ML20213A522

Summary of ACRS 315th Meetings on 860710-12 in Washington,Dc.Viewgraphs Encl

ML20213A522
Person / Time
Issue date:	01/28/1987
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2442, NUDOCS 8702030324
Download: ML20213A522 (406)

Text

h hb YYh TABLE OF CONTENTS ,

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3-MINUTES OF THE 315TH ACRS MEETING  !' i Il I U JULY 10-12, 1986 onp j// gf/f'/

WASHINGTON, D.C. T 'l* 7 I. Chairman'sReport(0 pen).......................................... 1 II. Davis-Besse Nuclear Power Station Unit 1 Restart (0 pen)........... 2 III. Subcommittee Report on R&W Owners Group Trip Reduction and Transient Response Improvements Program (0 pen).................... 9 IV. Reorgani za tion o f TVA (0 pen ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 V. Standardization Policy Statement and EPRI Advanced LUR Program (0 pen) ........................................................... 19 VI. Technical Specifications - Proposed NRC Policy Statement (0 pen)... 27 VII. ReactivationofDeferredorCancelledNuclearPowerPlants(0 pen). 30 VIII. BWR Containment Performance During Severe Accidents (0 pen)........ 33 IX. Report of Subconmittee on At xiliary Systems (0 pen) . . . . . . . . . . . . . . . . 34 X. Executive Sessions (0 pen)......................................... 35 A. Subcommittee Assignments....................................... 35

1. Additional Items Relating to ACRS Report on the Restart of Davis-Besse Nucler Power Station , Unit 1. . . . . . . . . . . . . . . 35

2. BWR Containment Performance............................... 36

3. Safeguards and Security................................... 37 B . Reports , Letters a nd Memo ra nda. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37

1. ACRS Comments on the Restart of Davis-Besse Nuclear Power Station, Unit 1........................................... 37

2. Commission Policy Statement on Technical Specifications... 37

3. ACRS Recommendations on the Hope Creek Generating Station. 37 4 ACRS Action on the Proposed Revisions to Sections 9.2.?

of the Standard Review Plan (SRP)......................... 37

5. ACRS Coments on the BAW Owners Group Safety and Perform-a n ce Inp rov eme n t P rog ram. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37

6. Letter of July 1, 1986 to David Ward, Chairman, ACRS. 38 h0 4 0701g9 Ti ' ~ D M W ACRS-2442 cect1tselDy_LLjAf -

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315TH ACRS NEETING 11

7. ACRS Views on Fire Protection Research and Fire-Related Systems Interactions...................................... 38 .

8. Additional Recommendations on the Development of a De Minimis Dose........................................... 38 l 9. Emergency Planning Requirements for NRC Licensees......... 38

10. Aptitude Testing.......................................... 38

11. TVA Reorganization........................................ 38

12. Regulatory Process........................................ 38 C. Future Agenda.................................................. 39

1. Future Agenda............................................. 39 i

l' 2. Future Subcommittee Activities............................ 39' i D. Bilateral Exchange of Nucl ear Informa tion. . . . . . . . . . . . . . . . . . . . . . 39 E. EPRI Advanced LWR Program...................................... 39

F. Nomination of ACRS Member...................................... 39 I G. Death of Admi ral Hyman G. Rickover. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 l H. I AEA Mee ting on Chernobyl Accident. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 3

1. Report of the Nominating Pane 1................................. 40 i

J. Move to Bethesda............................................... a0 K. Probabilistic Assessment....................................... 40 l

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TABLE OF CONTENTS APPENDICES TO MINUTES OF THE 31STH ACRS MEETING JULY 10-12, 1986 Appendix I - Attendees................................................ A-1 Appendix II - Future Agenda........................................... A-7 Appendix III - ACRS Subcommittee Meetings............................. A-9 Appendix IV - Toledo Edison Davis-Besse Presentation.................. A-17 Appendix V - NRR Staff Briefing on Davis-Besse ....................... A-25 Appendix VI - Maintenance Review at Davis-Besse....................... A-141 Appendix VII - NRC Reassessment of B&W Plant Designs.................. A-145 Appendix VIII - NRC Reassessment Program Scope. . . . . . . . . . . . . . . . . . . . . . . . A-147 Appendix IX - Key Dates Related to the B&WOG Program.................. A-149 Appendix X - Loss of Integrated Control System Power and Overcooling.. A-153 Transient at Pancho Seco on December 26, 1985 Appendix XI - NRC Status Report on TVA Review......................... A-194 Appendix XII - TVA Briefing to ACRS, Washington, D.C., July 10, 1986.. A-200 Appendix XIII - Staff Presentation on Standardization................ A-215 Appendix XIV - Statement of F. Sears, Atomic Industrial Forum Study... A-224 Group Appendix XV - EPRI ALWR P rogram Presen tation. . . . . . . . . . . . . . . . . . . . . . . . . A-239 Appendi x XVI - U.S . Advanced LWR Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-259 Appendix XVII - DOE Advanced Light Water Reactor Program.............. A-270 Appendix XVIII - Staff Review of EPRI-ALWR Requirenents Document...... A-272 Appendix XIX - Proposed Commission Policy Statement on Technical...... A-282 Specifications Appendix XX - Reactivation of Nuclear Power Plant Construction........ A-297 Projects Appendix XXI - Policy Statement on Deferred Plants.................... A-314 Appendix XXII - BWR Containment Performance Briefing.................. A-324 Appendix XXI!! - RES/DET Presentation to ACRS, July 10, 1986......... A-335 Appendix XXIV - Additional Documents Provided for ACRS' use.......... A-339

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'b3012 \ Federal Register/ \ V:1. 51. Nr.121 / Tuesd:y, June 24, 1980 / N:tices $

tha holder of specibc byproduct material IV Thursday./uly la 1986 icense No. 37-1M52-01 (the license) \ The licensee may show cause, within d:30 A.Af.-d:45 A.Af.: Report ofACRS ssred by the Nuclear Regulatory 25 days of the date ofissuance of this Chairman (Open)-The ACRSChairman (ommission (the Commission or NRC) Order, as required by section IV above. will report briefly regarding items of pprsuant to 10 CFR Parts 30 and 35.The b9 filing a written answer under oath or current interest to the Committee.

1.jense authorizes the use of affirmation setting forth the matter of radiopharmaceuticals to perform A A . equimmentsfor

/ diagnostic procedures listed in Groups I-l!J of Schedule A, to CFR 35.100, and cisa to perform in vitro stuyies. 'I he lic: hse was originally issued on June 4 fact and law on which the licensee telles 3

to d monstrate that prohibition of this indis dual from performance or supe ision of!! censed activities in not futum StandaMonts (OpenWe members will hear and discuss a report by resentatives of the NRC Staff and the rgectric Power Research Institute d TheI " regarding the joint effort to develop a set 1979kwas most recently renawed on pro id fr$ 10 C 2. 2 , by N:vember 30,1984, and is due to expire consen g to the entry o an order in d ca W uhmds fw futum en De ember 31,1989. Dr. Salvatore E. 8t8n 8ti2*d nu 88r power plants.

Imper ale is listed on the licenhe as an substant lly the form oposed in this Order. lf a licensee fails to fl!e an 0:15 A.AI-1230 P.Af. and 1:30 P.AL-cuthori ed user oflicensed mat rial. answer within the specified time, the 2:30RAf.:PmposedNRCPolicy -

e \ Director. Office ofInspection and StatementRegarding Standardized U

i \ Enforcemerh. may issue without further NuclearPowerPlants (Open)- 'Ihe As a result of an NRC inspectiorgand notice an Orgfer modifying the license as members will consider the proposed inv:stigation at hiercy llospital in described ab ve. NRC policy statement on standardized Wilkes Barre. Pennsylvania, where r. yg nuclear power plants. Representatives Imperiale is also emp!nyed as the of the NRC Staff will make hfedical Director of Radiology and th The licensee or any other person presentations and participate in the R:diation Safety Officer, the NRC adversely affected by this Order may discussion to the degree conridered det:rmined that Dr. imperlate knew that request a hearing within 23 days after appropriate.

a di: gnostic misadministnation had issuance of this Order. Any answer to 2:30 RAL-3 JORAL: WA Nuclear (ccurred at the hospital in hiay 1985 and this Order or any request for hearing Activities (O en)--The Committee will knew that the incident should have been shall be submitted to the Director Office hear reports from representatives of the reported to the NRC, but told his staff of Inspection and Enforcement. U.S.

Nuclear Regulatory Commission. NRC Staff and the Tennessee Valley n:t to do anything te Authorit 8 ding th 'd reporting of the misabardinginistration the Washington. DC 20555. Copies also shall rewganiza of the'IVA nuclear bec:use he did not think the incident be sent to the Executive I.egal Director at the same address and to the Regional organization to deal with nuclear power w:s that serious. p! ant problems.

Dr. Imperiale admitted this in an Administrator U.S. Nuclear Regulatory int:rview conducted under oath with an Commission. Region 1. 031 Park Avenue, 5:30P.Af.-d.30PAf. Nuc/carPower NRC investigator on August 7,1985 and King of Prussia, Pennsylvanla 19400. If a Plant AuxiliarySystems (Open)-The in o sworn statement dated August 15, hearing is requested, the Commisslorf members will hear a report from its 1985, provided to the NRC investigators. will issue an order designating the time subcommittee regarding provisions in During the interview, Dr. Imperiale also and place of any hearing. Ifsa hearing is nuclear power plants to provide stated that he did not recall all his held, the issue to be consideted at such protection against fires.

re:::ns for his decision. hearing shall be: \ Priday,fuly 1f,196d Whether, on the basis of the matters gig

\

The willful violation pf NRC set forth in this Order. lJcensebo. 37-18452-01 should be modified in \he manner set forth in Section IV of his d.30 A.Af-f 30 A.Af. Davis.Besse Nuclear PowerPlant. Unit 1(Open/

Closed)-The members will hear requirements by Dr. !mperiale, while Order, presentations from representative of the perf rming licensed actikilles at hfercy NRC Staff and the licensee as lospital. raises serious estions ,gy,j'd ,3,ahModa. Mar %and. da tr day appropriate regarding the corrective h:ther the licensee wil comply with For the Nuclear Regulatory Commissi action and restart of this unit following mmission requirements while Dr.

lamn M. Tador. the loss of feedwater incident on June 9.

Imperlate has any responsibility for the 1985, perf rmance or supervision'oflicensed Director. Offece o//nspection and Portions of th!s session will be closed I c. 1 F ed 6-23-66: 8 45 am] k'fgh',han sp lcable i c y.

I1:30 A.Af.-t.00 PAf.: Technical Acco ingly, pursuant to sec11ons 81-lAdytsory Committee onforReactor Specifications Nuclear Po wer P/ ants 16tb. ant ino of the Atomic Energy Act Safeguarde; Meeting Agenda (Open)-The Committee will consider a cf 1954, as mended, and the In accordance with the purposes of proposed NRC policy statement C:mmissi ' regulations in 10 te8arding the nature of technical 2'202 and Pa s 30 and 35. It is he +b sections 29 and 182b. of the Atomic

'P' ' ts.

ordered that t elicensee shall: ejgy ,

Ac 42 U. 2039 2232b), the Rep en b ? PRCbe p Show cauu. in mannerherein aft" Safeguards will hold a meeting on July brief the ACRS members regarding this maun.

IoulInothmod$ 17apbbi 10-12. tDan. In Room 104n.171711 Street.

Salv: fore M importa, from serving in an NW., Washington. DC. Notice of this 200 pal-JooPAL Puture ACRS cap:c6ty involving the rformance or meeting was published in the l'ederal Activities (Open/ Closed)-The superviuon of hcensed 'h itie s. Register on hf ay 19,1980. members will discuss anticipated ACRS

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4 i Fed:r:I Register / Vcl. 51. Nr.121 / Tu:sd!y Jun2 24,1986 / N tic:s

(' 23013 cctivity and proposed topics for October 2.1985 (50 FR 191). In JIte entire meeting will be open to -

• consideration. This session may include accordance with these procedures. oral c briefing regarding a postulated blic attep' dance. , ,.

or written statements may be presented scenario for the nuclent power plant by members of the public. recordings e agenda for the subject meeting ",

cccident at the Chernobyl Nuclear will be permitted only during those h be sa follows: ' , , ,. .

Station. portions of the meeting when a ' -

Portions of this session will be closed transcript is being kept, and questions

,e day JulyDses-4:30 A.M. . .

cs required to discuss classified may be asked only b members of the ""#IIIA008888 * * '

I fo a n related to matters being mmitt , acons ants, a d Staff.

W Mm e dW 3:00 AAf.-3:30 AAf.: BSW Nuclear statements should notify the ACRS discuss the following topics: (1) The .

Powerplants (Open}-The members Executive Director as far in advance as Standardizat Policy Statement. (2) will hear reports for representatives of practicable so that appropriate proposed cha to 10 CFR 50, and (3) the NRC Staff and the B&W Owners arrangements can be made to allow the the EPRI Advanc Light Water .

Group regarding proposed plans for the necessary time during the meeting for Requirements de ' nts, a <

cvaluallon of the long. term safety of D&W nuclear power plants, such statements. Use of still. motion -

Oral statementa ma 1., presented by picture and televisions cameras during the members of the pubHs with the De er or C c 1/e uc/ oh 8 ' d concurrence of the Subcouunittee ortl th e6 cte in i efing reg d ng f$ct s oI Y "

the m ob et as or his ** E * *

• c:nsidered in the reactivation of deferred or cancelled nuclear power ma be obtained by a prepaid tel calfto the ACRS Executive Director, s

one only during those portions of the W

plants.

R.F. Fraley, prior to the meeting. In view meeting when a transcript is bel kept.

6 JO E31.-7.00 RAf.: Nomm.atw.n of of the possibility that the schedule for and questions may be asked only y ACRS Alember(Closed)--The members ACRS meetings may be adjusted by the members of the Subcommittee,its g will discuss the qualificadon of candidates proposed for appointment to Chairman as necessary to facilitate the consultants, and Staff. Persens desiring the ACRS. conduct of the meeting, reons to make oral statements should notify g "

planning to attend shou check with the the ACRS staff member named below as.

This session will be closed to discuss far in advance as is practicable so that Information the release of which would ACRA Executive Directorif such  !

represent a clearly unwarranted rescheduling would resultin major appropriate arrangements can be made, invasion of personal privacy, inC nVenl8nC8-

• During the in!titel portion of the I have determined in accordance with meeting, the Subcommittee, along with Satunfoy. /uly 12. JA96 subsection 10(d) Pub. L.92-463 that it is any ofits consultants who may be '

8.30 A.Af.-12 coa Noon: Preparat/on of necessary to clou portions of this .

meeting as noted above to discuss present, may exchange preliminary ACRS Reports to the Nuclear views regarding matters to be -

Regulatory Commission (Open/ Proprietary information (5 U.S.C. .

$52b(c)(4)) applicable to the facilities considered during the belance of the .

Closed)--The members will discuss meeting.

proposed reports to the NRC regarding being discussed. information the release .

matters considered during this meeting. of which would represent a clearly The Saboommittee will then bear Portions of this session will be closed unwarranted invasion of personal resentations by and hold discussions

' privacy (5 U.S.C.552b(c)(6)) classified '

cs required to discuss Proprietary h representatives of the NRC Staff.

Information applicable to the matters data (5 U.S.C. 651b(c)(1)). Its nsultants, and other internted being discussed. Further information regarding topica pers e regarding this review.

tw P Af.-2:30 EAf.: Act/v/ ties of to be discussed, whether the meeting ACRS Subcommittees (Open)-ACRS has been cancelled or rescheduled, the Fur r information regarding topics subcommittee chairmen will report to Chairman's ruling on requests for the to be die seed, whether the meeting the Committee regarding the status of opportunity to resent oral statements has been celled or rescheduled, the designated subcommittee assignments and the time al otted can be obtained by Chairman's ling on requests for the including proposed revisions to NRC a prepaid telephone call to the ACRS Opportunity to resent oral statements Regulatory Culdes. NRC activities Executive Director. Mr. Raymond F. and the time all ed therefor can be regarding chilled water systems in Fraley (telephone 202/634-3265). obtained by a p aid telephone call to nuclear ower plants, and the reliability between 8:15 A.M. and 5.00 P.M. the cognizant ACR taff member. Mr.

cnd performance of nuclear power plant Dated:lune to. tese Iferman Alderman ( pohne 202/634-

' control room heating, cooling, and lohn C.lloyle' 1413) between 8.15 A. and 5.00 P.M.

v:ntilating systems. * " " * * * " " " * * * * * "I A dvisory Committoe Nonagement Officer.

2:30 AA1.-J:Jo AAf. Preporat/on of * "8 "*"

• ACRS Reports to the NRC(Open/ trR Doc. so-14230 Filed 6-23-46. &45 am) . named individual one or tw sys Closed)-The members will comptete 8"** CC'8 ** 8*

before the scheduled meetin be discussion of matters considered during advised of any changes in schedples, this meeting etc., which may have occurred.

Adviso ommittee on Reactor Portions of this session will be closed Safeguards, bcommittee on as required to discuss Proprietary Dated; lune 17. tena improved LW gna; Meeting Information applicable to the matters Morton W. Iaarktn.

being discussed. The ACRS Subcomm on improved A nistant Erecut/re DirectorforPro/ect on July It'"I' *-

i l Procedures for the conduct of and LWR Designs will hold a me participation in ACRS mcctings were 9.1980. Room tolo, 171711 Street

• V., (FR Doc. 80-14231 Filed 6-23-es: e.45 sml published in the Federal Register on Wa shington, DC. sumo coce reewi-as

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o UNITED STATES g, g [/"

fo -a g nlUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS ,

WASHINGTON, D. C. 206S5 '

Revised: July 7, 1986 SCHEDULE AND OUTLINE FOR DISCUSSION 315TH ACRS MEETING JULY 10-12, 1986 WASHINGTON, D. C.

Thursday, July 10, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.

1) 8:30 - 8:45 A.M. Report of ACRS Chairman (0 pen) 1.1) Opening Statement (DAW) 1.2) Items of current interest (DAW /RFF)

2) 8:45 - 11:45 A.M. Standardization Policy Statement & EPRI Advanced LWR (BREAK-9:50-10:00) Program (CJW/HA)(Open) 2.1) 8:45-9:30: Briefing by representatives of the NRC Staff 2.2) 9:30-9:50: AIF discussion on proposed Stand-ardization Policy Statement 9:50-10:00: BREAK 2.3) 10:00-11:00: EPRI discussion of Advanced LWR Program Plan 2.4) 11:00-11:30: NRR discussion of the EPRI Ad-vanced LWR Program Plan 2.5) 11:30-11:45: ACRS report to NRC 11:45 - 11:55 A.M. BREAK

3) 11:55 - 12:30 P.M. ACRS Views / Recommendations Regarding Future Plant Designs (0 pen) (CJW/HA)

, 3.1) SafetyGoal/SevereAccidentIssues(DO) 3.2) Accident Prevention (WK) 3.3 AccidentMitigation/ Containment (CPS) 3.4 Security / Sabotage (CM) 3.5 Revisions to Regulations (FJR) 3.6 OtherSafetyConsiderations(MWC) 12:30 - 1:30 P.M. LUNCH

3) 1:30 - 2:30 P.M. ACRS Views /Recomendations Regarding Future Plant Designs (0 pen) (CJW/HA)

• Continue discussions noted above

4) 2:30 - 5:30 P.M. Reorganization of TVA (0 pen) (CJW/RPS)

(BREAK-4:00-4:15) 4.1) 2:30 - 2:45: Report of ACRS Subcomittee i

4.2) 2:45 - 4:00: Presentations by NRC Staff /TVA 4:00 - 4:15: BREAK 4.3) 4:15 - 5:30: Discussion and ACRS consultants' reports

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315th ACRS Meeting Agenda t '

5) 5:30 - 6:30 P.M. ACRS Subcomittee Activities (0 pen) 5.1) Report of Subcomittee on Auxiliary Systems regarding fire protection features in nuclear plants and Standard Review Plan Sections 9.2.1 and 9.2.2 regarding service water systems (CYM/JCE/SD) 5.2) Presentation by members of the Office of Nuclear Regulatory Research, as appropriate s

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315th ACRS Meeting Agenda I

{' Friday, July 11, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.

m 8:30 - 11:30 A.M. Davis-Besse Nuclear Plant, Unit 1 - Restart

. 6)

'-(BREAK-10:30-10:45) (0 pen / Closed) (FJR/HA)

(Note: Portions of this session will be closed as necessary to discuss Proprietary Infonnation applicable to this plant.)

7) 11:30 - 1:00 P.M. Future Activities (0 pen) 7.1) 11:30-12:30: Briefing by R. Bernero, NRR, on BWR containment performance during severe

'+

accidents (WK/MDH)

J'- 7.2) Anticipated subcommittee activity (MWL) s 7.3) Proposed items for ACRS consideration (DAW /RFF) 7.4) Hope Creek Nuclear Plant - turbine overspeed testing (GAR /MME) 1:00 - 2:00 P.M. LUNCH

, 8) 2:00 - 2:30 P.M. B&W Water Reactors (0 pen) 8.1) Subcommittee report on B&W Owners' Grop Trip Reduction and Transient Response Improvements Program (CJW/RKM)

9) 2

30 - 4:30 P.M. Technical Specifications - Proposed NRC Policy State-(BREAK: 3:15-3:30) ment F 17 (0 pen) (CYM/JOS) Subcommittee report of July 1 P Operating Procedures Subconnittee 9.2) Briefing by NRC Staff representatives Reactivation of Deferred or Cancelled Nuclear Power

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10) 4:30 - 5:30 P.M.

Plants 10.1) Briefing by representatives of the Office of Nuclear Reactor Regulation

11) 5:30 - 6:00 P.M. Nomination of ACRS Member (Closed) 11.1) Report of Nominating Panel regarding

. candidates for appointment to the ACRS l

(HWL/ALN)

(Note: This portion of the meeting will be closed to discuss information of a personal nature, the release of which would represent a clearly unwarranted invasion of personal privacy.)

