIPS limit and modifying the sampling t

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Minutes of 314th ACRS Meeting plans. ' Apparently, the Trojan plant had not resubmitted their technical specifications until late in 1984-85. This was the first time that the snubbers had been tested. He pointed out there have been a number of other instances with large steam generator snubbers that failed to lock up and also continually locked up. Trojan is not the first plant to have encountered the problem. T. Chinn explained that the arrangement of the snubbers and their hydraulic lines at Trojan is such that the design intent is that if one snubber locks up then the flow would not go through the check valves which would activate that particular snubber. The four steam generator snubbers are arranged in a parallel arrangement such that the necessary fluid flow to permit motion could be transferred to the remaining three snubbers. A failure of one snubber would not prevent the necessary motion. J. C. Ebersole asked if there are any generic letters to warn people of this problem. T.

Chinn indicated that NRR and IE are working together to evaluate the generic implications of this occurrence. Trojan is one of very few plants which utilizes this particular control. valve and the control valves have been changed out to a new design at the Trojan plant as they have at many of the other utilities.

D. Single Failure of Miniflow Logic at Pilgrim J. C. Ebcrsolc cxplained that the Pilgrim Nuclear Power Sta-tion still retains a system called loop selection whereby a large size break on one side of the reactor in such a two flow system will result in difficulty delivering enough water for low pressure injection. The solution to the problem appears to be a cross tie with a valve which would send the water from the three pumps towards one but not towards the empty loop.

Such a modification necessitates an extremely fast transient DP gradient and the signal generation system would never have worked. Subsequent findings are that the low pressure system served more than just a spray system as it also constituted an inventory make-up function as well. It was decided to lock out the loop selection logic at that plant. What was found at Pilgrim was that the loop selection logic was still in exis-tence and its presence could cause a potential for certain valves to close that put all four of the RHR pumps against a closed discharge. On a single failure malfunction, one would be without core cooling pumps. The RHR pumps in that config-uration can lock up with zero flow and grind to inoperability permanently. D. Allison indicated that the Staff briefed the subcommittee on the fact that a single failure of the miniflow logic could disable all redundant RHR pumps during the smaller intermediate size break LOCA. J. C. Ebersole suggested that the Staff should pay particular attention to that selection

logic design. He asked how many nuclear plants there are like

Minutes of 314th ACRS Meeting this one. W. Hodges, NRC, indicated that all BWR-3 plants, including Dresden 2 and 3, Millstone 1, Pilgrim, Monticello, Quad Cities 1 and 2, plus two BWR 4 plants are like this one.

Duane Arnold and Fermi 2 also have this selection logic design. J. C. Ebersole wondered why the problem did not propagate to these other plants. W. Hodges indicated that .

those operators were able to demonstrate that they satisfied the regulations with the logic as it exists. They satisfied Appendix K and showed that the loop selection logic selected a proper loop. If it didn't select the proper loop g

for smaller

. breaks, they still didn't go above the 2200 F peak cladding temperature. C. Michelson suggested that the logic seems to work fine on paper, but it is in the hydraulics of the actual operation where the difficulty arose. The system has such high hydraulic noise that at any point in time it cannot detect on which side the break is. J. C. Ebersole suggested that the staff consider changing out this flaw for all other susceptible plants. The Committee discussed the possibility of requesting Staff action in an ACRS letter. E. Jordan explained that the Staff has notified all plants that are susceptible requesting _ that they assess the single failure vulnerability and submit a plan of action to the NRC.

Affected licensees have fonned an owners group to study the situation and derive a solution.

IV. Source Term for Nuclear Power Plant Accidents (0 pen)

[ Note: M. D. Houston was the Designated Federal Official for this portion of the meeting.]

W. Kerr mentioned the previous Committee letter in December 1985 on the draft report entitled, Reassessment of the Technical Basis for Estimating Source Terms, NUREG-0956. The letter listed a number of comments which the Staff has addressed as part of its review of public comrents. Significant changes have been made in the format i of the report. He indicated that the Subcommittee met on June 3, 1986, to discuss what was called " Review Copy of NUREG-0956" dated May 23, 1986. It was the consensus of the Subcommittee that significant improvements have been made in the report.

. M. Silberberg, NRC, discussed the state of progress on source term technology regarding the implementation of the Severe Accident Policy Statement. He discussed major changes made to NUREG-0956 (See Appendix XVII). The most important change was to add a considerable amount of redundant information directly into the i report to deal with the technical basis for the source term.

Considerable time has been spent describing the upgraded computer code package to augment the presentation in the draft report. The report mentions only some of the features of the code suite. Also presented are analyses of additional sequences performed with the

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f source term code package for the NUREG-1150 (Risk Perspectives Rebaselining) analysis. The chapter _on risk and contair. ment - has been removed to an appendix in response to _ the_ public coninents. A-chapter has _been added to reflect revision to the severe accident research plan in NUREG-0900. There is also an improved statement of conclusions. He stated that the Staff believes that major

advances have been. made in this technology since WASH-1400 (the Reactor Safety Study), particularly in the last five years. An important part of this progress is the development of an. analytical approach to source term estimation, the source term code package.-

He briefly discussed several areas of '_ improvement (see Appendix XVII). D. W. Moeller noted that~.a number of the major advances are also on the list of. areas of needed research. M. Silberberg did

not believe this unusual since there are still some gaps left in L the process of gaining a deeper _ understanding of severe accident'

phenomena and -understanding the true uncertainties present.

1 Further improvements can be expected from the research that is now in place. Nevertheless, he indicated that the Staff believes that it now has a sound technical basis from.which to move forward to use the new source information to. reevaluate regulatory practice as t

mandated by the severe accident policy. The process of using the

, new source term information is already in progress in the l NUREG-1150 risk rebaselining study now being concluded. The second application will deal. with the implementation plan for the Severe Accident Policy Statement and the regulatory use of new source term information. NUREG-0956 is an important element of the process of

[ moving forward in examining current regulatory practice with l respect to source terms.

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J. Mitchell, NRC, discussed specific changes to NUREG-0956 base 1 upon ACRS comments and recommendations in a December 12, 1985 ACRS report to the Cennission. She indicated that the ACRS asked whether the Staff finds a significant difference between severe ,

accident research program results to date and the Reactor Safety Study. She indicated that the final NUREG-0956 is less ambiaucus ,

than was the draft of the report. It now states that the Staff believes that there are not large systematic reductions. She mentioned that the ACRS questioned whether the selected accident sequences for the five reference plants provided sufficient tests of the capabilities of the computer codes. She indicated that while the Staff has shown that the codes execute and give physically reasonable results, there was now a program for validation of the codes and models. The Staff will compare the results with existing experiments and future research experiments.

D. W. Moeller asked if the South Texas Project is ;one of the five I referenced plants. J. Mitchell indicated that the five referenced plants are Surry, Peach Bottom, Sequoyah, Zion and Grand Gulf.

These are the plants discussed in NUREG-0956. After its first publication as a draft, material was added to NUREG-1150 on La Salle, a BWR Mark II. L. Soffer, NPR, indicated that one of the

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earliest applications of the source term code package and the revised methodology was a relook at accident risks in the South Texas Project where there was a discussion using the source term from WASH-1400 and the revised methodology so that the results could be compared side by side. D. W. Moeller indicated that he had found this comparison very helpful in terms of the ACRS review.

W. Kerr cautioned care in interpreting the results since it is not a safety document. -

J. Mitchell noted the ACRS' desire to know what effect each of the major improvements being made would have on the source term. She indicated that while this appears to be a reasonable request, it is not very easy to satisfy. The Reactor Safety Study methodology was used as a set of small stand alone codes that were woven together to provide results. The Staff believes that it is not really practical to look at the advances one by one compared with the old methodology. For those sequences that are comparable, the Staff has provided a comparison with a bottom line. That includes the effects of all of the advances including the framework. The Staff is evaluating the effect of the iodine chemical form assumption with a forward-looking chemistry package rather than looking backward at what the Reactor Safety Study has said.

In general, J. Mitchell agreed with most of what was said in the ACRS report. J. C. Mark referred to a quotation in chapter 4 of NUREG-0956 which refers to "during the multiple hydrogen burns" in a number of places. He indicated that he could not find any description of the assumptions made for a hydrogen burn. He asked what kind of a burn was assumed and under what conditions did it occur. J. Mitchell indicated that the assumptions in most of the cases were that the hydrogen concentration should be 8 percent. J.

C. Mark indicated that that appears reasonable as it comes straight from the- TMI 2 experience. J. Mitchell indicated that the steam moisture content should be below about 55 percent. In some cases, there might be steam inerting. J. C. Mark asked if the hydrogen is I coming from metal-water reaction or from core-concrete interactions. J. Mitchell indicated that it depends on the time of the accident. In vessel it derives from zirconium oxidation. J.

C. Mark indicated that in that case one gets pure hydrogen with steam. In concrete, one gets as much water and carbon dioxide as hydrogen plus carbon monoxide. J. C. Mark and J. Mitchell dis-

, cussed the course of a hydrogen burn accident including airborne I fission products and released fission products as a function of time.

V. Briefing Regarding IAEA Meeting on Chernobyl (0 pen) l l [ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.]

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H.. Denton, Director, NRR, reported on his briefing by Soviet i

nuclear experts at the May 21, 1986 meeting of the International Atomic Energy Agency's Board of. Governors. He-indicated that it appears that the Soviets will only participate in an IAEA format and will not engage in bilateral discussions with the U. S. .Some time during July, H. Denton said that Soviet Union will report to the IAEA on the causes of the accident in a post-accident review meeting. Another IAEA scheduled meeting is intended to produce binding international early warning and coordination agreements.

The objective will be to get all nations to sign such a binding agreement. H. Denton also indicated that the IAEA plans to assem-4 ble experts from around the world to propose ways to consider

additional safety features or measures to improve the safety of all nuclear plants. Finally, a conference of governments will meet to consider binding agreements from the three previous meetings and consider the expert's recommendations. He spoke of a unified U. S.

approach to Chernobyl through the National Science Advisory Board.

i He suggested that the ACRS should consider developing a factual

! report of what happened at Chernobyl.

. H. Denton indicated that the chief Soviet spokesman at the May 21 meeting was Boris Siminoff who described a postulated event se-

quence which began with an intensive evaporation.of cooling water,

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an overpressurization, and a hydrogen explosion. B. Siminoff claimed that the accident occurred in the reactor core as a result

of a sudden power surge from 7 percent to over 50 percent power. ,

! There was a thermally disruptive metal-water reaction, a hydrogen i leak and hydrogen release. A fire started. Based on what he learned at that meeting, H. Denton said that the first priority of the Soviet fire fighters at Chernobyl was to prevent the fire from spreading to the adjacent Unit No. 3. The Soviets are very con-cerned with water contamination and are now concentrating on keeping the apparently molten core from penetrating the suppression pool basemat by pumping concrete into the pool cavity. Their aim is to " entomb the reactor" while providing some internal cooling.

H. Denton explained that the IAEA meeting intended to discuss early notification of trans-boundary releases with the objective of signing a binding agreement will go into emergency response and discuss a strengthening of the incident reporting system. The IAEA may ask the United Nations organizations, WHO and UNSCEAR to review world dose contamination as a result of the Chernobyl event. H.

Denton indicated that he expects an increase in NRC resources allocated to the study of the Chernobyl event. He noted that little has been learned of a technical nature about Chernobyl since NRR last briefed the Committee in May.

W. Kerr asked if H. Denton had any additional comments regarding the event's scenario. H. Denton indicated that it appeared to have begun on the morning of April 26 and was not a slowly developing i

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I Minutes of 314th ACRS Meeting accident. He was not sure if there.was a positive void coefficient involved. The event was characterized by overpressurized pressure tubes and a rupture of. pressure tubes with the release of steam and disruption in the upper part of the core. A metal-water reaction i ensued from water which was released from the pressure tubes with a subsequent release of hydrogen which seeped into the graphite moderator. A fire ensued. D. W. Moeller asked what would have happened had the Chernobyl reactor had a large dry containment. H.

. Denton indicated that he did not know the nature of the pressures generated during the event and could not make any coment on the impact of containment. It was noted that it-is possible that the Soviets were_ doing some manipulations of an experimental nature at the time of the event. He also indicated that 300 individuals who were reported hospitalized were employees of the facility. No residents were involved, according to the Soviets.

. H. Denton mentioned international reaction to the Chernobyl event.

, The filtered-vented containment is now a principal issue in Sweden.

1 There is a also a sense of urgency to move on the IAEA meeting.'

The KWU Plant built in Austria may be scrapped in reaction to '

Chernobyl. H. Denton indicated that the U. S. must organize its

efforts by incorporating DOE and FEMA and participating through the IAEA process. He speculated that the states in the U. S. will want emergency planning regulations reexamined. IDCOR was asked by the Staff to speed up its efforts to examine containment performance to mitigate severe core damage without containment failure. The NRC intends to develop the outline of a report on Chernobyl and present it to the ACRS. A delegation of five individuals from the U. S.

will participate in the IAEA meeting when the Soviet report is given.

VI. Meeting with NRC Comissioners (0 pen)

[ Note: Comissioners present were: N. J. Palladino, Chairman; T.

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M. Roberts; J. K. Asselstine; F. M. Bernthal and L. W. Zech, Jr.]

Chairman Palladino indicated that the Comission intends to discuss ACRS views on the GESSAR II BWR/6 nuclear island design for future plants per the Comittee's January 14, 1986 report to the Comis-sion. The ACRS report indicates that although the GESSAR II design has improved safety features, there are questions whether the design satisfactorily addresses all concerns in the NRC's severe accident policy. He indicated that the concerns expressed by the ACRS bear on the question regarding the Comission's role with respect to ultimate approval of standard designs such as GESSAR II.

F. J. Remick discussed the ACRS findings and recomendations in its i January la letter. He indicated that the Committee believes that the GESSAR II design includes features that have the potential to

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provide a significant improvement in safety over current BWR designs. If this were an application for a construction permit for one or more plants of this design, the Comittee would not hesitate to recommend its approval. The Comittee, however, was unable to agree with the Staff for various reasons that the design satisfac-torily or completely addresses all the concerns described in the Comission's Severe Accident Policy Statement. The Comittee did approve the GESSAR II design provided that it was for a limited time such as five- years and provided that the Staff's review procedure not be viewed as a precedent for handling of future applications. In particular, the information provided to the ACRS in connection with GESSAR II would not be sufficient to support an application for a one step licensing process. He mentioned receipt by the Committee of NUREG-0979 Supplement 5, the fifth supplement to the GESSAR II FDA, but indicated that the Comittee had not yet had time to fully analyze it.

M. W. Carbon spoke about his additional coment appended to the January 14 letter. He indicated that he and C. J. Wylie believe that the GESSAR II design represents a definite improvement in safety over BWR designs that have been approved in the past and that the applicant has met all of the NRC requirements. He noted that many items are still open and considerable review will yet take place. Nevertheless, he indicated that he and C. J. Wylie support the Staff's plan to issue an FDA applicable to one step licensing. He indicated that he was personally not totally happy with GESSAR II as the design for a standard plant, but was en-couraged by the kinds of improvement that were made in this appli-cation. He thought a long term standard plant design which might result in the construction of many plants ought to be handled in a somewhat different fashion. Chairman Palladino asked for further clarification of that remark. M. W. Carbon indicated that the NRC should cue vendors who might submit such a potential standard plant design of some of the features that would be highly desirable in such a design. He did not think that it would be adequate for the Commission to wait until the vendors bring the design to the NRC.

It would be best to enter the process at an early phase so that the vendors are aware of some of the features the Comission thinks would be desirable in future standard plants. In answer to a question by Comissioner Asselstine, M. W. Carbon indicated that his willingness to sign off on the Staff's review was in part because of the fact that very few of these plants will ever be built. Comissioner Asselstine asked for examples of principal areas of improvement over BWR/6s built into the GESSAR design. M.

W. Carbon cited the ultimate plant protection system (UPPS) which he thought a definite step in the right direction.

Comissioner Bernthal thought it interesting that the same argument has come up in the past, that the Comission get involved in the early stages of review of what the industry and the vendors are

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• flinutes of' 314th ACRS Meeting developing in the way of advanced reactors. Chairman Palladino

, thought that it best to have essentially a complete design when the NRC begins its review. He asked if the UPPS is essentially com-plete. C. J. Wylie indicated that the UPPS is a design concept which does not have hardware and criteria associated with it as do some other parts of the plant. But the Staff has made a provision that it would be reviewed in the course of a construction permit application.

D. Okrent prescnted several questions he thought should be explored as the Commission reviews the GESSAR II matter.

What is an FDA? What comitment is the NRC making when it is-sues one? What commitments would it be making if it approves GESSAR II? How much detailed information should be provided by an applicant for an FDA?

What should be the level and depth of the PRA? Should it treat uncertainties as well as the state-of-the-art will per-mit? How should interface requirements with the balance of plant be specified in view of the fact that the PRA makes assumptions on the performance of the balance of plant?

What seismic fragility requirements should be established by the GESSAR PRA and by the Staff review?

What performance requirements for GESSAR II systems are estab-lished by the PRA, if any? What level of PRA evaluation and review is required of the NRC Staff for it to accept an FDA?

Should the Staff make use of mean or mode values which are evaluated to the state of the art?

What should be the quantitative safety objectives for a future plant?

Should there be some kind of containment performance criterion for a future plant?

How will a future plant design deal with a terrorist threat and sabotage?

How does one deal with cost benefit analysis for possible design improvements? How does one insure defense-in-depth at the same time?

How does one ensure that the frequency of challenges to safety systems is acceptable? -

D. Okrent explained that his' additional remarks at the end of the ACRS report did not imply a disagreement with the letter except

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Minutes of 314th ACRS Peeting that it would be better- to call it an interim letter to issue a  ;

limited FDA. He noted his original intent was to write an interim '

letter and discuss the matter with.the Commission. He also noted that the Staff and its consultants evaluated the seismic design and seismic PRA for GESSAR II and contended that it was inadequately designed for purposes of a full review. Missing were the seismic contribution to core melt, as well as a discussion of the elimina-tion of relay chatter. .The UPPS was really ill-defined. It has no seismic capability as proposed, and if seismic events turned out to be a major contributor, it would not! have alleviated this aspect of the plant risk. He mentioned his concerns regarding sabotage pro-tection and the inability of the Staff to require anything of GESSAR II beyond what is already in the existing regulations regarding access control and fences. He noted that the control room appears to be vulnerable to the kinds of terrorism that has caused severe damage in the Middle East. He also questioned the ability of the drywell to maintain its integrity should a core melt occur and much of the core got through the vessel to the concrete basemat. The concrete would heat up the vessel pedestal, and if the sacrificial shield failed, the vessel could tip over pulling the piping, such as steam lines, from their penetrations. He noted that the NRC Staff was not concerned by this risk and that Brookhaven National Laboratory thcught that this would be a case of late containment failure with a small radioactive release.

D. Okrent indicated that he was generally concerned regarding the quality of the NRC Staff review. Commissioner Bernthal wondered what this catastrophic core melt scenario described by D. Okrent would involve. The Coninission and the ACRS discussed various aspects of such a catastrophic core melt which might involve rupture of the drywell and bypassing of the suppression pcol. D.

Okrent suggested that the result. might be a small radioactive release or the release might be quite significant. What is im-portant is that the Staff should analyze the scenario before an FDA is issued or enough is known abcut the scenario to rule it out as a viable possibility. He stressed that in his own opinion future plants including GESSAR II should include the features that are included in his added remarks to the January 14 report. These features should include independent decay heat removal systems, as well as features for sabotage protection. Future plants should be surrounded by a containment designed especially for core nelt acci-dents.

J. C. Ebersole suggested that the practicality of the boiling water reactor gave it the potential even as far back as 1968 of being a potential workhorse power plant. Nevertheless, the boiler still '

has a completely inadequate reactivity control system with hydrau- ,

lic drives and a multiplicity of valves and complications which have led to a record of less than optimal performance. Perhaps relatively modest enhancements to the GESSAR design, such as a

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Minutes of 314th ACRS Meeting complete description of the UPPS, might be all .that is necessary.

l He agreed that the seismic area, the area of sabotage protection, and protection from fire, in addition to modifications- in the reactivity control system, are ' issues that need to be addressed.

Comissioner Asselstine spoke in favor of a high-quality design that is a near perfect plant that will not require' constant modi-fications and the constant addition of complexities. ..W Kerr i

suggested that his experience in the design of large. industrial systems is that they are subject to considerable surprises. He indicated that he was more comfortable with an evolutionary approach to changes with a goal in mind rather than pursuing a complete, perfect design. W. Kerr also expressed some nisgivings regarding the process of NRC review which was primarily associated with the use of the PRA in arriving at decisions about severe accidents. He did not think this review should represent a  !

complete review but hoped that the process could be -improved, i

Commissioner Bernthal noted that since there does not appear to be a rush to build one of these GESSAR-II plants, he thought it might be a good idea to insist that the Staff review be more thorough.