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315th ACRS Meeting Agenda Saturday, July 12, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.

ACRS Reports to NRC (0 pen / Closed)

12) 8:30 - 12:00 Noon 12.1) Discuss proposed ACRS reports to NRC regardingbavis-Besse restart (FJR/HA) 12.1-1) 12.1-2) Policy Statement on Tech. Specs.

(CYM/JOS) (Tentative) 12.1-3) Policy Statement on Standardized Nuclear Plants (CJW/HA) 12.1-4) TVAReorganization(tentative)

(CJW/RPS) 12.1-5) Auxiliary Systems, Fire Protection andSRP9.2.1and9.2.2(CYM/JCE/SD)

(Note: Portions of this session will be closed as necessary to discuss Proprietary Information applicable to this plant.)

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

ACRS Subcomittee Activity (0 pen)

13) 1:00 - 2:00 P.M. Davis-Besse, Reg. Processes (HWL/GRQ) 13.1) 13.2)

NRC Staff review of chilled-water systems in nuclear power plants (CYM/JOS) (tentative, if time pennits) 13.3)

NRC effort regarding control room habitability (DWM/JOS) (tentative, if time permits)

Complete preparation of ACRS reports (0 pen / Closed)

14) 2:00 - 3:30 P.M. (Note: Portions of this session will be closed as necessary to discuss Proprietary Information applicable to this plant.)

s F} lW[T MINUTES OF THE j ,

i ; Ei J 315TH ACRS MEETINr, l il f 7 JULY 10-12, 1986 b. jj jj N The 315th meeting of the Advisory Committee on Reactor Safeguards, held at 1717 H Street, N.W., Washinoton, D.C., was convened by Chairman D. A. Ward at 8

30 a.m.,

Thursday, July 10, 1986.

[ Note: For a list of attendees, see Appendix I.]

Chairman D. A. Ward noted the existence of the published agenda for the meeting, and identified the items to be discussed. He noted thet the meeting was being held in conformance with the Federal Advisory Committee 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 available for purchase from ACE-Federal Reporters, Inc., 444 North Capital Street, Washington, D.C. 20001.]

I. Chairman's Report (0 pen)

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

Chairman Ward made several announcements:

(a) The Comission has given approval for the Staff to allow restart o#

the Enrico Fermi Atomic Power Plant, Unit No. 2, which was shut down because of a recriticality problem.

(b) Note was taken of the fact that the new Commission Chairman, L. W.

Zech, Jr., is using a somewhat different organization than his predecessor, N. J. Palladino. Those interested in the details on how this orcanization will interact with the ACRS were encouraged to obtain a copy of a organizational chart from the ACRS Office.

(c) An IAEA briefing will be conducted by representatives of the Soviet Union on August 25, 1986 in Vienna, Austria. There is a move to expand the U.S. delegation which currently has only five members. An ACRS member may be part of the delegation and H. Denton, Director of NRR, is working on this aspect. D. Okrent suggested that Chairman Zech be asked to intervene regarding ACRS attenc'ance.

(d) Members should take note of a letter frcm Chairman Zech to the ACRS Chairman dated July 1, 1986 which outlines his policy regarding Agency activities.

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PROPOSED filNUTES - 315TH (e) Chairman Ward noted the death of Admiral Hyman Rickover, who died on Tuesday morning, July 8, 1986 (f) Chairman Ward announced that three nominees for appointment to the Committee would be present during this meeting.

(g) It was noted that F. J. Remick will be reappointed to the Committee for a second term.

II. Davis-Besse Nuclear Power Station Unit 1 Restart (0 pen)

[ Note: H. Alderman was the Designated Federal Official for this portion ofthemeeting.]

F. J. Remick provided background material regarding the June 9,1985 incident at the Davis-Besse Nuclear Power Station which experienced a partial loss of main feedwater while operating at 90 percent power. The ensuing reactor trip was followed by spurious isolation of the steam generators which initiated a chain of events involving a number of equip-ment malfunctions and several operator errors. Ultimately, all feedwater was interrupted for a very short period of time. By the time operators were able to restore feedwater flow, both steam generators had dried out.

The Director of the Office of Nuclear Reactor Reculation issued a 50.54(f) letter precluding restart of the Davis-Besse facility without NRC approval. The letter requested information from the Licensee on the cause of the failures that occurred during the incident and a draft of corrective action to be taken before restart of Unit 1. The 50.54(f) letter, issued on August 14, 1986, superseded a June 10 confirmatory action letter frem Region III which asked the Licensee to establish causes and determine corrective actions for the incident, to evaluate the condition of the reactor vessel and the steam generators, and to perform confirmatory testing subject to Region III concurrence before restart.

l The NRC Staff reviewed the Davis-Besse course of action report which was issued on September 10, 1985 with other supporting material. The Staff's report, NUREG-1177 (June 1986), is entitled " Safety Evaluation Report Related to the Restart of Davis-Besse Nuclear Power Station Unit 1 Following the Event on June 9, 1985." The Staff's SER addressed l management restructuring, maintenance procedures and training, the l results of specific investigations of various aspects of the event including evaluation of plant modifications to be made, and an evaluation of a systems review and test program to review 34 of the major systems on decay heat removal reliability. The Staff's SER concluded that no issues related to the June 9th event remain to be resolved. The restart of Unit 1 is currently estimated for early November 1986, but the lead item on the critical path is replacement of four feedwater drive shafts.

J. Williams, Senior Vice-President, Toledo Edison, noted that Toledo Edison's objective is to commence startup on October 25, an event con-trolled by replacement of the main cooling pump shafts. He indicated l

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PROPOSED MINUTES - 315TH that one of several programatic steps being taken to improve the manage-ment of Davis-Besse has been to move all engineering on site. Toledo Edison is expanding the engineering force to eliminate reliance on consultants to the maximum degree possible. All documentation is being moved to a new records control document control organization at the plant. He mentioned revamping of the Nuclear Review Board with addition

of members outside of Toledo Edison, as well as the setting up of an independent safety engineering group. The Nuclear Licensing Director was '

i elevated in the organization chart. A more rigorous drug screening program has been added for incoming applicants, and a program of random drug testin has been instituted for the entire organization (see Appendix IV)g .

i C. Michelson asked if all Appendix R modifications will be completed before startup. J. Williams indicated that this would not be the case

. since Toledo Edison had a previous agreement with the NRC Staff on a schedule for completion of Appendix R modifications in 1989.

i S. Smith, Assistant Plant Manager for Maintenance at Davis-Besse, dis-

! cussed extensive improvements made to the maintenance program in the area i

of organization and staffing, training, maintenance activities, and

engineering interface and support (see Appendix IV). He discussed the backlog of corrective modification work orders indicating that the maintenance organization is now completing about 95 percent of all 4

scheduled preventive maintenance every month. He mentioned a labor strike of the chemistry and health physics maintenance personnel in Local -

245 of the International Brotherhood of Electrical Workers. G. A. Reed 4

asked if the strike had any relationship to the random drug testing effort, or the application of advanced management techniques such as aptitude selection testing. S. Smith indicated that there has been controversy regarding the drug testing program, but the basic issue of '

the strike is health benefits.

! J. Wood, Toledo Edison, discussed an equipment investigation effort to determine the root cause of equipment malfunctions in order to implement an appropriate and effective corrective actions program. A main feed pump turbine overspeed trip was the initiator for the June 9 event. A failure in the speed control circuit board was found which has since been replaced. He reviewed 13 areas which impacted seven systems, noting the failures that occurred, the reasons determined for the failures, and the corrective actions taken. It was noted that the pilot-operated relief valves operated three times during the transient and may have stuck open the third time. D. A. Ward asked if Toledo Edison has an explanation as to why the valve may have failed to close. J. Wood indicated that there were no obviously broken parts of control systems that were found. The most probable cause at this time appears to have been foreign material i lodged in the pilot area that subsequently was cleared. D. A. Ward asked if anything has been done to alleviate this sort of problem in the future. J. Wood indicated that measures have been taken, such as enhancement of the control room aspects and testing of the block valve.

PROPOSED MINUTES - 315TH G. A. Reed pointed out that the Licensee should investigate and try to understand some of the complexities of boric acid environments in the top of the pressurizer (boron vapor and hydrogen) with respect to pilot-operated relief valve function. J. Wood indicated that this matter is being investigated and Toledo Edison has the capability to put a separate loop seal on the pilot. He also mentioned Toledo Edison's concerns with the loop seal of slug flow through the valve and the effect on the pipe, the problem of cracking of the reactor coolant pump shafts, and failure rates of various components at Davis-Besse over the years since the start of operation. G. A. Reed wondered if the failure rates are similar to the industry average or substantially different from them.

The problem regarding axial cracking in the reactor coolant pump shafts was discussed. J. Wood indicated that while these pumps are similar to the Byron Jackson pumps at Crystal River, it is not clear what caused the cracking at Crystal River. The axial cracking at Davis-Besse appears to be thermal fatigue, not mechanical cracking.

Sushil Jain, Director of Nuclear Engineering at the Davis-Besse plant, discussed improvements made to the auxiliary feedwater system and the steam and feedwater cooling system. D. Okrent asked if there was information on failure rates of various components for Davis-Besse over the years since the start of operation. He asked whether it is possible to determine if the failure rates at Davis-Besse are similar to the industry average or substantially. different for certain kinds of equipment. S. Jain mentioned an auxiliary feedwater reliability analysis in 1980-81 and a 1985 study of specific Davis-Besse failure rates for different components. It was found that the failure to open or failure to close of motor-operated valves was the greatest outlier in comparison to published data. In 1980 steps were taken to improve the motor-operated valve reliability, as well as the governor which was also an outlier for the auxiliary feedwater pump turbine. J. C. Ebersole noted that the Davis-Besse station is dependent on the functioning of valves in the case of an high energy line break. S. Jain indicated the consequences of the high energy line break for the worst possible areas were analyzed using the RELAP code regarding temperature and humidity conditions. J. C. Ebersole suggested that there might be a partial break which would not be detected by the pressure detection system which would not close the valve. S. Jain agreed that one would have to rely on regular surveillance or walkdowns in that area. C. Michelson asked if Davis-Besse is taking advantage of the leak-before-break concept to eliminate arbitrary intermediate breaks. S. Jain indicated that Toledo Edison is not taking credit for the leak- before-break philosophy in the case of the arbitrary intermediate break. D. Okrent asked if the NRC Staff can answer the question regarding an incomplete failure of a high energy line. J. C. Ebersole explained that it is generic and applies to air pressure or DC voltage as well. J. Stolz, NRC Staff, indicated that the Staff does not evaluate these intermediate states.

S. Jain described important changes made by Toledo Edison to the auxiliary feedwater system at Davis-Besse, including the installation of

s PROPOSED MIflUTES - 315TH the emission valves to alleviate the condensation problem found in the overspeed trip of the auxiliary feed pump on June 9,1985. The full capacity auxiliary feedwater flow, motor-driven feedwater pump powered by the nonessential busses has been installed. It can be powered by the emergency diesel generator. This pump essentially eliminates all of the high energy line concerns that existed with the old smaller startup feedwater pump. Because of the separation of the new motor-driven feedwater pump from the other auxiliary feedwater pumps, those high energy line concerns are mitigated. The pump is in the turbine building with independent wiring and controls. J. C. Ebersole asked to what degree it is safety grade. He noted that it has a number of safety-grade characteristics. S. Jain indicated that Toledo Edison has made every effort to upgrade the pump to bring it as close as possible to a safety-grade pump. He noted that it is not seismically qualified, but a technical specification will be submitted for this pump prior to restart which is very similar to that for the other auxiliary feedwater pump. J.

C. Ebersole asked regarding the vulnerability of this pump to fire and the hostile environment. D. Okrent asked if it is subject to flooding by pipe breaks in the system that cool the condenser. S. Jain indicated that Toledo Edison has looked at the pump rupture that could cause inoperability of this pump, as well as other flood conditions in a hazards analysis.

S. Jain discussed improvements to tne stean and feedwater rupture control system (SFRCS). He indicated that the SFRCS power supply performance is being improved by providing additional cooling within the cabinets. The logic of the SFRCS is also being changed so that on a low-level signal there would no longer be isolation of the main feedwater and the main steam lines. The logic has also been changed to delete isolation of auxiliary feedwater from the steam generator that is the last to depres-surize. J. C. Ebersole asked why the motor-driven feed pump was made l

i manual instead of automatic start. S. Jain indicated that automatic actuation complicates the control system.

S. Jain discussed the bleed-and-feed cooling evaluation done following the June 9 event which assumes that the plant was running at full power.

In answer to a question by D. Okrent, S. Jain indicated that the analysis assumes the loss of offsite power and, therefore, the reactor coolant pumps were not operating. In answer to questions by D. A. Hard, J.

Williams indicated that if the hot leg temperature hits 600 F, the opera-tor must initiate bleed-and-feed by new procedures. S. Jain indicated i

that the analysis showed that the operator would have 10 minutes after the hot leg temperature reaches 600 F before bleed-and-feed would be ineffective. D. A. Ward noted that Toledo Edison does not know what happens if bleed-and-feed is not successfully initiated within 10 l minutes. S. Jain stated that the analysis indicates that the operator l has 37 minutes following complete loss of feedwater to restore any type of feedwater to the steam generators. This can be done using either the t

auxiliary feedwater pump, main feedwater pump, or the motor-driven i

feedwater pump, if possible, before loss of water level at the core l

PROPOSED MINUTES - 315TH occurs. H. Etherington asked if Toledo Edison is maintaining pressure with the PORVs, or are the PORVs open during the bleed-and-feed operation. S. Jain indicated that after the operator has taken this action, the PORV is continuously open.

S. Jain discussed longer term decay heat removal reliability improvements. He indicated that Toledo Edison is proposing to provide the means of blowing down the primary side to the point to where one can use intermediate head, high pressure injection pumps. The conceptual design is to provide trains of an emergency depressurization system from the upper hemisphere of the steam generator on the reactor coolant system side, piping into a "T" with two valves in series. The system will be single failure proof, and redundant. C. Michelson asked if the PORVs 'are qualified against the possibility of slugs of water. S. Jain indicated that the PORVs are designed to take a water-vapor mixed flow. They are not designed for slug flow at present. F. J. Remick asked when Toledo Edison plans to install the emergency depressurization system at Davis-Besse. J. Williams indicated that it would be installed during a December 1987 outage. C. Michelson asked if the reactor vessel level instrumentation is in place. S. Jain indicated that Davis-Besse has a hot leg level monitoring system, but not a reactor vessel level measurement system. C. Michelson expressed concern that Toledo Edison will not have a measurement of water level in the reactor vessel since the hot leg indication is not on the vessel. C. Michelson expressed his concern regarding overflow of the steam generator because of the lack of vessel level indication.

P. Hildebrandt, Toledo Edison, discussed the system review and test program which reviews 34 systems important to safe plant operation, a range of design, maintenance, and other types of problems to determine whether corrective actions were required prior to restart. He mentioned certain recurring problem areas, including inattention to heating, ven-tilation and air conditioning requirements, inoperable nitrogen regula-tors, inadequate maintenance for hydro-motor actuators, etc. The Commit-tee briefly discussed a problem regarding Raychem splice terminations and shrink sleeve installations.

( D. Okrent asked Toledo Edison at what level within the plant management I would one find a good knowledge of the results and the phenomena involved in PWR PRAs. He explained that his concern was for a knowledge of the dominant sequences and the phenomena involved in core melt, as well as containment behavior. L. Storz, Davis-Besse Plant Manager, indicated that there is a high level of understanding regarding accident recovery in the organization and in the training programs, which are now designed to look at symptom-based procedures. There is not a high level of expertise in the plant organization. P. Hildebrandt indicated that S.

Jain has now been placed in charge of the nuclear engineering group with a specific intent of providing PRA and similar types of technical evaluation support for engineering. D. Okrent asked if there is a good i

knowledge of the phenomenology in PRAs in other parts of the manaaement

4 PROPOSED MINUTES - 315TH organization. J. Williams indicated that he had a good understanding of the scenarios and sequences, but he relies heavily on staff for the details. D. Okrent thought it important for the plant manager tn have that knowledge. He wondered what action the Staff believes would have been taken from the results of the systems evaluation that has just been done if the Davis-Besse plant had not suffered the loss-of-feedwater event. C. McCracken, NRC Staff, indicated that prior to the event on June 9 a' configuration management program for systems and a reevaluation of maintenance were ongoing, in addition to several other programs, with the intent of improving performance. Should the June 9th event not have occurred, some performance improvement would have occurred over a protracted period of time. The system review and test program would probably not have been as extensive or comprehensive as it currently is.

The event was an initiator which focused their attention, as well as the attention of the NRC Staff in a more concentrated time frane. J. Stolz indicated that concerns generated not only by Davis-Besse, but also by the transient at Rancho Seco, have spawned the assessment by the the B&W Owners Group (BWOG). While the Staff did not have the benefit of a Davis-Besse PRA to review and focus on the valve issue (the central issue in connection with this loss-of-feedwater event), the Staff certainly has focused on that issue in a generic sense for all of the B&W plants. D.

Okrent suggested that the Staff's opinion is that the BWOG effort is not sufficiently comprehensive. He noted that the Davis-Besse self review was components / systems oriented but not plant oriented. He asked if Toledo Edison had any comment on this matter. P. Hildebrandt suggested that the Davis-Besse effort has been an intersystem review which focused particularly on decay heat renoval. J. Williams thought an auxiliary feedwater reliability study undertaken by Toledo Edison took the whole plant into account. S. Jain indicated that a first level PRA is underway which is plant oriented.

C. McCracken cited the Staff's findings. He indicated that the Davis-Besse plant has been restored to its licensing basis, and contingent upon implementation of Licensee commitments may resume operation (see Appendix V). J. C. Ebersole questioned that the licensing basis for Davis-Besse was adequate before the June 9 event. He stated that the Staff did not require diversity in the feedwater system, but contended that the design configuration was adequate and that the June 9th event resulted from poor maintenance. C. McCracken indicated that the issue of diversity in the auxiliary feedwater system was being pursued for the last six years, and the Licensee had already committed to installation of the electric motor-driven feed pump prior to the June 9 event.

C. McCracken discussed several problem areas which led to major changes at the Davis-Besse plant.

Motor-Operated Valves Main Steam Safety Valves and Atmospheric Vent Valves

l

l PROPOSED MINUTES - 315TH

• Safety Features Actuation System Safety Significant Human Engineering Deficiencies Single Failure Considerations He indicated that the Staff considers the Davis-Besse maintenance program and system review and test programs extremely important to addressing the major issues identified the the 50.54(f) letter, and in the IIT report.

C. McCracken explained that the Staff's concern regarding the issue of motor-operated valves was that the auxiliary feedwater valves failed to reopen during the June 9,1985 event. It was determined that controls were improperly set. The Staff is concerned that other motor-operated valves may not function, and the Licensee, in response to this concern, reworked all valves and safety systems and systems important to safety using MOVATS. The Licensee has agreed to test representative valves for each design type to ensure that the new method of setting up valve con-trols will work.

C. McCracken indicated that Toledo Edison experienced main steam pressure

, fluctuation anomalies during the June 9th event caused by improper opera-tion of main steam safety valves and atmospheric vent valves. The atmo-spheric valve control system has been repaired and the main steam safety valves have been removed, reworked, and retested. A long-term maintenance plan for testing and repairing these valves is being developed. He mentioned a concern that the safety features actuation system did not have sufficient independence to meet the single failure criterion. Modifications have been incorporated by Toledo Edison to meet channel independence criteria of IEEE Standard 279-1971. Twenty-nine j safety significant human engineering defects were identified prior to the ,

June 9 event. Because of some of the problems which occurred in the control room during the incident, i.e., the isolation of both auxiliary feedwater valves, the Staff thought that these safety-significant human 1 engineering defects needed resolution prior to restart.

The Committee discussed significant commitments by the Licensee related to major problems identified by the IIT. D. Okrent asked if the Staff has systematically looked at the lessons learned from Davis-Besse and has identified which are generic not only to B&W plants but to other nuclear plants. C. McCracken indicated that the Staff has been concentrating on the generic implications for B&W plants, such as the issue of motor-operated valves which was handled through a bulletin. Information notices were sent out on a number of other issues. The generic implica-tions for other plants will be examined later.

The Committee discussed its belief that the Commission made an explicit commitment to pool the lessons learned from Davis-Besse into a single SER. C. Michelson indicated that he expected to see in the Davis-Besse SER closure on all items brought forth by the IIT report. H. W. Lewis

PROPOSED MINUTES - 315TH discussed the promise from the Commission for a coherent presentation of root causes which was not in the IIT report. C. McCracken indicated that it is in the Staff's SER, C. Michelson noted that it had been resolved that the Davis-Besse plant will not be required to have reactor vessel level indication. He asked why Davis-Besse was singled out for an ex-emption. A'. Deagazio, NRC, explained that Davis-Besse was granted an exemption for a period of time to meet a requirement for a- vessel head vent. An order was subsequently issued on instrumentation which required an addition to the hot leg level measuring system. It required a head vent and reactor vessel measuring system for the head. Toledo Edison then proposed to put in an continuous vent from the vessel head to the top of the hot leg. C. Michelson asked whether the hot leg level indication will work after the primary system has been emptied and the vessel is refilled. J. Stolz stated that Davis-Besse is unique with respect to continuous venting. C. Michelson indicated his concern regarded lack of level indication upon refilling the vessel. R. Hernan, NRC, indicated that no B&W plants have level indication below the bottom of the hot leg. Below the bottom of the hot leg, B&W plants will have only thermocouple indication. C. Michelson thought that the Comittee~

ought to study this matter.

S. Persenko, NRC, indicated that two maintenance surveys were conducted by the Staff in September 1985 and March 1986 (see Appendix VI). He cited a list of weaknesses found during the September survey which was followed up during the second survey in March of 1986. The March survey was conducted primarily to address modifications to the maintenance program to address the nine weaknesses found in the September survey.

The Staff confirmed considerable progress and that the program was functioning as intended with no major identifiable programmatic weaknesses. He noted that the Staff review was a programmatic one, and not an in-depth review.

F. J. Remick indicated that contingent upon the implementation of the Licensee's comitments identified in the SER, the Subcomittee agreed l with the Staff that the Davis-Besse Nuclear Power Station could resume

! operation. He indicated that if any members of the Committee disagreed with that Subcomittee position, the ACRS could write a letter. C.

Michelson mentioned some reservation about level indication, as well as the primary system blowdown proposal by Toledo Edison relative to the l

PORVs. H. W. Lewis indicated reservations with respect to closure of the j IIT, but noted that this would not affect the restart decision.

III. Subcommittee Report on B&W Owners Group Trip Reduction and Transient l Response Improvements Program (0 pen) l l [ Note: R. K. Major was the Designated Federal Official for this portion of the meeting.]

C. J. Wylie indicated that the ACRS Subcnmmittee on B&W Reactor Plants met June 25, 1986 for two purposes:

1

e PROPOSED MINUTES - 315TH To hear the results of the Incident Investigation Team's review of the loss of ICS instrumentation and overcooling transient at Rancho Seco on December 26, 1985 To hear a briefing by the NRC Staff and the B&W Owners Group on reassessment of B&W nuclear power plants C. J. Wylie indicated that the NRC made a decision to reassess the B&W plants' design following the Davis-Besse and Rancho Seco incidents. In January 1986 the ED0 contacted the B&W Owners Group urging them to take a lead role in this assessment process to be completed in CY 1986 (see Appendix VII). In February 1986 the Owners Group agreed to the Staff's request and documented its program in a May 15, 1986 submittal to the NRC.

C. J. Wylie presented the objectives of the NRC reassessment program (see Appendix VIII). He outlined the scope of the NRC program, much of which involved review of the B&W Owners Group's work. He noted that the B&W Owners Group derived from the ED0's January 24, 1986 letter that the primary concern was that the frequency of complex transients is too high for B&W plants (see Appendix IX). H. Tucker, Duke Power Company, and also Chaiman of the B&W Owners Group, indicated in a letter to the E00 that the Owners Group would take the lead effort to reduce trip frequency and improve transient response. The program would reduce complex transient frequencies. C. J. Wylie noted that the Subcommittee had reservations regarding a letter written by the Staff which contended that the BW0G program was generally on target. The Staff at that time did express concerns that the program primarily addressed trip reductions and post-trip response and was not sufficiently focused to respond to the ED0's letter and the Staff's reassessment program. A broader based effort was needed that reviewed systems in detail and anticipated future problems. The Staff was also unclear on how the BWOG would address management considerations as a safety issue.

C. J. Wylie noted that, at the Subcommittee meeting, following the presentations by the Staff and the BWOG, the Subcommittee was unanimously of the opinion that while the BWOG program, as presently described, was a good start, it was not adequate to answer the question of whether B&W plants as presently designed and modified can be operated safely enough.

C. J. Wylie noted that the BWOG requested a delay in their presentation to the full Committee until September 1986, to reexamine its program in l

light of the Subcomittee's and Staff's comments. Since the June 25 Subcommittee meeting, BWOG has renamed its program the " Safety and Performance Improvement Program." D. A. Ward suggested that the BWOG program does not really focus on the basic safety issue. This issue is that an usually large number of severe operational events are occurring in two-thirds of B&W plants; the other one-third of the population cf the plants has not had severe transients over the period of record. He thought this was statistically significant. The Staff shculd follow up this matter to determine if there is some subtle difference in the design

PROPOSED MINUTES - 315TH of the troublesome plants or whether it is the way the plants are being operated. G. A. Reed indicated that he was pleased to see that the program was being reoriented and put on a broader footing. W. Kerr suggested that the operating organizations may be the problem, not the plant design. He indicated that he would like to see more emphasis on the contribution of poor morale to the operating organization in some of these plants. M. W. Carbon cautioned that while one organization might do a particularly good job of running B&W plants, one should not overlook the fact that there is a disproportionately large number of serious operating events in plants with B&W systems.

C. Wylie noted that the BWOG intends to brief the full Committee at its September meeting. Major alteration to the program will probably cause slippage in the Staff's projected schedule. The Committee discussed the merits of writing a report on this issue at this meeting.

F. Hebdon, Deputy Director, AE0D, briefed the Committee on the loss of integrated control system power and overcooling transient at the Pancho Seco Nuclear Generating Station, Unit 1 on December 26, 1985. He noted that the integrated control system (ICS) DC power supply played a central role in the event that occurred at Rancho Seco. He indicated that the power supply monitor spuriously tripped two switches. This is a trip circuit designed to trip the system if the voltage to two 24-volt buses becomes erratic. J. C. Ebersole asked if there is an analysis that describes what happens when the system trips. F. Hebdon indicated that the ICS is a nonsafety-related system, and the analysis would not be part of the licensing basis. J. C. Ebersole thought that the chain of events that would occur after a trip of this system deserves consideration by the licensee. F. Hebdon indicated that the question of the ICS and nonnuclear instrumentation is an issue that has been of concern to the Staff for several years. This matter was discussed in considerable detail in the IIT report.

F. Hebdon explained that a transient occurred during the 25-minute period that the switches were open. J. C. Ebersole pointed out the " ungraceful" nature of the trip. F. Hebdon agreed that the incident was significant because a single failure in the ICS, which is a nonsafety-related system, subjected the plant to an undesirable overcooling transient (see Appendix X). The fundamental causes for this transient were design weaknesses and vulnerabilities in the ICS and in the equipment controlled by that system. These weaknesses and vulnerabilities were not adequately compensated by other design features, plant procedures, or operator training and the weaknesses and vulnerabilities were largely known to the Licensee and the Staff by virtue of a number of precursor events after related analyses and studies. Information was available and known that could have prevented this overcooling transient, but in the absence of adequate plant modifications the incident should have been expected.

1 F. Hebdon noted some undesirable features of the transient including the fact that the reactor coolant system cooled down 180 F in 26 minutes d

PROPOSED MINUTES - 315TH (limit is 100 F per hour). The pressurizer emptied and a bubble formed in the reactor vessel head. The plant entered the pressurized thermal shock region and did not have safety features actuation. Water overflowed from the steam generators into the main steam lines (plant does not have main steam isolation valves). The sensitivity of this particular type of plant to the ICS was known to the Staff and known to the utility; however, modifications were not made to ensure that either the event did not occur or that it would be adequately mitigated.

F. Hebdon indicated that the IIT report reached 15 principal findings and conclusions and 11 miscellaneous other findings and conclusions. The first principal finding was that .the overcooling transient was initiated by the failure of a single module in the nonsafety-related ICS (the spurious tripping of the power supply module interrupted all plus/minus twenty-four volt DC power). He noted this was the actual initiator of the event. It was found that a bad electrical contact in that line created the resistance that was necessary to cause the power supply monitor to act erratically. This caused the power supply monitor to trip the two switches. F. Hebdon pointed out a design weakness in that the line where the voltage is sensed is also the line that supplies power to the monitor. This is a design weakness. C. Michelson indicated that he thought the IIT report was rather comprehensive and clear. H. W. Lewis noted the Committee's continuing interest in following the progress of the IIT process. He agreed with C. Michelson's appraisal of the Ranch Seco report.

IV. ReorganizationofTVA(0 pen)

[ Note. R. P. Savio was the designated Federal Official for this portion of the meeting.]

C. J. Wylie indicated that the ACRS Ad Hoc Subcommittee on TVA met on June 12-13, 1986 to discuss the TVA management reorganization and TVA's plans for restart of its nuclear plants. He mentioned the Staff dis-cussion of TVA's corporate nuclear performance plan and plant-specific concerns. The Staff endorsed the concepts outlined in the TVA corporate nuclear perfonnance plan, but concluded that there was not yet sufficient information regarding its implementation to judge its effectiveness.

Plant-specific issues identified for resolution were equipment qualifica-tion, welding design, plant modification controls, operational readiness, resolution of employee concerns, and fire protection and seismic design which were identified as issues specific to Browns Ferry.

C. J. Wylie mentioned TVA presentations regarding the root causes which led to the shutdown of the TVA nuclear plants, highlighting the lack of experienced nuclear managers to provide leadership and the proper direction for TVA's nuclear activities. He indicated that the restart approach for TVA's nuclear plants is contained in the revised Corporate Nuclear Performance Plan dated March 10, 1986. TVA described, in detail, activities being taken, such as assembling a capable senior nanagement team, consolidating nuclear activities within a single organization,

PROPOSED MINUTES - 315TH centralizing direction and control from headquarters, establishing clear responsibility for functional areas, establishing employee concerns programs, increasing management awareness, improving management systems and controls, and programatic improvement. The evaluation of TVA's difficulties is being conducted by a Task Force lead by former Admiral S.

A. White, appointed in November 1985. He mentioned two reports by ACRS consultants which will be discussed later in this session.

R. H. Wessman, NRC, discussed the status of Staff actions on TVA facil-ities (see Appendix XI). He mentioned TVA's public announcement that Sequoyah Unit 2, which was to be the lead unit to come back into opera-tion, would be ready in January 1987, with Watts Bar ready for licensing on about May 1987. He noted that there are no definitive schedules regarding resumed operation for any of the Browns Ferry units. The Staff's initial evaluation of intimidation and harassment issues was forwarded to TVA on June 2,1986. He mentioned activities associated with the Sequoyah restart. Equipment qualification inspections are largely complete. Still open are issues involving the splicing of cable manufactured by Raychem, and an issue concerning high-energy line breaks.

He noted TVA's development of a Design Control and Reverification Program at Sequoyah which will probably become the model for Browns Ferry and Watts Bar. The Staff has essentially completed its welding inspection activity at Sequoyah with no major welding issues currently outstanding.

Major activities at Watts Bar will involve welding inspections, design control activity, and considerable review of employee concerns.

B. J. Youngblood, NRC, presented the four areas of Staff concern in the corporate area as outlined in the Staff's 50.54(f) letter to TVA from the Commission on September 17, 1985 (see Appendix XI). J. C. Ebersole and D. Okrent expressed their concerns regarding the disregard by the TVA board for the caliber of their nuclear program. C. Michelson expressed concern regarding the nuclear qualifications of the TVA Board members.

B. J. Youngblood indicated that the NRC has expressed concerns that the Board be cognizant / aware of actions taking place although the Staff has not gotten into the area of the qualifications of the individual Board members. He presented the major elements of the revised TVA Corporation Plan submitted to the Staff on March 10, 1986. He indicated that the Staff's preliminary review finds the approach acceptable as it provides centralized leadership and direction and places responsibility and accountability with one individual unencumbered with nonnuclear matters.

C. Michelson asked if the Staff believes that TVA has a focal point for safety considerations. He suggested that public health and safety considerations are dispersed among a number of different organizations.

J. Youngblood agreed that there does not appear to be a single focal point for safety unless one talks about licensing and engineering. C.

Michelson noted that it was pointed out at the Subconmittee meeting that the licensing group is not the single focal point for safety. J. C.

Ebersole suggested that Admiral White is the fccal point for everything, including the investigation of abstract safety issues which may be in conflict with demanding schedules. J. Youngblood indicated that the I

1 PROPOSED MINUTES - 315TH Staff is awaiting the latest TVA submittal of its Corporate Plan.

Present concerns involve the span of management control, the training of nanagers, and the transition from contract employees to TVA personnel.

D. Okrent asked if the Staff has evaluated its own shortcomings over the last few years regarding the TVA experience. R. H. Wessman indicated that the Staff has not yet conducted a self-evaluation regarding its handling of the TVA activities. The Staff's efforts at this time have been primarily directed toward review of TVA activities. The Staff expects to start an internal exanination of the NRC Staff's performance later.

C. C. Mason, TVA Deputy Manager of the Office of Nuclear Power, discussed the elements of the revised TVA Corporation Nuclear Performance Plan (see Appendix XII). He explained that Admiral White has concluded that TVA's most pressing problem is to improve its nuclear management. The root cause of the overall nuclear program problems has been identified as a lack of experienced nuclear managers who can provide leadership and direction, the lack of ability to hire, develop, and retain experienced nuclear managers, the fragmented and cumbersome organizational structure, and outmoded and incomplete management control systems appear to be the primary reasons for the decline of TVA's nuclear program. He explained that the organization was characterized by a lack of communication and coordination among various nuclear departments. This problem was ad-dressed by bringing all nuclear matters under one control focus, removal of nonnuclear matters, establishing clear simple lines of authority and responsibility, and developing consistency across the entire nuclear effort. He discussed the evolutionary process of reorganization of the Office of Nuclear Power. He noted that Admiral White did not find that TVA nuclear managers lacked technical skills or professional knowledge.

Rather he found a lack of commitment to and responsibility for achieving excellence in performance. The organization lacked aggressive and inquisitive managers who would reach out for problems, make decisions, and take control. D. A. Ward thought a management development program was missing from TVA's Corporate Nuclear Performance Plan. C. C. Mason acknowledged that fact and indicated that TVA is working on one for the total management complement of about 1600 management-scale employees. D.

W. Moeller asked if TVA has recognized the need to have indicators of performance to determine whether the changes being implemented are working. C. C. Mason indicated that TVA is participating in the INP0 performance indicator reporting system. He arided that TVA also monitors its regulatory performance regarding such things as quality assurance violations and radwaste incidents. D. W. Moeller asked for additional information regarding the role of INP0. C. C. Mason indicated that INP0 has agreed at the request of the TVA Board of Directors to visit annually and do a corporate evaluation until TVA's performance is back up to industry standards. C. Michelson asked about the general plan for replacing contractor employees with TVA personnel. C. C. Mason indicated that it will be on a case-by-case basis when Admiral White is convinced that the TVA deputy is qualified. Note was taken of the question

o PF,0 POSED MINUTES - 315TH regarding conflict of interest and legality of contractor employees supervising TVA personnel.

C. C. Mason discussed improvements in the management control system highlighted by formation of an integrated system of corporate procedures as well as a process of development of new corporate level standards and directives which can be used to evaluate and revise all existing nuclear procedures, as necessary. C. Michelson hoped that deviation reports for Watts Bar will be rechecked at Sequoyah and Browns Ferry in a systematic fashion to preclude such deviations at those plants. C. C. Mason indi-cated that he expected increased upper management awareness of nuclear activities, especially in the case of recurring or persistent problems with TVA's nuclear program.

J. C. Ebersole mentioned the fact that the process of handling a sub-stantial design alteration or modification was often not brought to closure, with design, construction, and operation taking a portion of the commitment and causing fragmenting of the process. He asked if this matter has been corrected. W. Orotleff explained the process by which a plant modification package will now be handled by TVA (see Appendix XII).

He explained how a change control board is attached to a stand-alone plant modification charged with following such a modification from detailed engineering to installation. The modification would be kept in a package configuration from inception to completion and would not be issued piecemeal to each discipline as in the past.

D. Okrent asked the location of individuals at TVA who do work with probabilistic risk assessment. C. C. Mason indicated that they are located in TVA's Office of Engineering in Knoxville. D. Okrent asked C.

C. Mason how much knowledge he had both of the quantitative results and the phenomenology of PRAs that have been done for TVA plants. C. C.

Mason indicated that he had very little knowledge of them, but indicated that his staff is looking into that and expects to provide briefings for him and Admiral White in the next week or so. He noted that Admiral White has about the same knowledge as he on PRAs. D. Okrent asked who was the first person below C. C. Mason to have intimate knowledge of PRAs. W. Drotleff indicated that the manager of the nuclear engineering branch would be that individual. He noted that he was not an expert in PRA areas. D. Okrent suggested that there is a deficiency in management knowledge. Knowledge of the import of these studies is missing from level after level of management.

D. Okrent asked if there is one group within the TVA organization that has no other responsibility except following the issue of safety, whether it be design, construction, or operation. C. C. Mason indicated that TVA does not plan to have an individual or a single group with that specific charge. The corporate nuclear safety conscience is provided by the Nuclear Safety Review Board. C. Michelson noted that it was once envi-sioned that there would be an advisory board to the TVA Board composed of five to seven members. C. C. Mason indicated that it has been decided

PROPOSED MINUTES - 315TH that TVA no longer needs an advisory board to the Board of Directors.

TVA will rely on the annual corporate INP0 evaluation, the Inspector General's reports, and other information that the Board gets. It is expected that all of the reports of the Nuclear Safety Review Board recocinendations will be provided to the TVA Board of Directors.

D. Okrent expressed concern that TVA was not properly prepared to deal with surprise problems that will surely arise. C. C. Mason indicated that TVA is doing investigations of allegations but one cannot assume that every problem has been found. D. Okrent nentioned a weld fabrication and weld wire problem which arose out of the last INP0 evaluation of TVA plants. C. C. Mason indicated that TVA has investigated allegations about welding and done an exhaustive investigation of the welding program at Sequoyah. He did agree that INP0 was extremely dissatisfied with the welding program at Matts Bar.

C. Michelson asked if TVA was aware of a problem associated with the quality of welding rods used. C. C. Mason indicated that he did not know of any problems associated with the records of weld rod quality. J.

Houston, TVA, indicated that there have been a nunber of allegations concerning the control of weld wire filament material. These have been extensively investigated as part of the Watts Bar special employee concern program called the Weld Evaluation Project. So far, the investigations have not shown any major problems with the control of material or the traceability of the weld. He noted that all concerns identified at Watts Bar which have generic applicability are being explicitly evaluated relative to the other plants within the TVA system in an organized systematic way. D. Okrent asked the results of the last INP0 review at TVA sites regarding the degree of satisfaction. C. C.

Mason indicated that the November 1985 INP0 corporate evaluation of TVA's nuclear program was extremely dissatisfied with Watts Bar and Browns Ferry, as well as the general corporate structure.

H. Hagedorn, ACRS consultant, summarized comments from his TVA-related memorandum submitted to the ACRS on July 2,1986. He spoke of the pros-pect of the ACRS doing substantially more work in the field of organi-

ation and management in the years to come if it accepts the following four assumptions:

Human factors play an inherent role in nuclear power plant safety.

The useability of control and monitoring equipment ~can never be separated from its design, construction, calibration, and maintenance.

A nuclear power plant cannot be kept sufficiently safe if its manage-ment's concern with safety is limited merely to complying with NRC regulations.

Appropriate management skills, attitudes, and activities can be identified and should be demonstrated by electric power utilities in the management of their nuclear power plants.

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PROPOSED MINUIES - 315TH

Only adequate use of technical competence and technical information permits rational consideration of the safety aspects of management decisions in nuclear power utilities. j H. Hagedorn suggested that the basic organization and management questions have been identified in the case of the TVA organization. The basic issue is the effect of corporate and power plant organization arrangements on TVA nuclear power plant safety. Does the organization of TVA hinder, permit, or foster safe operation? He indicated that TVA has given the accountability for nuclear power generation to Admiral White.

This accountability has assembled in one job what was distributed previously among several managers and several different functional organizations. He suggested that the main question regarding this organization is whether the restructuring works by providing better decisions more quickly.

H. Hagedorn suggested that a manager in charge of a large organization will become heavily burdened if he has more than four or five managers reporting directly to him. He pointed out that Admiral White's organi-zation chart shows twenty-six individuals reporting directly to him. He wondered whether White's reporting burden has become so great that it cancels any improvements made by collecting all of the reporting channels in the TVA nuclear power enterprise into his one accountable job. He thought the ACRS and the NRC Staff ought not to judge TVA's very flat organization too harshly as he speculated that it might be an interim organization and one set up primarily to encourage a closeness to the new boss. Furthermore he suggested that the flat organization has less structure than a more articulated organization and offers less chance of making a mistake in structure before further refining the present struc-ture if that becomes desirable. There is considerable flexibility built into this very flat arrangement. He noted that there is a possibility of j a backlog of people waiting to see Admiral White causing a possible

! problem regarding Mr. White's access to information. He wondered what i contingency plans there are for dealing with the situation if bogging I

down begins to occur.

l H. Hagedorn wondered about the organizational integrity at the plant level. He wondered whether the combination of site director and plant

! manager would have enough clout to assure the coherence of operations or l would other channels to headquarters diminish the operational integrity at the plant level. If operational integrity was to be assured through collective accountability, how would this work? The answer to this j question would be found in the revised job descriptions for middle and

senior level TVA managers. Because of the publicity accorded tc~ employee l concerns, he thought that the NRC would be well advised to seek I assurances that a reasonable degree of employee confidence in management has actually been restored before allowing the first plant or two to be restarted. He suggested some kind of employee survey to acquire detailed l information on this subject. D. W. hoeller asked if H. Hagedorn had l looked at any of the indicators of performance used by INP0 and had l

PROPOSED MINUTES - 315TH decided whether they truly measure whether management is adequate. H.

Hagedorn indicated that he did not know of performance indicators which would allow an easy before and after comparison of an organization.

P. Barton, ACRS consultant, agreed with TVA's assessment that the root cause for shutting down their nuclear plants was the lack of a sufficient number of experienced nuclear operating personnel. Based on his experi-ence with the TVA organization through the years, the two basic reasons why TVA was unable to develop and retain nuclear managers was first the salary structure which has the top pay limited by statute and a failure to develop a satisfactory performance evaluation program and management development program. He noted that the differential for five levels of management is but $6,000 per year and experienced nuclear plant managers can get offers of 25 to 50 percent more than what TVA can possibly offer.

The nuclear performance plan that TVA proposes to follow is heavily staffed for the next two years with contract personnel. To fill these positions with trained TVA personnel now on board or hire experienced people from the industry will require an extensive and well-organized management development program to motivate and install confidence in the present middle- and lower-management personnel. He expressed concern that considerably more information needs to be presented concerning how the bootstrap management development is to be accomplished in time to replace the contract managers.

P. Barton stated that he was not satisfied that the proper relationship and division of responsibility has been established between the Director of Nuclear Engineering and the Director of Nuclear Safety and Licensing.