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1 D. A. Ward agreed that the ACRS thinks that the Staff has come up

! short in defining the GESSAR-II design as an appropriate standard design.

I Chairman Palladino thought it unfair to hear only one side of the issue. He thought that the Comission ought to hear from the Staff

, regarding the process and their conclusions. Then the Comission i should decide for itself what its role is in the FDA process. It has never been defined. Comissioner Bernthal speculated on what-would happen to the FDA assuming that the Comission approved it.

He wondered what GE intends to do with the FDA. Comissioner Asselstine did not believe that GE would go to licensing hearings in the U. S. He speculated that GE might be helped in the export I market.

H. Denton indicated that the Comission approved the policy state-ment that provided certain provisions for the CESSAR and GESSAR plants which had been under review for some time. The Staff l attempted to review those applications against the standards l established and found that they met the regulations. He noted that ,

i he could not speak to the use GE plans to make of the FDA. He I thought it best that GE address that question. He noted that GE has also under development an advanced boiling water reactor which he presumed the Comission would want the Staff to review.

Chairman Palladino took note that the Staff reviewed the GESSAR-II application under Comission direction and indicated that it is

' incumbent upon the NRC to process proposals such as this one from industry. It is the job of the Comission to assure that it is a i good, safe design that meets the regulations and the objectives set forth by the Comission. D. Okrent pointed out that there is a difference of opinion between the NRC Staff and the majority on the i

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ACRS as to whether in fact the Comission severe accident policy was properly implemented in this review. C. P. Siess noted that the GESSAR-II design is a future plant in the sense that if one is built it will be built in the future but it is not a future design.

He suggested that the health and safety of the public would not be endangered if one or two of these plants were built. The ACRS main concern was with the review process.

D. A. Ward took note of the retirement of N. J. Palladino as Chair-man of the NRC and congratulat'ed him on his record and accomplish-ments in the last five years. He pointed out that his experience and judgment would be missed by his fellow Comissioners and the Staff.

VII. Executive Sessions (0 pen)

[ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.]

A. Subcomittee Assignments

1. ACRS Review of the Hanford N-Reactor /Chernobyl Reactor A Subcomittee consisting of D. A. Ward, D. W. Moeller, W. Kerr, and F. J. Remick as Chairman was directed to become familiar with the design and the course of the recent accident at the Chernobyl Nuclear Plant. This Subcomittee will act in an information gathering capac-ity to follow the course of ongoing studies including the DOE Special Comittee and the National Academy of Sci-ences review of this matter. The Safety Philosophy, Technology, and Criteria Subcomittee (D. Okrent, Chair-man) was asked to consider the implications of the Chernobyl accident to reactor safety in the United States.

2. Membership on the ACRS Management Comittee A proposal by H. W. Lewis regarding the composition o the Management Comittee was deferred until the August meeting when the Procedures and Administration Subcom-mittee will consider the question of a two-year term for the Chairman as well as the membership of the Management Comittee. Procedures for selection of the ACRS Chairman should also be discussed at this time.

3. Subcommittee Assignments Distribution of advanced water reactor reviews between the Advanced Water and Advanced hon-water Reactor

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flinutes .of 314th ACRS lieeting Subcomittees needs further discussion by the ACRS Chairman and the Subcomittee Chairmen (M. W. Carbon and C.J.Wylie).

It was agreed that the review of Improved Technical Specificaticns will be handled by the Subcomittee on Plant Operating Procedures (C. Michelson) rather than the Subcomittee on Operating Reactors (J. C. Ebersole).

4. Report of the ACRS Management Comittee The Chairman reported on the June 4, 1986 meeting of the ACRS Management Comittee (Note: Several of the specific recomendations/ decisions are reported elsewhere in this list. The remainder are as follows:

Members were asked to provide comments to the ACRS Executive Director regarding the Status Report on Implementation of the Recomendations of the Panel on ACRS Effectiveness (distributed along with the June 4,1986 meeting summary). Members were asked to devote particular attention to the status / action proposed for implementation of the recomendations The Chairman noted that several specific requests for reassignment of specific tasks and additional resources by ACRS chairmen and subcomittees were considered and have been addressed in memoranda to the proposers (see specific list in the sumary of the June 4, 1986 Management Comittee meeting).

These items were dealt with as follows:

(1) Steam generator overfill should be assigned to the Safety Philosophy, Technology, and Criteria Subcomittee (D0). Reassignment was proposed by the Management Committee provided a real risk frcm this event can be shown to exist. C.

P. Siess and P. G. Shewmon should be added to the subcomittee for this review i

(2) Themal hydraulic bases for E0Fs to the Plant Operating Procedures Subcommittee. It was agreed that this should be reassigned l (3) Functioning of isolation valves under accident I loadings to Subcommittee on Reliability Assur-ance. It was agreed that this should be

reassigned provided a problem can be shown to exist l

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(4) Lack of adequate E0Ps for severe accidents (e.g., should cooling water be turned on or not after the core is molten?) to. the Class 9 Subcomittee. The Management Comittee agreed to this reassignment Proposed addition of a subcommittee per discussion during the 313th ACRS meeting - For example, D. Okrent's sugges-tion for a Containment Subcomittee - D. Okrent will be queried by M. W. Libarkin regarding the specific tasks for such a group since a Subcomittee on Containment Requirements already exists Assignment of additional resources to generic subcomit-tees as requested by Subcomittee Chairman:

(1) Waste Management (DWM) request for 59 person-days vs. 40. It was agreed that this reassignment should not be made at this time (2) Severe Accidents (WK) request for 4 subcomittee meetings vs. 2 allocated. This request will be revisited in 6 months (3) G. A. Reed regarding a 1-day meeting for IE programs and 3-4 subcomittee meetings for WAPR review. One subcomittee day will be allocated for review of IE programs. This should include discussion of IE's proposed use of PRA to identify important areas for attention. The WAPR review now has 1 subcommittee meeting assigned. Remaining meetings should be deferred until firm schedules from the NRC Staff are available.

(4) C. Michelson request for Auxiliary Systems review (2 Subcommittee meetings) of the Fire Protection Provisions in nuclear plants. T. G. McCreless was directed to have an ACRS Fellow examine the situa-tion regarding fire protection including the work of Dr. Apostolakus.

The numerical ranking of ACRS reports considered during 1986 by 10 ACRS members produced the following results:

(1) The highest ranking was 2.8, the lowest was 0.6, and the average is 1.7 (2) Generally, ACRS reports with high ratings (above 1.5) have been sent although this is not true in all cases as noted below I

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(3) Highly rated reports (State of Nuclear Safety and Safety Considerations in Future Reactors) have not been sent, primarily because of difficulty in preparing them. A means should be considered to reactivate these reports (4) Reports on waste management, although sent by the Comittee in most cases, generally had lower rank-ings (1.5-1.8 range)

(5) Reports with rankings below 1.0 were not sent. Only one report with a ranking below 1.4 was sent (6) The move to Bethesda got very low rankings [1.0 (sent) and 0.7 (not sent)]

Monideep De provided a more statistically meaningful evalua-tion regarding the means and standard deviations _of member ratings.

The members did not agree that difficult reports such as the reports on the state of Nuclear Safety and Safety considera-tions in Future Reactors should be reactivated as a result of this poll.

B. Reports, Letters, and Memoranda

1. ACRS Coments on the NRC Safety Research Program and Budget for Fiscal Year 1988 The Comittee prepared a report to the Comissioners on its review of the proposed program and associated budget for the NRC Safety Research Program for Fiscal Year 1988.

2. ACRS Report on South Texas Project, Units 1 and 2 The Comittee prepared a report to the Comissioners of its review of the application of Houston Lighting and Power Company (HL&P, the Applicant), acting on behalf of '

itself and as agent for the City Public Service Board of San Antonio, Central Power and Light Company, and the City of Austin, for a license to operate the South Texas Project, Units 1 and 2.

3. ACRS Coments on NUREG-0956, " Reassessment of the Techni-cal Bases for Estimating Source Terms -- Review CoDy/

Final The Comittee prepared a report to the Comissioners of its review of the final version of NUREG-0956, " Reassess-ment of the Technical Bases for Estimating Source Terms."

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Minutes of 314th ACRS Meeting 4. ACRS Recomendations on Hope Creek The Comittee prepared a report to the Comissioners noting its concern regarding proposed resolution of recommendations made in its December 18, 1984 Hope Creek Operating License report concerning a structured turbine over-speed test program and habitability requirements for the Hope Creek control room.

5. Proposed NRC Policy Statement on Standardized Nuclear Power Plants The Comittee authorized a memorandum to H. J. Palladino, NRC Chairman, (Attention: N. Haller, Executive Assistant) from the ACRS Executive Director regarding the pending Comission action on publication of a proposed NRC Policy Statement on Standardized Nuclear Plants. (Ref. . V.

Stello, EDO menorandum of May 14, 1986 to the NRC Comis-sioners regarding Standardized Nuclear Power Plants.)

6. ACRS Status H. W. Lewis proposed a letter to the Comissioners regarding changes by the Comissioners in a cable invit-ing representatives of the Soviet Union to the Wingspread international meeting on reactor safety. The ACRS decided not to send the letter.

C. Future Agenda

1. Future Agenda The Comittee agreed on tentative agenda items for the 315th ACRS meeting, July 10-12, 1986 (see Appendix II).

2. Future Subcommittee Meetings A schedule of future subcomittee activities was distributed to members (see Appendix III).

D. B&W Program on Trip Reduction F. J. Remick explained that Duke Power Company plans to set up a review group with four outside members as a part of the B&W Owners Group Stop Trip Program (curtail unnecessary trips) ano has asked for his participation on the review group. Several

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c Minutes of 314th ACRS Meeting members pointed out the conflict-of-interest possibilities involved when any B&W plant is reviewed by the ACRS. The consensus of the Comittee was against participation by F. J.

Remick and he agreed to decline the invitation.

E. Memorandum of Understanding with the EDO It was noted by the Comittee that on occasions when the NRC Staff comes before the ACRS with a preliminary position on a proposed NRC rule or policy statement and the ACRS declines to comment, the NRC Staff has bypassed the Comittee during later review stages. It was decided that subcomittee chairmen should set out a course of action for anticipated ACRS review.

The ACRS Executive Director, R. F. Fraley, was directed to send a memorandum to the ED0 regarding this proposed course of action indicating ACRS interest in later review -of the preliminary Staff position. Copies of the memorandum should be sent to the cognizant Staff members at the branch level and project level.

F. Proposed Amendment to ACRS Bylaws The Comittee approved a change in the ACRS Bylaws which provides a mechanism for individual members to express their personal views in meetings with individual Commissioners. The ACRS Executive Director was directed to inform the Comissioners of the change in such a way as to allow an opportunity for coment as to whether this change is responsive to their needs.

The Comittee approved a proposed Bylaw change regarding the authority of the ACRS Chairman to hold up a completed ACRS report if he discovers that it contains a serious error or misstatement which was not evident during its preparation.

The Comittee approved guidelines (as a change to the Bylaws) regarding preparation of added (minority) coments to ACRS reports. Changes proposed by H. W. Lewis which were part of the draft made available during the 314th meeting are to be incorporated. Any coments regarding these changes should be directed to T. G. McCreless as soon as possible.

G. Conduct of Members The Comittee discussed the recent coments by P. G. Shewmon to Dr. M. B. McNeill, NRC/RES regarding the NRC research program at Battelle, Columbus, and suggested that it would be most appropriate to preface such comments by individual members with a standard disclaimer. This disclaimer should stress that the coments are the member's own personal views

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-O liinutes of 314th ACRS Meeting and do not necessarily represent a position of the Comittee.

If the intention is to interpret a Committee position, a member should first seek guidance from the full Comittee.

H. Testimony by H. W. Lewis H. W. Lewis noted his intent to testify during hearings of the House Subcommittee _on Energy and the_ Environment regarding the l viability of the nuclear power program in the United . States.

His coments will represent his own personal opinions regarding subjects to be addressed rather than the opinion of the Comittee. The Comittee agreed to support this effort logistica11y.

I. Implications of Chernobyl A letter from the State Department was circulated during the meeting _which stated that certain specific infonnation regarding the Chernobyl accident is to be considered classified. R. F. Fraley was asked to determine the authority under which such matters shculd be 6;clared classified.

J. Report of the Nominating Panel The ACRS panel regarding nomination of candidates for appointment to the ACRS produced a list of six potential candidates for final consideration by the Comittee. An additional name was added to the list and two candidates are to be invited to the 315th ACRS meeting (July 1986).

K. Reappointment of ACRS Member The Ad Hoc Subcomittee set up to consider the reappointment of F. J. Remick whose term ends in September 1986 recorrended the reappointment of F. J. Remick to the Committee. The ACRS endorsed the Ad Hoc Subcommittee's recomendation and the ACRS Executive Director was directed to inform the Comissioners of i

the Comittee's recomendation.

l L. Agenda for the Wingspread International Meeting i A proposed agenda and outline for this reeting was l distributed. Arrangements for the meeting were discussed.

Members who plan to bring their wives were asked to advise the ACRS Office (T. G. McCreless).

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o Minutes of 314th ACRS Meeting 11 . Member Complaint Regarding Lack of Documentation in Safety Evaluation Reports G. A. Reed expressed his unhappiness with the lack of plant layout and system schematic drawings in the Staff's SER for the South Texas Project, Units 1 and 2 which made the OL review of that facility more difficult. Since the South Texas Project appears to be the last OL review to come before the Committee for some time, it was suggested that, for future ,

plant reviews, ACRS Staff engineers prepare a package of drawings from the applicant's Safety Analysis Report prior to Committee consideration of a plant. G. A. Reed was asked to provide a list of the drawings in which he is interested.

N. Topics for the 315th ACRS Meeting The members agreed to devote considerable time (3-6 hrs) during the 315th (July) ACRS meeting to the consideration of the proposed NRC Policy Statement on Standardized Nuclear Plants. D. A. Ward and C. J. Wylie were directed to work out a detailed agenda for this meeting. D. Okrent suggested that topics which have been set aside (e.g., General Design Criteria update) should be considered as well as those issues included in the proposed policy statement.

The 314th ACRF neeting was adjourned at 1:20 P.M., Saturday, June 7, 1986.

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$ APPENDIXES T0 f MINUTES OF THE 314TH ACRS MEETING

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APPENDIX I I NRC ATTENDEES AT 314TH ACRS MEETING NRC ATTE 4

O Thursday, June 5, 1986 314TH ACRS OFFICE OF NUCLEAR REACTOR REGULATION l S. Long B. Mann i

R. Hernan -

J. E. Rosenthal K.S. West J. L. Milhoan '

R. P. Goel D. Choppa R. Caruso .

D. Scaletti R. Martin l N. Kadambi 0FFICE OF NUCLEAR REGULATORY RESEARCH i A. Datta REGION IV l

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Thursday, June 5, 1986 WESTINGHOUSE ELEC.TRIC CORP. .0ECHTEL M' DER COR'd.

B. S. Monty D H. Ashton '

G.E. Lang 5. N. Letouirneau B. D. Losen. J. K. Atwell W. J. Johnson J. Litehiser W.R.Spezialetti A. Zaccaris' ,

A.G. Dai '

K. P. Slaby '-

.s F. J. Twogood M. J. Hitchler , NEWMAN&' HOLT 7]NSER M. Beaumont ' -

, S. Goldberg <

HOUSTON POWER & LIGHT COMPANY A H..Ge.tterman M. E. Powell J. W. Bailey J. M. Dew R. L. Balean E. Dotson _

O M. R.Wisenburg J.E. Geiger I.Crawford, III E.A. Alexander R. A. Frazar J. R. Pendland J. Newn J. H. Goldberg M. A. McBurnett D. Cody W. Kinsey J.G.Dewease A. O. Hill R.C. Munter ,

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Friday, June 6, 1986

_J. Trotter, NUS Corp.

E. Fotopoulos, SERCH Licensing, Bechtel L. Conner, DDA J. Nurmi, Qatel ,

'J. Kuemin, Consumers Power Co.

G. A.Zimmerman, Portland General Electric Co.

C.R. Klee,Bechtel Power J. A.Gieseke, Battelle ~

R.S. Denning, Battelle D. Runkle, Morgan Assoc.

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APPENDIX II FUTURE AGENDA n

APPENDIX A FUTURE AGENDA JULY MEETING Cavis-Besse Nuclear Plant -- Subcommittee report concerning i hr request from Commissioner Asselstine regarding the regulatory processes associated with review, approval, and operation of this plant and an auxiliary feedwater system that does not meet current safety standards TVA Reorganization -- Discuss TVA reorganization to 3 hrs resolve QA, design, construction, etc., problems at TVA plants B&W Water Reactors -- Status report regarding review of 21 hrs long range safety of B&W Reactors with briefings by '

representatives of the NRC Staff and the B&W Owners Group l (

l \ <

Davis-Besse Nuclear Plant -- Proposed restart of this 3 hrs plant following loss of feedwater incident on June 9, 1985 NRC Regulator Guides -- ACRS coments regarding Regulatory i hr Guide revisions for Regulatory Guide 1.35, Rev. 3, ISI of Ungrouted Tendons in Prestressed Concrete Containments and Regulatory Guide 1.35.1, Determining Prestressing Forces for Inspection of Prestressed Concrete Contain-ments EPRI Recuirements for Standardized LWRs -- Briefing by li hrs EPRI anc NRC Staff representatives regarding status of activities to develop requirements for standardized LWRs l

Design, Maintenance and Testing of Safety-Related Check i hr Valves - Briefing by IE Staff Auxiliary Systems -- Report of ACRS Subcommittee regarding I hr provisions for fire protection in nuclear power plants Proposed NRC Policy Statement on Standardized Nuclear 6 hrs Power Plants -- ACRS comments on proposed NRC policy statement and proposed NUREG to identify topics that are important to the implementation of this policy Meeting with Director, NMSS -- Discuss matters of I hr mutual interest

i i

b Control Room Habitability Improvement Effort -- Discuss I hr related NRC program and ACRS coments as appropriate Staff Reviews of Chilled Water Systems -- Discuss related I hr NRC program and ACRS coments as appropriate Reactivation of Deferred and Cancelled Nuclear Plants -- li hrs Briefing by representatives of the NRC Staff regarding factors to be considered Proposed NRC Policy Statement on Technical Specifications -- li hrs ACRS comments are requested

- New Members -- Discussion regarding nominations of 3/4 hr candidates to fill 'the vacancy on the ACRS

~

NRC Long Range Plan -- Discuss proposed outline for deferred preparation of a long range plan to .

August Restart of San Onofre Nuclear Plant, Unit 1 -- Review of

• deferred corrective action at San Onofre Unit 1 to correct check to valve problems and resulting water hamer August Containment Performance Design Objective -- ACRS coments deferred to August /

September Fitness for Duty Requirments --Status report by NRC Staff deferred to August Safety System Functional Inspections -- Status report by deferred

. NRC Staff l

I 0

. . - _ _ . ~ _ - _ _ . .. - . - - . ._ .- . - .

APPENDIX III i ACRS SUBCOMMITTEE MEETINGS j ACRS SUBCOMIT1 O Ad Hoc Subcomittee on TVA, June 12, 1986 and 13, 1986, Chattanooga,TN(Savio),

8:30 A.M. The June 12 discussion will be held in TVA's Chattanooga Office Complex, Tennesee River Room,1101 Market Street, Chattanooga, TN, and the June 13 discussion will be held at the Sequoyah Nuclear Plant, Managers Conference l Room, Daisy, TN. The Subcomittee will discuss TVA reorganization and related i technical and management issues. Attendance by the following is anticipated, and 4

reservations have been at the Holiday Inn Downtown, 401 West Martin Luther King i Blvd., Chattanooga, TN for the nights of June 11 and 12:

Mr. Wylie Mr. Ward i Mr. Ebersole Mr. Hagedorn Mr. Michelson Mr. Barton j

, Mr. Reed '

Long Range Plan for NRC (CLOSED), June 17, 1986, 1717 H Street, NW, Washington, DC (Major), 8:30 A.M., Room 1046. The Subcommittee will review the proposed EllC Five Year Plan and prepare to address Comittee comments to the Comission.

Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of June 16:

. Dr. Carbon STATE PLAZA Mr. Wylie (tent.) NONE l

Dr. Remick NONE f Babcock and Wilcox (B&W) Reacter Plants, June 25, 1986, 1717 H Street, NW, i Washington, DC (Major), 8:30 A.M. Room 1046. The Subcomittee will consider the B&W Owners Group plans to reassess the long-term safety of B&W reactors, includ-ing the implications of operating experience on the adequacy of B&W plant de-signs. The Subcommittee will also be briefed on the NRC Staff's Incident Inves-tigation Team's (IIT) findings related to the 12/26/85 loss of integrated control system power and overcooling transient at the Rancho Seco nuclear power. plant.

Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of June 24:

Mr. Wylie NONE Mr. Michelson DAYS INN

. Mr. Ebersole CARLYLE Mr. Reed DAYS INN

Dr. Kerr LOMBARDY Mr. Ward NONE j Metal Components, June 25, 1986, Pittsburgh,PA(Igne). The Subcomittee will

review the status of NDE of cast stainless steel, and changes in steel-making i practice. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Shewmon Dr. Bush Mr. Etherington Dr. B. Thompson i Dr. Mark (tent.) -

4

Auxiliary Systems, June 26, 1986, 1717 HStreet,NW, Washington,DC(Duraiswamy),

8:00 A.M. - 12:45 P.M., Room 1046. The Subcommittee will discuss: (1) the status of the Appendix R compliance, (2) differing technical views among the Staff, (3) proposed research and associated budget for FY 1988 and 1989 in the fire protection area, (4) updates on the pro experimental program on fire protection, (5)gress being activities inspection made in the Sandia to determine compliance with the Fire Protection Requirements, and (6) recent experiences associated with inadvertent actuation of fire protection systems. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of June 25:

Mr. Michelson DAYS INN Dr. Shewmon MILLERS Mr. Ebersole CARLYLE Mr. Wylie NONE Mr. Reed NONE Regulatory Policies and Practices, June 26, 1986, 1717 H Street, NW, Washington, DC, (Quittschreiber) 8:30 A.M., Room 1167. The Subcommittee will review the Tegulatory process as it relates to the June 9,1985 Davis-Besse event using the Davis-Besse Ad Hoc Report as te basis for the meeting. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of June 25:

Dr. Lewis HYATT Mr. Michelson (P.M.) DAYS INN Dr. Remick NONE Mr. Wylie (P.M.) NONE Dr. Siess (A.M.) ANTHONY Gas Cooled Reactor Plants, June 26, 1986, 1717 H Street, NW, Washington, DC (McKinley), 1:30 P.M., Room 1046. The Subcommittee will review the applicability of NRC requirements for equipment qualification and cable testing to Fort St.

Vrain, an HTGR. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of June 25:

Dr. Siess ANTHONY Dr. Shewmon- MILLERS Mr. Ebersole CARLYLE Mr. Ward NONE Mr. Reed NONE Davis-Besse, June 27, 1986, 1717 HStreet,NW, Washington,DC(Alderman),

8:30 A.M., Room 1046. The Subcommittee will review start-up activities for Davis-Besse. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of June 26:

Dr. Remick NONE Dr. Siess ANTHONY Mr. Reed (p/t A.M.) NONE O

Joint Occupational and Environmental Protection Systems and Auxiliary Systems, June 27, 1986, 1717 H 5treet, NW, Washington, DC (Schiffgens/ Dura 1swamy),

8:30 A.M., Room 1167. The Subcommittees will: (1) review a draft AEOD report on the effects of ambient temperature on I&C Systems, (2) be briefed on the status of various control room HVAC Systems problems and the Staff's control room habitability . improvement effort, (3) discuss with the Staff the 1 mrem /yr

" cutoff" dose rate for the calculation of collective population doses, and (4) be briefed on the Staff's evaluation of the Shearon Harris Chilled Water Systems.

Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of June 26:

Dr. Moeller CARLYLE Dr. Shewmon MILLERS Mr. Michelson DAYS INN . Mr. Wylie NONE Mr. Ebersole CARLYLE Dr. First NONE Dr. Mark LOMBARDY Mr. Kathren NONE Mr. Reed (p/t, P.M.) NONE Mr. Till NONE Plant Operatinc Procedures, July.1,1986,1717 H Street, NW, Washington, DC (Schiffgens), E :30 A.M. , Room 1046. Ths Subcommittee will review " Proposed '

Commission Policy Statement on Technical Specifications." Lodging will be announced later. Attendance by the following is anticipated:

O Mr. Michelson Mr. Ebersole Mr. Reed Dr. Remick Mr. Wylie Metal Components, July 1 and 2, 1986, Columbus, OH, (Igne), 8:30 A.M. The Subcomittee will review the degraded piping program of RES and NRR and visit its primary testing facility at Columbus, OH. Lodging will be announced later.

Attendance by the following is anticipated:

Dr. Shewmon Mr. Ward Mr. Etherington Mr. Bender Dr. Lewis Mr. Rodabaugh Dr. Okrent Dr. Hutchinson Improved LWR Designs, July 9, 1986, 1717 H Street, NW, Washington, DC (Alderman),

l 8:30 A.M. (A.M. Only), Room 1046. The Subcommittee will be briefed and discuss the following topics: (1) the Standardization Policy Statement (2) proposed changes to 10 CFR 50, and (3) the EPRT Advanced Light Water Requirements documents. Loding will be announced later. Attendance by the following is anticipated:

Mr. Wylie Mr. Michelson

, Dr. Carbon Mr. Reed l Mr. Ebersole Dr. Siess Dr. Kerr 315th ACRS Meeting, July 10-12, 1986, Washington, DC, Room 1046. ,

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U Human Factors, July 15, 1986, 1717 HStreet,NW,Washinqton,DC(Schiffgens),

8:30 A.M., Room 1046. The Subconnittee will review: (:,) 5ECY-86-153, industry f and staff comments on proposed fitness for duty Policy Statement, (2) SECY-86-70, proposed rulemaking on degree requirements for SR0s at nuclear power plants and (3) SECY-86-119, the annual status report on implementation of the Connission Policy Statement on training and qualification. Lodging will be announced later.

Attendance by the following is anticipated: .

Dr. Remick Mr. Reed Mr. Ebersole Mr. Ward  ;

Mr. Michelson Mr. Wylie Waste Management, July 21 - 23, 1986, 1717 H Street, NW, Washington, DC (Merrill), 8:30 A.M., Room 1046. The Subcommittee will review: (1) NUREG-0518, Final Environmental Statement pertaining to the salvaging of contaminated smelted alloys, (2) the broader generic question concerning residual radiation limits and the disposition of land, buildings, metals and equipment resulting from the decontamination and deconnissioning of nuclear power plants and fuel facilities, and (3) various nuclear waste topics. Lodging will be announced later.

Attendance by the following is anticipated:

Dr. Moeller Dr. Shewmon O Dr. Carter Dr. Carbon Dr. Mark Dr. Orth Dr. Remick Dr. Steindler Naval Reactors, (operation of a nuclear-powered submarine), July 28, 1986 (Boehnert). The Subcommittee will observe the activities of a nuclear submarine i- crew. Attendance by the following is anticipated: -

Dr. Kerr -

(majority of ACRS members)

L Westinghouse Reactor Plants, July 30, 1986, 1717 H Street, NW, Washington, DC l (Houston), 1:00 P.M. - 5:00 P.M., Room 1046. The Subcommittee will continue discussion and connent on NRC Staff actions taken with respect to the SONGS-1 water hammer / loss of AC power event. This will be a follow-up Subcommittee meeting to the February 12, 1986 meeting on the same subject. Lodging will be announced later. Attendance by the following is anticipated:  !

Mr. Reed Mr. Michelson Mr. Ebersole Mr. Wylie Dr. Kerr Dr. Catton

. Waste Management Subcommittee Visit to WIPP and NTS Facilities, July 30 -

August 1, 1986 (Merrill). The Subcommittee will be briefed and take surface and

, underground tours of the DOE Waste Isolation Pilot Plant (WIPP) near Carlsbad, M and the DOE Nevada Test Site (NTS) Facilities near Las Vegas, NV -- G-Tunnel, Climax,' Jackass Flats (E-MAD), and Yucca Mountain. The purpose of these visits is for the members to gain a better understanding of the problems associated with the design, construction and operation of underground facilities similar to the geologic repository for High-Level Radioactive Wastes. Attendance by the following is anticipated, and reservations have been made for them at the AMFAC l

Hotel, 2910 Yale Blvd., SE, Albuquerque, NM for the nights of July 29 and 30.

, Reservations have also been made for them at the Tropicana Hotel in Las Vegas, NV for July 31 and August 1 and 2 :

, Dr. Moeller Dr. Shewmon Dr. Carbon Dr. Donoghue Dr. Remick Dr. Krauskopf Scram Systems Reliability, July 31, 1986, 1717 H Street, NW, Washington, DC, (Boehnert), 8:30 A.M.. The Subcommittee will discuss the status of the ATWS Rule implementation effort. Lodging will be announced later. Attendance by the l following is anticipated:

[ Dr. Kerr Mr. Wylie Mr. Ebersole Dr. Davis Dr. Lewis Dr. Lipinski l Mr. Reed Metal Components, August 4 and 5, 1986, Hanford, WA (Igne). The Subcommittee will visit and review steam generator integrity program and visit its facilities.

In addition, the integrated FM/NDE program will be discussed. Attendance by the following is anticipated:

Dr. Shewmon Mr. Ward (tent.)

Mr. Etherington Dr. Bush Dr. Lewis Mr. B. Thomspon Reliability Assurance, August 5, 1986, 1717 HStreet,NW, Washington,DC(Major),

8:30 A.M. (A.M. Only), Room 1046. The Subcommittee will review the final resolution of USI A-46, " Seismic Qualification of Equipment in Operating Plants."

Lodging will be announced later. Attendance by the following is anticipated:

Mr. Wylie Mr. Michelson Mr. Ebersole Dr. Siess O

Safefluards and Security, August 5, 1986, 1717 H Street, NW, Washington, DC (Sch'ffgens), 8:30 A.M., Room 1046. The Subcommittee will review echnical

, Assistance Program on the " Evaluation of Methods of Reduction of Vulnerability to Sabotage (Generic Issue A-29)" with the NRC Staff. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Mark Dr. Okrent Dr. Carbon Mr. Reed Dr. Kerr Mr. Ward (tent.)

Dr. Moeller Mr. Wylie ,

Extreme External Phenomena, August 6, 1986 (tentative), 1717 H Street, NW, Washington, DC (Savio), 8:30 A.M., Room 1046. The Subcommittee will conduct a .

workshop to review the importance of seismic risk to nuclear power plants.

. Seismic hazard will be the principal topic to be discussed. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Okrent Dr. Lewis Dr. Carbon Dr. Siess (all other ACRS Members, as available)

L 316th ACRS Meeting, August 7-9, 1986, Washington, DC, Room 1046.

Maintenance Practices and Procedures, August 13, 1986, 1717 H Street, NW, Washington, DC (Alderman), 8:30 A.M. (A.M. Only), Room 1046. The Subcomittee will review the report on Phase I of Maintenance Program Plan. Lodging will be announced la,ter. Attendance by the followir.g is anticipated:

Mr. Reed Dr. Moeller

Mr. Michelson Mr. Wylie

! Decay Heat Removal Systems, September 24, 1966, 1717 H Street, NW, Washington.

j DC (Boehnert), 8:30 A.M., Room 1046. The Subcomittee will continue its review

oT NRR's proposed resolution position for USI A-45, " Shutdown Decay Heat Removal Systems." Lodging will be announced later. Attendance by the following is anticipated:

Mr. Ward Mr. Reed Mr. Ebersole Dr. Catton l

Mr. Michelson Mr. Davis i

Wingspread International Conference (CLOSED), October 19-23, 1986, Racine, WI l

(McCreless). Represeritatives from the ACRS, RSK, GPR, and Japan will exchange j infomation on nuclear reactor safety.

Instrumentation and Control Systems, Date to be determined (July), Washington, DC,(El-Zeftawy). The Subcomittee will review the Westinghouse RVLIS level Instrumentation. Attendance by the following is anticipated: .

Mr. Ebersole Mr. Michelson '

Dr. Kerr Mr. Wylie Dr. Lewis

1 l

Nuclear Plant Chemistry, Date to be determined (July / August), Washington, DC (Alderman). The Subconunittee will discuss fission product source terms, aerosol

, behavior, emergency planning, etc. Attendance by the following is anticipated:

Dr. Moeller Mr. Reed Mr. Ebersole Dr. Shewmon Mr. Etherington - i Westinghouse Water Reactors, Date to be determined (July / August).

Washington, DC (El-Zeftawy). The Subcomittee will begin the PDA review of the Westinghouse Advanced Pressurized Water Reactor (RESAR SP/90).

Attendance by the following is anticipated:

Mr. Reed Dr. Shewmon Dr. Kerr Mr. Wylie Mr. Michelson Mr. Davis Spent Fuel Storage, Date to be determined (July / August), Washington, DC t (Alderman). The Subcommittee will continue its review of ID C R Part 72 and Monitored Retrievable Storage (MRS). Attendance by the following is anticipated:

, Dr. Siess Dr. Remick

! Dr. Kerr Dr. Shewmon Dr. Moeller

! Decay Heat Removal Systems, Date to be determined (mid-August),1717 H Street, NW, Washing;on, DC i Action Plan to address ns coiicer(Boehnert).

with the reliability of.The certain Subcommittee plants' AFW will revi systems. Attendance by the following is anticipated: -

Mr. Ward

• Mr. Reed

, Mr. Ebersole - Dr. Catton I

Mr. Michelson Mr. Davis

{ Thennal Hydraulic Phenomena, Date to be determined (mid-August),1717 H Street, '

NW, Washington,DC(Boehnert). The Subcommittee will continue its review of the RES-proposed revision to the ECCS Rule (10CFR50.46 and Appendix K). Attendance ,

by the following is anticipated:  ;

i Mr. Michelson Dr. Catton '

, Mr. Ebersole Mr. Schrock l Mr. Reed Dr. Sullivan l Mr. Ward Dr. Tien l '

AC/DC Power Systems Reliability, Date to be determined (August), Washington, DC

(El-Zeftawy). The Subcomittee will review the proposed Station Blackout rule (SECY-85-163). Attendance by the following is anticipated

Dr. Kerr Mr. Reed Mr. Ebersole Mr. Wylie Dr. Lewis -

i l

, Regional Operations, Date to be determined (August-September), Chicago, IL 1 (Boehnert). The subcommittee will begin its reivew of the activities of the NRC Regional Offices. This meeting will focus on the activities of the Region III l Office. Attendance by the following is anticipated:

Dr. Remick Mr. Reed Dr. Carbon Mr. Wylie Mr. Michelson Seabrook Units 1 and 2, Date to be detennined (late summer /early fall),

Washington, DC (Major). The Subcomittee will review the application for a full power operating license for Seabrook 1 and 2. Attendance by the following is anticipated:

Dr. Kerr Dr. Moeller Dr. Lewis Mr. Michelson Structural Engineering, Date to be determined (late 1986), Albuquerque, M (Igne). The Subcomittee will visit 'and review containment integrity and Category I structures, facilities, and programs. Attendance by the following is anticipated:

h U

Dr. Siess Mr. Ebersole Dr. Shewmon Mr. Bender Dr. Kerr Dr. Pickel Dr. Okrent -

ProbabilisticRiskAssessment,Datetobedetermined(September / October),

Washington, DC (Savio). The Subcomittee will review the probabilistic risk assessment for Millstone 3. Attendance by the following is anticipated:

Dr. Okrent Mr. Michelson Dr. Kerr Dr. Siess Mr. Ebersole Mr. Ward Dr. Lewis Mr. Wylie Dr. Mark O

W1 l -i L O The meetings listed below have been changed or added to the list of Subcommittee meetings previously issued at the full Comittee meeting.

e CHANGEDIII The Long Range Plan meeting previcusly scheduled for June 17, 1986 has been POSTPONED to JULY 9, 1986.

Long Range Plan for NRC (CLOSED), July 9, 1986, 1717 H. Street, NW, Washington, DC, (Major), 1:00 P.M., Room 1046. The Subcommittee will consider the NRC Staff's Five Year Plan and discuss the guidelines for the --

review of a long range plan. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Carbon Mr. Remick Dr. Moeller Mr. Wylie O ADDED!!!

Management Committee (CLOSED), July 9, 1986, 1717 H Street, NW, Washington, DC (Fraley), 1:30 P.M., Chairman's Office. Specific topics have not been selected as yet. Lodging will be announced later. Attendance by the following is anticipated:

l Mr. Ebersole Mr. Ward Dr. Lewis

! Procedures and Administration, August 6, 1986, 1717 H Street, NW, Washington, Tane(Fraley), 1:00 P.M. The Subcommittee will consider the Effectiveness DC l l's recomendations regarding ACRS Officers' terms. Lodging will be I

. announced later. Attendance by the following is anticipated:

Mr. Ward Dr. Moeller Mr. Ebersole Dr. Remick Dr. Lewis Dr. Siess l

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CP&L 25.2 HL&P 30.8 %

e q Austin 16%

San Antonio 28 %

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San Antonio 59 35 Victoria sus '

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Corpus Christ V 521 l2.

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l l SITE LOCATION STP

l BRIEF HISTORY e June 1973 - Plans to build STP announced

- Brown & Root named as A/E constructor i e August 1975 - LWA granted i

• December 1975 - CP issued e August 1979 .

- Unit 1 NSSS Components set e November 1979 - Special NRC inspection commences e December 1979 - Stop work on complex conc. rete placement e April 1980 - Stop work issued on welding e April 1980 - Order to show cause issued i e October 1980 - Welding restarted i e January 1981 - Restart on complex concrete .

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• May 1981 - ASLB hearings on OL commence l

. STP  !

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i BRIEF HISTORY (Continued)

e. September 1981- Brown & Root terminated, Bechtel hired ,

e February 1982 - Ebasco named as new constructor e June 1982 - Non safety-related construction resumed e August 1982 - Safety-related construction resumed e March 1984 - ASLB Partial Initial Decisio.n e March 1985 - First systems turned over to Start-up e May 1985 - Energization e July 1985 - ASLB Phase 11 Hearings

• December 1985 - Commenced NSSS Flush i

e March 1986 - ASLB Phase lli Pre-hearing

e April 1986 - SER issued l

STP l

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i STP STATUS - MAY 1986 ~

Percent Complete Key Dates Fuel Commercial Scheduled Actual Load Operation Construction .

Unit 1 Power Block and BOP 90.3 89.9 6/87 12/87 Unit 2 Power Block 59.8 60.4 12/88 6/89 Total 77.6 77.5 I

Engineering 93.3 92.9 1

STP o- - . _ _ _ . - ,-

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SOUTH TEXAS PROJECT MILESTONES l

e Secondary hydro July 1986 e Primary hydro August 1986 -

. Hot functional test January 1987

. ILRT / SIT Ma'rch 1987 e Fuelload June 1987 i

STP i

1

l 0 CJ O PROJECT STAFFING - MAY 1986

~

o Bechtel Construction Management 695 I EBASCO manual 5,410 EBASCO non-manual 1,451 j Bechtel Engineering and Home Office 540 HL&P* . 1,669 Other 362 '

Total . 10,127 l

• Includes all testing personnel sTP 1

l .

a .._ . ... m.._._- .m,, - . 2._. - -.2.-_ __ - a _.- . ._-. m __..

---

.._g . _ ,, % . ..- .. . . ._.

, " g - ' ,.%

j O O O~

i i

CONSTRUCTION ORGANIZATION, ~

PHILOSOPHY AND STATUS 1

8 5

C l

50UTHTEMR5 O PROJE,..C,_T 5

- , _ .. _ g=g BA .

"; E 5 -

45R

_ 59< l E

I:

p

. 8 5a 5

l C 0 -

CP HOUSTON LIGHTING & POWER COMPANY Senior Executive Organization .

Board of

. Directors l

l l

Chairman and Chief Executive Officer D.D. Jordan l

I '

y .

Pres,and Chief rating .

cer D.D. Sykora .

I I I I I

! Group V. Pres. Group V. Pres. Group V. Pres. V. Pres. Senior V. Pres.

Admin.and External Power Finance Gen. Counsel G'O VP $ support Affairs C0'porate Nuclear Operations Comptroller h

E.A. Turner RJ.Snokhous D.E. Simmons J.S. Brian H.R. Kelly

D Q^

MANAGEMENT PHILOSOPHY e To build and operate the plant in full compliance with regulatory ,

requirements e To require the person performing a task to accept total responsibility for performing that task in a quality manner e To require Quality Assurance to independently confirm the quality of activities being performed e To perform management oversight of project activities to ensure compliance with applicable program requirements e To report in a timely forthright manner, all matters required by regulatory authorities -

STP

l -, c ,~

O _

O, -

O^

MANAGEMENT PHILOSOPHY (Continued) i e To protect the environment and the health of the public and our

employees i

I e To utilize proven equipment and techniques in design, construction  ;

and operation to assure reliable operation of the plant i I

e To keep abreast of industry occurrences and apply worthwhile experiences to improve programs i

i

! e To ensure that the concerns of employee's are heard and properly l aded upon -

, e To learn from our mistakes by determining underlying causes and to take necessary actions to preclude recurrence e To plan for, develop, and retain qualified and trained personnel l

8" l .

> b .i l

l HOUSTON LIGHTING AND POWER ~

l ORGANIZATION Chief Executive Officer i

~

~

. l Group Vice Pres,i dent '

i I

l l l l l l 1 Nuclear Nuclear Nuclear Project Nuclear Engineering Special  !

Operations Licensing Assurance Engineering- Assurance Assignments ,

l 1

STP l

k _ _

O O CT STP ORGANIZATIONS '

.!1

. HL&P Project Manager

~

l - . Bechtel -

Architect / Engineer &

Construction Manager

. Ebasco -

Constructor

. Westinghouse -

NSSS y l 4,$

i m

i.-

-_.m __-.

f K.

• y4

~

s-M

,p*

- .i

~ , .

~

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~

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. ~- . .

-;.  ?

~

SOUTH TEXAS PROJECT

,i~ m

,c '. ,i t,

j - .,

_- ~ _

, q .

l' HL&P -

j / - PraJe -

1 -

Paenneer - 1

] ,

i l l l l l l HL&P Project Project HL&P HL&P HL&P Cost / HL&P I - QA Manager Construction Engineering HL&P Startup gg HL&P Project Nuclear Meneger Bechtel Manager Manager Meneger Manager Meneger i I l

s l l

' I l

l Bechtel '

i

3 ,

0"*"'Y

- - -l- - " Construction Meneger NSSS w- - ," ---

l Bechtel Ebesco Construction

- ouenty - - - - - - Meneger Assurance Ebesco PROGRAMMATIC DIRECTION

--- ADMINISTRATIVE DIRECTION 1

I l, -

..._s2.: . _. - - . - . _ . - . m- - - ...-__a . _ _ .# m-,- , - - - . - . _ _ . _ . - _ - -_-...m. .. _._ ._..- ._..m- -._- _u-_ . ..m_m.- ,_ m 0

1 APPENDIX VI PLANT LAYOUT 1

O _

l 1

1 3

< O y

o 5 r Z

. <C i

CL .

b i

---~w--v+-~ 6mw- ~-mm-va-_www, m-+-,,a wxrem,,,-,-~sw__

C i O 0:

Site Boundary

\

s =, Unit No.1

" Railroa or [MeteorologicalTower .

N -

Essential Cooling Pond Colorado Swi hyard- 3

/ " River m Unit No. 2 %

\

Circulating /

j Water "

~

l Discharge  :

/ Structure Circulating FM S21 Water intake Structure Embankment MAIN Nah"eup

[ COOLING """"f --

Discharge RESERVOIR m

f RESERVOIR Structure MAKEUP .

PIPEUNES Reservoir f Makeup

) Pumping u Facilities i s Spillway / k I Spillway

. Blowdown l Site Facilities Dischar e Boundary Channe

"^ =# =

'=

owjown ,

SITE MAP STP

, ^

\'

High Voltage

Transmission Lines DieselGenerator Building

)

e g

0 f

ll _ _ _

1

"'gp ssential y

E g, = -

1. Cooling Water

! -'- In e 5tmctum West -

l Gate : ',.

j , EssentialCooling Pond i House r F

{  %.

,) ,,,4, y 7 d2_"N '

l 0 O f .2  ? 9 '

~

l TG a p ast EGate Guard House

\

i Security >

Fence g> - ,

.l g .1 NAdministration Building

!, x ,

vi_EAss .

1. ' RCB

! \

] ' [Qsj[?

jMA8 r

, Fire Pump House

) Railroad ,

, d((}$ ,

5 g 5 I I '"

f E } I ]

4: ,

r3 '

i AB i

tsse, JL3 O' o - maneu, oemine,aiim House -

l '

j AFsi = Auxiliary Feedwater storage Tank j 7, " $*cgCM,%

, FuelOil Storage Tank t FHS = Fuel H Suilding AB = Auxiliary r TG8 = Turbine Generator tuilding '

L^A. : W'M^"L'"*E7.d!J"

. ng PLOT PLAN i STP

! I i

i l

n o

i O .

O'

.N .

r m .

Ci I j,

• N _

Steam Generator Personnel & Equipment Reactor Air Lock Access ,

Reactor

( "

/

l \s Containment L

Building -

ud$ >

L Reactor Coolant :mg v 1)N ' El. Operating Floor Pump '

68' - 0 "

g,jp ,

$-g z.8 '

!) ~

i

~

T - Fuel Storage Pit Reactor Pressure -p .'

Vessel ,,

x ,

x p ,3 g- -

Residual Heat '

4 2 -

) u, '

Removal Purnp

/ -

(p

' F? ,

?

- STP

, , . r Reactor I Containment ,,,7 }C m-Building , .-

, 4

, [. y, . .. *. $o**o**

,. Post Accident Sampling System 0

l h

l I

% l

~

~[

SpentJFuel_ Pool

$1, f;",0 . o-c ;'- -

Shipping Bay e

p+:

y i, ,.

+ ..:.- .

Low He Safety injection Pump i High Hea'd Safety injection i Pump Containment .

Spray Pump Fuel Handling Building or,,

O

4 commaa wa's N

j y,*rg=*r Electrical Mechanical Auxiliary Building Auxiliary Buildin l l l (Elev. 35 ft. O in.) (Elev. 41 ft. O in.)g Recycle Holdup Emergency Entranc Only j\p * -+

f anksg T

Equ t

{

_ T E R iew.t.,r-_]ks r (-

(vl l ,

.ci,

+

E A_ss 8*'""

^

u E' supervgi- i - --

c. t, R _

l,_11' O ,-

q' a F

~

'1'tr~ lh I p L 0 -

I. L vol-e ..

4i i> f

%_l- ControlTank E-

• [

l n-m k-l U

EL -

-.,mers p e, _

i l l "

l A -

Mpe Aree Refueling Water  ;

Storage Reactor Makeup Tank I water storage Tank Reactor Conta.inment Building V

MECHANICAL AND ELECTRICAL AUXlLIARY BUILDING .

STP

. . o I

l

! s s

4 n O N

Reactor '

Containme'nt Relay Room h

w ComputeN l Room

- oqtrol Room

" Switchgear Room l

• (C loor Elev. 35' - 0" ,,,,

.. Equipment Removal D 4; -

Hatch

,' 7 , ,

'po  % Grade Elev. 28' - 0" y y

/

\ , .7:. c # '

'~

Auxiliary Shutdown Panel Elev.10' - 0"

/ '

Access Control Elev. 41.'.0"

/

Mechanical and Electrical Auxiliary Building STP

m w ,-,. .m-9 -- ,- v - -- -- ,

)g.. -

n LC. 1 N

O p

U2 li iap '""b Q o 0 0b 0 0 M "n;ll'

=

1 D, u V .

m o _

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- ic sim -I "'"""' m F9 m

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m g

C niarT 1 .$L o

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rotareneG .W y leseiD ybdnatS /

)f EEE rr & , ~

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f ,'

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/. 'o 5 1, 1, 2

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m o 3 m L E

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n! o R m o'

3 m E82 o m 3 n-i

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.m i* l li s  ! !I i y: :  ;

! II; $

l -

ii

l.  !-

ri I

i i l '

l i i l

.i 1

I i

i Monorail -

r i

Stack N . . . .

I C

g l

Walkw - - - -

y -

M.S. lsolation Valve l l I l 1 l

N i 1.F is ifE s

l h "

R

) -

g _

j y r

' Main Steam Lines Auxiliary Feedwater Feedwater Storage Tank u v fHydraulic Actuator i

y Module n n 0 D#

m l

3 yyy, w Grade 3 El. 28 ft. O in.

l

_ic D. R. A A

?@ 92

Auxiliary

{ Feedwater Pump ISOLATION VALVE CUBICLE AND AFWST

- .. _ _ _ _ rm

1

. l t - n ,,  ; . ,

~

i 1

l l!

) i Safety Reliefs -L J ,

r i  ! T T l

Main Steam isolation Valve _____ _

lj-.

y a [%v n

I

)

FF ' "

ii

. 1 - -

3 _ t_r P g Atmospheric Steam Turbine Generator + Dump Power Operated Building -

@ [ Relief Valve

/ Main Steam Pipe j ' Platform , , , ,, ,, 'p ,

1 "

~

j ll "" l -FeedwaterIsolation Valve s}:s l , ,: s Feedwater Pipe

, N l Grade m HydraulicActuator El. 28 ft. 0 in. " Module t

~ Auxiliary Feedwater -

! [ Pump p Rhactor Containment j, Building

-i .'

CROSS SECTION -ISOLATION VALVE CUBICLE i STP i i

I . O o ON 6

i 1

-l THREE TRAIN DESIGN i A r 50UTHTEMAS >

W4

,,,,.P.RO.. J.E.C,_T g=;o S5 i 5~

=

i

9- 3 9: :

SOUTH TEXAS PROJECT  ;

i i

1 i

• e Three-train ESF systems l ,

j - Physically segregated

- Electrically independent j . >

9 STP

i O' . ~O~ O?

l i

1

~

Spray

!!!N$.I CCW Surge Tank n inside iOutside -

7 T i

ReactoriReactor J g f

.........::.. A- - i8 i C>

j Containment jContainment / \

l Q.g

] Accum CS Pump .

COId '

Leg HHSI Pump Essential

, ca g;,g 1

RHR HX.

I Pump  !

! LHSIPump 3( ' '

1 _ i RCS l l) [ ,

~

CCW HX. .

q I - -

i gt - x

-i 8 RHR

CCW Essential Pump Sump .i Pump Cooling

-+ Pond Discharge g Structure I

j l "n Diesel Generator HX'S, Essential Chillers I

CCW Pump Supplementary Coolers C, .i I  !

RCFC's '

INTERRELATIONSHIP OF ESF SYSTEMS IN '

THE THREE-TRAIN CONCEPT

! (Typical of Three) STP i

i

,I i

o o ca

~ """

l 5

ECCS

'  !!gg!

. RWST'! CCW SurgeTank l  !!26!!

i

=n- M N N N

!!enctoriineacierii =3 mi ;wnii b.k!

it: -

(A iBi C )

! MdMj,  !!! iiM !4 :

gy j.:f,'

. . . . . . . . . . . . . n

!!!MNN! h!b  ! !C5iPusty hii!

""~""

a

!RC5!!._  ;;is i '

i,:

iC6ld" i

l @ dj" ?" g j '" .".

"~

!7[

!!HMA485$DF

...M = = :w Essential

] "~  !!

!!g!

1 i Cooling i . . . . . . . . . . . . . . . . . . . . . .

- L h iJ . . .

I Pump .

j -  !

!!g

~

u .3 M =i! !!!!

3.it:::

i!

!.D':i! !

I I )[ ii

!!{4 1 -

, %. mii CCW HX.

1 RCS Hot  ;; gii

!!rmmii!li!!

~jjjjj;,jj i;

iiji ygi.

I f

I89  !!p" iii!! m=r CCW Essential 1  !!

RHR iijii i:"i:i":-

Pump Cooling i Pump $Wap - - -

--+ Pond l

jj![' Discharge j Structure l l

! I Diesel Generator HX'S, Essential Chillers I  ! CCW Pump Supplementary Coolers m m ,

I-  !

RCFC s INTERRELATIONSHIP OF ESF SYSTEMS IN

THE THREE-TRAIN CONCEPT

. l (Typical of Three) .

i ,

STP

1 i

I .

Spray Add i Tank RWST c

CCW SurgeTank n inside iOutside' -

7 ReactoriReactor J g f 8

(A iBi C )

ContainmentjContainment / \

Accum i ~ CS Pump RCS.

• i

...f...

" " " ~ " " " .......

Celd !FW "N. -

t,g n.m a*  !

Essential i

{ HHSiPump

{  ! -

Cooling j Pump ,

RHR HX.  !

LHSIPump N ' '

i ..+ ii's '

l l RCS RHR I

5 1

l 1...  :.mis i

1 _ N_

CCW Hx.

Hst .X; , o -s , ,

le9 "~":.':::......... i mm l " CCW Essential l  !!kHR!!

i Pump Cooling .i

!!i,hiisi@"i!!.

- =.." Sump - -

--> Pond

[ Discharge Structure .

i 1 -

v ag

!  ! Diesel Generator HX'S, Essential Chillers I  !

L -

CCWPump Supplementary Coolers ,

t i r- i I  ! ,

RCFC's ' '

) ,

INTERRELATIONSHIP OF ESF SYSTEMS IN t

THE THREE-TRAIN CONCEPT L

l (Typical of Three)

STP i

Ci  ! O'  :

O' i

Spray Add.

][h Tank s RWST CCVW54srgdiiiill p insideiOutside -

j 7 k~ $l.lll5llllll$!lhf; ReactoriReactor g ContainmentjContainment )(

s C ' ,

Accum i CS Nmp RCS gg  !

4 '

Essential HHSIPump C ng RHR HX.  ! LHSIPump /

7 RCS 1.l p ~l~.l l.".l[ " !

t . W=*

T

- \) h - "*$- '

M " '

~! ~

iiggi.! Essential RHR Pump Sump .;

i N *: _ _ "gI pg;; Discharge

l::f.f  ;;l:;;

q 7 hili

~ Diesel Generator HX's, Essential Chillers

ijgg lj] CCW Pump Supplementary Coolers CD +

+

550 lh h? .

INTERREthTIONSHIP OF ESF SYSTEMS IN  ?

THE THREE-TRAIN CONCEPT (Typical of Three) .

STP

~ ~

i i

~

4  ! #

Ydd bbb  ; . Tani . RWST CCWSurgeTank j n insideIOutside -

7 l

' ReactoriReactor J

)( 8

{A isi C ) = =

Containment jContainment RCS Ccid '

gf Lcg ijfshiiWitgit

HHSIPump yliiiggg'"

g RHR HX. 8 LHSIPump A, " M'""Riii ""'

i I

l I '

l 2 L'Dl rw .

_h- """';....

CCW HX. !

i RCS  !

l;. . .

l Hot M - ,J  :

Leg i l

l RHR

-m

"' CCW l!IM '

Pump i Pump ncii$UeII$k!j!

Sump x=N5de y

" !!iip $ttwtMte!!'isEssis "j g is i!!

l

!  ! Diesel Generator HMI, Essential Chillers I  ! CCWPump Supplementary Coolers l r- r-  ! '

l' I  !

RCFC's INTERRELATIONSHIP OF ESF SYSTEMS IN

, THE THREE-TRAIN CONCEPT l

(Typical of Three)

STP i-

l 4

0 6  ; CP SOUTH TEXAS PROJECT i

l 1

i Benefits:

e Proven equipment size I

e Greater margin provided j e Greater protection for plant investment j -Single train for small break LOCA

-Single train for normal shutdown

- Redundant paths for fire protection i

. Simpler piping '

e i

! STP 1

~

. . ~ ._

.i l

I 1

i

~

SOUTH TEXAS PROJECT THREE TRAIN SYSTEMS i -

{ Minimum number of trains required Trains Normal Operation / Large Breaks System installed Other Shutdown Accidents Diesel 3 0/0 2 1 Generators i Essential 3

' 1/1 2 Cooling Water 1 Component Cooling Water 3 1/1 2 1' Reactor 3 i Containment (2 RCFC 2/2* 2 1 Fan Coolers Units per (3 RCFC Units)

] train) l Safety injection 3 0/0 2** 1 Containment 3 0/0 2 0 Spray .

2 Residual 3 0/1 2 1 Heat Removal j Heat Exchangers 1

Auxiliary '

4 0/0 2 i Feedwater 1 i

Normally supplied by RCB chilled water

• • Share RHR exhanger- RHR pumps not required 4

STP

C 6 3 STP HVAC THREE TRAIN SYSTEMS Minimum number of trains required rains Normal Operation / Large Other Accidents System Installed Shutdown Breaks with Load Shedding l Essential Chilled Water

• 3 2/2 2 1 EAB HVAC 3 2/2 2 1 i Control Room HVAC 3 2/2 -

2 1 FHB Exhaust 3 2/2 '2 1 ECW Ventilation

• 3 1/1 2 1 DGB Ventilation ESF* 3 0/0 2 1 IVC ESF Ventilation

• 3 0/0 2 1 ECCS ESF Cubicle Cooling

• 3 0/0 2 1 CCW ESF Cubicle Cooling

• 3 1/1 2 1 Charging Pump Coolers

• 3 1/1 0 0 Chiller Cubicle Cooler

• 3 2/2 ,

2 1 Cubicles are provided with individual 100% cooling and ventilation system sw

9' 9 98 SOUTH TEXAS PROJECT FIRE PROTECTION FEATURES e Provides two Ahutdown pathways assuming a fire in any fire are'a VERSUS Only one safe shutdown pathway as required by Appendix R.

• Complies with the requirements of Appendix R and provides equal or better alternatives to the criteria of Appendix A to branch technical

~

position APCSB 9.5-1 ~

e Safe shutdown capability provided outside the control room for all three trains with capability to maintain cold shutdown i  ! STP j.

rF 9

O

. O~ O

! SIMPLIFIED ELEC AUXILIARY BUILDING - .

VERTICAL SEPARATION

, .o .

Train C Penetration Area Train C Train B f A Penetration Area Train B r m Train A Penetration Area Train A

, r ,

i, .

g. g g U anda containment R"a'sdJ containment AUXILIARY FEEDWATER SYSTEM FourTrain Concept 8 , v

• V\ c' m

^

A .

j ,

"A" Actuation Signal O kM , e

1. Q v~

j r, n~

8 m ,,,

Actuation Signal

[ v ,' Auxiliary

^

r' n~ ^

s E Feedwater Storaae "c" TanE Actuation O W '

4  :

Signal

1. T -

j #'

v

^ T n [i Actuation "A" ,N

() Signal STP

(7 i

i O,  !

D i .

i i i a

! CONTROL ROOM INTEGRATION I

l

! TEMAS '

l EG e

o 5

+

.i. . _ _ _.t. . _ . _ .

a l~o o' d' i CONTROL ROOM DESIGN REVIEW l CONTRIBUTING FACTORS - 1982 .

i j e Safety related control panels <20% complete l

i e Final evolution of TMI criteria .

, e W OG ERG's available

. Active participation of HL&P Operations STP e

r- - - - - - -

)

o o o'

. CONTROL ROOM DESIGN REVIEW l INTEGRATED DESIGN CONCEPT 1

1 .

i Develop integrated criteria to address:

l . Human factors -

I

. Post-accident monitoring (including R.G.1.97) l

. Safety parameter display

. Emergency operating procedures ,

I . Safety grade cold shutdown

. Bypass / inoperable. status monitoring .

e Annunciator and alarm prioritization

, STP

^

I

,n, .

O O L '

O.. -

CONTROL ROOM DESIGN REVIEW ,

l INTEGRATED DESIGN. CONCEPT (Continued) e

! Constructed full-scale mock-up / utilized simulator ,

e Performed CRDR -

. Performed re-layout .

j -: Total re-layout of 6 panels including ESF pan 61s

- Upgraded layout of remai.ning 4 panels 2

~

b i _

- - =

i STP e / -

1 . . .

x;' ' W ls g- .

~; -

_.  ?-[

i +

1 ,

~

). ,

J , ~ , ~ ' -

./

' ^

' ~, . ~

-MILESTONES  :

1 CRDR'

~

r. - ~

Xs_. _ .

j i , - rn , _

j_ .

.1 -

CRDR started l

~. 1 8/82 .

0< m

! issue program plan / plan report

~ '

1$8Y  : I Issue implementation plan report

~

3/83 NRC in-process audit 5/83 Panel re-design released to fabrication 9/83 l NRC audit report 10/83 l Issue executive sum. mary report 4/84 l

Panel delivery / installation 6/84 .

l NRC site visit 10/84 l Issue executive summary report addendum 4/85 l

l j STP i

l

.- ~ .

-?

w -

.o .

m '

REMAINING WORKITEMS -

CRDR -

! . Checkout of labels and scales i e

l . Complete surveys dependent on control room completion f

! - Lighting / indicator visibility i

- Sound / annunciator horns

- Computer displays

! - Workspace t

- Commertications

. EOP validation 1 STP I

i. ,.._

0 . .

O. .

O'

- ALTERNATIVE SHUTDOWN INTEGRATED DESIGN CONCEPT

. Appendix R criteria ,

. Safety grade cold shutdown criteria l . CRDR/ human factors criteria for this auxiliary I shutdown panel - no fabrication started

. Cable routing design essentially not started O

i i

i j

STP

i i

j ALTERNATIVE SHUTDOWN - MONITORING 1 Main -

! Control Room -

To j f , - Protection~

l System f

QDPS _ _

l . e ., e,A u # Field

! N l Sensor

' Auxiliary l

ASP Shutdown Area -

e s 6 . < a i g ,

! o d cr ALTERNATIVE SHUTDOWN - CONTROL N

[ Main XFER SWGR p

4 Control Room A l Room l

l

/ '

SWGR l

Room B X.FER 1 , (Typical for PNL

3 Trains)

ASP Auxiliary Shutdown Equ.ipment Area -> Control Center l

l XFER SWGR I

. PNL Room C l' ' I g

4

o g .

3 l

i ,

t f

h QUALIFIED DISPLAY .