Additional infonnation needs to be presented concerning the responsibility of nuclear engineering as it relates to system design and modifications, probabilistic risk assessment and its results, and safe operating procedures as they relate to system design and industry experience. He wondered who was responsible for nuclear fuel management and thought that it ought to be the responsibility of nuclear engineering and not nuclear services. He noted that the new nuclear engineering organization in Chattanooga has not presented fully the division of responsibility and coordination between the headquarters group located in Knoxville and the engineers located at each nuclear site. There appear to be four levels of management between the lead engineers en-site and the Director of Nuclear Engineering. He expressed concern regarding the authority of the site engineer to make plant modifications and assure that these modifications are in compliance with all nuclear commitments, regulations, codes, standards, and nuclear safety requirements.

P. Barton applauded the appointment of Admiral White and agreed with the decision of the TVA Board of Directors to place the entire nuclear program under the management of one man with full authority. He questioned, however, the effectiveness of the flat organization with twenty-six people reporting to Mr. White. As TVA begins to resume power i

operations he speculated that if the organization did not break down to l

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PROPOSED MINUTES - 315TH )

four or five individuals reporting to Mr. White it would by natural gravity break down into that type of organization. He suggested that high morale of nuclear plant operating personnel is one of the most important ingredients for plant safety. Restoring operation to Sequoyah would have a significant positive effect on morale .for the entire TVA organization and restore confidence in the present management.

D. W. Moeller speculated why the Sequoyah plant should be the first to restart and considered in better shape than the Watts Bar or Browns Ferry plants. He wondered what the basic reasons for the shutdown of Sequoyah were originally. C. C. Mason explained that recurring problems with the environmental qualification program at Sequoyah raised some serious questions as to the adequacy of the qualification program. Because of potential serious deficiencies in this area the plant was shutdown to investigate them in depth. P. Barton suggested that a nine-month period appears to be too long to organize a management development program since TVA needs to quickly train sufficient managers to replace the contract people. He thought that they would find it particularly difficult to accomplish the mission of the program unless the program were put into' operation much more quickly. C. C. Mason suggested that this management development program is to handle all of the 1500 to 1600 managers from the general foreman level up to help them improve their ability to manage and supervise. TVA will not rely on this particular program to develop replacements for the contract people who are in key positions in TVA's nuclear organization.

The Comittee discussed interest in and the requirements for writing a letter on the TVA revised Corporate Nuclear Performance Plan. It was decided that a draft letter would be presented later in the meeting.

V. Standardization Policy Statement and EPRI Advanced LWR Program (0 pen)

[ Note. H. Alderman was the Designated Federal Official for this portion of the meeting.]

C. J. Wylie reported that the ACRS Advanced Light-Water Reactor Design l Subcommittee met on July 9, 1986 to discuss the EPRI advanced LWR l requirements program to accomplish key prerequisites to the development of an advanced light-water reactor for the U.S. as an attractive energy

generation option for the time frame 1990-2000.

D. Scaletti, NRR, discussed the development of the current policy on standardization, which brought together the four concepts of reference i system, duplicate plant, manufacturing license, and replicate plant in one policy statement in 1978 (see Appendix XIII). He cited reasons for the need to revise the 1978 Standardization Policy Statement which included incorporation of provisions of the Severe Accident Policy Statement and the provisions of the draft Nuclear Power Plant Licensing I and Standardization Act, which would allow one-step licensing. He l presented a chronology of recent policy development which culminated in i

- . , . - - . . , _ . .--- - -.,,-,.,, ,--- -,,. . n n - ,n-, . - . . - , - , ~ , , , - - - - - - - , - - , , , , , , - - - - .-

PROPOSED MINUTES - 315TH the latest version of the revised draft policy which was sent to the Commission in May 1986. He identified the goals of the proposed Standardization Policy as an essentially complete plant and essentially final design detail and a reference system design certification. A concurrent NUREG iy to be published. J. C. Mark noted that the general issue of sabotage is not mentioned in either the Policy Statement or the accompanying NUREG which will contain implementation guidance for the Policy. D. Scale.tti agreed that sabotage is not included. F. Miraglia, NRC, indicated that the Policy Statement is an outline of Commission goals and objectives in the long term for standardization. It identifies a process by which the goals can be obtained but does not deal with the specifics of the process. He suggested that the Severe Accident Policy Statement has a provision which says that generic issues and unresolved safety issues have to be considered in the development of a new design.

One or two of these generic issues are related to sabotage (Unresolved Safety Issue A-29). J. C. Mark agreed that the Committee should not go into the details of how sabotage considerations should be implemented.

Nevertheless he thought it belonged in the policy statement.

H. Etherington was bothered by the term complete design and wondered whether to expect detailed drawings. D. Scaletti indicated that the design had to be sufficiently complete to support an operating license review. F. Miraglia acknowledged that defining the final design level of detail has been a concern among the Staff, Commission, and the industry.

In order to issue an FDA approval the industry would have to come to an accommodation with the Staff sufficient for the Staff to support the design in a certified process for 10 years. The license could be condi-tioned on specified inspection hold points, additional information hold points, and verification hold points. C. J. Wylie asked if that kind of detail would be in the NUREG that has been proposed. F. Miraglia indicated that the NUREG will have generalities, such as types of information that can be provided, as was in the AIF document proposed to the Staff in November 1985. Data to support the complete designation will be outlined in the NUREG but the Staff cannot expect site-specific data in the package from an applicant.

J. C. Ebersole asked and was informed that a standardized plant being

, submitted for approval at this time would have to meet only the current l

sabotage requirements and address USI A-29. D. Scaletti that, while a design would be approved for 10 years, if it were to be renewed at the end of that period it would have to meet the regulations current at that time. Any changes in the regulations within the 10-year certification period would have to be handled as a backfit. D. Okrent noted that some l regulations are front-fit only.

D. Scaletti discussed the certification process for a reference system design. He indicated that once this process is complete and a final design approval issued the design would be subject to one of fcur rule-making options:

i l

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PROPOSED MINUTES - 315TH

• Notice and coment

• Notice and comment with the right to request a legislative-type hearing Notice and coment with a mandatory hearing (also a legislative-type hearing)

• Formal on-the-record proceeding held by a representative of the Comission D. Scaletti indicated, in answer to a question by F. J. Remick, that the representative of the Comission could be the Comission itself, if it so wished.

D. Scaletti discussed the preliminary outline for the NUREG that supports the Standardization Policy (see Appendix XIII). He noted that the schedule for completion of the NUREG will be the middle of September 1986. D. A. Ward asked to what extent the NUREG will clarify the completeness of design scope in detail issue. D. Scaletti indicated that it is subtle. From the standpoint of design certification the applicant is required to comply with 10 CFR 50.34(b), the operating license information requirement.

D. Scaletti discussed conforming changes to regulations. He noted that 50.34(b), the CP/ML rule, definitely needs to be updated since it is inconsistent with the requirements of the Severe Accident Policy State-ment. This regulation currently applies only to five or six applications that were in place at the time of the TMI-2 accident.

D. Scaletti highlighted the proposed federal legislation on standardized designs, legislation specifically addressing the change to authorize one-step licensing, early site reviews and reviews of designs, certified standard designs with a 10-year approval, and an allocation of fees to users. It provides a congressional endorsement for stability of the licensing process. F. J. Remick asked if the Congress would have to approve this legislation before the Standardization Policy could go into effect. D. Scaletti explained that the Policy could be implemented through one-step licensing or two-step licensing, and that legislation 4

i would have to be approved before the Staff could use the one-step licensing process, as well as allocate fees to the users. This is different from the current 20 percent payrent each time a design is referenced for five periods. F. Miraglia clarified that the legislation proposed by the Comission is the same legislation that was proposed in 1982 and not acted upon. It was subsequently revised in 1985.

F. Sears, Vice President of Nuclear and Environmental Engineering for Northeast Utilities, and representing the Atomic Industrial Form (AIF) i Study Group on Practical Application of Standardized Nuclear Power Plants in the United States, discussed the prospect of nuclear energy as a generation alternative. He took note of the potential increase in demand for electricity in the 1990s and noted that many utilities are beginning to plan for new generating capacity. Unfortunately, nuclear energy is

PROPOSED MINijTES - 315TH not viewed as a viable generation alternative from an economic point of view. Part of the perception is an unstable and unpredictable licensing process which needs to be changed. Nuclear power is also very expensive and much too long a period of time is required to license and construct a plant. He explained that the AIF study group has derived certain proposed revisions which they believe will improve the viability of nuclear power, as well as enhance nuclear plant safety. These revisions specifically address a nuclear plant design certification process and a combined construction and operating license.

F. Sears indicated that designs for standard plants would be developed by a consortium. of utilities, vendors, turbine-generator vendors, architect engineers, and constructors (see Appendix XIV). The objective of design certification is to develop a detailed plant design which completely satisfies regulatory requirements before construction begins. It would then be licensed and the design frozen. The regulatory requirements would be thoroughly defined and incorporated and public input would be considered before the construction of the plant. F. Sears discussed the meaning of a complete design, indicating that small piping might not be included but the general cable layouts and piping runs would already be done before the start of construction. It was noted, in answer to a question by M. W. Carbon, that the PRA methodology would be decided before the design certification was issued by the NRC, but the PRA would not be completed until just prior to the start of construction. Public discussions of the adequacy of the plant design itself will occur at the time of the design certification application review and approval. Only site-specific issues would be considered at the construction and operating license stage. After the construction and operating license has been issued and construction begun, confirmatory orders would be conducted by both the NRC and the owners. These orders would ensure that the plant is being constructed in accordance with the approved requirements. lipon completion of construction the NRC would finalize its audited results and the owner will begin plant operation. C. Michelson asked if the industry plans to go directly to the design certification process or intends to continue to go through the FDA process first and then get that design certified. F. Sears indicated that it is the belief of the industry that the best way is to go directly to the design certification. C. Michelson noted that, at the time of the certification, it would appear that there was 'very little left to do in terms of what is needed for the design. He asked how many drawings would remain, or the kinds of drawings that would remain, in order to actually construct the plant from the design package. F. Sears indicated that at the time of the construction and operating license over 70 percent of the drawings would be completed. No systen would be started until it was about 95 percent complete.

F. Sears discussed some of the benefits of the design certification process. He indicated that the primary benefit, if it is successful, will be that nuclear power will again be competitive with other power generation alternatives. Reductions in cost will occur from two

PROPOSED MINUTES - 315TH significant factors: The cost of borrowed. funds will be reduced because the investors and the lenders alike will perceive that project progress will be predictable and the investment reduced. The time from start of construction to plant operation will be significantly reduced. The result will be reduced re-work, optimized design, and approved construction management. The proposed design certification / construction and operating license process will enhance public participation in the licensing process since the public has available the full information on the plant design. The public has a second opportunity to question the design on technical issues when the site-specific issues are taken up.

One is able to separate the design of the plant from local considerations.

F. Sears, Chairman of the Utilities Steering Committee for the Advanced Light Water Reactor Program (ALWR) of the Electric Power Research Insti-tute (EPRI), indicated that there are four major elements to the EPRI ALWR program in the U.S.:

Determine the set of regulatory requirements for the next generation of light water reactors Generate a set of utility-approved and NRC-certified plant requirements documents for advanced LWR plants Develop a detailed engineering design and obtain NRC licensing certi-fication for the next generation light water reactors Develop a conceptual design for a less-than or equal-to 600 MWe advanced PWR/BWR F. Sears indicated that the 600 MWe plant concept is being considered because one can make step changes due to the small size and rely more on i passive features. G. A. Reed expressed concern regarding the economic

! viability of a 600 MWe plant. While he noted that there are advantages 1 to a smaller-sized nuclear plant for some customers, overhead of the regulatory process as well as the need for some minimum level of plant

, staff should limit the attractiveness of smaller size. F. Sears

contended that there are simplifications planned that will make a smaller l plant more economical. D. Okrent asked what EPRI's targets are for public safety in the requirements document.5 F. Sears spoke of a core damage frequency target of less than 10- per reactor year and an exposure to the public at the sgte boundary during an accident of 25 rem at a frequency of less than 10- per reactor year (see Appendix XV). J.

C. Ebersole pointed out that there is no guidance in the current regulatory structure about what to do beyond general identification of the issue of sabotage. F. Sears characterized the subject as one of various unresolved safety issues which are being addressed in the industry's requirements documents. C. Michelson wanted to see how EPRI addresses the issue of fire protection. K. E. Stahlkopf, EPRI, indicated that fire protection has not yet been addressed in the requirements document.

K. E. Stahlkopf differentiated between the 13 chapter utility requirements document which is centered around an 1100 MWe plant and the

PROPOSED MINUTES - 315TH small plant design development effort which involves proposals from four teams representing the four major light water reactor vendors in the United States. The small plant program has undergone two interim reviews which resulted in the dropping of the Combustion Engineering proposal as well as the Babcock & Wilcox and UE&C team. The two teams remaining are one BWR team composed of GE/Bechtel/MIT and one PWR team composed of Westinghouse / Burns & Rowe. Both conceptual designs lean heavily toward passive safety features.

K. E. Stahlkopf discussed submission of the EPRI requirements document for NRC approval. He indicated that the NRC will be issuing a draft SER for each chapter, with a final overall SER when the 13th chapter is completed. Memoranda of Understanding (MOU) exist between DOE, EPRI, and the vendor who is doing work toward design certification. He also indicated that there are MOUs with Taiwan Power for their participation in the utility requirements document, with Japan Atomic Power Company for their participation in the development of small reactors, and under negotiation is an MOU with Korea Electric Power and Kansai of Japan for their participation in this program. He noted that the program will not be tilted to meet the needs of the foreign countries. The MOUs make it very clear that these foreign utilities are participating to gain education in the American system.

K. E. Stahlkopf discussed progress on the remaining unresolved safety issues, noting that 588 issues were identified in 1983, and that, with new issues arising at about the rate of 30 per year, in 1986, there were 703 total issues that needed to be addressed. He indicated that to date EPRI has been able to resolve all but 46 issues.

D. Noble, EPRI, indicated that the utility requirements document repre-sents those features that the U. S. utilities desire in advance. It is an extensive compilation of design, construction, and plant performance requirements for an advanced plant. It reflects NRC goals and criteria.

EPRI views this document as a starting point for subsequent detailed engineering. The document establishes conformance to utility require-ments, and provides the basis for the development of standard plant designs.

D. Noble discussed the substance of the utility requirements document, noting that Chapter 1 is a statement of the overall requiyents and goals. D. Noble spoke of an overall basic requiremer.t of 10 for core melt frequency. D. Okrent cautioned that one should be able to present reasonable assurance with regard to uncertainties that this objective can be achieved. He also thought the issue of containment ought to be specifically addressed with regard to possibly achieving improved safety.

D. Noble mentioned that an 87 percent overall availability is called for in the requirements document.

i D. Okrent roted a 92 percent annual plant availability in Japan where at least some components are subjected by vendors to specific tests that go beyond the design basis. D. Noble indicated that EPRI has not included,

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PROPOSED MINUTES - 315TH ,

!. as yet, specific requirements for component testing, but plans to discuss

. the performance of specific components in future chapters. G. A. Reed f

challenged the use of a two-year refueling interval for the new design.

He noted that this implies a spectral shift of some variety which would add complexity and an increased number of control rods. He thought it certainly difficult to achieve 87 percent availability in the first 10 years. J. C. Ebersole asked if EPRI intends to invoke reliability re-quirements in equipment specifications. D. Noble indicated that the

approach is not to produce an equipment specification but to set

< difficult requirements that the suppliers of equipment and systems will have to meet. J. C. Ebersole wondered how EPRI will avoid the low-bidder problem. D. Noble indicated that EPRI intends to establish the j . availability reliability requirements in performance specifications. C.

J. Wylie asked if EPRI will specify seismic fragility levels, noting that such levels can be very important to work on PRAs. D. Noble indicated that EPRI would be specified where necessary to establish the design requirements in meeting overall levels.

4 D. Noble indicated that waste generation treatment systems for the new plant will be designed to produce no more than 2,500 cu. ft per year--a j significant improvement over experience with current plants. In the area l

of occupational exposure,100 man-ren per year is established as the

requirement. The requirements document will have goals in the cost area, an attempt to achieve 6.5c per kilowatt-hour (40 percent improvement over 4

plants licensed and in operation since 1985). A 54-month con-structability schedule was advanced from the start of structural concrete through the startup and testing process to initial operation.

D. Bunche, DOE, indicated that it is DOE's goal to force the development and certification of simpler, safer and more reliable light water

reactors for the future (see Appendix XVII). DOE intends to closely

coordinate its activities with the EPRI industry ALWR program. He

!. indicated that DOE has been encouraged to support a review of utility

! requirements and a matchup of the products with the NRC's needs. There 1 are two product lines for which DOE has specific schedules, one of which is the General Electric Hatashi-Toshiba advanced boiling water reactor

that has been under development in Japan and a Combustion Engineering system-80 program for which there is an FDA.

D. Bunche spoke about mid-sized reactor development programs which are to

be handled as a two-stage process. The first stage will put the gut issues on the table, such as basic criteria, and a second stage will look

at implementation such as a certification process. G. A. Reed suggested a reorientation of nuclear regulation to a concept called a " safety enve-

. lope." The concept would be to protect the health and safety of the public and prevent core melt. This would be accomplished by strict regulation of a set of safety systems to consist of the decay heat l removal system, a high-integrity reactor pressure vessel, a containment structure of high integrity, and a subcriticality system. D. Bunche agreed that this was a good idea and that it had already been explored in

PROPOSED MINUTES - 315TH '

the early 1980s as part of a process of restructuring regulatory operations. He thought the concept had a lot of promise, but also saw considerable difficulty without inclusion of a lot of development and testing.

D. Moran, NRC, discussed the Staff plan for review of the EPRI ALWR requirements document. The Staff plans to spend a nominal six months per chapter on the 13 chapters to be transmitted, based on EPRI's ability to transmit. The 13 chapter titles were presented (see Appendix XIII). He listed the Staff disciplines required as including site and environmental, reactor systems, plant systems, materials , plant structures, and electrical systems / instrumentation and control.

D. Okrent asked where the Staff plans to bring PRA, severe accidents, and sabotage into that set of disciplines. D. Moran indicated that the PRA requirement has to be considered after the design has been generated.

D. Okrent wondered about containment performance. D. Moran indicated that existing regulations will be used to measure whether EPRI's document has incorporated those regulations in their specifications, as well as a resolution of safety issues which are ongoing. J. Kniel, NRC, indicated that EPRI expects to address the implementation of Severe Accident Policy requirements and the NRC will be reviewing the requirements document in light of the Severe Accident Policy.

D. Moran discussed future interactions with the ACRS. He noted that the Staff will keep the ACRS informed of the program status, including fur-nishing copies of requirements document chapters. If key topics of significance arise, the Staff will schedule discussions with the appropriate subcommittee. The Staff also wishes to report from time to time on groups of chapters and, when the final SER is ready in final draft, the Staff intends to have a final meeting with the ACRS prior to issuing the complete SER. C. J. Wylie asked the Committee if it thought the Staff's procedure is acceptable. D. Okrent suggested that at some intermediate point the full Committee should look at the more significant aspects and discuss them before the complete package is set. R. Hernan, NRC, indicated that, within the budgetary constraints of the ACRS, the Staff would appreciate written coninents when appropriate, and certainly at the conclusion of the Staff's review and the ACRS review of Chapter 1.

The Committee discussed a possible letter on the Standardization Policy Statement. C. J. Wylie thought that the issues of " essentially complete," which was questioned today, the omission of a statement in the policy statement regarding sabotage and security, and some discussion regarding the terms of certification (10-year terms) should be included in the letter. D. Okrent pointed cut that there appear to be significant differences in the current version of the Standardization Policy Statement and the Safety Goal Policy adopted recently by the Commissioners. He also had some question concerning containment capability which should be addressed.

s PROPOSED MINUTES - 315TH VI. Technical Specifications - Proposed NRC Policy Statement (0 pen)

[ Note: J. O. Schiffgens was the Designated Federal Official for this portion of the meeting.]

C. Michelson indicated that the Plant Operating Procedures Secommittee held a meeting on July 1 to consider the Commission Policy Statement on Technical Specifications Improvement. The purpose of the Policy Statement was to clarify the scope and purpose of what technical specifications should be and was to serve as a basis for industry and the NRC to implement a voluntary technical specifications improvement program. The commission paper discussed alternatives to making technical specifications more useful and considered methods for reducing their number and detail. He noted that the ACRS is obligated to provide its written views on the Policy Statement.

E. Butcher, NRC, indicated that phase one of the NRR Technical Specifica-tions Improvement Program, or initial studies, was a problem identifica-tion and recomendations study which began in 1982 (see Appendix XIX).

The second phase of the program involves implementation. Initial efforts by the Staff were a proposed rule in 1982 which spoke to two classes of specifications, real specifications which required prior Staff approval before being changed and a second category called supplemental specifica-tions--which could be changed by licensees without prior Staff approval, using a process similar to 10 CFR 50.59. This program was never imple-mented because of difficulty in de#ining the specific criteria for making judgments between the real technical specifications and those that would go into the supplemental document.

E. Butcher mentioned a second major study on testing requirements in technical specifications published as NUREG-1024. He indicated that H.

Denton, Director of NRR, commissioned a Technical Specification Improve-ment Project, a group to study all issues relevant to technical specifications. At the same time, an industry study was initiated by the Atomic Industrial Forum. The current policy statement focuses primarily on the results of the AIF and Technical Specification Improvement Project reports, and details about 25-30 different recommendations. An almost l universal comment from the studies was the lack of well-defined criteria l

for technical specifications which has resulted in many requirements that are not necessarily of prime safety significance and documents. This tends to deflect attention from the items that are of importance and ought to be in the specifications in many cases. A second general conclusion was that there was a reluctance by both licensees and the NRC to use documents and control mechanisms other than technical specifications for setting forth detailed requirements to be followed by a licensee. There was a reluctance to rely on the FSAR, procedures, the QA program, the security program, or the fire protection program. The language in the technical specifications tends to be confusing--there are format problems. Test frecuencies are often too high, and there is often too much testing on a given component that cannot be justified on a regular basis. Allowed average times in action statements are often i

PROPOSED MINUTES - 315TH completely inconsistent with those rigorous risk evaluations which have -

been done. Many unnecessary requirements tend to distract attention of the operators.