! .. PROCESSING SYSTEM (QDPS)

I

]

e i i j

- . .  : - i _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

i j INTEGRATED DESIGN CONCEPT QDPS

. Optimize instrumentation design to address overlapping regulatory guidance

-TMI .

- Appendix R

- Safety grade cold shutdown

. Optimize cable routing

. Minimize modifications to existing equipment STP

INPUTS TO QDPS e Post-accident monitoring parameters .

- Inadequate core cooling instrumentation '

l - RG 1.97 category 1 variables

! e Safe shutdown monitoring and control 1

parameters -

rQDPS e Complementary post-accident monitoring, control, and protection system parameters j (to enhance display implementation) .

i e Advanced design modification para' meters "

- SGWLCS .

- TAS .

sre

., b 4

I' OUTPUTS FROM QDPS e Qualified PAMS display in Control Room

[ e Critical Safety Function parameter displays in Control Room QDPS e Safe shutdown system displays in Control Room and ASP e Isolated data links to SPDS  ;

e Control (modulating) for safe shutdown i valves -

e Protection system (SGWLCS and TAS)

STP I

s *

.,t ,

J . .

DIGITAL SYSTEM ADVANTAGES -

QDPS 3 .

. Graphic displays support operating procedures

, . Reduces control panel clutter - fewer panel

! indicators '

i e Relieves operator of cross channel checking burden

. Performs quality checking of inputs signal e Simplifies implementation of signal distribution via data links

. On-line diagnostics and self-calibration

, STP S *

i. _ _ _ . _ _ _ _ ._ . __ _ _ _ _ _ _ _

I Q O 03 .

VERIFICATION AND VALIDATION PROGRAM  ;

j QDPS

. STP V&V plan based on established technology (RESAR 414, ANSI /IEEE-ANS-7-4.3.2)

. Performed by an independent team of verifiers ,

i

! . Consists of exhaustive functional and/or

! structural testing of software l

1

. Program will be completed prior to fuel load .

.. A minimum of four audits will be performed by the NRC staff l

STP i

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OFFSITE/ONSITE ELECTRIC POWER SY. STEM l

l SOUTHTEMAS NE I

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g l g 2 14 138 KV to River Pumping Station

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, , y NewIh Bus g ,[ - 345 KV Switchyard g, , \ \

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( Emergency TRF

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( Steel Tower

. (Typical)

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Standby Standby y >5 O Transformer i*E4 O Transformer

,, , 4 #2 #1 g 00 ,

g 00

.. h Turbine j .

Turbine h o Gen Bide

1. 2 Q 1384.16 KV G*n Bld9 81 O 13sa.1s Kv O O Transformers DD Transformers ty;;;iT e Aux ty;;;g o Aux Gen DESF Gen DESF RWS 0 TRFS .

P.!i!9. O TR FS O O Reactor Reactor

' O Mech & Elect 81 Mech & Elect gg Auxiliaries [p Auxiliaries s Blda

_73 Bido

. $1 E

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STP 345 KV SWITCHYARD .

AFTER TWO UNIT OPERATION 130 KV TO 8LES$4NG (CPL)

Blessing Lon Hill Skyline Hill Country Holmen Velasco No. 2 d' D L) (CPL) (CPSI (CPS) (COA) (HL&P)

& M 83*"

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g i Sout Texas Prosect 345 KV Switchyard <

9 r

To Unit 2 13.8 KV Emers Supply Unit Aux. Unit Aux I" lu / Xfer I "A / Xfmr 44ain XM ""T" "ar" Xfm a"r"n av"n

%a Standby u a Standby Gen. Skr.[ ] T T xw 2 Gen. Skr.t j T T "" '

' ' I To Unit 2 1 r 1 r 1 r g  % , r , r ,, 13.S KV 8uses Gen.

(f),r ~

4 Gen.

Unit 2 1r , To Unit 2

, Unit I 1r f --

To Unit 1 To Unit 2 13.8 KV Buses i3,3 g y g ,,,,

13.8 Aus Buses

6

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i CP t SITE DISTRIBUTION . .

i Summary: _ ,

. Nine transmission circuits /four grids

. Generator breaker

~

. Transformer capacity

. Independent ESF feed ,

Ali power system alignments from control room l

l STP I

____y._____ ..,. ,_ ,--.-----.,-( .

.. ___.__-.____y .u, .- ... . .._.-. m - _. .. .. .. -_ ._ . . _ . . , _ _. _ -____ _ _ ,-____ _ _

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C) l DRAFT REGULATORY GUIDE i

e STP would fall into the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> category to withstand a station

' blackout e STP can withstand at least a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> station blackout i

sTe l .

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! STATION BLACKOUT .

Study results:

. 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> auxiliary feedwater available

. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> class 1 E battery available~

. RCP seal leakage 25 gpm maximum j e RCP seal cooling available from BOP diesel

. Additional power ~possible

- 5 BOP batteries

- 5 BOP diesels onsite

D .

CP f .

DIES L GENERATOR FUEL OIL STORAGE TANK SK1mc Gemagnes panom l

i..

O O O'

^

DIESEL FUEL OIL STORAGE TANK o Each train of diesel fuel oil storage tanks for the three 1 sprinklers trains are located ~in C .

. separate rooms enclosed .

~

by 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated fire

. ;l barriers

. g

• Fire protection is provided

^^^

by automatic foam-water '

sprinklers. Continuous ventilation will be stora$ "

-, Provided under normal  ;

Tank Conditions.with fans wastes $ eIi "f' ~ === powered by ESF electrical

- u u u u u buses, w,th i damper closure 7"" '* '

cennNEF .

under fire conditions e Room drains will effect '

removal of foam-water g oiesei-senerator

~

and fuelleakage to the -

___ oily waste system. Tank i$bN!k.Ih3

• sump to oily waste jyd;; drainage is provided for by -

a separate connect, ion O e f STP I

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l TANK ROOM .

l l

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!  ;. Continuously ventilated;from ceiling to floor to remove potential fumes

. Vent fans on 1E bus e Foam-water suppression

. Watertight, locked doors l

. Tank level monitored l

STP 6

it

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MAIN FEEDWATER SYSTEM l

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l MAIN FEEDWATER SYSTEM i IVC RCB

) --> Class 2 Retirc. for  %

1 cleanup to y 8 AFW conden w ^ 8 Steam lg U$ N gQ Gen.

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EUROPEAN EXPERIENCE .

Equipment European Plants e 14-foot Fuel Design D, T, P, 5 e Model 100 RC Pump D,T,K e Model E Steam Generator D, T e Rapid Refueling Head Package D, T Desian Aspect e Three Train Systems D, T e Qualified RHRS Inside Containment D, T, P, S e Backup Power for CVCS PD Pump -

P,5 e Steam Generator Sludge Lance Ports at Preheater D, T Legend: D-Doel 4/ Belgium

~ T-Tihange 3/ Belgium P - Paluel/ France 1,2,3, & 4 5-St. Anton/ France 1 & 2 K- Krsko/ Yugoslavia STP 9

.o 6 CV l.

GOOD OPERATIONAL E.XPERIENCE DURING i DOEL 4 STARTUP e Rapid refueling performed e Moisture carryover testing shows low carryover s Natural recirculation test completed - no problems e Physics tests - results as predicted l

s STP 1 , , .

STEAM GENERATOR MODIFICATIONS i

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WESTINGHOUSE STEAM GENERATOR COMPARISON '

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e Tube expansion to'160 tubes:in preheater Expansion complete, NDE complete Acceptable startup and power run at DOEL 4 In tube accelerometers at DOEL 4 show low vibration levels that are well within guidelines

)- -

No long-term wear problems predicted e Moisture Separator Modifications Completed DOEL 4 carryover tests good '

• Sludge lancing ports added to preheater area .-

l -

Similar design as DOEL 4 Similar locations as DOEL 4

• Tube sheet expansion area has been stress relieved using

Rotopeening Pro. cess l -

This is same process used at DOEL 4 Primary stress corrosion will have reduced potential STP-

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ADDITIONAL DESIGN FEATURES

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! ADDITIONAL STP DESIGN FEATURES ~

l e Full flow deaerator l

~

l e Separate startup FW pump i

e Full flow polishers /precation and mixed bed j .

e Feedwater and auxiliary feedwater anti-waterharnmer design features L e Elimination of copper materialin feedwater/

l condensate train including. Titanium condenser i tubes e Rapid refueling e 14-foot fuel design STP

/

---,e -- -_ - - - - - , - _ _ _ . _ _

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, NUCLEAR '

l ASSURANC. E 4

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soumiums P R.O...J E...C,_T

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NUCLEAR ASSURANCE .

! e Operations QA l

e Independent Safety Engineering Group (ISEG) e Safeteam o Fitness for duty STP

c OPERATIONS QUALITY ASSURANCE .

Nuclear Group Vice President Manager Nuclear Assurance .

I Plant Interface Operations

~

i Manager QA Manager Coordination .

1 I I Quality Quality Eng,ineering ,

Control e Procedures e inspection / Witness e Audits /Surveillances e Trend Analysis STP

.c 4

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i L OPERATIONS QUALITY ASSURANCE l

i Nuclear Group Vice President l

l Manager l Nuclear .

l Assurance i I i .

I Plant Interface Operations Manager QA Manager l Coordination

! l l l .,

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i i

Quality .

ouality Control f

Eng.ineering '

e Inspection / Witness e Procedures e Audits /Surveillandes e Trend Analysis STP s

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! OPERATIONS QUALITY' ASSURANCE i

Nuclear Group ,

Vice President l

_ i l Manager

Nuclear l Assurance i

Plant interface Operations

]

Manager Coordination QA anager .

I

! I I

! ouality Eng.ineering Quallty '

Control i ,

e Procedures e Audits /Surveillances e Inspection / Witness l e Trend Analysis ,

i STP

2 t

OPERATIONS QUALITY ASSURANCE i

Nuclear Group i

l Vice President -

I .

Manager Nuclear ~

Assurance I

Plant interface Operations Manager QA Manager Coordination interface / Coordination: Communication channel to assure consistent interpretation and implementation of HL&P quality philosophy STP i ,

, ., . . ~

l OPERATIONS QUALITY ASSURANCE l

Nuclear Group Vice President ,

l Technical Services I  ! e Planning for major modifications /

Manager - outages l

Nuclear e QE / QC effort for major Assurance modifications / outages e Design office quality assurahce l I , o Procurement quality assurance Plant Interface Operations e Vendor control activities Manager O^ """U

Coordinatioin I

i i I i

Quality Quality Eng,neermg i Control i

j e Procedures e inspection / Witness j e Audits /Surveillances i

e Trend Analysis, i } STP

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i OPERATIONS QUALITY ASSURANCE i

l  ;

! Nuclear Group '

Vice President -

l -

I Manager Nuclear

] Assurance

! I Plant interface Operations Manager QA anager Coordination

! l

! I I

Q Quality Eng,uality ineering Control e Procedures e inspection / Witness e Audits /Surveillances e Trend Analysis STP

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l OPERATIONS QUALITY ASSURANCE i  ! i Operations QA Manager = 23 years nuclear QA experience 10 years operations experience 1 l

j professional engineer 32 People = 101 years operations experience

! 16 of 32 have be'e n reactor operators J

. 6 Supervisors = 32 years operations experience l 4 of 6 are degreed 4 of 6 have been military plant operators .

STP I .

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) STP NUCLEAR ASSURANCE 1

Nuclear Group

) Vice President I

Manager  ;

) Nuclear j Assurance .

l

. Safeteam I ISEG i

I I I Operations Quality Technical -

Assurance Services STP 4

j

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INDEPENDENT SAFETY ENGINEERING GROUP t

(lSEG).

{ .

l e To be created in 1987 i

e Staffed by 5 senior level, experienced engineers l . Responsibilities

-Provide continuing systematic and

, independent assessment of plant activities,

) including maintenance and modifications 4

i - Perform observations of plant operations and maintenance activities for additional

) verification of proper conduct 1

STP l I

l O' .

O O

INDEPENDENT SAFETY ENGINEERING GROUP

(ISEG) - .

t .

e Perform review / analysis of selected modifications

~

e Perform review / analysis of

~

4 se!ected problems from other plants i

! e Perform root cause analysis of j selected problems at STP .

STP I

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STP NUCLEAR ASSURANCE .

i t

l Nuclear Group l Vice President

Manager '

Nuclear 4

. Assurance .

i -

I Safeteam (1987)

ISEG i

I I

i l .

I I -

Operations Quality Technical Assurance Services I

STP

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O' O*

I j SAFETEAM PROGRAM

! Definition:

i .

An administrative program for the purpose of providing a forum for STP employees to .

identify any concern they may have in the areas of nuclear safety or quality STP l

a

, -v- - - __ _ _ _ _ _ _ _ _ _ _ . _ . _

c.

1 O O O' i

SAFETEAM PROGRAM 1

Objective:

To encourage employees to come forward -

l with their concerns now, so that we may l

investigate the concerns, correct any deficiencies, and report back to them

Confidential- Anonymous l

l 1

STP i ,

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. OPERATIONS QUALITY ASSURANCE Nuclear Group

{ Vice President i

1 I I

i Manager Nuclear l Assurance

i. l i

Plant Interface Operations Manager O^ ""*9

Coordination I

! .I i

l I

Quality Quality Control Engineering e inspection / Witness e Procedures e Audits /Surveillandes e Trend Analysis i STP I

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ENGINEERING ASSURANCE PROGRAM

~

i' Purp'.ose: Provide on-going real time review of design ac't ivities -

e To influence future design activities

~

i l e To confirm the adequacy of the design and design process l

e Established 1982

e Performed first review in 1983 e Substantial participation by Stone and Webster Engineering Co. personnel e 35,000 manhours expended STP I

O O O' 1 i

ENGINEERING ASSURANCE-Complete

. - Soil-structure interaction analysis and seismic design

- Design control and design verification '

- ASME pipe stress analysis .

- Containment analysis

- Environmental qualification of equipment

{ - A'SME Ill pipe support design l - Control room HVAC system

{ - Component cooling water system l

- Offsite AC power and medium voltage AC & DC battery power l supply systems -

l - High-energy line break analysis l - Class l ASME pipe stress analysis including handling of  ;

l Westinghouse loads by Bechtel l

1986 -

- Separation and fire protection criteria

- System walkdowns to assess system interactions, seismic il/l, etc.

l STP i  !

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O' ENGINEERING ASSURANCE ,

4 i

Conclusion:

i i

1 ,

i An additional step taken by HL&P to ,

increase our level of confidence in the adequacy of the design and design process for the STP .

Implemented during Construction Phase; .

carried forward into Operations Phase STP l 8 .  !

F

O O G

! STP NUCLEAR ASSURANCE I .

Summary

! e Operations QA staff.in place -

e Technical Services provides support to l implement a comprehensive QA program e Safetearh in place .

e ISEG plans formulated and scheduled e Plan to integrate the Engineering Assurance function into Quality Engineering underway STP

- c.

l

! O O. C,Y f FITNESS FOR DUTY

)! ,

f e Based on Edison Electric Institute .

(EEI) guide

~

! e Ten key program elements j

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! ~ FITNESS FOR DUTY -

i l . Written policy . Union briefing l . Top management support e Contractor notification  :

\

i e Policy communication . Law enforcement liaison

. Behavioral observation e Chemical testing '

l training i I

e Implementation training . Employee assistance programs STP

l 'o '

d CP J. FITNESS FOR DUTY l

{ .

l e Written policy e Union briefing l e Top managementsupport e Contractor notification ,

i l e Policy communication e Law enforcement liaison

! e Behavioral observation

. Chemical testing

! training l e implementation training e Employee assistance

! programs 1

STP 9

e .

l O O O" FITNESS FOR DUTY .

. Written policy e Union briefing l e Top management support e Contractor notification l

o Policy communication e Law enforcement liaison e Behavioral observation . Chemical testing I training -

! e implementation training e Employee assistance

. programs 4 .

b

, STP t .

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.i . FITNESS FOR DUTY TOTALS .

j .

.MAY,1986

) Total Number of Personnel Sampled 5,343 ,

, . / - 1 l ,-. c- .

Total Number of Personnel Failing

'130 .

e F i ,

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SUMMARY

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l . e Based on eel guide

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l e Implemented late 1985 i

o Chemical testing began 1-1-86 d

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3 OPERATIONAL PHASE .- .

ORGANIZATION

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4 HOUSTON LIGHTING & POWER COMPANY

! Sen.ior Execut.ive Organ.izat. ion i .

i  ! Board of .

l; I Directors i

i i

I I , Chairman and l '

Chief Executive l Officer 1

i D.D. Jordan

! I

' l .

Pres, and Chief Operating Officer I D.D. 5ykora I

I I I I l Group V. Pres. Group V. Pres. Group V. Pres. V. Pres. Senior V. Pres.

! Admin. and External Power Finance Gen. Counsel Gro" 8*

Support Affairs Operations Comptroller Cor orate .

N a

! E.A. Turner RJ.5nokhous D.E. Simmons J.S. Brian H.R. Kelly l J.H. Goldberg .

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NUCLEAR GROUP ORGANIZATION .

1 4

Nudear Group 4

l l l l l l ___1 l

Nuclear Corporate l Nudear Nudear Engineering SPecial

! Licensing NSRB Matrix Operat, ions Assurance and Assignments Support j Construction and Services 1

STP i

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NUCLEAR GROUP ORGANIZATION Nuclear Group

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l 1 I Co rate Nuclear ,

Nuclear ucensing J"','a'*<'e """"/"'

Ass g ents NSRB $

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Operations l

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1 I l Nuclear Nuclear Nuclear Plant Tra. .ining Secur,ity Operations 1

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NUCLEAR TRAINING DEPARTMENT Vice President Nuclear '

Operations I

Manager Nuclear Training I .

I I I Operations StaffTraining s Progra Design Traming Division Division Evaluation Section v

e Licensed Operations Training a GeneralTraining a Program Academic e Non-licensed Operator e Technician Training Soundness Training e INPO Accreditation

e Maintenance Training eSTA e instructor Certification e Engineering Training s SimulatorTraining e HRD Coordinator e Requalification Training

, , STP i

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Computer C Lab ................... Room tr /1 1 ~'.................::!!"i C - - e- l 3 n -

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[ l Mechanical i */SimulatorCentrol'd g sf1] 37 Room Simulator Offices l Lab "-j--------------------.

,8 i..i ij 1&C Lab 3  :: l i j ; fl

:  :: 11 _

, t.......... ....... .....'  ::

M n II !! m h C

7 l' 2 c

l' I Classrooms

" "" i Recept. ion g _

i Area Chem.is try Lab. #b 3 Library ,

6^ n.n.n so nn , ,

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5 Lunchroom Offices , y l

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oo NUCLEAR TRAINING FACILITY I _ _ _ _ . _ _ _ _ ____ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _

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1-j HL&P NUCLEAR TRAINING PROGRAMS l 1. Non-licensed Operator 11. Requalification*

-Plant Operator 12. Supervisory Skills

• l

-Chemical Operator 13. Simulator Training *

2. Reactor Operator (RO) 14. General Employee Training

3. Senior Reactor Operator (SRO/SS) 15. Instructor Certification

4. Shift Technical Advisor (STA) 16. Fire Brigade Leader S. I & C Technician l 17. Performance Technician i

6. Electrical Maintenance

18. Emergency Plan Training

7. Mechanical Maintenance

19. QA Training j 8. Chemical Technician

20. Erigineering Support Training i 9. Radiation Protection Technician

! 21. RMS Tra. .ining '

10. Technical Staff and Managers

22. PWR Fam.l.i iarizat. ion

• Will be accredited with licensed

. programs STP

(-

i i

NUCLEAR SECURITY DEPARTMENT i

i

. i i

V.P. Nuclear Operations Manager Nuclear -

Security Physical Protect, ion Safeguards Services Services l STP i

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l NUCLEAR GROUP ORGANIZATION i

i-i Nuclear j Group -

y  ;  ;  ; ----- q Nuclear Corporate Nuclear Nuclear Nuclear l Engineering I Special Matrix Op,erations Assurance

and Assignments NSRB

, Suprt aM l.ICenSing j l construction services i

l ,

e NRC and State of Texas licensing activities and interfaces .

I i

e Licensing commitment tracking j e Operational experience review l e Maintenance of FSAR I

e Dissemination of new or revised licensing requirements

• e Preparation of comments on proposed rules, regulations,

! and policies of the NRC l .

)

I i

j STP

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NUCLEAR GROUP ORGANIZATION Nuclear j Group i

! l l l l I I I l -~~~~~l I- Nuclear Corporate Nuclear Operations Nuclear Licensing Nuclear . Engineering and Special Assagnments NSR8 Matrix Su"" rt and Assurance construction 5ervices i

! I I i

Quality Safeteam ISEG Assurance (Employee Concerns) i STP l

I I

i

~

l NUCLEAR GROUP ORGANIZATION Nuclear Group I

i l 1- l l l l l ----- l Nuclear Nuclear Nuclear Nuclear speciai "5""

(gr,*t*,

^55i9am*at5 Operations Licensing Assurance Engineering & suced'*""

Construction I

i i I I I

! Engineering & RMS / DC & Nuclear Adm,inistrat, ion Construction Information Engineering

. Management Processing e Physical design e Records e Reliability analysis e Site operations management e Plant analysis support e Document control e Fuel planning & supply e Construction e informa, tion e Core physics &

management processmg performance j e Spare parts support e "Q" list .

} e Decommissioning i -

1 ,

- _m - .-

i 1

1 I

NUCLEAR GROUP ORGANIZATION t

. Nuclear Group

! l l l I l l l  ; ------]

Nuclear Nuclear Nuclear E ng Corpor, ate operations ucensing Assurance aM SpeCial NSRS Matnx l

construction Assignments '"EN v ""

I i

e i

l I

i l

i STP

l -s -

1 i

NUCLEAR GROUP ORGANIZATION Nuclear Group I ---------

Nuclear Corpor, ate Nuclear Nuclear Engineering Special Matnx Nuclear tand 5 Operations Licensing Assurance and Assignments NSRB sugC'5 j Construction l

2 e independent review and responsibilities are specified in l tech specs i

! e Will be established in early 1987 e Will consist of a full time director, members from senior management, and consultants as necessary to provide expertise specified in FSAR STP t

i l __ _ _ - -_ _ _ ______ _ _ _ _ _ _ _ _ _ _

I; O -

. Y O i

i i

NUCLEAR GROUP ORGANIZATION Nuciear i Group

. I i

I I I I I I _________l Nuclear Nuclear Nuclear E r og Special Corporate Operations Ucensing Assurance and Assignments NSRB Matrix Support and Services i

j .

• Purchasing l . e Stores l e Accounting i e Human Resources l

1 STP I

{ r. .

l.

j g, i i i

i  !

1 l  : STAFFING PLAN NUCLEAR OPERATIONS PHASE t

2000 Legend:

Unit 1 Unit 2 Fuel Load Corporate Matrix Sup & Ser 1800 -

Fuel Load -

i i V Y Nuclear Assurance & License i

1600 -

1 1 k 3 l t i

, Nuclear Eng & Construct. ion '

I -

1400 -

l jjjjjjjjjj NuclearOperations l j I

I I l

e 1200 -

I I

E M

! o I M -

5 1000 -

~ '

a. l 3 4, ,

h!!!!!.

s ewssi i ,'  :

l  :!;g ... ;;j b.

~

600 l ,

l j

! l!@::j i

f lllll:l

.i l

l

! l '

i i  !!  ! l  !  !  !  !  !  !

! 400 -

I ,!.! ! ! .

!! l l l

i i  :

i  !  :  !;;i : -! j l  :::  :

1 i ll

! li  : li:::::  !

l 200 - l  ;

j l

l.

I l .

!l l

l flf-;'f:!

l. .

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, l l  !!!!:!!!!!  !; li:!:!:!:!

l 0

1986 1987 1988 1989 4 i

l i ,

STP 1

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1. e O O Cr

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! OPERATIONS DEPARTMENT i

i I

50U1HTOM5 O

-- PR.C. HE..CT S 2o

+-

mg RE s*

3*

=

) -

I

')

l I

i NUCLEAR PLANT OPERATIONS DEPARTMENT i

i V.P. Nuclear

• i Operations I

, Mont Manager i

South Texas Project t

i I

't I

i Plant Superintendent ,

l I I I I Reactor Che ca Technical Management Facilities Health & Safety gs Maintenance Outage SeMces Operations gTn,alysis Support Manager Serwces Services Manager Manager Manager Manager Manager Manager Manager

)

i i

) Total Department Staffing 847

Unit One 0.L Staffing 652 Current Staffing 491 STP ,

~

C .

G TO KEEP ABREAST OF INDUSTRY OCCURRENCES AND .

APPLY WORTHWHILE EXPERIENCES TO IMPROVE PROGRAMS e Studied organizations of other utilities e Reviewed studies and reports addressing '

weaknesses and strengths l e Developed organization providing strength ,

and flexibility  ;

Line responsibility backed by support 1 organization l -

Production unit Support divisions STP 1  !

I - - ,- -, - - - - - - - - _

! O '

O Q)

PLAN FOR, DEVELOP, AND RETAIN QUALIFIED AND TRAINED PERSONNEL t 1

HL&P recognized the need: .

e To develop its operating staff early to involve employees in the construction and testing program to gain experience i

e To ensure that those employees who would I operate the plant would have sufficient time to develop the procedures they would use-and sufficient time to learn the plant e To minimize dependence on contract l personnel -

STP

l .-

O O Cf NUCLEAR PLANT OPERATIONS DEPARTMENT i

V.P. Nude.r Oper.tions l

M.nt M.n.ger SouthTex.s Project i

I Plant Superintendent

! l l I I I I 7 (ha3(*'

l Reactor gr.'.*$,s 4,, ,, ugntenance ugt rges oo g He.ith fety Operations u.n.}er u. .ger N' u.n. ,er u.n. ,er ** u.n. ,er l Manager i

STP

p. .

REACTOR OPERATIONS DIVISION .

Responsible for the operation of:

. Nuclear steam supply system

. Safe' guards systems -

. Turbine-generator

. Support auxiliaries l STP l .,

\ .

- 1

?

\ O .

0 O .

l. i

{

I REACTOR OPERATIONS DIVISION l.

Reactor Operations i Manager  ;

s I I I I Unit 1 Unit 2 Support Outage Operations Operations Operations Operations Supervisor Supervisor Supervisor Supervisor I I I I Shift Shift Shift Shift i . Supervisors Supervisors Supervisor Supervisor Total Division Staffing 125 Unit One O.L Staffing 106 Current Staffing 97 i

I STP J

l i l i

l o' -

d~ c?

SHIFT COMPLIMENT CHART '

No. of One Two i Unit Units Position Shifts Shift Supervisor l' Licensed SRO) 6 1 2 Unit Supervisor (Licensed SRO) 6 1 2 Reactor. Operator (Licensed RO) 6 3 6 .

Reactor Plant Operator 6 4 8 Administrative Aide 4 1 2 shiftTechnical Advisor .

1 1 STP -

I j i

' \ k O O O'

~

i -

l REACTOR OPERATIONS DIVISION l STAFF EXPERIENCE i

! NUCLEAR EXPERIENCE: 281 Years i

! COMMERCIAL EXPERIENCE: 73 Years NUCLEAR NAVY EXPERIENCE: 30 of 39 .

4 Candidates l

STP

o. o o' REACTOR OPERATIONS DIVISION PREVIOUS LICENSES -

Division Manager / Operations Supervisors

.

• SRO License on Large PWR -- 3 Shift Supervisor Candidates e SRO License on Large PWR -- 6 Unit Supervisors e RO License -- 4 e SRO Certification -- 1 e RO Certification -- .

3 e SRO License Research Reactor-- 1 STP

d d CT' t

I NUCLEAR PLANT OPERATIONS DEPARTMENT:

- t

^- k V.P. Nuclear  !

Operations, i

I Mont Manager .

- SouthTexas Project e

I , ,.

' ~

Plant -

-J ~

Superintendent -

r I I -Q I Reactor 7"hai(*i M*"*9"a'a' > F8(iii'i'5 Health & Safety Operations Chemical - Support Maintenance SeavKe5 Sdfvices outage Ser m s Ma"*ew Manaeer Manager Operat. ions & Manager u ,,,,,, u ,,,,,, u ,,,,,

Analysis -

Manager STP

,6

c:. - i 4

CHEMICAL OPERATIONS AND ANALYSIS

~

DIVISION

~

r p ,T T

\

Responsible for the operation of the: .

e Water production plant e Condensate polishers and regeneration .

systems e Radwaste processing systems e Wat.er production support systems e Waste processing support s9 stems Responsible for analyzing and maintaining chemical specifications for all plant systems STP

x .

l STATION RADWASTE GROUP

~

1

' t x.

! Responsible for ensuring that: '

4 e Radwaste systems are' operated optimally .

e Other plant systems are operated by plant

personnel to minimize radioactive l

I contamination and production of radwaste '

I i

1 i STP

,,_.,.,.r ,,_ , - . , - - .

w .a m -- - c-- -

--- - ee-- -' =-- - " " - -- - ------

.! ..7  ;

2 -

.O- D l D,~  : ,

~

! CHEMICAL OPERATIONS & ANALYSIS '

~

I DIVISION ANALYSTS

! Responsible For:

i

e. Monitoring the chemistry. parameters of all .

plant systems e Providing recommendations to the reactor

plant operators and chemical operators on j maintaining systems within specifications

! e Successfully completing three-year training program l e Preparing procedures '

.e Performing tests without aid of contract j employees -

STP i  ;

J' CHEMICAL OPERATIONS AND ANALYSIS l

DIVISION SUPPORT

! Key Activites: '

. Support CO&A in program development and l system operation -

! . Program development and operation of the radiochemistry counting room i

. Development of computerized chemistry i

parameter monitoring and trending program Effluent release program -

. Ha'zardous chemical control program

. Spill prevention program

. No contract employees

STP .

l' s . . _. _ . _ _ _ .

i I NUCLEAR PLANT OPERATIONS DEPARTMENT I i

{ '

V.P. Nuclear Operations

. I

, Plant Manager south Texas Project J

l l

l l Plant Superintendent i

I

I I I I l Reactor Chemi, cal Management Facilities Health & safety Opsrations g nafysYs'P Technical "d"a'n'a"g'e"/* services services ,fa"n'a*8!, services l "*"*8

Manager Manager "*"*8 ""*8

Support _ _ _

Manager STP

0 l

@ CD SYSTEMS PERFORMANCE SECTION i

Responsible for:.

. Monitoring plant performance through:

, - Direct testing l - Observation of normal operating '

parameters through plant tours-

- Review of plant maintenance work .

requests

! .. Trending plant problems

. Monitoring equipment performance

. Conducting the plant surveillance testing program Aiding the determination of corrective actions for malfunctioning equipment STP ss ,

i

., a.

1 i ' REACTOR PERFORMANCE SECTION l Responsible for:

e Routine monitoring of core performance j e Preparation and performance of special tests l e Phase Ill startup testing program including fuel j load and subsequent tests i

Shift technical advisors will hold a senior reactor operator license

, STP 1 , ,,

o ,

O O' PERFORMANCE SUPPORT SECTION .

Engineers e Electrical systems i

. HVAC systems l

e Fire protection e Snubbertesting e Vibration monitoring

. Operations Experience Review Program Performance Technicians e Performance testing e Surveillance testing '

STP

b .

0 '

i i  ! OPERATING EXPERIENCE REVIEW PROGRAM i

l Key elemerits:

l e NRC Bulletins, Notices and Circulars since 1972

) e INPO SOER's, SER's and O&MR's since 1980

) .

e Screened by Nuclear Licensing Department l

l e Each item addressed by written Plan of Action l e Required action is tracked via computer l e QA Department verifies closure of each Plan of Action .

I STP 1 ;a i i .. _ _ _ _

1 .,

COMPUTER SUPPORT SECTION Responsible for the startup and operation of

! the plant computers:

e Plant process computer e Radiation monitoring computer e Emergency response facilities data acquisition and display system computer e Plant security system computer l

STP 1

1 1

1 . . < .

o cf o1 l NUCLEAR PLANT OPERATIONS DEPARTMENT V.P. Nuclear Operations I

Plant Manager South Texas Project i

I Plant l Superintendent l l l l l l Managemer.t Facilities Health & Safety l Chemical Technical Outa e S'"i"5 Reactor 5 S'"i"5 S'"i"5 Manager Optrations gPdna ysYs Support Manager Maintenance Manager Manager Manager Manager Manager STP l

3

, I ,

l i

i l .

i MAINTENANCE DIVISION 1

s

! 1 i

Maintenance Total Division Staffing 294 '

Manager Unit One O.L. Staffing 228 Current Staffing 155 i

~

l i

l l l l l i Mechanical Electrical Instrument Maintenance Metrology j Maintenance Maintenance and Control Support Laboratory l

Supsrvisor Supervisor Supervisor Supervisor Supervisor ,

! I

! I I I I

! Maintenance Maintenance Maintenance i Machanical Electrical Instrument Su ort Support Su

Maintenance Maintenance instrument and Control Techni Metrology and Control Technical Superv, cal Tetbortital .

i Technical Technical Technical Supervisor isor Supervisor Laboratory Supervisors Supervisors Engineers Supervisors Specialists

, Operations ater.ia l Administrative

o 6 C'

l M AINTE N AN C E .P E RS O N N E L. Q U ALIFICATIO N S NUCLEAR EXPERIENCE 1

l

! e Journeyman with commercial nuclear ,

i experience in each craft l

l e 26 Electricians with over 190 years nuclear '

experience e 35 Mechanics with over 206 years nuclear experience e 36 I&C Te~ c hnicians with over 180 years nuclear exp~erience STP

l: 0 -

O O 4 MAINTENANCE PROGRAM HIGHLIGHTS  :

e Utilize prepared procedures for major maintenance (1400) e Corrective maintenance work requests i

! e Safety related work reviewed by QA j e Broad-based preventative maintenance

program

, STP i .

6 &

- n .

0 0 .

MAINTENANCE PROGRAM HIGHLIGHTS -

l e Root cause determination

e Material control .

e Work experience i

j e Work quality STP l'

.i f

o 6

~

G I I NUCLEAR PLANT OPERATIONS DEPARTMENT V.P. Nuclear Operations

! l l

Plant Manager i South Texas Project

! l

! l -

i Plant l Superintendent

! I I I I I Chemical Technical Reactor SIg

, Maintenance Management Facilities Outage Health &

S

! 0*/n*l' e"' Ou hnapi5 r d"'n@r "*"*9*' Services Services Manager Safety l

Manager Manager Services j Manager i

j STP I

4 .

l 1

6 G

~

O -

~

NUCLEAR PLANT OPERATIONS DEPARTMENT V.P. Nutlear Operations I

Plant Manager South Texas Project l

I l r

l -

l Plant Superintendent I .

I I I I Reactor Chemical Technical Management Facilities Outage Health & Safety g, Maintenance

• M*"*9" da$'ge'r a7ge'r " *"*9" SERVICES u!'n$be"r' Ohn'affr' 5"a"nUr Manager

- STP It

s. m f l

O . .

O -

O'.

i l RADIOLOGICAL LABORATORY SECTION Responsible for:

e Radiological environmental j monitoring program s' Dose assessment monitoring program i

i i- Key Activities:

! e Environmental sampling program o Offsite Dose Calculation Manual e Laboratory and dosimetry program STP F 4

. e s . - _ _ - _ _ _ _ - _ _ _ _ _ _ _ _

RADIOLOGICAL PROTECTION SECTION Responsibilities: '

! e Whole Body Counting. Program e Respiratory Protection Program l

, e Radiation Work Permits l

e Surveys .

i l- e Calibration of Portable Monitoring l Instrumentation i

i f

1 STP i

).

i

( l . ,, .

e p.

O .

O O' AS LOW AS REASDNABLY ACHIEVABLE ALARA 5 Rem / year is maximum dose any individual will receive at STP e Review of engineering design .

Re' view plant design to assure that engineering organization has incorporated features that will reduce doses to workers Perform walk-downs to verify that design is carried out properly in construction and to find ways to minimize system interrelations that would increase dose levels j e Effective work practices l -

Pre-job planning Exposure reduction Exposure usage accountability

! Post-job review

.l STP 3 .

t

o.

O C

1. 6 i s ,

O l NUCLEAR OPERATIONS PERSONNEL ~

i EDUCATION i

o Department Bachelors Degrees Advanced Degrees l Technical Support 41 2 Reactor Operations 7 0 1 Chemical Operations i

and Analysis 12 1 l Health Physics 11 4

! Maintenance 5 0 l Management Services ' 19 3

~

Managers 2 0 TOTAL 97 10 STP

..m.  :

0' ,

6 G NUCLEAR PLANT OPERATIONS DEPARTMENT EXPERIENCE

, Current Total .

I Organization Staffing Nuclear Experience j Plant Management 2 14 Reactor Operations 94 2.95

! Chemical Operations & Analysis 67 206 Technical Support 66 161 j

Maintenance 155 464 Health & Safety Services 40 142 l TOTAL 424 1282 1

STP l

0 6 6

KEY ELEMENTS e Licensed operators on six-shift rotation e Other critical positions on five-shift rotation e Maintenance performed primarily on two shifts, five days / week e Chemical analyst coverage 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> / day, 7 days / week -

, 1 s

0 I

e

' ~ APPENDIX.XI-NRC REGIONAL INSPECTION PROGRAM STATUS 4

c.

SOUTH TEXAb l'KUJtLi ACRS MEETING ,

I JUNE 5, 1986 i INSPECTION PROGRAM STATUS CONSTRUCTION PRE 0PERATION INSPECTION RESULTS RIV CAT SALP ALLEGATION STATUS 1

CURRENT OBSERVATIONS i .

PRESENTED BY:

l LES CONSTABLE, CHIEF REACTOR PROJECTS SECTION C REACTOR PROJECTS BRANCH USNRC, REGION IV l

l  %

O .

STP INSPECTION STATUS TOTAL NRC SITE INSPECTION / INVESTIGATION MANHOURS 1976 - 1986 21,731 HRS CONSTRUCTION INSPECTION UNIT 1 = 10,032 UNIT 2 = 4,404 INVESTIGATIONS = 1,768 STARTUP/0PERATIONS = 1,479 OTHER = 4,048' CONSTRUCTION INSPECTION PROGRAM STATUS UNIT 1 = 70-80% COMPLETE UNIT 2 = 40-50% COMPLETE PRE 0PERATIONAL INSPECTION PROGRAM SYSTEM TESTING 15%

PROCEDURE REVIEWS 10%

TECHNICAL SPECIFICATIONS 20%

ORGANIZATION / STAFFING 50%

TRAINING 30%

' TRANSITION ENGINEERING OVERVIEW & MISCELLANE0US

PERFORMANCE CATEGORY-12/82 - 11/83 12/83 - 6/85 FUNCTIONAL AREA A. -S0ILS AND FOUNDATIONS 3 2 B. CONTAINMENT AND OTHER SAFETY-RELATED STRUCTURES 2 2 C. PIPING SYSTEMS AND SUPPORTS 1 2 D. SAFETY-RELATED COMPONENTS 2 2

E. HVAC NA 2 F. FIRE PROTECTION NA 2

, G. ELECTRICAL POWER SUPPLY AND DISTRIBUTION 1 2 l H.- INSTRUMENTATION AND NA NA CONTROL SYSTEMS I. LICENSING ACTIVITIES 1 2 J. PHYSICAL SECURITY NA 2

! K. TRAINING NA 1 L. CORRECTIVE ACTION REPORTING 3 1 l

M. DESIGN AND DESIGN CHANCE CONTROL 2 1 l

N. MATERIAL CONTROL 3 2 0,. QUALITY PROGRAM AND ADMINISTRATIVE CONTROLS 3 2 ,

~

l P. PRE 0PERATIONAL TESTING NA NA

I

~

O 1 I

ALLEGATIONS i CURRENTLY OPEN 19 NEW ALLEGATIONS JUNE 85 - MAY 86 24 4

O

a

~

NRR STAFN PRESE 1 a APPENDIX XII

{1

'l/1RR PRESENTATION ON SOUTH TEXAS

,,7

- i t

(

s, g )

,a o A

y.

SUBJECT:

SOUTH TEXAS PROJECT, UNITS 1 & 2 8

'+

JUNE 5, 1986 DATE: -

PRESENTER: N. PRASAD KADAMBI I A

,y i

es 6

PRESENTER'S TITLE / BRANCH /DIV: . PROJECT MANAGER

?ROJECT DIRECTORATE #5

'DrQSION OF PWR LICENSING-A PRESENTER'S NR,C TEL. NO.: (301) 492-7272 s .

t SUBCOMMITTEE: FULL COMMITTEE i -

i' ~

Y t

l 1._ _ _ .__ ..,___ _ _ _.,.,_ . -

s ,

\

4 , - 4. $ '

ACRS COMMITTEE MEETING, SOUTH TEXAS PROJECT

, di

\

SUMMARY

dF SUBCOMMITTEE MTG

'e 5- , .C

(' e THE STAFF' PRESENTATION CONSISTED OF REVIEWING

("v THE ISSUES ASSOCIATED WTTH LICENSE CONDITIONS, OPEN ITEMS AND CONFIRMATORY ITEMS,

>9

, g STAFF WAS REQUESTED TO PROVIDE CLARIFICATIONS AND/0R ADDITIONAL INFORMATION ON FIVE QUESTIONS, h

.,[ . _

, , STAFF WAS REQUESTED TO PROVIDE CLARIFICATIONS AND/0R ADDITIONAL INFORMATION ON FIVE QUEST!0NS, s

4  % .

g; '"40 OTHER COMMENTS WERE RECEIVED ON THE SAFETY EVALUATION REPORT FOR SOUTH TEXAR (NUREG-0781).