E. Butcher discussed four specific recomendations to be implemented by the policy statement:

Define scope and purpose of technical specifications Revise standard technical specifications to correct current human factors and technical weaknesses Continue development and use of PRA for technical specifications Upgrade the use of other tools including 10 CFR 50.59.

E. Butcher explained that the new system which will have regulatory requirements documented in more than just one place will be pared down to

- about 60 percent of the total current technical specifications. C.

Michelson wondered if the Staff is contemplating transfer of the remaining 40 percent of the technical specifications only as a first step.

into some other document not controlled directly by NRC. The utility could make the judgment to discard the technical specifications under a 50.59-type analysis. E. Butcher agreed that a good deal of the 40 percent would be dropped from this phase and ultimately disregarded l altogether. He noted that the combination of programs the Staff has in place to upgrade the use of the tools that the utilities will be exercising to dispose of some of these technical specifications will 1-assure that nothing of significance is eliminated. Certain short-term improvements will be made using PRA in parallel with the development of the new standard technical specifications. These will be items such as

changing the allowed outage times and surveillance intervals. The Staff also intends to consolidate and simplify limiting conditions for operations (LCO).

I D. Fischer, NRR, indicated that the policy statement was developed based upon recommendations of earlier studies, such as NUREG-1024. Criteria were validated by trial-use studies and rf W evaluations, and fine tuning was done during NRC/ industry working crou; meetings. Both NRC and AIF i attempted to apply the criteria to t% N Creek and Limerick Technical Specifications separately. The cSnt. :+;s the policy statement, other than a short summary and some bad.grou..d discussion, are a subjective statement of the scope and purpose of technical specifications and objec-

, tive criteria for determining the content of the revised technical speci-i fications. There are three design-basis accident-based criteria consis-tent with the existing regulatory framework and licensing basis. Four additional systems are included as well as a discussion of plant-specific and generic risk. J. C. Ebersole asked how the technical specifications will address the need for feedwater on BWRs. E. Butcher indicated that auxiliary feedwater specifications will stay in the technical i

specifications. D. Fischer explained that since the implementation of the Policy Statement will result in the relocation of some requirements to other licensee-controlled documents, for example, procedures, FSARs, or programatic documents like the QA plan. Control mechanisms will be l

PROPOSED MINUTES - 315TH ,

needed to indicate how the licensee can decide to use a specification,

. dispose of it, or prepare reports under it. The Staff will use the model of 10 CFR 50.59, the unreviewed . safety question mechanism, as one possibility. The Policy Statement will state that requirements relocated to procedures must be accompanied by an administrative control section to assure that the procedures are not. changed improperly. A multidisciplinary review must be made of technical specification changes before they are implemented.

D. Fischer discussed the purpose of technical specifications and the three criteria the Staff have used to decide which items should remain in the Technical Specifications and which should be relocated (see Appendix XIX). Notwithstanding these three criteria, the Staff has identified four systems that don't meet these three criteria, but, based on operating experience and on risk, need to stay in the technical specifications. Those systems are the reactor core isolation cooling system or isolation condenser, residual heat removal system, standby liquid control system, and the recirculation pump trip system.

'D. Fischer indicated that the Staff anticipates a 40 percent reduction in the number of LCOs as a result of the Policy Statement. In addition, reports on relocated technical specifications will no longer be reported as LERs. Some of the lost reporting data might be captured under plant

shutdowns that might be reported otherwise. F. Hebdon indicated that situations might trigger other criteria that are in the LER rule and result in the situation being reported. D. W. Moeller asked if the Staff anticipates some reduction in the number of LERs. F. Hebdon indicated that about 30 percent of the LERs arise basically from the technical specifications, and since about 40 percent of the technical specification i requitements are being relocated there ought to be on the order of a 10-12 percent reduction, which is about 250-300 LERs per year. D.

Fischer indicated that the change to the technical specifications should

, be risk neutral since the relocated items are still enforceable. On a generic basis, anything relocated which is significant to risk will be captured in the Policy Statement.

The Committee discussed safety systems which meet the criteria at Wolf l Creek and Limerick. They discussed limiting conditions for operations l

which do not meet the criteria and currently have action statements that limit reactor pcwer. It was noted that when an LCO does not pose an inanediate threat to the public health and safety, such as remote shutdown instrumentation or ordered chemistry specifications , these LCOs were removed from the technical specifications. C. Michelson pointed out that '

all of fire protection has been taken out of the technical specifications. This means that licensees will no longer have to submit LERs on the problems they are having with their fire protection system. ,

C. Michelson expressed concern that data will be lost on the failure, malfunction, and general reliability of fire protection equipment. E.

l Butcher agreed that he was not aware of a separate reporting requirement on fire protection equipment. C. Michelson noted that very little is

PROPOSED fiINUTES - 315TH known about the reliability of that equipment. It is a subject which requires greater attention. E. Butcher explained that all of the requirements to maintain the fire protection system were made license conditions that specify the conditions under which the licensee can change that program and provides a set of criteria somewhat like 50.59.

Enforcement was actually escalated by this change. Alternate shutdown systems and the rooms that house them were made part of fire protection and remain a part of the fire protection program. C. Michelson noted that there were a number of potential accidents, such as pipe breaks outside of the primary containment which would not be accommodated under these criteria and would be removed because that particular break is no longer officially a design basis. E. Butcher indicated that the Staff will take another look at pipe breaks outside ~ containment as an event, j and see whether anything was taken out of the technical specifications.

D. Fischer indicated that the Staff intends to submit the Policy Statement to the Commission in August 1986, for public coment in October, and for final issuarce in January 1987.

VII. Reactivation of Deferred or Cancelled Nuclear Power Plants (0 pen)

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

11. B. Spangler, Special Assistant for Policy Development, NRR, briefed the ACRS on the scope of the preliminary study of LWR reactivation, issues, NUREG-1205, entitled " Reactivation of Nuclear Power Plant Construction Projects" (see Appendix XX). He explained that five different tasks were assigned to him by H. Denton, Director of NRR:

1. Prepare a tabulation of current status of cancelled or deferred plants

2. Identify safety and environmental issues arising from significant new information

3. Identify regulatory or policy issues and information useful to a determination of the adequacy of existing rules or policies and the desirability of improvements ~

4. Identify regulatory options and decision criteria for dealing with the above issues

5. Identify decision considerations that would determine Staff require-ments He indicated that the Atomic Industrial Forum in a letter in June 1986 to H. Denton clearly indicated that they did not wish any changes even if they would lead to a more cost-effective regulation on the grounds that improvements would be destabilizing. He expressed caution regarding

FR0 POSED itIMUTES - 315TH identification of decision considerations that would determine Staff requirements noting the interrelationship between NRC decisions and the decisions of utilities. The NRC must be careful not to implement policies that would discourage reactivations by utilities.

M. B. Spangler indicated that there are some potential safety and environmental issues that derive from the Staff's requirements for staffing. They are not only influenced by the number and timing of utility decisions to reactivate but also by the number of safety and environmental issues to be dealt with, particularly those arising from new safety information. He presented a list of regulatory questions for achieving cost-effective measures of equipment maintenance and quality preservation of deferred and cancelled plants. J. C. Ebersole asked if the NRC has placed any requirements on cancelled plants to hold the condition of that plant in any given state. Of particular interest was a requirement on maintaining the status of the unused equipment.

M. B. Spangler indicated that he did not believe there was any such requirement. C. Michelson asked if any utilities have mothballed their nuclear plants to keep them in a state of cleanliness for a relatively long-term situation. M. B. Spangler indicated that Washington Public Power Supply System had a very ambitious program. It was so ambiticus that the cost per year was a target of criticism about the need to cancel a plant rather than to keep mothballing it. As a result, they now have a program for a minimal preservation and maintenance of this equipment that would still accomplish the task but at less cost. J. C. Ebersole asked if any utility has an agreement with the NRC regarding preservation of the equipment when they abandon a plant. T. Michaels, NRR, indicated that this only applies if it a deferred plant; if it is a cancelled plant the construction permit is withdrawn and the NRC has no authority over the utility. The Committee discussed the fact that some utilities have cancelled plants and attempted to sell some of the equipment at a fraction of its acquisition value. C. Michelson expressed concern that the NRC keep track of this equipment to make sure that the proper

" pedigree" has been preserved and controlled. M. Spangler indicated that IE does not inspect the equipment on cancelled plants.

M. B. Spangler discussed five proposed objectives of the LWR reactivation policy. One objective, of particular significance to cancelled plants rather than deferred plants, was to ensure that NRC's review process will continue to provide appropriate public access for expressing significant concerns regarding LWR reactivation. With regard to deferred plants, if the utility has not had an OL review a public forum would still be avail-able. He suggested, with regard to cancelled plants, a rulemaking action to simplify the treatment of generic rulemaking on various issues that would provide a more effective way of assimilating public input into adjudicatory hearings. If there are unique aspects in a plant or site, adjudicatory hearings at that given site or plant design might have considerable value. But, discussing generic issues is a waste of re-sources, even intervenor resources. M. B. Spangler discussed options for improved regulatory review of LWR reactivation projects for plants with 1 i

PROPOSED liINUTES - 315TH cancelled construction permits. These options included a generic rule-making and a one-step licensing CP/0L.

T. Michaels, NRC, discussed the Policy Statement on Deferred Plants developed in response to Commission directives on the subject. A Staff task force consisting of members of NRR, IE, RES, Administration, and the Regional Offices was formed to consider and develop the following consid-erations (see Appendix XXI):

Documentation, maintenance and preservation requirements for deferred plants Applicability of new regulatory Staff positions for deferred plants being reactivated Procedures for reactivating deferred plants Identification of regulatory improvements and research initiatives Population and status of deferred and terminated plants T. flichaels characterized certain derived definitions. A deferred plant was characterized as one where the licensee expects to reactivate in time, termination has not been announced, the construction permit is still in effect, but construction has either ceased or been reduced to a maintenance level. The terminated plant is one where the licensee has decided to terminate construction permanently, but the construction permit is still in effect and has not been withdrawn by the NRC. A cancelled plant is one where the construction permit has been withdrawn by the NRC. He noted that the regulatory issues associated with cancelling plants and the reactivation of cancelled plants have much in common with the revised Policy Statement for the improved treatment of regulatory issues for new plants. One of the more important aspects of the Policy Statement is the identification of the applicable regulations and guides dealing with documentation, maintenance, and preservation requi rements. These documentation, maintenance, and preservation requirements apply to plants under construction as well as to plants with deferred status. The licensee may choose to modify existing commitments during extensive construction delays by developing a quality assurance plan that is commensurate with the expected activities and length of delay. The position being taken within the Policy Statement is that plant-specific backfits of new Staff positions will be considered in accordance with the backfit rule. Generic backiits will either be implemented through rulemaking or generic issue resolution through a ,

backfit analysis similar to 50.109. He noted that the Severe Accident Policy will apply. It was noted that the Policy Statement covers the situation where a licensee wishes to terminate a plant and have the construction permit withdrawn. If there is a hearing board in place, the hearing board must be satisfied that site-stabilization criteria have been met, such as grading requirements and attention to other environ-mental issues. The plant must also be rendered inoperable as a l

l

PROPOSED HINUTES - 315TH '

i nuclear plant. F. J. Remick agreed that the NRC would want to maintain i

control until environmental considerations are satisfied before the construction permit is withdrawn.

T. Michaels explained that there are no safety implications in this

Policy Statement. It essentially consolidates existing requirements and

has been coordinated with the nuclear industry. He expected submission ,

to the Commission on July 21, 1986.

_ VIII. BWR Containment Performance During Severe Accidents (0 pen)

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

, R. Bernero, NRC, indicated that there are 24 either currently operating

! or planned BWR nuclear plants with a Mark I containment and 9 BWRs with a Mark I.I containment. In accordance with the NRC Severe Accident Policy,

future plants will be _ subjected to a review similar to that for a standard plant such as GESSAR where a conventional review and a PRA

should be prepared. Existing plants are currently considered to be safe

enough under the Policy Statement. The IDCOR process is reeded for existing plants as part of the review, and individual plant evaluations i will be done. There will be generic treatment for generic natters (see-Appendix XXII). Generic matters affect the pressure suppression l containments. A subset of this affects the Mark Is.

R. Bernero discussed the development of emergency operating procedures for prevention and mitigation. He indicated that the emergency procedure guidelines (EPG) are reviewed by the NRC Staff. These result in a proce-dures generation package (PGP) which eventually become the emergency operating procedures (EOP), which are audited by the Staff. He noted that there is a chaotic situation at present in the industry regarding these emergency operating procedures. As a result, the NRC Staff wishes now to concentrate on review of the latest version of the EPG (Revision 4). W. Kerr asked if the Staff is promulgating a containment performance criterion. R. Bernero indicated that what is proposed for a criterion is that there be substantial assurance of no release from the containment for a core melt. He indicated that the IDCOR individual plant l evaluations methodology is before the NRC for review, and about six individual plant evaluations (IPE) are now available. He noted the

characteristics of a BWR are that it accepts release of primary coolant

< everywhere in the plant and makes all pumps dynamic decay heat removal

devices.

R. Bernero indicated that the Staff is concerned about E0P strategies.

There does not appear to be a clear policy regarding containment pressure i management. The Staff is also not sure whether the IDCOR individual

plant evaluation is a quantitative assessment or an evaluation to decide whether changes are worthwhile. The Staff is not sure whether the IDCOR i IFEs will resul t in any generic solutions or evident strategies. He

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

PROPOSED filNUTES - 315TH recalled that WASH-1400 assumed passive plant operators and omitted the characteristics of the nuclear machine.

R. Bernero explained that his proposed generic letter would be a five-element approach: 1) in-place hydrogen control; 2) sprays which exploit backup water sources and pumps; 3) management of containment pressure to avert uncontrolled overpressure failure--use of a controlled release path (scrubbing) [of import is what the appropriate threshold pressure for action is]; 4) reduced likelihood of failure by direct attack from core debris; and 5) integrated clear symptom-based training and procedures strategies.

R. Bernero discussed the fact that the threshold pressure for containment venting is an open issue subject to further examination. J. C. Ebersole suggested that it is a design deficiency that should be repaired. The licensee must vent the containment before he is no longer able to vent.

R. Bernero noted that the torus in the Mark I containment is susceptible to failure by direct attack from core debris during a core melt accident.

He indicated that the Staff is looking for substantial assurance that the containment will mitigate a core melt accident by a combination of pressure control and spray quenching of the core debris. He put forth a suggested approach and advance notice of a proposed backfit to solicit public comment. Implementation would be made with the IPE and EPG, Revision 4 W. Kerr expressed concern that the Staff approach would lead to the elimination of the containment as a barrier. R. Bernero indicated that was not so. He stated that IDCOR is doing a review of this suggested approach and the results will be available to the Staff by August 1986. He indicated that the Staff plans to develop a generic letter and mesh that with the IDCOR results. The Staff would then return to the ACRS. D. Okrent remained unconvinced that the five-element approach was a complete set.

D. A. Ward indicated that the ACRS Subcommittee on Containment Requirements would be prepared to handle the Staff submittal. He suggested that the steam generator tube rupture issue regarding the controlled release of radioactive materials was a PWR analogy to the containment venting issue for BWRs. He suggested that the bleed-and-feed concept for PWRs may put the core in danger of melting.

IX. Report of Subccmmittee on Auxiliary Systems (0 pen)

[ Note: S. Duraiswamy was the Designated Federal Official for this portion of the meeting.]

The Chairman of the Subcommittee on Auxiliary Systems reported regarding fire protection concerns in nuclear plants. C. Michelson and J. C.

Ebersole spoke on the decimation of the completion of full-scale room test fire protection programs in the NRC research program. J. C.

Ebersole suggested that the Ccmmittee recommend that the work being done at the Sandia National Laboratories on fires in large areas be

9 PROPOSED MINUTES - 315TH reinstated. He noted that there are Differing Professional Opinions among NRC Staff members concerning interpretation of Appendix R requirements. There is a question as t, whether a piece of equipment should be allowed to be considered on'j partially destroyed in a fire.

The assumption of partial destruction would result in an assumption of partial competence of such a piece of equipment. He raised questions regarding the adequacy of backup control centers, convergent circuits which are protected for one hour. He suggested that a backup control center does not have the ability to handle a local fire, and an assumption that it can is a nonconservative interpretation of Appendix R.

C. Michelson suggested that Appendix R should be strictly implemented to assure that backup control centers will be competent.

The Committee discussed the fact that, in all but three nuclear plants in the United States, if the control room and cable-spreading room are lost to fire, the operators must evacuate through a tunnel to the alternate shutdown panel. If there is no reliance on three-hour fire barriers, but only on one-hour fire barriers, the only barrier to safe shutdown is the alternative 20 feet of separation between different cable trains. He' noted that the research terminated was on fire propagation based on the 20-foc t separation. Research is needed on the 20-foot separation question and on the reliability of fire protection equipment. The Staff has decided to rely only on Appendix R. C. P. Siess suogested that it might be desirable to take time out to reassess what is really needed.

C. Michelson indicated it is a question of safe shutdown capability in the event of fires. D. A. Ward asked C. Michelson if he wanted to draft a memorandum to the ED0 requesting restoration of the funds. C. P. Siess pointed out that the office directors of Research and NRR have made a decision that there are more imporcant projects that need to be funded.

A. Datta, RES, identified the phase one fire protection research programs that have been abandoned (see Appendix XXIII). He discussed the impacts i of the loss of this research. He indicated that component failure threshold data will now be unavailable and this was the main focus of the program.

i X. Executive Sessions (0 pen)

[ Note: R. F. Fraley was the Designated Federal Official for this portion i

of the meeting.]

A. Subcommittee Assignments

1. Additional Items Relating to ACRS Peport on the Restart of I Davis-Besse Nuclear Power Station, Unit 1 During the discussion of the report entitled "ACRS Comments on j

I the Restart of Davis-Besse Nuclear Power Station, Unit 1," C.

Michelson proposed that three additional safety items be included. It was decided that a list of these items be famarded l

l

1 9

PROPOSED MINUTES - 315TH to Chairman Ward for consideration during future appropriate subcommittee meetings. They are as follows:

1) Hot Leg Level Indication -- Lack of Reactor Vessel Level Indication The viability of using hot leg level indication during the draining and refill portions of an accident should be inves-tigated. Included should be such subjects as loop-to-loop oscillations and the susceptibility of the level measuring arrangement to the blowdown and refill process, disruptions such as flashing, and the frothing process in the vertical pipe. [C. Michelson suggested the Instrumentation and Control Subcommittee.]

2) Blowdown Valve Operability The ability of blowdown valves such as PORVs or the special blowdown valves proposed for Davis-Besse to accommodate water slug flow during the blowdown process without jeopardy to valve operability has not been established. This should be explored at the subcommittee level. [C. Michelson suggested either the Reliability Assurance or Thernal Hydraulics Subcommittees for exploring this matter.]

3) Environmental Qualification The pressurizer PORVs will be ultimately exposed to the environment created by their operation. It is not clear how the valves will be environmentally qualified. This should be reviewed. [C. Michelson suggested the Reliability Assurance Subcommittee.]

2. BWR Containment Performance During a briefing by R. Bernero, NRR, on the performance of pressure suppression containments (principally BWR Mark Is),

there was discussion of the development of a generic letter by the Staff to deal with the consequences of serious core damage

[e.g., H control, pressure control / venting, containment (dry-well)2 of heat (e.g.

molten debris removal

, into(e.g., reliability the wetwell )of sprays),

, operator E0Ps transport for severe accidents]. D. A. Ward indicated that the Containnent Performance Subcommittee will review the generic letter when it is available.

PROPOSED MIriUTES - 315TH 3. Safeguards and Security J. C. Mark suggested cancellation of the August 5,1986 meeting of the Subcomittee on Safeguards and Security which was to review the Technical Assistance Program on the " Evaluation of Methods of Reduction of Vulnerability to Sabotage (Generic Issue A-29)" because the A-29 program has been proceeding more slowly than anticipated. The Committee concurred.

B. Reports, Letters, and Memoranda

1. ACRS Comments on the Restart of Davis-Besse Nuclear Power Station, Unit 1 The Committee prepared a report to the Comissioners of its review of the actions taken to resoive concerns raised as a result of the loss of feedwater event at the Davis-Besse Power Station on June 9, 1985.

2. Comission Policy Statement on Technical Specifications The Committee prepared a report to the Commissioners of its re-view of the proposed Comission Policy Statement on Technical Specifications (the second draft) as transmitted via memorandum dated June 30, 1986.

3. ACRS Recomendations on the Hope Creek Generating Station The Committee prepared a letter to the EDO regarding resciution of its recommendation that a structured test program for evalu-ating overspeed protection of the turbine he prepared and sub-mitted to the NRC Staff for review and approval before full power operation of the Hope Creek Generating Station.

4. ACRS Action on the Proposed Revisions to Sections 9.2.2 of the Standard Review Plan (SRP)

The Committee prepared a memorandum to the ED0 regarding the proposed revisions to the SRP Sections 9.2.1, " Station Service Water Systems," and 9.2.2, " Reactor Auxiliary Cooling Water Sys-tems."

5. ACRS Comments on the B&W Owners Group Safety and Performance Improvement Program The Committee prepared'a letter to the EDO regarding its discus-sion of the B&W Owners Group Safety and Performance Irprovement Program.

j PROPOSED MIllUTES - 315TH 6. Your Letter of July 1,1986 to David Ward, Chairman, ACRS Tne Comittee prepared a letter to the ED0 regarding continuing NRC support of ongoing research provided by the ICRP, the NCRP, and the NAS concerning the biological . effects of exposures to various types of radiation.