, C-s s

i

I Of _ . ' ,.., 1 ,

s_

i.

>g) ,

. * ,\ < ,

_- Y*

1

,~ '( ~

\

_N Listing of open items ~ [f ,/

(V 4 y Item,i y .

N SER section'

/

(d Untirria1\ flooding analysis <N.

3.4.1,9:2.7,9.3.5 (

s. '

'~ ~

t . >

'(2) gliitunaf aissiles analysis .7 -

3:5.1, 10.4.9 4 y \% ~

'(3)' . Staff review of jet impinge ent from 3.s.15 )

high energy pipe failures

• p' (4) Equipment qualification i .

s

./, j A (a) Seismic and dynamic qualification) N 3.10.1 Q y (b) Pump and valve operability assurance v$ 3.10.2 3.11.3 (c) Environmentalequjpmentqualification.

(5) Preservice inspection /iriservice' inspection 5.2.4, 6.6.1

,_ 3  : proggm review 1

j j ' '

(6) Design, verification, and validationto.f 7.1.2 qualified display proce'ssing system 1 -

1

[f

.1 3 7.3.2.5 '

] i(s ),',i Acceptability

.and non-safety systems of isolation between .i <

safety .' ,'

, /r.. - < .' , ,,

(8) . Confctmanceito RG 1.97 -

3 . 4a 7.5.2.4

)

s f(,

1 4 ,

Q ;, +

(9 , Test results of aluminum-sheathed and J w 8.3.3.3 1'

<S[copperisheathcdcable~f ,

b ' ,

t

/1Gh Maximum available fault currents at 8.3.3.5 h J,f' electrical penetrations 4, 8o (10 Safe and alternate shutdown systems

. 9.5.1 i .

sS~ ,s

( (12) . Auxiliary (dedwater system reliability 10.4.9

. c.

., study l,f'.

r/ ,

M (13) . Eriergency pianning ' l, +

13.3 X; ~'

.y , a (14) Industrial security 13.6

,. (15) Analysis for boron dilution event x . 15.4.6 c 3 during modes 4'and 5' (16) -t/se of TREAT code for small-break 15.6.5, 6.3.5 /

, los's-of-coolant-accident analysis (17)i Review of submittals on Generic. 15.8.2 -

! Letter 83-28 <

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]j ' s

'),

. ' Ab 1

/ ',

RELIABILITY OF FIRE DAMPERS

\._,)) >-

t

/

gp THE STAFF IS AWARE OF RELIABILITY PROBLEMS (INF0.

NOTICES 83-69, 84-31 AND 10 CFR 21 REPORT FROM RUSKIN) gp THE APPLICANT HAS INFORMED THE STAFF THAT ONLY RUSKIN DAMPERS USED AT SOUTH TEXAS.

() gp THE APPLICANT FILED REPORTS UNDER 10 CFR 50.55 (E)

DESCRIBING DEFICIENCIES AND CORRECTIVE ACTIONS AT SOUTH TEXAS.

gp THE APPLICANT HAS CONCLUDED THAT THE PROBLEM AS BEEN RESOLVED. THE IMPLEMENTATION IS SUBJECT TO NRC INSPECTION.

t g

L- . .

DIESEL FUEL OIL STORGE AREAS I

g ACRS RAISED QUESTION IN SEPTEMBER, 1975 AND STAFF SERS IN OCTOBER, 1975 AND APRIL 1986 ADDRES9ED THE ISSUE

. . g STAFF HAS FOUND TO BE ACCEPTABLE DESIGN CHANGES RELATED TO PROXIMITY OF FilEL STORAGE AREAS TO THE CONTR'0L ROOM AND THE FIRE PROTECTION IN THE STORAGE AREAS, .

O

~

g ADEQUATE FIRE PROTECTION HAS NOW BEEN PROVIDED IN THE STORAGE ROOMS, e

O .

1 l

- - - - + - , - + w- ,.--w - , , --.w-----,-- ,------_---v-- 2-m,- ,, -- - - - , , . - - - - .-e,- ,., , - -.. , . - , - - . -- , - -- . - ,-- -

SEPARATION OF BATTERY ROOMS FROM BALANCE - 0F'- PLANT, r

i e SEPARATION FOR FIRE PROTECTION WHICH IS REFERRED TO IN SER RELATES ONLY TO AREAS WITHIN EACH TRAIN, l

e THE TERM " BALANCE - 0F - PLANT" AS llSED DOES NOT RELATE TO SECONDARY SYSTEMS OR TURBINE - GENERATOR SYSTEMS, .

LO l

\

O .

- -+--v-m-- ew-,r rvy,m--w,,-eeg---, - + w g- ee- ,,,,v,w%mewww-r-- *---w=.ee- + -e- wev=--ww------Tve- - - -- , +-'w-rT**ae *w----rr==r = = = - - 8,===--ser*-*w*** *'=-r'"" - - ' ' ' ' = ' ' ' ' ' * - - - ' -

COMBUSTION. TESTS ON IEEE-383 O OUALIFIED CABLES 9

g QUESTION: DID STAFF CONSIDER SANDIA TESTS IN EVALUATING THE FIRE POTENTIAL FROM THE CABLES, g ANSWER: THE STAFF DID TAKE THE APPLICABLE TESTS 4

INTO ACCOUNT, g RECENT TESTS IN CABINETS NOT APPLICABLE TO CABLE

TRAYS, i

e 1

l l

,- ,w-, , - - - --. - - ~ . , _ - - . ,m.,,.-,.--w. ,-,--,,m.,,we-__.

4 m RELEVANCE OF SAN ON0FRE EVENT TO SOUTH TEXAS DESIGN e

g- THE SAN ON0FRE EVENT OF NOVEMBER 21, 1985 WAS CHARACTERIZED BY WATER HAMMER PHENOMENA AND CHECK VALVE FAILURE, g STRONG DEFENSE AGAINST WATER HAMMER PROVIDED AT

' SOUTH TEXAS (SER SECTION 10,4,9) BY SEPARATE -

LINES FOR FW AND AFW, g FEEDWATER LINE WITH ESF ACTUATED ISOLATION VALVE, TN

ADDITION TO THE CHECK VALVE, PROVIDES ADDITIONAL DEFENSE.

O .

. , - - . - +-n.-. m--- -- ... - . _ - ,, , , _ _ , - - ,,,...__.,,,_..y-...,___.,,yg_yw._-- m, _ , _ _ _.x,_ ,_y, ,_, y-,_ _ . _ , , - -- - . _ _ _ _ _ _ _ - - . , -

---

+-.n -.usw w nu. w s v m .a m e - ~u--

8 POSSIBLE A'IWS EVENT AT LASALLE 00lNTY This preliminary notification constitutes EARLY notice oiSTATION, m e 6. v.UNIT 2

.. .n n.t sarety or~ 'ublic p

interest significance. The information is as initially received without verification or 4 evaluation, and is basically all that is known by the Region III staff on this date.

o Facility: Comonwealth Edison Co. Licensee Emergency Classification:

m LaSalle Unit 2 Notification of an Unusual Event

[V )

Marseilles IL 61341 , . . .

. ... _ w ..... ~ , - -. p~

n Alert Site Area Emergency h Docket No: 50-374 ,. . . . . General Emergency 9,$~0b. '

Not Applicable

Subject:

POSSIBLE ATWS EVENT o

At 4:21 a.m. (CDT), June'1, 1986 ,rh11e operating at about 83 per cent power, both feedwater pumps tripped during a surveillance test, causing the reactor water level to decrease.

AyM preliminary review of this incident indicates that the water level may have decreased to C about four inches above instrument zero (which is 13.7 feet above the top of the fuel), but which is below the automatic scram setpoint of 12.5Triches. Yet, no scram occurred (the normal operating water level is 36 inches above instrument zero). Control room operators,

, who apparently did not recognize the potential " Anticipated Transient Without Scram" (ATWS restarted the feedwater pumps in about two minutes.

The possible ATWS was not identified until about two hours after the event when an oncoming shift engineer apparently~ noticed an abnormal trace on one of the reactor water level recorders. After analyzing the situation, the licensee initiated a shutdown at 2:40 p.m. (CDT),

June 1, 1986, and declared an " Alert" under its emergency classification system on the basis of the potential ATWS at about 6:30 g,m. (CDT) June 1, 1986. ..

LaSalle Unit 1 is currently shut down for refueling.

The licensee is investigating the event to determine whether there may have been an ATWS, or whether instrumentation indicating the low water level may have been faulty.

Region III (Chicago) will issue a Confirmatory Action Letter documenting the licensees Os comitment to obtain the Regional Administrator's concurrence before restarting the unit. An Augmented Inspection Team (AIT) composed of regional inspectors including the Oyster Creek resident inspector (the Oyster Creek design is similar to LaSalle) and a representative from the Office of Nuc?sar Reactor Regulation has been dispatched to the site and is expected to l arrive in the early afternoon of June 2.

The State of Illinois will be notified. Region III first learned of this event at 5:48 p.m.

(CDT), June 1, 1986. This information is current as of 12:00 p.m. (CDT), June 2,1986. -

l CONTACT: G. Wright W. Guldemond -

FTS 388-5695 FTS 388-5574 DISTRIBUTION:

H. St. EDO MR E/W Willste Mail: ADM:DMB Chairman Palladino PA IE NMSS DOT:Trans only Com. Zech ELD OIA RES Com. Bernthal AEOD Com, Roberts .

Com. Asselstine SP Regional Offices ACRS SECY INPO NSAC CA RIII Resident Office PDR Licensee: (Corp. Office-Reac

&_Lic.Only) -

C s=Mo G

lemovetom ARSTce

' Mde acWx

- APPENDIX XIV REACIUR SCR#f AT PALISALES PIMr

O PALISADES PLANT BACNGROUND: .

SALP CATEGORY 3 - MAINTENANCE, SURVEILLANCE, QUALITY PROGRAM LACK OF AGRESSIVE CORECTIVE ACTION POOR MANAGEENT 00NTROLS ,

CYCLE 5 ECURENT EQUIPENT PROBLEMS - 1985 SAFETY INJECTION TANK SYSTEF6 (SIT)

MARCH 1986 STARTUP FROM REFUELING / MAINTENANCE OllTAGE TWO 0F FOUR PRIMARY COOLANT PLPPS WITH SEAL PROBLEMS PCS LOOP CECK VALVE LEAKAGE SIT SYSTEM VALVE LEAKAGE CVCS DIVERT VALVE LEAKAGE APRIL 10, 1985, SHUTDOWN - PCS LEAKAGE APRIL 11, 1986, DERATING - CONDENSATE PLPP PACKING FAILUE APRIL 23-29, 1986, - VALVE LEAKAGE PROBLEMS IN PCS MAKEUP SYSTEM i-I

! s i

O PALISADES PLANT - EACTOR TRIP 0F MAY 19 1986

[

9 PROBLEMS:

PtJLTIPLE FAILVES TURBIE BY-PASS VALVE FAILED TO OPEN 1 STEAM DlFP VALVE FAILED TO OPEN .

BACKPESSUE EGULATOR IN LET-DOWN LIE FAILED CLOSED 4

PESSURIZER SPRAY VALVE FAILED TO FULLY CLOSE 4

VARIABLE SPEED CHARGING PLPP TRIPPED 5 TIES ROD BOTTOM LIGHT FAILED TO INDICATE OIE R0D FULL IN i TURBIE LIFT PLFPS FAILED TO START AUTOMATICALLY l

EXISTING OUT__0F__ SERVICE EQUIPENT CONDENSATE ECIRC VALVE AUTO OPERATOR INOPERABLE BANK 0F PRESSURIZER HEATERS INOPERABLE SIGNIFICANCE Uf@ECESSARY CHALLENGES TO SAFETY EQUIPENT INCREASED BURDEN ON OPERATORS TO COPPENSATE FOR FAILED OR

. DEFICIENT EQUIPENT IIPLICATIONS CONERNING TE QUALITY OF FAINTENANCE At0 POST-t%INTENANCE TESTING

--_=

SEGUENCE OF EVENTS .

PM ON TURBINE VALVE CONTROL CABIET FANS 14:15:14 TURBIE VALVES CLOSED REACTOR TRIP ON HIGH PESSURIZER PESSUE

-TURBIE TRIP l

FIRST ATmSPHERIC DlFP VALVE OPENED, AFW PLFP P-8A STARTED 2ND ATmSPHERIC DlPP VALVE OPENED 3RD ATPOSPHERIC DlFP VALVE OPENED CHARGING PlFP P-55A STARTED (55B & C ALEADY RUNNING)

PRESSURIZER LEVEL LOW LAST LETDOWN ISOLATED CHARGING PLFP 55A TRIPPED; THIS PLFP WAS ESTARTED 4 MDE TIES TRIPPING 30 SECONDS LATER AFTER EACH START 14:22:15 PRESSURIZER LEVEL NORMAL O '

PLANT PARA ETERS:

PRESSURIZER PESSUE MAX 2245 PSIA, MIN 1689 PSIA T/ HOT MAX 594*F, MIN 535'F T/ COLD MAX 557'F, MIN 535'F S/G PRESSUE MAX 1025 PSIA S/G LEVEL DROPPED FROM 70 TO 12 KRCENT NRC RESPONSE:

EGION III ISSUED A CONFIRMATORY ACTION LETTER EQUIRING ,

LICENSEE CONDUCT THOROUGH INVESTIGATION INTO CAUSE AND IMPLICATIONS OF TE MAY 19 TRIP .

EGION III APPROVAL PRIOR TO ESTART O

w 1

CONCLUSIONS:

PERFORMANCE OF PLANT OPERATORS AE TE OPERATION OF OTER MAJOR OR SAFETY-RELATED SYSTEMS EE AS EXPECTED AND DESIGED CONSIDERING THE EQUIPENT FAILUES THAT OCCURRED.

• SIGNIFICANT EAkNESSES EXIST IN MAINTENANCE FUNCTIONS OF blAGNOSTICS, i

. REPAIR, POST-MAINTENANCE TESTING. TESE EAKESSES WERE CONTRIBUTORY TO M)ST OF TE EQUIPENT FAILUES.

EQUIPENT FAILUES AND DEGRADED EQ IPENT HAS PLACED VARYING LEVELS OF ADDITIONAL BURDEN ON Pl, ANT OPERATORS. FOR MAY 19,1986, TRIP, EQUIPENT FAILUES DISTRACTED OPERATORS BUT DID NOT SIGNIFICANTLY ,

JEOPARDIZE PLAhT SAFETY.

O

• PLANT OPERATORS HAVE SERIOUS CONCERNS EGARDING TE ADEQUACY OF MAINTENANCE ACTIVITIES AND EQUIPENT ELIABILITY.

l O

APPENDIX XV SNUBBER FAIIDRE AT TPG AN TROJAN - REPFATED SNLE ~

JLE,1985 (T. CH g

U PRORLEM: STEAM GENERATOR HYDRAULIC SNLEBERS LOCKING UP WHEN NOT DESIRED

. SIGNIFICANCE: OVERSTPESSING PORTIONS OF THE RCS PIPING BACKGR0lND:

1985 FEBRUARY - ISSUANCE OF SNLEBER TECHNICAL SPECIFICATI0FS APRIL - SNUBBERS ItsitU FOR THE FIRST TIE, 2 SG HYDRAULIC SNUBBERS Ttsitu; BOTH FAILED, ALL 16 SG SNUBBERS DECLARED INDPERABLE, FAILURES ATTRIBUTED TO INAPPROPRIATE ACCEPTANCE CRITERIA FOR CONTROL VALVES APRIL - HOT LEG (T0 SG "B") PIPE WHIP RESTPAINT LATERAL IUBER

^

WAS FOLND PULLED FROM THE WALL 1986 JANUARY - LER 85-13 STATES THAT SNUBBER LOCKUP MIGIT HAVE CAtEED OVERSTRESSING OF "B" SG HOT LEG ELBOW, O* APRIL - OllTAGE INSPECTION REVEALS 11 0F 16 SG SNUBBF.RS FAILED TESTS; ATTRIBUTED TO CONTROL VALVE DESIGN DEFICIENCY FOLLOW-tP:

PT PERFORWD ON "B" S6 ELBOW TO PI WELD, NO INDICATIONS F0lFD LICENSEE, NRR AND REGION 'I WALKED DOWN RCS PIPING; EVIDENCE OF RESTPAINED THERMAL GROWTH OBSERVED UT PERFORE D ON ALL 4 HOT LEG ELBOWS, NO IF0ICATIONS FOUND SNUBBER CONTROL VALVES REPLACED WITH A NEW DESIGN LICENSEE TO MONITOR THERMAL GROWTH DURING HEAT-UP AND OPERATION NRR TO PEVIEW LICENSEE'S sixESS AND FATIGUE ANALYSES FOR RCS PIPING ,

O

APPENDIX XVI RBCENT SIGNIFICANT EVINIS .

I i

~ l Agenda for ALns

- (b -

Meeting on June 6, 1986 1:00 p.m.

Room 1046, H Street 7;

. RECENT SIGNIFICANT EVENTS Presenter /0ffice Date Plant Event telephone M 5/19/86 Pilgrim Single Failure Could Disable E. Weiss, IE 2 All Redundant RHR Pumps 492-9005 6/85 Trojan Repeated Snubber Failure T. Chan, NRR' 5 492-7136 5/19/86 Palisades Reactor Scram W. Hehl, Reg III #" # A AIT on site as of 5/22/86 312-790-5552 ^*4*> e 1

O i

s

-e , - , , . , - - - - , . - - . - , _ --~,-,,,,-,,,,,nar- ,mm,w.,_, _~w-,--~enm,- --*-----~erw -v

PILGRIM - SINGLE FAILURE COULD DISABLE ALL REDUNDANT RHR PUMPS MAY 19, 1986 (ERIC WEISS, IE)

PROBLEM:

SINGLE-FAILURE OF MINIFLOW LOGIC COULD DISABLE ALL REDUNDANT RHR PUMPS DURING SMALL OR INTERMEDIATE SIZE BREAK LOCA SIGNIFICANCE:

POTENTIAL SINGLE FAILURE CAUSES LOSS OF MULTIPLE SAFETY FUNCTIONS P0TENTIAL FOR NO LONG TERM COOLING FROM SAFETY SYSTEMS CIRCUMSTANCES:

LICENSEE REVIEW (PROMPTED BY INF0 NOTICE 85-94) DISCOVERED THAT SINGLE FAILURE OF EITHER MINIFLOW SWITCH COULD PREVENT ALL AUTOMATIC LOW FLOW PROTECTION FOR ALL RHR PUMPS; PUMPS COULD BURN UP IF MANUAL ACTION NOT TAKEN IMMEDIATELY DURING SOME ACCIDENTS OR SPURIOUS ACTUATIONS, RHR PUMPS WOULD BECOME DEAD HEADED FOR EXTENDED PERIOD CURRENT MINIFLOW LOGIC DESIGNED TO BE CONSISTENT WITH LOOP

SELECT LOGIC FOR LPCI

/}*

WHEN FLOW DETECTOR $ IN EITHER LOOP SENSE ADEQUATE FLOW, BOTH RHR MINIFLOW LINE VALVES CLOSE CONSEQUENCE OF RHR PUMP LOSS IS LOSS OF LONG TERM COOLING WITH RHR HEAT EXCHANGERS, AND OTHER FUNCTIONS INCLUDING:

-SUUTDOWN COOLING MODE

-d9W PRESSURE COOLANT INJECTION

-HEAD SPRAY (REMOVED FROM PILGRIM) .

-CUtiTAINMENT SPRAY

-TORUS SPRAY

-SUPPRESSION P0OL COOLING WHICH EVENTUALLY WOULD CAUSE LOSS OF:

-LOW PRESSURE CORE SPRAY 1 -HIGH PRESSURE COOLANT INJECTION i -REACTOR CORE ISOLATION COOLING

• GE FIX IS TO ELIMINATE "CLOSE" SIGNAL TO MINIFLOW VALVES; COULD INCREASE PEAK CLAD TEMP 50*F IN SOME BREAK SIZES; NRC CONSIDERS THIS TO BE INTERIM ACTION FOLLOW-UP l

()*

• IE BULLETIN 86-01 ISSUED 5/23/86 IE AND GE ARE DETERMINING GENERIC SIGNIFICANCE

• NRR WILL REVIEW RESOLUTION FOR PLANTS WITH PROBLEM, INCLUDING ..

TECHNICAL SPECIFICATION ISSUES ,;[

e SIMPLIFIED DIAGRAM OF PILGRIM MINIMUM' FLOW FOR RHR PUMPS (J .

F" 7 v

A VM VM RX LPCI PRESS. V LPCI INJECT VESSEL dlNJECT 4

_O'V'1 RECIRC.