7. ACRS Views on Fire Protection Research and Fire-Related Systems Interactions The Committee prepared a report to the Commissioners of its con-sideration of planned reductions in the fire-related portions of the NRC safety research program.

8. Additional Recommendations on the Development of a De Minimis Dose The Comittee prepared a report to the Comissioners of its con-sideration of procedures that might be used in the development of a de minimis radiation dose value for regulatory use.

9. Emergency Planning Requirements for NRC Licensees The Committee prepared a memorandum to the EDO regarding consid-eration by the NRC of the recomendation that nonnuclear power plant licensees, who must develop plans for coping with acci-dental radionuclide releases or must justify the reasons that the development of such plans is not needed, utilize in this process the screening model recently published by the National Council on Radiation Protection and fieasurements.

10. Aptitude Testing Time did not permit discussion of a proposed report to the Comission regarding use of aptitude testing in selection of nuclear plant personnel. This will be carried over for discussion during the 316th ACRS meeting.

11. TVA Reorganization A proposed report was discussed briefly. Mr. Wylie was asked to redraft the proposed report for further discussion during the 316th ACRS meeting.

12. Regulatory Process The Comittee discussed its concerns regarding the fact that the existing system of NRC regulations and enforcement does not provide an appropriately high quality of nuclear plant performance, as clearly evidenced by the recent experience with

/

PROPOSED MINUTES - 315TH Davis-Besse, the incident at Rancho Seco, and the problems with the TVA nuclear power plants. Members were asked to provide comments to W. Kerr on a draft report he prepared at the 315th meeting. The subject will be considered again at the 316th meeting (August).

C. Future Agenda

1. Future Agenda The Committee agreed on tentative agenda items for the 316th ACRS meeting, August 7-9, 1986 (see Appendix II).

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

D. Bilateral Exchange of Nuclear Information Chairman Ward reported interest by Professor M. Cumo, President of the Technical Comittee for Nuclear Safety and the Safeguarding of Health from Ionizing Radiation, a nuclear safety organization in Italy, which advises on all cases bearing upon nuclear safety and the safeguarding of health from ionizing radiation, in an exchange of nuclear reactor safety information with the ACRS. The Committee agreed that Chairman Ward should write to Professor Cumo and suggest a bilateral meeting of representatives from the Technical Committee and from the ACRS.

E. EPRI Advanced LWR Program The Committee was briefed by representatives vf the NRC Staff regard-ing the schedule for review of the 13 chapters of the EPRI Advanced Light Water Reactor Pequirements Document. Chapter 1, Overall Re-quirements, is expected in July 1986 and is to be reviewed by the Staff in the next six months. The Staff expects to discuss its completed review with the full Committee and recuests ACRS comments and recomendations at that time. Dr. Okrent suggested that it may be appropriate for the Committee to provide interaction on additional chapters of this document since it is an important document.

F. Nomination of ACRS Member it. J. Clausen, Technical Assistant to Chairman Zech, introduced one of the candidates for an ACRS member vacancy to Chairnan Zech. R. F.

Fraley, ACRS Executive Director, was asked to inform ftr. Clausen that this acticn constituted a violation of the ACRS nomination process.

O PROPOSED MINUTES - 315TH G. Death of Admiral Hyman G. Rickover The Comittee designated R. F. Fraley as its representative at the memorial service for Admiral Hyman G. Rickover at the Washington Cathedral on Monday, July 14, 1986.

H. IAEA Meeting on Chernobyl Accident W. Kerr was informed that he has been provided a slot as ACRS observer on the team that will travel to the IAEA meeting this August in Vienna where representatives of the Soviet Union are scheduled to make a technical presentation regarding the Chernobyl nuclear acci-dent.

1. Report of the Nominatino Panel The Committee accepted the report of the Nominating Panel regarding candidates for appointment to the ACRS to fill the vacancy which arose when R. C. Axtmann left the Committee. Two names, an expert in waste management and a reactor expert were recommended to the Commission for their consideration.

J. Move to Bethesda Chairman Ward noted plans to move the ACRS to Bethesda during February 1987, although this date may slip.

K. Probabilistic Assessment Dr. Okrent noted that IPEs for the front-ends of approximately six nuclear plants have been completed consistent with proposed IDCOR methodology and requested that copies of these evaluations be made available to the Comittee.

The 315th ACRS Meeting was adjourned at 2:55 p.m., Saturday, July 12, 1986.

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APPEf1 DICES 10 THE 315TH ACRS f1EETIllG JULY 10-12,1986 ,

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APPENDIX I

, ATTENDEES 4

NRC STAFF ATTENDEES i 315TH ACRS MEETING

Thursday, July 10, 1986

, OFFICE OF NUCLEAR REACTOR REGULATION

! Tom King i F. Miraglia j H. Berkow

! D. Scaletti l D. Lynch

!' R. Hernan i D. Moran l L. Crocker 4

B. J. Youngblood D. E. Hickman

! C. R. Stahle H. E. Gilpan 0FFICE OF NUCLEAR REGULATORY RESEARCH A. Datta I

4 L

i 1-ce j .. '

i I

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. _ . . - . . - - . - . - . . . - . - . - - . . - - - - - - . . = . - . . - . . - -

I 4

4 4

i j r i  !

I I

NRC STAFF ATTENDEES 4

l 315TH ACRS MEETING j

Friday, July 11, 1986 1

1 0FFICE OF NUCLEAR REACTOR REGULATION i

j R. Hernan

P. T. Kuo l

L. Kelly i E. B. Tomlinson R. Bernero C. Thomas M. L. Wohl S. Shankman i M. B. Spangler

! T. S. Michaels R. F. Hershman I. Jackiw, Region III 4

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DIV. OF HUMAN FACTORS TECH.

I. Recarte i

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(  !

INVITED ATTENDEES

. 315TH ACRS MEETING  !

t i.

1 )

j Thursday, July 10, 1986 i TENNESSEE VALLEY AUTHORITY James Huston ,

i' R. Gridley D. Lumbert W. T. Cottle '

C. C. Mason

. W. C. Drotleff .

l D.L. Williams  ;

I C. Crawford j t  ;

i i  !

i i

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i l i

I i

l INVITED ATTENDEES -

l 315TH ACRS MEETING l Friday, July 11, 1986 i

j TOLEDO EDISON I S. Piccdlo j L. Lingenfelter i J. Hirsch j J. Haverly ,

1 S. C. Jain  !

! J. K. Wood

! L. F. Storr

, T. Myers J. Stusdovant

R. F. Peters

! S. J. Smith j J. Williams P. Smart  ;

! J. Fay l D. C. Hilderbrandt

! R. Wiel i

i I

l BECHTEL l E. J. Ray i

N. Kalyanam I i l

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Q' PUBLIC ATTENDEES 315TH ACRS MEETING Thursday, July 10, 1986 K. Arn, Bechtel Power Corp.

L. P. Bupp S. Gibbon, PE Co.

E. A. Kennedy, Combustion Engr.

R. L. Smith, IEAL R. P. Vijuk, Westinghouse M. Beaumont, Westinghouse J. Fuitord, NUS L. Neal, General Electric F. Sears, Northeast Utilities B. Reing, DuPont G. Brown, Stone & Webster J. Nurmi, QATEL J. Knubel, GPU Nuclear L. A. Johnson, Virginia Power R. Borsum, Babcock & Wilcox R. S. Boyd, KMC M. Acleel, General Electric L. Cuolo, Fried, Frank D W. F. Rolf, Comonwealth Edison Co.

L. Connor, DSA R. L. Mitt 1, PSE&G C. L. Tully, AIF K. Stahlkopf, EPRI D. Noble, EPRI C. W. Fay, Wisconsin Electric Power R. Engel, S. Levy, Inc.

J. R. Humphries, C.O.

R. A. Szalay, Atomic Industrial Forum R. A. Bruce, Westinghouse J. F. Bray, Media News Serrice M. Biricik P. D.Krippner, American huclear Ins.

I. Bray, M. News T. Chonan, Energy Report H. Hagedorn, Arthur D. Little S. Jkullte, Capital Broadcast News R. Minor, Capital Broadcast News G. Brown, Stone & Webster B. O'Driscoll, Gannett News Service J. Thurber, Battelle R. Powelson, Knoxville News Sentinel A-5

PUBLIC ATTENDEES 315TH ACRS MEETING i Friday, July 11, 1986 l E. Fotopoulos, Serch Licensing, Bechtel

J. Nurmi, QATEL

! J. Hannah, Associated Press

' K. R. Campbell, NUS C. Angelini, ANB T. Channer, Energy Report L.S. Gifford, General Electric 2

L. Connor, DSA R.Borsom, Babcock & Wilcox K. D. Kirby, SAIC R. J. Stevens, Florida Power & Light j Andrew Hennendinger, Potomac News J. F. Pearson, Jr. Babcock & Wilcox R. Leachman, Lockheed Missiles & Space J. J. Field, SMUD R. Ashley, SMUD D. B. Ingmire, Management Consultants, Inc.

B. Kahn, Georgia Tech l

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APPENDIX II FUTURE AGENDA p APPENDIX A - .

g FUTURE AGENDA AUGUST ACRS MEETING Meeting with Director, NMSS -- Discuss items of mutual li hours interest San Onofre Nuclear Generating Station, Unit 1 -- 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Restart of plant following loss of feedwater event on November 21, 1985 Scram System Reliability -- ACRS Subcommittee report I hour ,

on status of ATW5 implementation effort i Human Factors Issues -- ACRS coments on (1) Proposed li hours NRC policy statement on fitness for duty; (2) proposed NRC rule on Educational Requirements for Senior Reactor Operators (3) proposed revision of Regulatory Guide 1.114, Guidance to Operators at the Controls and to SR0s in the Control Room USI A-46, Seismic Qualification of Equipment in 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> Operating Plants -- ACRS coments on proposed final USI resolution package Recent Events at Operating Nuclear Power Plants -- 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> V Report of ACR5 Subcomittee and representatives of the NRC Staff regarding recent incidents and accidents at nuclear power plants Continue discussion of ACRS comments on Advanced LWR 3-4 hours Designs -- Discuss proposed positions on related issues Long-Range Plan -- Report of ACRS Subcomittee regarding 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> proposed guide for a long-range plan Meeting with NRC Comissioners -- Discuss ACF.S report deferred to on proposed NRC Policy Statement regarding standardized September nuclear plants Completion of Maintenance and Surveillance Program deferred Plan, Phase I -- NRC briefing on results of Phase I of this program Continue discussion of NRC Standardization Policy 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Statement -- ACRS coments ,

TVA Reorganization -- ACRS coments regarding proposed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> reorganization NRC Regulatory Process -- Discuss proposed report to li hours

'O the Comission

p Report of Subcommittee on Waste Management to discuss I hour the following:

1) proposed ACRS comments regarding residual radiation limits for the disposition of land, building, equipment, and metals (including contaminated smelted alloys)

2) various issues pertinent to the storage of high-level radioactive wastes in geologic repositories

3) various issues pertinent to the disposal of low-level radioactive wastes

4) visit to WIPP and NTS facilities ACRS Management Subcomittee Meeting on July 9,1986 3/4 hour Containment Performance Design Objectives -- Staff deferred to preliminary position paper October ECCS Rule (10 CFR Part 50.46? Revision -- ACRS coments deferred to requested on final rule and Appendix K September Aptitude Testing -- Discuss proposed ACRS coments re- i hour garding use of aptitude testing in selection of nuclear O plant personnel Subcommittee Activities -- Reports regarding status li hours of ATW5 implementation effort, and ACRS procedures and administration l

L O w l

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' APPENDIX III ACRS SUBCOMMITTEE JUL 12 W MEETINGS nemtu O ~

ACRS SUBCOMMITTEE MEETINGS

(

Human Factors, July 15, 1986, 1717 H Street, NW, Washington, DC (Schiffgens), 8:30 A.M., Room 1046. The Subcommittee will review: (1)

SECY-86-153, industry and staff comments on proposed fitness for duty policy statement, (2) SECY-86-70, proposed rulemaking on degree require-ments for SR0s at nuclear power plants, (3) SECY-86-119, the annual statu; report on implerrentation of the Commission policy statement on training and qualification, and (4) the proposed Reg. Guide. 1.114. Rev. 2 Guidance to Operators at the Controls and to Senior Operators in the Control Room of a Nuclear Power Unit." Attendance by the following is anticipated, and reservaticns have been made at the hotels indicated for the night of July 14:

t Dr. Remick NONE Mr. Ward NONE Mr. Ebersole CARLYLE Mr. Wylie DAYS INN Mr. Reed DAYS INN Waste Management, July 21 - 23, 1986, 1717 H Street, NW, Washington, DC (Merrill), 8:30 A.M., Room 1046. The Subcommittee will review: (1) EPA's development (with NRC Staff's cooperation) of residual radiation limits and the disposition of land, buildings, equipment and metals (including contaminated smelted alloys -- NUREG-0518, Final Environmental Statement) resulting from the decontamination and decommissioning of nuclear power plants and fuel facilities; (2) the following High-Level Radioactive Waste

( (HLW) topics being addressed by the NRC Division of Waste Management (DWM)

Staff: (1) sorption and solubility Generic Technical Positions and Draft LetterReportontheSorptionWorkshop,(b)DWM's5-yearplan,(c)the NRC-proposed Federally Funded R&D Center (FFRDC), and (d) the status of their review of DOE's Final Environmental Assessments (EAs) for the candidate repository sites nominated for site characterization; (3) the following subjects being investigated under DWM's Low-level Radioactive Waste (LLW) Program: (1) alternatives to shallow land burial, (b) radio-active wastes that are below regulatory concern, and (c) mixed radioactive andhazardouswastes;(4)thefollowingwastemanagementresearchtopics under consideration by RES: (a) the development of field data on the movement of radionuclides within the environment and the associated impact of heat-water-rock interactions, (b) the predicted performance of reposi- '

tory systems under realistic field conditions, and (c) preview of setting priorities for waste management issues subject to rulemaking. The Subcen-mittee will also be briefed on DOE's Waste Isolation Pilot Plant (WIPP) and Nevada Test Site (NTS) in preparation for its July 31-Aug. I visit to those

sites. Attendance by the following is anticipated, and reservations have

been made at the hotels indicated for the nights of July 20, 21 and 22:

Dr. Moeller CARLYLE Dr. Carter ANTHONY Dr. Kerr LOMBARDY Dr. Orth ANTHONY l Dr. Mark LOMBARDY Dr. Steindler LOMBARDY '

i Dr. Remick NONE l

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V Naval Reactors (CLOSED), (operation of a nuclear-powered submarine),

July 28, 1986 (Boehnert). The Subcommittee will observe the activities of a nuclear submarine crew. Attendance by the following is anticipated, and hotel reservations have been made for the nights of July 27 and 28:

'Mr. Ebersole Dr. Remick Mr. Etherington Dr. Shewmon Dr. Kerr Dr. Siess Mr. Michelson Mr. Ward Dr. Moeller Mr. Wylie Westinghouse Reactor Plants, July 30, 1986, 1717 H Street, NW, Washington, DC (Houston), 1:00 P.M. - 5:00 P.M., Room 1046. The Subcommittee will continue discussion and coment 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.

Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of July 29:

Mr. Reed DAYS INN (7/30) Mr. Michelson DAYS INN Mr. Ebersole CARLYLE Mr. Wylie DAYSINN(7/30)

O Dr. Kerr LOMBARDY (7/30) Dr. Catton DUPONT PLAZA V Waste Management Subcommittee Visit to WIPP and NTS Facilities (CLOSED),

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, NM 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 geologic repositories for the underground storage of High-Level Radioactive Wastes. Attendance by the following is anticipated, and reservatiens have been made at the AMFAC Hotel, 2910 Yale Blvd., SE, Albuquerque, NM for the nights of July 29 and 30. and at the Tropicana Hotel, Tropicana and Las Vegas Blvds., Las Vegas, NV for the nights of July 31 and August 1:

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., Room 1046. The Subcommittee will discuss the sIatus of the ATWS Rule implementation effort. Attendance by the following is anticipated,and reservations have been made at the hotels indicated for the night of July 30:

Dr. Kerr LOMBARDY Mr. Reed DAYS INN Mr. Ebersole CARLYLE Mr. Wylie DAYS INN Dr. Lewis HYATT Mr. Davis NONE 1

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Reactor Operations, August 4, 1986, 1717 H Street, Washington, DC (Alderman), 1:00 P.M., Room 1046. The Subcommittee will review recent events at operating reactors. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of August 3:

Mr. Ebersole CARLYLE Mr. Reed DAYS INN Dr. Lewis HYATT Dr. Remick NONE Mr. Michelson DAYS INN Mr. Wylie DAYS INN Metal Components, August 4, 1986,Hanford,WA(Igne),8:00A.M. The Subcommittee will review the steam generator integrity program. In addi-tion, the integrated Fracture Mechanics / Nondestructive Examination program will be discussed. Attendance by the following is anticipated:

Dr. Shewmon Mr. Ward Dr. Mark (tent.) Dr. Bush Reliability Assurance, August 5, 1986, 1717 H Street, NW, Washington, DC (Major), 8:30 A.M. - 12:00 N00N, Room 1046. The Subcommittee will review the final resolution of USI A-46, " Seismic Qualification of Equipment in Operating Plants." Attendance by the following is anticipated, and res-O ervations have been made at the hotels indicated for the night of August 4:

Mr. Wylie DAYS INN Mr. Michelson DAYS INN Mr. Ebersole CARLYLE Dr. Siess ANTHONY Safeguards and Security, Awgust-S -1986, i CANCELLED Extrerre 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.

Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of August 5:

Dr. Okrent ANTHONY Mr. Michelson DAYS INN Dr. Carbon STATE PLAZA Dr. Moeller CARLYLE Mr. Ebersele CARLYLE Mr. Reed DAYS INN Dr. Kerr LOMBARDY Dr. Siess ANTHONY Dr. Lewis HYATT Mr. Ward NONE Dr. Mark LOMBARDY Mr. Wylie DAYS INN

O --

Procedures and Administration (CLOSED), August 6, 1986, 1717 H Street, NW, Washington, DC (Fraley), 3:00 P.M. - 5:30 P.M., Room 1010 (Chairman's Office). The Subcommittee will discuss proposed membership of the ACRS Management Committee and the length of the term for ACRS Chairman. Attend-ance by the following is anticipated, and reservations have been made at the hotels indicated for the night of August 5:

Mr. Ward NONE Dr. Moeller CARLYLE Mr. Ebersole CARLYLE Dr. Remick NONE Dr. Lewis

• HYATT Dr. Siess* ANTH0NY

(* as available) 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),1:00 P.M. , Room 1046. The Subcommittee will review the report on Phase I of Maintenance Program Plan. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of August 12:

Mr. Reed DAYS INN Mr. Wylie DAYS INN Mr. Michelson DAYS INN 1&E Programs, August 14, 1986, 5th Floor Hearing Room, East West Towers -

West Building, 4350 East West Highway, Bethesda, MD (Boehnert), 8:30 A.M.

The Subcommittee will review I&E Inspection Programs with focus on the Safety System Function Inspection (SSFI) Program, and the risk-related inspection methodology. A tour of the I&E Incident Response Center is also planned. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of August 13:

Mr. Reed DAYS INN Mr. Michelson DAYS INN Mr. Ebersole CARLYLE Nuclear Plant Chenistry, August 26, 1986, 1717 H Street, NW, Washington, DC (Alderman), 8:30 A.M., Room 1046. The Subcommittee will discuss various topics relevant to plant chemistry, i.e., Na0H in containment spray, suppression pool scrubbing, H, water chemistry, etc. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Moeller Mr. Etherington Mr. Ebersole Mr. Reed Decay Heat Removal Systems, August 27, 1986, 1717 H Street, NW, Washington, DC (Boehnert), 1:00 P.M., Room 1046. The Subcommittee will review NRR's Xction Plan to address concerns with the reliability of certain plants' AFW systems. Lodging will be announced later. Attendance by the following is anticipated:

Mr. Ward Mr. Reed Mr. Ebersole J fa Dr. Catton Mr. Michelson //,,/ N Mr. Davis

~~

v Thermal Hydraulic Phenomena, August 28, 1986, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M., Room 1046. The Subcommittee will continue its review of the RES-proposed revision to the ECCS Rule (10CFR50.46 and Appendix K). Lodging will be announced later. Attendance by the following is anticipated: .

Mr. Michelson Dr. Catton Mr. Ebersole Mr. Schrock Mr. Reed Dr. Sullivan Mr. Ward Dr. Tien Management Subcommittee, September 10, 1986, 1717 H Street, NW, Washington, DC (Fraley), 8:30 A.M. - 5:30 P.M. , Room .1010. The Subcommittee will iiiscuss topics and/or a policy statement for ACRS activities. Lodging will be announced later. Attendance by the following is anticipated:

Mr. Ward Mr. Ebersole j Dr. Lewis 317th ACRS Meeting, September 11-13, 1986, Washington, DC, Room 1046.

(O U

) Containment Perforr;ance, September 23, 1986, 1717 H Street, NW, Washington, DC (Houston), 1:00 P.M. - 5:00 P.M., Room 1046. The Subcommittee will Te' view a draft position paper on containment performance design objective as an addition to the Safety Goal Policy. Lodging will be announced later.