M RECIRC.

MP P.U.g. P P.U.3-i

____ ________________, c________________ l

/ . sC. _____I OR GATE FLOW INDICATOR ji l

f TORUS I O l TORUS f I

L FLOW INDICATOR

~

1 MIN. FLOW ______ __ ____1____ __ ,_____ MIN. FLOW BYPASS VALVE BYPASS VALVE

- - - -- es s u s s u s e s 1 r i f RHR S r RHR PUMPS 1 r '1f PUMPS "B" "A"

. .

' ~

. y r M r RHR - - RHR HX "B" HX "A" - - - -

• 'EITHER SENSOR DETECTING FLOW WILL CAUSE MINIMUM FLOW VALVE TO GO CLOSED 3

TROJAN - REPEATED SNUBBER FAILURES JUNE,1985 ( T GNAN , NRR)

O PROBLEM:

STEAM GENERATOR HYDRAULIC SNUBBERS LOCKING UP DUE TO DESIGN INADEQUACY.

SIGNIFICANCE:

DAMAGE TO HOT LEG PIPE WHIP RESTRAINT (1985)

OVERSTRESSING.0F HOT LEG ELBOWS

-PREVIOUSLY UNACCOUNTED FOR MOVEMENT IN THE PRESSURIZER SURGE LINE (1982-1985)

CIRCUMSTANCES:

NRC RECENTLY. LEARNED THAT RCS HOT LEG PIPE RESTRAINT HAD

-PULLED FROM WALL IN 1985 LICENSEE, NRR, AND REGION V WALKED DOWN RCS PIPING DYE PENETRANT TEST PERFORMED ON "B" SG ELBOW. NO INDICATIONS FOUND, H PERFORMED UT ON ALL 4 HOT LEG ELBOWS AND FOUND NO INDICATIONS -

CRUSHED GRAPHITE SHIMS FOUND ON 3 0F 4 HOT LEG PIPE WHIP RESTRAINTS INDICATING HOT LEG TO RESTRAINT BINDING 11 0F 16 SG SNUBBERS FOUND TO HAVE FAILED AGAIN IN SAME WAY BACKGROUND:

1982 - LICENSEE REMOVED THE THERMAL SLEEVE ON THE PRESSURIZER SURGE LINE; HOWEVER, SURGE LINE DID NOT SETTLE OVER NEXT FEW CYCLES, AS HAD BEEN EXPECTED IN ,

H ANALYSES; MOVEMENT CONTINUED 1985 - LICENSEE HIRED IMPELL TO REVIEW THE SURGE LINE MOVEMENT; UNABLE TO ACCOUNT FOR CONTINUED MOVEMENT 1985 - A HOT LEG (T0 SG "B") PIPE WHIP RESTRAINT HORIZONTAL SUPPORT WAS FOUND PULLED FROM THE WALL O .

. f

TROJAN - REPEATED SNUBBER FAILURES JUNE,1985 ( I CHAN, NRR), (CON'T.)

O BACKGROUND, (CON'T.)

1985 - SNUBBERS TESTED PER NEW TS REQUIREMENTS

- 2 0F 16 SG HYDRAULIC SNUBBERS WOULD NOT RES. POND TO 100 KIP LOAD; SHOULD HAVE RESPONDED AT < 10 KIP;

~

ALL 16 WERE DECLARED INOPERABLE AND REBUILT '

- SNUBBER FAILURE ATTRIBUTED TO CLOGGED HYDRAULIC LINES; CLEANED WHEN ASSUMED THAT ALL SG SNUBBERS WERE IN0PERABLE,IMPELL ANALYSES WAS ABLE TO ACCOUNT FOR THE SURGE LINE MOVEMENT AND THE DAMAGE TO THE PIPE WHIP RESTRAINT THE. LICENSEE CLAIMED (1985) THAT ALTHOUGH HOT LEG STRESSES EXCEEDED ASME SECTION III ALLOWABLES, STRAIN IS WITHIN 1%

.- LIMIT,WHICHWASNRC-APPROVEDLIMITFORSONG}-1ONSEISMIC CRITERIA AND METHODOLOGY FOLLOW-UP:

SNUBBER CONTROL VALVES TO BE REPLACED WITH NEW DESIGN LICENSEE TO PERFORM PRE-STARTUP WALKDOWN OF RCS IN A HOT CONDITION NRR TO REVIEW RCS PIPING STRESSES AND APPLICABILITY AND ACCEPTABILITY OF LICENSEE'S ANALYSIS ,

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i ACRS EETING ON l

NUREG-0956 FINAL REPORT JUNE 6, 1986 .i

+

l M. SILBERBERG l

b s .

J. MITCHELL .

m s

g1 0FFICE OF NUCLEAR REGULATORY RESEARCH t

\ w .

\ . . . . . .. . . _ .- - . _ _ _ 5L - _ '

v-O O O '

i MAJOR OiANGES IN NUREG-0956

1. ADDED TECHNICAL INFORMATION

2. UPGRADED CODE SUITE

3. PERFORPED NEW SEQUENCE ANALYSES II. REMOVED MATERIAL ON RISK AND CONTAINMENT ,

5. ADDED DISCUSSION OF PUBLIC C0FfiENTS

6. REFLECTED REVISED SEVERE ACCIDENT RESEARCH PLAN

7. IMPROVED THE STATEMEtlT OF CONCLUSIONS l

I

- a '

1 Table ES.1 Major advahces in' source term technology

~

i since ' the W4Sff-1400 Reactor Safety Study

~

, P:

i c

Area of Improvement _

+

l i

1. Treatment of chemical forms of iodinetand other fission products

2. Mechanistic analysis of fission product retention in reactor coolant ,

system -

3. Improved data base for in-vessel melt:Iprogression, hydrogen generation, and control rod behavior

4. Mechanistic treatment of aerosol behavior in containment, including the .

effects of suppression pools and ice (compartments

5. Greatly enlarged data base for in-ves'sel fission product release from fuel

! 6. Data base and mechanistic treatment of core-concrete interaction and related radionuclide release a

! 7. Improved models for analysis of containment pressure loads '

h O

[,. ** STAFF RESPONSE TO ACRS IEITER OF

- UNIT DECEMBER 12, 1985 J' o NUCLEAR REGU t .. I

(; - *'

s , .v /

ADVISORY COMMITTEI msmo I I

• • • • ~

December 12, 1985 Honorable Nunzio J. Palladino Chairman U.S. Nuclear Re Washington, DC'gulatory 20555 Commission

Dear Dr. Palladino:

SUSJECT: .

ACRS CONMENTS ON NUREG-0956, " REASSESSMENT OF THE TECHNICA BASES FOR ESTIMATING SOURCE TERMS -- ORAFT REPO During its 306th meeting, October 10-12, 1985, the Advisory Comittee on

. Reactor Safeguards discussed NUREG-0956 with representatives of the NRC Staff, and5-7, Oecember we1985. completed cur deliberations during the 308th meeting, This report had previously been reviewed by a Sub- @

ccmmittee in meetings on May 2. August I and 2, and September 27, 1985.

We also in the report had Reference the benefit 6. of the documents referenced in 1-5 and disc We conclude that:

(1) Although the report is a useful description of progress that has been made in the NRC's Severe Accident Research Program, it deciding whether and how to restructure existing reg deal with accidents beyond the current design basis accidents.

(2) Since much of the motivation for the severe accident research program came from observations made after the THI-2 accident, some of which led several investigators to conclude that source terms previously used to describe severe accident consecuences

were much too large, we believe the report should either state that infomation developed to date indicates a significant difference compared to the predictions of WASH-1400, or that no

' significant difference is now believed to exist. The report is ambiguous on this point.

(3) The report is cast in a framework which depends on the use of a suite of codes to describe the course of severe accidents.

Reference the results that is the made codes to predict.the considerable No guidance uncertainty is given as to that ex how to take this uncertainty into account in making decisions related to licensing or regulation. Since* dealing with this .

uncertainty is one of the more difficult parts of the ' decision process,with dealing more it. attention needs to be given to approaches for (a)

( The dealssuite with contatnment of codes that in aforms rather.much of thefashion.

preliminary basis for the repo It Tl3

Honorable Nunzio J. palladino  :

, December 12, 1985 *

( .

appears to us that a much less ambiguous method for taking account of containment performance is needed, especially in

, light of the operating wide variety of containment types that exist in-plants.

(5)

Many of the phenomena and the processes described in this report

. have also been studied in some detail by those responsible for jrN the. Industry Degraded Core Rulemaking (IDCOR) Program.

. be valuable, in censidering the results of and the conclusionsIt would W drawn from NRC's research programs, to have some discussion of thethe by differences IDCOR group and the similarities compared of this to those of thereport.

conclusions reached Additional given in whatcomments follows. on these points and other features of the report It was be civen recognized, following the TMI-2 accident, that more attention must considered as design basis accidents.to the risk posed by accidents 'beyon It was also known that new informa-WASH-1400, Reactor SafetyAccordingly, Study. tion and new understanding ha the NRC Staff undertook to collect, evaluate, and publish in NUREG-0772, Technical Bases for Estimat-i ing Fission Product Behavior, the best infomation then available concern-corefission ing damaging product release and transport during and following a severe accident.

j On the basis that wculd be needed of that collection, by the NRC Staff and of an evaluation as it prepared to deal with ofthethe info

. severe accident issue, the Office of Nuclear Regulatory Research (RES) formulated a research program aimed at imoroving the accuracy with which j! the source term could be predicted. NUREG-0956 reports the results of that research.

l The report was described by staff members of the RES Office as contai t

the scientific bases from which source tem calculations could be made.

It places major emphasis on the assembly of a set of computer codes which have been used for computing source tems for five reference plants.

Several steps were taken to improve the validity of the codes: a vali-dation study National of the constituent computer codes done at the Oak Ridge Laboratory, a quantitative uncertainty study performed by the

• Sandia National Laboratories, and an independent review of the results of the NRC's source tem research by a study group of the American Physical Society. *

. Much of the research that foms the basis for this report was stimulated h by the investigations associated with the TMI-2 accident. Several inves-tigators concluded, primarily as a result of the radioactive iodine ectimated to have been released to the containment atmosphere during the

- accident, that the source tems calculated and used in WASH-1400 were much -

u .

, 7W

t Honorable Nunzio J. Palladino December 12, 1985

)

l of fission' products in the light of a more careful investig s chemistry and the physics of the various processes involved, I-Many of those who concluded that the source terms used in the WASH-14 calculations were too large also predicted that when more appropriately chosen could besource shownterms to be were used, the calculated risk from severe accidents %

several previously calculated. orders of magnitude smaller than those contain newly some develogied set conclusions of codes concerning the risks to be expected w incorporating the new data resulting from an extensive research program, a,re applied to the analysis of severe reactor accidents.

Conments in the report on this question are at best tentative.

For example, the report notes in the section on Risk Insights that a "com-parative risk appraisal" (using WASH-1400 accident frequencies, but source

,. terms The calculated from the new set of codes) indicates a reduction report concludes that the reduction (early fatalities are about a in risk.

divided between that resulting from a change in the treatm product release and transport, preach to describing containment behavior.and that resulting from a different ap-are ambiguous. In other cases the comments

{,' selected accident sequences for five reference plants that re reactor containment types in operation in the United States. These selected sequences the computer codes." have provided a sufficient test of the capabilities of What was the " sufficient test"? How was the ade-ovacy of the codes developed? One attempting to judge the merits of the code set, or to ascertain'whether the risks predicted in light of the new .

information that information helpful. has been developed are indeed smaller, would find more On the basis of our examination of the report, and of cur extensive discussions with the Staff, we conclude that the report can best be characterized conclusion. as a status report for a task well begun but far from

-- the question of how and for what parpose the material in the be used.

Several possible applications were mentioned;-but we were told that details of usage will be developed by those who are to use it -- that this its not report contains primarily tt.e science that has been developed, and application.

This response is understandable, given the compart-mentalization of the Staff that exists, but as a result, as uses are developed, questions will arise that are likely to require further inves-tigation or additional explication of the material that has been gathered.

We commend for consideration of the Staff the proposition that applied research is not completed until it is used.

e

. 7YE

'  % \.

(

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• Honorable Nunzio J. Palladino {

Dechr 11,1985 (

i ca 4 q

p

, y y .

o Q,

We conclude' that this repor't,

- e required in order to make its appifcation to' severe accident a feasible, can identifiable be understood only as part of a package made up of several components.

This report is one of those. It includes, or refers to, the new information thatdas been developed cortcornin 4 product release from fuel (both . in' and outside nf the vessel),gand[its fission 1 transport into containment. J3t also incorporates'the suite of cnesi

developed (as described in SM 2104/ for model1Jng itna course of severe .

{

accident sequences following"the onset of core damage.

• S

~

The new risk csiculations$o be carriedaut for six s(lected plants a be reported in NUREG-1150, Risk .Per;scectives And Rebaselining,. form i another component. The accident secuence initiator" frequencies to be'used ?

'. in this set of calculattdns ' will presumably ccme from the Accident k.' i Sequence Evaluation Program. Presumably the modelling of containment ' '

performance to be used in the calculations will come from the : Severe Accident Risk Reduction Program, although this is not clear.  ;

4 The incorporation, yet to, be accomplished, into one coherent . method, of the various approaches being developed' to descrpe cactainee,it performance

- is another, and an extremely important comp 5nent./ That (onnulation of methods for carrying out a detailed severeMecidemt anS1ysis for each k~ operating i other. ~ plant, cited in the Severe ' Accident' Policy Statement, is an- ,

r '

Judged in this context we believe the report is a uAeful addtion to theh earlier information on fission produ:t release and transport, and to the i

methods that have been used in the 'past to model the behavior and the j consequences of severe accidents. However, we conclude that the codes. it their present form, should not be given mucn weignt in making decisions. '

For example, the ' report observes that considerable uncertainty exists in j

the results to be expected when the constituent codes are employed.

i Reference is made to further work to be done in defining uncertainties.

Hqwever, no guidance is given tc the prospective iser on how to account for or how to deal with uncertainties. Nor is there any conenent on  :

l whether the uncertainty to be expected from eatployment of the suggested new approach is greater than or less than that which might be expected if, say, the WASH-1400 approach is used.

. More information on the effects of 'i the identified uncertainties is needed. Guidance on how to deal with existing uncertainties shovid be provided if the results of the report are to be used for making decisions. Furthermore, the description given to us ,

! by the Staff, of work which is planned to provi.de more nearly cuantitative ' <

estimates of uncertaintv, leads us to bef feve that what is proposed would be cetter described as a sensitivity analysis.

One of the " Sour:e Ters Insights" given in the report is that, "For most

/E accident sequences, the largest single factor affecting source terms is

% containment behavior. ~ A delay of several hours in containment failure I

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Honorable Nunzio J. Palladino December 12, 1985

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will reduce source terms significantly." .

(~ , However. the guidance on containment We agree with both statements.

behavior modelling

' Appendi,i A gives some general discussion of containment, typesis and of confusing.'

their behavior in accident situations.

of a Sandia National Laboratories' report which treats " Containmen Analysis."

It is intended to " provide a containment matrix' for the risk perspective foranalyzed" the other plants the Surryinplant and to discuss the containment behavice of BMI-2104 i

h evaluations are preifminary. conclusions are f aced with the caveats, an at variance with other NRC work related to containment For behavfer.T-example, in Appendix 8, in several places, there is reference to in-vessel -

steam explosions in a context which indicates that they are thought by the Staff to be failure.

containment a possible significant contributor to the likelihood of early sion ' Review Group However, the report of a review by the Steam Explo-convened consensus that the likelihood by the NRC Staff (NUREG-1116) indicates a of early containment l in-vessel steam explosion is so low as to be negligible. rupture caused by

(' There is also a 2

l an important containment bypass mechanism.concent in' NUREG-0955

' to deal with it. No guidance is given as to howl rp to containment system performance, and in view of the prelim 6 of current models, much more work is needed in this area. We emphasize,

decision making, that development of more elaborate co the only way ser even necessarily the best way to proceed. Some well defined method for describing containment behavior is needed.

s Bearing in mind that early coments concerning the contribution of fodine .

s" occurred, and observing that the report points to4. better fis chemistry as one of the major improvements that has been produced, we asked in the waywhatfodine changes in risk could be identified as a result of the changes is treated.

tempted to identify these changes.We were told that the Staff had not at-of the designation of the report as a scientific document,We efforts to'suggest that, identify the changes in risk due to differences in the treatment of a few utility of the resultskey contributors would add considerably to the understanding and to discuss areas .of agreement 'and of disagreementWe also believe it wu with more dis-cussion of the latter) betwern the source Tens Package rep (orted upon here and other relevant work, the IDCOR approach, for example.

~ There are several key areas in the modelling of severe accident pro-grossion as described in the report, about which we have some reser 'p vations.

l The transport and the retention of radionuclides in the primary -

system system. are tightly coupled to the temotrature distribution in the pria(ry This in turn in driven recirculation is the likely to be asystem.

primary strong function of the buoyancy ,

b This phenomenon is not -

747

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e Honorab1e Nunzio J. 11adino , _.

December 12, 1985 <

treated in the models ' described 1 in NUREG-0956.

suggests primary system. that it could have an important bearing on temperatures in For example, some investigators have sugoested that it might lead to transport and condensation of fission products in the ste generator pressure tubessequences.

accident sufficient toItproduce tube rupture in some postulated is also predicted by some that mechanism may lead to a sufficiently high temperature of thethis upper lev components of the primary system pressure sequences, before, .for example, the in PWRs, that rupture will occur, in molten core material severe containment from heating, atmosphere the bottom occurs.of the reactor vessel, leading mode of heat transfer deserves further investigation.This possibly important Fission product release frera the fuel Core Even code. melt inprocression its present form and enre melt temoeratures are based on th g tation of the physical processes,it is meant to predict.the molten , core temperature is subject codethe As a result, provides l uncertsinty is reflected in calculations to considerable uncertainty. + This' of Better! understanding of the resultant uncertainties is needed. fission, product re

1. Ex-vessel release of fission products from the melt is strongly dependent upon the meltinteraction.

core-concrete temperature, and this in ' turn is highly dependent on the @

Scme investigators interpret of Germany) to indicate that the heat transfer to con -

thatpackage.

this predicted by the code used to model the core-concrete interaction in containment fation % currently calculated to come from releasedfurther deserves during corv e ncrete interactions, this possible . discrepancy investigation.

i

(

The not some report readilyisavaf basedf able. upon work Because of thedescribed importance ofin a large numbe a thorough understanding it te of the bases of the results reported and conclusions drawn vital that care be taken to identify the documents to which a user

- c:n to to obtain further information. We emphasize the importance of' complete documentation of the foundation reports from which NUREG-0956 is drawn.

• i We have consnented in a letter to the Executive Dire dated August 13, ,

1ations to be carried out and to be reported in NUREG-1150, as well as the methods of external developed initiators.for analysis of individual plants, should take account, We express our appreciation to the Staff for providing us with thorough.

I well organized presentations on this report, and for their efforts in ,

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' Honorable Nunzio J. Palladino December 12, 1985 -'

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responding to a number of questions which we posed during the course of "

our discussions.

Sincerely,

• David A. Ward l Chairman

References:

1.

U. S. Nuclear Regulatory Comission, " Reassessment of the Technical i

Bases for Es'timating Source Tenns - Draft Report for Coment " USNRC Report #UREG-0956, dated July 1985 <

, 2. U. S. Nuclear Regulatory Comission, " Reactor Safety Study - An Assessment of Accident Risks in U. 5. ~ Comercial Nuclear Power i I

3. Plants " USNRC Report WASH-1400 (NUREG-75/104), dated October 1975 U. S. Nuclear Regulatory Comission " Technical Bases for Estimating Fission 0772, Product dated JuneBehavior 1981 During LWR Accidents," USNRC Report MUREG-4

( 5.

Battelle Columbus, "Radionuclide Release Under Specific LWR Accident Conditions," Vols. I-VII, BMI-2104, dated July 1983 - February 1985 U. 5. Nuclear Regulatory Comissien, "A Review of the Current Under - '

standino of the Potential for Containment Failure from In-Vessel Steam Explosions," USNRC Report NUREG-1116, dated June 1985

6. U. S. Nuclear Regulatory Comission " Risk Perspectives and Rebase-lining for Six Reference Plants," USNRC Report NUREG-1150, to be published O

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APPENDIX XIX ADDITIONAL DOCl2 B CS PROVIDED IVR ACRS' USE im ADDITIONAL DOClfERS PROVIDED IUR ACRS' ___ _

1. Testimony of H.11. Lewis, Subconnittee oli Energy and the Environment, House Interior Cannittee, June 10, 1986

2. Letter, J.C. Ebersole, ACRS Chairman to N.J. Palladino, IRC Chainnan, ACRS Report on the Hope Creek Generating Station, December 18, 1985.

3. NUREG-0979, Supplement No. 5, Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island Design, May 1986.

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