Attendance by the following is anticipated:

Dr. Mark Dr. Okrent Mr. Ebersole Dr. Siess Dr. Kerr Mr. Wylie Severe (Class 9) Accidents, September 24, 1986, 1717 H Street, NW, Washington, DC (Houston), 8:30 A.M., Room 1046. The Subcommittee will review the f.RR Implementation Plan for Severe Accidents and the IDCOR Methodology for Individual Plant Evaluation. Lodging will be announced later. Attendance by the following is anticipated:

Dr Kerr Mr. Ward Dr. Carbon Mr. Bender Dr. Mark Dr. Catton Dr. Okrent Dr. Corradini Dr. Shewmon Dr. Davis Dr. Siess w

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Decay Heat Remeval Systems, September M, 1986, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M., Room 1046. The Subcommittee will continue its review of 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 Mr. Michelson Mr. Davis cRS International Operating Experience, September M, 1986, 1717 H Street, NW, Washington, DC (Major), 8:30 A.M., Room 1046. The Subcommittee will track and evaluate information on the Soviet nuclear accident at Chernobyl and consider implications for U.S. reactors of similar type. Lodging will be announced later. Attendance by the following is anticipated:

Mr. Remick Dr. Siess Dr. Kerr Mr. Ward Dr. Moeller Wingspread International Conference (CLOSED), October 19-23, 1986, Racine, WI (McCreless). Representatives from the ACRS, RSK, GPR, and Japan will (n J

) Fichange information on nuclear reactor safety.

Instrumentation and Control Systems - Cancelled Spent Fuel Storage Date to be determined (July / August), Washington, DC (Alderman). The Subcommittee will continue its review of 10 CFR Part 72 and Monitored Retrievable Storage (MRS). Attendance by the following is anticipated:

Dr. Siess Dr. Remick Dr. Kerr Dr. Shewmon j Dr. Moeller Westinghouse Reactor Plants (CLOSED), Date to be determined (August),

Washington, DC (El-Zeftawy). The Subcommittee 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 AC/DC Power Systems Reliability, Date to be determined (August),

Washington, DC (El-Zeftawy). The Subcommittee will review the proposed .

Station Blackout rule (SECY-85-163). Attendance by the following is anticipated:

I Dr. Kerr Mr. Reed Mr. Ebersole M r. Wylie Dr. Lewis ~~ h

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

Dr. Remick Mr. Reed Dr. Carbon Mr. Wylie Mr. Michelson Probabilistic Risk Assessment, Date to be determined (September / October),

Washington, DC (Savio). The Subcommittee will review the probabilistic risk assessment for Millstone 3. Attendance by the following is antici-pated:

Dr. Okrent Mr. Michelson Dr. Kerr Dr. Siess Mr. Ebersole Mr. Ward Dr. Lewis Mr. Wylie Dr. Mark Seabrook Units 1 and 2, Date to be determined (late summer /early fall),

% Washington, DC (Major). The Subcommittee 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, NM (Igne). The Subcommittee will review containment integrity and Category I structures, facilities, and programs. Attendance by the following is anticipated:

,4 Dr. Siess Dr. Shewmon Mr. Ebersele Mr. Bender Dr. Kerr Dr. Pickel Dr. Okrent i

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Decay Heat Remeval Systems, September , 1986, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M., Room 1046. The Subcommittee will continue its review of 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 Mr. Michelson Mr. Davis InternationalOperatingExperience,Septemberh, 1986, 1717 H Street, NW, Washington, DC (Major), 8:30 A.M., Room 1046. The Subcommittee will track and evaluate information on the Soviet nuclear accident at Chernobyl and consider implications for U.S. reactors of similar type. Lodging will be announced later. Attendance by the following is anticipated:

Mr. Remick Dr. Siess Dr. Kerr Mr. Ward Dr. Moeller Wingspread International Conference (CLOSED), October 19-23, 1986, Racine, WI (McCreless). Representatives from the ACRS, RSK, GPR, and Japan will (9 eichange information on nuclear reactor safety.

\)

Instrumentation and Control Systems - Cancelled Spent Fuel Storage, Date to be determined (July / August), Washington, DC (Alderman). The Subcommittee will continue its review of 10 CFR Part 72 and Monitored Retrievable Storage (MRS). Attendance by the following is anticipated:

Dr. Siess Dr. Remick Dr. Kerr Dr. Shewmon Dr. Moeller Westinghouse Reactor Plants (CLOSED), Date to be determined (August),

Washington, DC (El-Zeftawy). The Subcommittee 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 AC/DC Power Systems Reliability, Date to be determined (August),

Washington, DC (El-Zeftawy). The Subcommittee will review the proposed Station Blackout rule (SECY-85-163). Attendance by the following is anticipated:

j Dr. Kerr Mr. Reed Mr. Ebersole Mr. Wylie Dr. Lewis jf-/f

r APPENDIX IV-TOLED0 EDIS0N DAVIS-BESSE PRESENTATION O

ACRS/ Toledo Edison Meeting Agenda July 11,1986 Joe Williams, Jr. Update of Mission Senict Vice President, Activities Nuclear Steve Smith Update of Maintenance AssistantP! ant Activities Managot; Maintenanco John Wood Summary of Eventinvestigation l NuclearPlant ReactorCoolant Pump Status Systems Director SushilJain Auxiliary Feedwater System O NuclearEngineering Director Modifications and Decay Heat Removal Phil Hildebrandt System Review and Test Program NuclearEngineering Group Director Joe Williams, Jr. Closing Remarks A11 l

Davis-Besse Restart Schedule JUN P6 jut. Of. AUG65 SEP06 OCTe6 NOV 89 2: I 2s s I i2 l i, l 2s 2 Is I is Inl 20 s I is l 20 l 2r 4 l: l is [2s i I sl s l 22_

6/25'88 RF IURN DH LOOP f 2 TO SERVICE 6/25 06 DH LOOP #I OUT OF SERVICE FOR NON-RAVCHLM WORK 6/30/96 DH LOOP #1 IN SERVICE 6/30/96 90TH DH LOOPS IN SERVICE 7/8/06 BEGIN SFAS TESTING 7/10/96 MATERIAL DELIVERY RCP 1-1 7/11/96 COMPLETE SFAS TESTING 7/14/06 DRAIN RCS TO BEGIN RCP WORK 7/24/06 MATERIAL DELIVERY RCP 2 2 t/8/06 M ATERI AL DELIVERY RCP 1-2 l s/15/86 RCP WORK COMPLET ED PUMP 1-1 s/15/06 RCP WORK COMPLET[D PUMP 2-2 10/5/86 "H LOOP e1 OUT OF SERVICE FOR RAVCHEM WORK t

10/10/96 RCP WORK COMPLETED PUtIP 1-2 10/10/86 COMPLETED ALL CONTAINGIENT WORK 10/23/96 DH LOOP f15ACKIN SERVICE AFTER RAVCHEM [

10/23/86 BEG;N HEATUP FOR HOT TESTING 10/25/06 ENT ER MODE 4 FOR HOT-FUNCTIONAL TESTING 10/27/06 STARTUP-ENTER MODE 3 11/6/96 RX CRITICAL-ENTER MODE 2 11/6/96 STARTUP-ENTER MODE 1 11/17/06 REACTOR AT 100% POWER

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l Manning a Manning Status 644 July 1 818 June 30,1986 185 Vacancies 114 Contractors seconded 58 Of those contractors seconded are engineers.

a 1986 Acceptances through June 23 65 Management 51 Non-management a Recruiting Assessment Process Security background check Drug screening Psychological testing and evaluation Education check Assessment and careerdevelopment testing Personalinterviews l

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Nuclear Mission Staffing i

Approved:1003  :

I 1100-1 j 1050 -

f 1000-950- ____ _ _ _ _ _ __ _ -~~

930 by year-end i 900-l 850- 818 802 I L 79' O

769 779 I

T 754 750- 739 ' l 70s 679 i

IO g 657 656 .

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June Jufy Atg Sept Oct Nov Dec l Jan Feb Mar April May Jure July Auj Sept Oct (J0< Dec l Jan Feb Ms Aptd May June July 1985 1986 1987

Nuclear Engineering Group .

Approved: 214 1

220 - .

210 -

200 -

190 -

180 -

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l 170 -

! 160 -

! 150 - T 140 -

132 125 130 -

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Nov Dec l Jan feb Mar Apr May June July Aug Sept Oct Nov Dec l Jan Feb Mar Apri! May June July Aug Sept 1985 1986 1987 e .

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Dmg and Alcohol Program Schedule:

Dec.1985 -Jan.1986: Introduction of Company Drug and Alcohol Policy byletter and employee meetings.

Feb.1986 Random drug screening began for Davis-Besse employees March 1986 Random drug screening began for contract employees Results to Date:

O 127 Davis-Besse employees tested (13% of  ;

workforce) 1 Employee tested positive (1%)

121 Contract employees tested (8.3% of workforce) 4 Contract employees tested positive (3.3%) -

All positive tests were marijuana. l All disciplinary actions to date have resulted in j termination of employment t i

O A-33 L

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Drug and AlcoholProgram Fitness for Duty Examinations:

a Newjob applicants s Annual physical examinations for Nuclear Security officem a Bi-annual physical examinations for licensed opemtors a Any employee absent more than 30 days in a 12 month period O a Employees arrested "off-work" for drug activity a Employees who have completed the Employee Assistance Program for drug and alcohol abuse a Employees transferring to Davis-Besse from other areas of Toledo Edison u Random drug screening a Behavioral observation a Postindustrialaccident onsite

Davis-Besse Procedures Effort Required For Appnwed Additional Restart 6/27/86 in 1986 NuclearMission Procedures 14 14 63 Division Procedures -

Station Administrative 21 21 90 Maintenance 198 198 500 Operations 133 90 304 TechnicalSupport 1 1 Chemistry / Health Physics 35 EnvironmentalQualification 21 11 Sui > Total 374 321 979 Engineering 41 Environmental Quellfication 12 12 Administrative 8 5 Test 112 102 '

Sut> Total 132 119 41 Quality Assurance 10 7 1 NuclearThining 3 16 Nuchar Services 4 90 Nuclear Projects 2 16 NuclearSafety & Licensing 4 1 1 Information "~r+;+T-it .

11 11 39 IndustrialSecurity 4 5 PASS Project 24 16 Other PCR Affected Procedures 86 48 Total 668 537 1,251 Complete 537 -

In Process 92 -

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Maintenance improvement Program Update Additional changes have been implemented in these areas:

• Organization and Staffing aTraining

' Maintenance Activities a Engineering Interface and Support O

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. O Organization New Management Personnel e Assistant Plant Manager, Maintenance ai & C Superintendent a Mechanical Maintenance General Foreman increased supervisory personnel for each discipline:

a Superintendent a General Foreman a Lead Engineer a Foreman improved supervisor / craftsman ratios:

a Mechanical,1/10 (was 1/23) e Electrical,1/6(was 1/26) eI & C,1/7(was 1/19)

Total Maintenance Department manning a Toledo Edison:199/185 a Contractors: 90-150 1

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O Training u Each discipline has a designated Daining Foreman a Raining shift concept has been impleir=M e Raining Councils formed in each discipline a Outside organizations or facilities are utilized to provide training a Raining Recon:Is Review has established levels of qualification for maintenance personnel a INPO accreditation process started Q a Site accreditation visit successfully completed on June 20,1986 a Awaiting scheduling of accreditation board meeting

O Corrective and Modification Work Orders Backlog:

1339 Corrective work orders open on June 9,1985 1109 Of those closed as of June 27,1986 230 Current number open 33 Requiredforrestart 111 Facility Change Requests open on June 9,1985 57 Of those closed as of June 27,1986 54 Currentnumberopen 8 Required for restart Current:

! 7740 Corrective work orders issued since June 9,1985 5413 Of those closed as of June 27,1986 2327 Currentnumberopen 1020 Requiredforrestart 1220 FOR's issued since June 9,1985 776 Of those closed as of June 27,1986 444 Currentnumberopen 319 Requiredforrestart 1380 Totalrequiredforrestart 9 A -30

O Preventive Maintenance Bacidog:

405 Work orders open on June 9,1985 '

398 Of those closed as of June 27,1986 7 Current number open Allwill be closed prior to restart Current:

2017 Work orders open since June 9,1985 1692 Of those closed as of June 27,1986 325 Currentnumberopen At restart - no bacidog PM work orders will be O

o i i aoias 332 Totalrequiredforrestart 1

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l Engineering Interface and Support '

s Engineering attendance at Plan of the Day meeting a Duty Manager RosterSystem a Station Technical Support Section

l e Request for Engineering Assistance Process NMP and Engineering procedures have been issued a Engineering support programmatically required in preventive maintenance and post maintenance testing O

O Equipment investigation Summary .

Purpose:

Determine root cause of equipment malfunctions to implement appropriate i and effective corrective actions.

Scope

13 Areas impacting 7 systems Main Feedpump Turbine SFRCS Auxiliary Feedpump Turt>ines Auxiliary Feedpump Turbine Trip & Throttle

Wives Auxiliary Feedpump Wives AF599 and 608 Pilot Op..M Relief Valve O Main Steam Headers Main FeedwaterStartup ControlWlve Auxiliary Feedwater Pump #1 Suction Supply Main Steam Wlve MS-106 Nuclear Instrumentation Neutron Source Range Detectors i Turt>ine Bypass Wlve i Safety ParameterDisplaySystem .l Corrective Actions Prior to Restart: i 15 Design Modifications 13 Maintenance /,4-!sce.T.rit activities l 5 Proceduraland/ortrainingissues O w

O I

Reactor Coolant Pumps (RCP's) e Crystal River Failure occurred at Crystal River in "A" Coolant Pump on January 1,1986.

Investigations found shaft was severed near top of

[ Joumal bearing.

l Also found crack in same 48" region on "B" pump.

Possibly also problems with "C" and "D" pumps.

Broken bolts holding on main impeller found on "A" and "B" pumps at Crystal River.

Pump cover was found to have axial cracking in Q lower region.

m Davis-Besse and Arkansas-1 have same design Byron-Jackson pumps as Crystal River.

m Ultrasonic testing (UT) at Davis-Besse concluded that shafts in all four pumps have cracks at 52" mglon.

m impeller bcits cannot be evaluated without pump disassembly.

m Procuring four additional imp.'seir.- it assemblies.

Deliveryin July and August.

O n-31

!O .

Reactor Coolant Pumps (RCP's)-

(Cont'd) e Removed shaft from RCP 2-1 for further examinations.

Examination at B&W Lynchburg Research Center showed axial cracidng at 48" s 5 mils deep, no i crack at 52".

One bolt broken, two bolts cracked, one bolt and all pins intact.

Dye penetrant exam showed no pump cover cracidng.

m Installed spare rotating assembly in RCP 2-1.

O Pi aaiao -g =e.T.=>t o' - iaias '=i tias assemblies July-September.

m B&W investigating anomalous UT indications.

O A-35

O AFW Reliabilityimprovements a Provide hot steam lines to AFW pumps.

m Install PGG govemor on AFPT-1.

• Provide time delay in AFW pump suction transfer i

scheme.

a install loca! AFPT Trip Throttle Valve indication.

• Eliminate deaeratorsuction path.

• Depower suction valves from CST.

m Valve motor operator improverrents.

l e Remove AFW pump suction strainers.

m Resize strainerfrom CST.

I 0 A-36

Davis Besse Auxiliary and Motor Driven Feedwater Systems Start-Up Configuration O

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Installation of Motor Driven Feedwater Pump New pump design features:

a Provides > 100% capacity auxiliary feedwater flow.

e Pump discharge aligned to the auxiliary feedwater headers during normal full power operation.

m Pump suction normally from the Condensate Storage Tank.

m Pump manually started from the Control Room.

m Pump discharge valve Control Room operated.

O eum, moiar can he sungiied from either emergency diesel generator following a loss of offsite power.

m Can be manually realigned to feed the Main Feedwater System. This will be the normal alignment during low power operation. Pump suction in this alignment will be from the DemeratorStorage Tank.

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m Eliminates high energy line break concems associated with existing start up feedpump.

4 0 A-57

O SFRCS Improvements a Filter existing steam generator level signals.

e improve SFRCS power supply p;.-feinence, a Remove main steam and main feedwater isolation on SG low level.

s Raise ICS low levellimit.

m Disable feed isolation to last steam generator depressurized.

e Provide seal-in manual reset for SFRCS full trip alarm.

, a SFRCS manual initiation improvements.

!O O .

1 1

1 O A -40 i

-~~~~ ~~~~

O Feed and Bleed Cooling Evaluation TOTAL LOSS OF FEEDVRTER FROM 100% POWER Assumptions: Power 1.02 x 2772 MWT Feedwater Main, Auxiliary, Motor driven and Startup Feed-Unavailable ReactorTHp 15 Seconds on high pnessure OperatorAction RCS Hot Leg Og initiator temperature at 600*F Makeup Flow 2 pumps initiated at operator action time PORV Blocked open at operatoraction time High Point & Vent opened at Pressurizer Vents operator action time Results: '

Core Mixture Level 15 feet (Collapsedlevel) 11.9 feet l

O A-4l ,

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O .

Longer Term Decay Heat Removal Reliabilityimprovements a Pnwision of primary system blowdown valves for enhancement of feed and bleed capability.

m Restoration of existing startup feedwater pump and provision of Control Room capability for associated valves.

• Provision of Control Room capability for the motor driven feedwater pump discharge valves to the AFW header.

m Further AFW valve line up modifications to make AFW flow initiation into steam generators Q independent of AC power.

e improve AFWlevel control.

e improve margin between SFRCS and ICS low level setpoints.

l m SFRCS logic revision to further minimize isolation.

a Control Room " mimic" panel for finalized AFW/SFRCS.

L l

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, O A -4h l

Primary Side O Decay Heat Removal (Long Term)

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O O us

O System Review and Test Program For 34 systems important to safe plant operation:

a identify important and recurring design, maintenance and operations problems and determine whether corrective actions are required prior to restart or can be taken over long term.

• Evaluate scc pe of existing periodic and surveillance testing to identify any additional l O testing needed to ensure required functions V will be performed.

m Conduct test program to assure these systems are functional. Testing will also be performed to verify adequacy of system modifications completed during outage.

This program will be completed prior to restart of Davis-Besse.

1 j

O  !

System Review and Test Program Approach a " Team" for each system headed by Toledo Edisjon engineer supported by highly-qualified industry personnel.

m Review by independent Process Review Committee-membership combined broad background of nuclear industry experience with anac!fic knowledge of Davis-Besse plant design.

m Detailed test procedures approved and implemented by Joint Test Group and restart test organization.

O

-___.A m W..+.

O .

)

Problem Areasidentified During System Reviews l

8 Forthe 34 systems l Approximately 150 problem areas require resolution priorto restart.

Approximately 200 problem areas require  ;

resolution overlong term.

m Resolution ranges from engineering evaluation to :

hardware changesin plant.

m Facility Change Requests, Maintenance Work Requests or Requests for Engineering Evaluation have been prepared to address all problems that O must be resolved priorto restart.

O A -44

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Determination of Corrective

, Actions Required Priorto Restart a Conective actions necessary to ensure safe and reliable plant operation.

m Based on collective E-p;;ience and judgment of Independent Process Review Committee considering a Plant design basis described in the Updated Safety Analysis Report.

eTechnicalSpsci1"ic uon requirements.

m Reliable plant operation.

O a Protection of personnel, systems and equipment.

m Prudent engineering practice.

1 0-0 0 A47

O Recurring Problem Areas

' Inattention to heating / ventilation / air conditioning requirements.

8 Inoperable nitrogen regulators.

s Inadequate maintenance for hydramotor actuators.

• Thmidng and replacement of limited life components (e.g., seals, elastomers, electrical compor.:.ib).

m Valve pacidng leakage.

• Steam trap maintenance.

8 l&C preventive maintenance / calibration.

l l

O Examples ofimportant Problem Areas Found in System Review

$ Control Room Emergency Ventilation Sistem inoperable (LER 85-018).

• Auxiliary feedwater discharge piping

overpressurization (LER 85-017).

m Startup strainers found installed in HPi pump suction (LER 85-006).

1. inadequate ventilation in service water pump room (LER 86001).

a Essential 4160V bus voltage high (LER 8&OO3).

O e improper terminal wire wrap technique in Steam Feedwater Rupture Control System (LER 8&009).

e improper installation of fire protection boot seals (LER 8&005).

m inadequate post LOCA service water flow to conta!nment air cooling system (LER 85-002).

m improper installation for fire dampers located in

.through wall ducts (LER 86-010).

m PORV not operationally qualified for environment during feed and bleed.

O H1

i Examples ofimportant Problem Areas Found in System Review (Cont'd) e Potential flooding of pit containing decay heat removal valves (DH-11 and 12 motor operated valves).

$ Lack of configuration control in Safety Features Actuation System.

e W'4=d station and instrument air system leaks and dependence on temporary diesel air

compressor.

a improper throttle position setting on containment O spray discharge valves based on allowable pump degradation.

1. Inadequate time-motion study for PASS (LER 86-020).

s M

I O

Examples of Problem '

Areas Found in ,

Test Review a incomplete testing of SFAS logic (LER 85-021).

! a incomplete testing for SFAS actuation of HPl system valves (LER 86-004) and DHR system valves (LER 86-023).

m Lack of venting of HPi system high point in containment (LER 86 012).

$ Inadequate p;.bisiance testing requirements for safety related heat exchangem.

a Lack of performance basis for acceptance criteria O for safety related pump testing.

e inadequate leak check requirements on some check and isolation valves (both testing method

and allowable leakage).

e incomplete accep;mnce testing for RCS hot leg level instrumentation.

m incomplete operability checks of standty flowpaths (e.g., decay heat removal pump /high pressure injection pump " piggyback" alignment).

e inadequate testing of the AFW level control system (LER 86 014).

$1nadequate testing of MSIV and AW (both tested with instrument air available).

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l O

Examples of Problem Areas Found During System Testing

$ 1mproperty wired containment air cooler fan and less than designed flow capacity (LER 86-007).

e inadequate flow and distribution of ventilation for j

service water pumps (LER 86-001).

m improper lower air pressure limit for multiple starts of emergency diesel generator (LER 86-002).

a lmproper setting of underfrequency protective relays on 13.8KV busses.

Ol i a inadequate maintenance and testing to assure operability of hydramotor actuators on safety  !

i related ventilation systems (LER 86-019).

l # Improper orifice size in MFPT hydraulic control oil system.

m Service water, containment air cooling and Control '

Room HVAC systems not properly balanced.

m Agastat time delay relay settings not repeatable.

m Inadequate maintenance on electrical load breakers, d-SL l

O Restart Test Activities Test Procedums Mode Required Approved (Test Procedures Only) (By Plant Manager) 5 100 97 4 1 1 3 10 10 2 0 0 1 1 1 Total 112 109 Tests O Mode 5

To Be Performed (ST's and TP's) 217 Completed 106 i

4 18 0

! 3 23 0 l

2 16 0

_0 1 10 Total 284 106

O Raychem Splice / Termination Evaluation Programs Problem:

Potential improper installation of Raychem shrink sleeve installation. .

Program:

The Raychem issue is addressed in the System Review and Test Program. IPRC concurrence with i the problem identification and program has been obtained.

Objective:

e A pilot program for one system to inspect Raychem O installations and electrical hardware.

m Conduct testing of selected existing Raychem installation at an independent laboratory.

m Repair as required.

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4 1

d O n-51 l

1 O

Raychem Splice / Termination Evaluation Programs Progress to Date l

u Decay heat loop 2 selected as pilot program l includes CCW, EDG and service water for train  !

operability. Inspections and repairs complete.

Currently

e inspecting decay heat loop 1 and miated trains of i

CCW, EDG and service water, a inspecting containment penetrations by area.

m Laboratory testing of selected samples in progress.

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13 Areas of Equipment Concem O

O A-57

O Equipment Concems a Main Feedpump Turbine j u Steam Feedwater Rupture Control System a Auxiliary Feedpump Turbines

e Auxiliary Feedwater Turbine THp & Throttle Wives e Auxiliary Feedwater Wives AF E99 and AF 608 m Pilot Cp. ^d Rollef Wlve '

s Main Steam Headers .

m Main Feedwater Startup Control Wlve  !

• Auxiliary Feedwater Pump #1 Suction Supply a Main Steam Wlve MS 106  ;

a Nuclear instrumentation Neutron Source Range Detectors aTurbine Bypass Wlve a Safety Parameter Dag. y System l

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- _ . . . - . _ _ . - - - - _ _ - - _ - - _ - - - - - \

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5 Main Feedpump Turbine (MFPI)

Concem: Overspeed tripping of MFPT 1-1 initiated a plant runback.

Findings

Failed circuit board capacitorin General Electric controlsystem.

Corrective 1. Replaced faul'ad board.

Actions: 2. Checked and will test control circuits for both MFFT 1-1 & 1-2.

I Generic Non: ixet::iiis specific to MFPT implications: control circuits.

l

_ ._--_- =_____ - _ . . .

O

Steam Feedwater Rupture ControlSystem (SFRCS)

Concem: Spurious SFRCS actuation cloeod both main steem isolation valves and isolated steam to main feedpump turbines.

i Findings: Turbine trip caused pressure i oscillations which SFRCS detected as

low steem generatorlevel. Lavel i Pressure tap was made more sensitive due to transmitterchangeouts.

1 Conective Added electronicmering to signals.

Action:

Generic increasein sensitMty/ response can implications: result due to transmitter changeouts.

installed filtering in Fleactor Protection system flow transmittercircuitry.

O Auxiliary Feedpump Turbines '

Concem: Both auxiliaryfeedpump turbines 1 tripped on overspeed - this prevented j supply of waterto steam generators.

i Findings: Condensation in long steam inlet lines j disrupts properturbine control.

l Corrective 1. Installed new steam admission Actions: valves close to turbines to keep lines hot with steam to greatly reduce waterformation.

} 2. Increased steam trap capability.

! Generic None-no otherquick start steam l Implications: driven turbines.

I

l l

O i

l Auxiliary Feedpump Turbine Trip and Throttle Valves Concem: Operators expedenced problems resetting the valves - delayed initiation of auxiliaryfeedwaterto steam generators.

Findings: Procedures and priortraining not sufficient.

Conective 1. Provided placards and local Actions: indicators on T&TV to help operators.

l 2. Enhanced communications between l pump roomsand from pump rooms l

O 'a c at ' aaa -

3.Will provide hands on hot training priorto restart.

Generic Othercrucialoperatoractions '

implications

performed locally. Covered by Operator i Actions review.

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~

O Auxiliary Feedpump Valves AF 599 and AF 608 Concem: Wivesfailed to open on demand after closing earlier - would have pnevented auxiliary feedwaterflow.

! Findings: Motoroperators on valves were not I properly adjusted allowing valves to

" torque out".

Corrective 1. Reedjusted AF599 and AF608.

Actions: 2. Evaluating and reed lusting other .

nuclearsafety reisted motor '

operated valves priorto restart.

4 O 3. Testing selected valve with full dinerentialpriorto restart.

t 4.Provided new maintenance

! Procedures.

Generic Applicable to othermotoroperated i implications: valves.

O us

O Pilot Operated Relief Valve (PON)

Concem: Dudng transient PORV failed to close

! property after third opening - closure of the blockvalveisolated the PORV

and it reensted.

Findings: No pir, ice: evidence found to explain improperclosure-foreign

material in pilot cannot be ruled out -

performance similartoindustry i ex+:f.;-:+

. Conective 1. Tested valve.

Actions: 2. Added acoustic monitorflow indication light on PORV control panel.

j

3. Changed PORV annunciatorlight from white to red.

4.impnwed panellabeling of sosenoid open/close switch.

5. Provided for PORV exercising dudng shutdowns.

. 6. Rebuilt PORV with new parts. ,

7. Quallflod solenoid foroperating erwironment.

Genedc Now valves of similardesign. ,

implications: '

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i O

Main Steam Headers

)

Concem: After closure of main steem isolation valves, pressure controlproblems were +:-4:d: aced in the main steam l headers.

Findings: Manualactuation of.L.W.:.ic vents valves (AWV)causedlarge pressure dropin header #1 AWV controlcircuitryon header #2is a lesserconcom. Switch contacts conoded on ICS module.

! Corrective 1. Fullcheckout and adjustment of l Actions: AWV control circuitry.

2. Tested and refurbished all18 main i

steam safety valves.

3. Flefurbished ICS modules for AW circuitry.

Generic Switch contacts being evaluated and l Implications: refurbished on otherICS modules.

l l

!O Main Feedwater Startup ControlValve Concem: Operators were uncertain of status of controlvalve SP7A due to blown light bulb.

Findings: Valve operated property - technician inserted inconectvoltagelamp during event.

! Corrective Provided additionalinformation to Action: operators.

Generic None-no significant findings.

implications:

f O

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1

/) -46

O i

Auxiliary Feedwater Pump ,

! #1 Suction Supply Concem: Pumpsuction transfoned fmm nonnalto backup watersupply about 20 minutes after reactortrip.

Findings: Noimpact to steam generator-l transientlow suction pressure caused transfer.

Corrective 1. Revised strainer arrangement.

Actions: 2. Revised transfer switch setpoints.

3. Provided time delay.

Generic Other pump suction transfer systems.

l Implications:

i 1

O A-67

O .

Main Steam Valve MS-106 Concem: Wlve position indication recorded as closed to not closed to closed in about one third the expected time - this valve is used to admit sisem from steem generator #1 to auxiliaryfeedpump j turbine #1.

l Findings: Motoroperatoron valve was not properly adjusted. Wedge was found to be steem cut.

Corrective 1. Reed lusted torque and limit switches.

Actions: 2. Installing new wedge prior to restart.

3. Testing valve with steem flow prior to i restart.

Generic Othermotor operated valves. '

Implications:

O A-48

O Nuclearinstrumentation Neutron Source

, Range Detectors Concem: Prior to event NI-1 was ir-:-;: st'; and '

NI-2 failed during transient - previous problems had been +:-T:f.;.-c+1

Findings

Ni-1--inadequate grounding of shield l found at preamp due to paintandlack ofstarwashers. '

NI-2-intermittent failure of containment;: .a-:}: i cable center conductor.

O in c ai coaaecior i o'ouad degradedin each detectorstring.

Corrective 1. NS-1--proper ground established.

Actions: 2.N62--replaced penetration /

module.

3. Replaced / refurbished connectors as required. l Generic Preventive maintenance program implications: needed forsource range,

~

intermediate range, and power range l connectors. -

4-59

O i

Turbine Bypass Valve Concem

Pneumaticactuatorassembly cracked and failed during cooldown i operationsseveralhours following niactor trip.

i Findings: Intemalvalve components became disengaged and caused hammer blow forces which damaged actuator.

t Corrective 1. Repaired damaged valve.

Actions: 2. Repaired steam traps and drains.

3. Refurbished otherturbine bypass valves. -

4. Revised operating procedure to i assure properdrainage of headers.

Generic Applies to both turbine bypass valve implications: headers.

O A-70 l

O Safety Parameter Display System (SPDS)

Concem: Both SPDS Control Room W._/

devices were inoperative during event -

theyareintended to be used bythe operators during transients.

Findings: Bad fiberoptic cable and faulty ,

terminations on data transmission cable.

Corrective 1. Utilhed spare cable.

Actions: 2. Corrected terminations.

3. Replacing obsolete terminal prior to O Generic None-no otherfiber optic systems.

l Implications:

g ---v-= - -

O System Review and Test Program Summary of Corrective Action items for Davis Besse O

l o va

o L 1 l

Reactor Coolant System PriorTo Restart a Test operability of PORV actuations solenoid for extended feed and bleed environmerdal conditions.

m Repair / replace and calibrate PORV discharge line temperature element.

einstallPORVloop sealdrain line.

m Modify PORV statusindications.

m improve stroke time on containment isolation valve.

m Repair leakage from RCS to nitrogen system.

m Realign hot leg Rosemount RTDs to TSAT meter.

Long Term a Repair end plugs on out-of-service pressurizer hestorelements.

m improve maintenance capability for RC drain tank.

a Repair / replace nitrogen regulator for RC drain tank. .

s Repair / replace Tave digital readout.

m Evaluate RC pump seal or motor parameters to be used forsecuring pumps.

• Calibrate RC pump vibration instrumentation.

u tmprove core exit temperature measurement capability.

m Modify pressurizer heat bank operation to handle larger hostlosses.

O A-Y3

~

4 O

Qualification of PORV e Successful completion of operability qualification for PORV solenoid for conditions e;+7=,d during extended feed and bleed operation.

eTest Conditions

- Testing environment at 220T and 15 psig.

- Initial 20 minutes on/off cyclic op;.aiion.

- 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> continuouslyenergized.

- Solenoid mounting bracket connected to thermalsource at 6509.

- Solenoid plunger loaded to represent valve t operation.

O High Pressure injection System PriorTo Restart s Confirm HPl pump capability at high flow, low head conditions.

m Confirm HPl pump capability at high suction temperature.

e investigate standing water fmm unidentified sources in several areas (e.g., ECCS room #1; containment vesselannulus floor).

m Remove startup strainers from HPI pump suctions.

m Modify HPl pump suction check valves (HP10 and O HP11) to facilitate proper disk sosting.

m Verify proper operation of several HPl pump component cooling water stop check valves.

m Revise plant documentation to reflect higher design pressure capability of section of piping downstream of discharge check valve HP23.

e investigate increasing tend in inboard bearing vibration for HPl pump 1-1.

O High Pressureinjection System (Cont'd:

Long Term a Evaluate need for design modification to HPl pump l component cooling water stop check valves to l preclude " sticking". ,

a Calibrate HPl pump motor temperature '

instrumentation.

m Eliminate low flow nuisance alarm on HPl pumps.

m Replace improper fkw measurement orffice in HPl pump 1-1 minimum recirculation line.

a Relocate control cable for AC lube oil pump for HPl O pump 1-1 per Appendix R requirements.

m improve communications capability between ECCS rooms and Control Room.

m Resolve means of leak testing for back-to back check valves in HPl discharge valves.

m Evaluate cyclic life of % inch elbowlets attached to HPl dischargeline.

O Core Flood System PriorTo Restart a Confirm location of level taps on Core Flood tanks.

m inspect / repair nitrogen regulator for Core Flood tanks.

Long Term a Prepare calibration procedure for Core Flood tank levelinstrumentation.

• Evaluate required purge time for sampling of Core Flood tanks.

• Calibrate sampling purge flow meter.

O A17

O l

Decay Heat Removal System PriorTo Restart e install level indicating capability for DH11 & DH12 valve pit.

e install modified pacidng in DH11 & DH12.

a Modify procedures to preclude overpressurization )

of decay heat pump suction.

a lmprove cold weather operation of BWST levelinstruments.

m Reinstall mis assembled studs on decay heat pump.

a Calibrate boron dilution flow transmitter instrument 1

! strings.

6 .

O O A-7f l

O Decay Heat RemovalSystem (Cont'd)

Long Term a Evaluate means of eliminating sealed pit design for valves DH11 & DH12.

m Evaluate possible system modifications to preclude decay heat pump suction overpressurization.

m Ensure spare parts availability for BWST recirculating pump and heater.

m Add temperature alarm on SWST; evaluate improved temperature control.

m Evaluate altamative means of precluding over-O ranging of low range decay heat pump suction pressure gauges.

m Evaluate more easily read oil level indicators for decay heat pumps.

m Evaluate means of reducing leakage from cyclone separators on decay heat pumps.

m improve preventive maintenance for pneumatic valves DH13A & B.

e investigate means of improving disk seating for checkvalves DH76 & DH77.

O A-79

O .

Containment Spray System PriorTo Restart a Verify torque switch and torque switch ty/ pass settings for two Containment Spray System valves (CS1530 and CS1531).

" Determine specific operator response required when containment emergency sump level indicator lights areilluminated.

Long Term a Evaluate adequacy of oil level sight glasses for containment spray pumps.

] a Evaluate need for containment spray pump discharge pressure indication in Control Room.

9 h-80

O Containment Emergency Ventilation System PriorTo Restart a Provide weather shield for EVS fan controller sensing line to ensuring operability.

8 Replace seals on hydromotor actuator for ventilation damper.

Long Tenn a Provide protection for instrumentation controls for fans.

O Q A-8/

O Containment Air Cooling System

and Hydrogen ControlSystem PriorTo Restart a Repair backdraft dampers for Containment Air Cooling System fans.

m Confirm proper operation of fans when shifted from high to low speed (overioad indication is being received).

m Bench test hydrogen dilution system relief valve.

Long Term a Confirm flow balancing of Containment Air Cooling System.

O m Regiace faiied r tum nend on cooier ci-4.

• Evaluate installation of flow meter in purge test line to facilitate p. Liming surveillance tests.

m Install soft seat for valve CV210 (containment isolation checkvalve).

m Confirm hydrogen recombiner is compatible with Davis-Besse system and can be made operable within required time.

/t-81

O Makeup and Purification System PriorTo Restart a Repair or replace failed containment isolation valve MU33.

, a Remove any startup strainers in system. i e Confirm correct valve trim in MU32.

i a Repair or replace lealdng reactor coolant letdown pressure reducing valve (MU6) and controls.

m Provide indicator of makeup flow for use during  !

feed and bleed operations. l e Remove inoperable and unused reactor coolant pump (RCP)sealleakageindication.

9 k-83

O Makeup and Purification System (Cont'd)

Long Term a Repair / replace lealdng reactor coolant batch makeup flow control valve, MU39 (currently W-a Repair / replace batch controller.

m Repair minor oil leaks on makeup pumps.

m Evaluate and repair leaking valves MU19 (RCP seal injection valve) and MU216 (RCP seal injection bypass valve).

m Evaluate and repair boronmeter.

e Perform review of problems with hydrogen system t

O (for maintaining makeup tank overpressure).

a Repair MU1903 (Cation domineralizer inlet isolation valve).

m Repair /r@ seal injection stop check valves.

m.9;/sceletdown block ortfice.

e improve communications capability between makeup pump room and Control Room.

m Evaluate intended function of boron permissive light.

• Perform review of failure modes of makeup system equipment (e.g., power supplies) and ability of operators to recognize problem and take corrective action.

k- $$

O 250/125 Volt DC System PriorTo Restart aNone Long Term a improve temperature control in battery room.

e imprwe ground fault detection and location.

m Evaluate design change for low voltage relays to reduce failures.

O i

O 4160 Volt AC System PriorTo Restart a Confirm operability of several breakers with original levering-in device.

s Replace CVE synchrocheck relay with different design.

8 Resolve tap setting for 4160/480 Wit transformers and reset alarm accordingly.

Long Term a Review consequence of paralleling bus tie transformers out-of-phase.

O m Visually inspect all 4160 Volt breakers in "Q" cubicles for"E'4fng druTage.

m Control nona 'Q" 4160 Volt breakers to ensure they are not utilized in essential applications.

m Provide improved control of circuit breaker and relay setting records.

l l

l l

O i l

480 Volt AC Distribution System PriorTo Restart a Confirm operability of switchgear cabinet door and stab withdrawal interlocks to ensure 480 VAC breaker operability.

Long Term a Evaluate removal of stab withdrawal interlock feature.

m Evaluate attemative means of providing ground fault protection for 480 Volt AC motor control centers.

O s Establish program for tracking limited life components in motor control centers.

I e improve preventive maintenance on swittigear cabinet door hardware and gasketing.

i F

a

)

1 O i 13.8 KV System PriorTo Restart a lnvestigate cause of fast transfer bmaker failures and calibrate relay timing.

Long Term a lmprove method for racking in/out of 13.8 KV breakers.

a Repair small oil leak in Auxiliary Transformer.

O l

O h-88

O

~

Emergency Diesel Generators PriorTo Restart a Elminate electrical noise problems associated with diesel govemor to imprtwo stability.

e improve temperature control for diesel generator room by changes to control / alarm system and maintaining ventilation damper actuators.

m improve reliability of lubricating oil soak back pumps.

Replace set.

Check condition of filters and strainers.

O 1.si muiiigie air siari ca,abiiiiv of diesei .eneraior.

m Calibrate low cooling water flow alarm switch.

e investigate cause of SCR diode failure nuisance l alarm.

m Minimize ice buildup on diesel generator air intake.

A-09

O Emergency Diesel Generators (Cont'dJ Long Term a Evaluate improvements to diesel g;acider air start system and improve reliability of air compressors.

m implement impic.;.T.;.ih for emergency diesel fuel oilsystem.

m Replace cooling waterflow gauge, a Perform evaluation of overall impie.:.T.;ab to diesel generator air intake configuration.

O a Improve diesel generator speed and electrical frequency control capability.

m Conect erratic beadng temperature indications for dieselgenerators.

e improve Control Room / diesel generator room communications capability.

i l

l 1

I kk0 l

l _

0 120 VoltInstrument AC Power PriorTo Restart e install larger power rating resistors in essential ,

inverters YV1, YV2, YV3 and YV4. '

Long Term a install ventilation fans for inverters.

s Evaluate installation of static transfer switches and/or tank circuits to reduce potential for losing an inverter when a ground fault occurs.

O O k4/

O Anticipatory Reactor Trip System PriorTo Restart a Evaluate adequacy of no actuation of annunciator alarms when ARTS cabinet door is open.

l Long Term i

e Modifylamp test circuit.

m Provide labeling to minimize problems in conelating channel and breaker designations.

m Evaluate separating ARTS channel signal inputs to computer to facilitate determining which parameter initiated an ARTS trip.

i O R vi.w main turbine sio, vaive insiino as gotentiai source of spurious low pressure ARTS trips.

1 O 4-e*

O Control Rod Drive Control System PriorTo Restart u Determine improved power supply fuse size and design; p.L... inrush current and current waveform test.

e improve cleanliness of CRDCS cabinets to reduce contact fouling problems.

m Ensure adequate forced air cooling of reactor '

service structure.

O l

l l

O A -93

O Control Rod Drive ControlSystem  !

(Cont'dJ Long Term a Evaluate attemate control rod direction error circuit design.

m Evaluate control rod motion momentary interrupt circuit.

m Review overall service structure cooling design.

a lmprove power cable mating and handling procedures.

m Evaluate use of higher temperature silicon power cables.

]

a inspect all control rod leaf spring anti-rotation devices at each refueling outage.

m Evaluate improved control rod drop time test techniques.

m To preclude low voltage problems, check voltage output of CRD transformers frequently; clean and inspect inductrol voltage regulator and lubricate .

motor-generator set every refueling outage.

m Determine long-term resolution of CRDM nozzle flangeleaks.

l e Evaluate means of minimizing occurrence of low insulation resistance in CRD stators.

O AM

.. .._ - .- ---..-.----------%------------ - ~ ' -

O .

Incore Monitoring System PriorTo Restart a Determine proper correction factors for neutron I

detectors.

Long Term a lmprove reliability of incore neutron detectors for 15% to 30% reactor power.

m Evaluate need for two incore instrumentation multipoint recorders.

. O O A-fr

O Reactor Protection System PriorTo Restart a Repair / replace defective compor.;aio in NI Source Range string contributing to erratic noise and loss of signalproblems.

m Install electronic filters on reactor coolant flow transmitters to reduce flow turbulence noise.

Long Term u Evaluate Technical Specification change to permit placing a defective channel in manual bypass.

e Evaluate providing more reliable power range r

4 O signal to Integrated Control System.

m Eliminate noise spikes in NI Source Range channels apparently due to door alarm switches on RPS cabinets.

L f

- F

-.--v --_..,y y-. - - - . - -- ..

O Steam and Feedwater Rupture Control System (SFRCS)

PriorTo Restart a Provide filtering of steam generator level transmitter signal.

m Modify SFRCS to preclude isolation of main feedwater and main steam on low water level in steam generator.

m Modify SFRCS to isolate only first steam generator '

forwhich low pressureis detected.

m Modify SFRCS such that a^ u sissitieric vent valves are closed by a full SFRCS trip (actu#Jon) rather O ta a w tr'a-a Modify SFRCS to provide open signals to MS106A and MS107A for all SFRCS actuation conditions.

m Rearrange manual SFRCS actuation switches and provide protection against inadvertent actuation.

e improve Control Room annunciator indication of which steam generator has been source of SFRCS actuation.

m Remove and replace all wire wrap terminations on logic boards and card racks.

O .

Steam and Feedwater Rupture Control System (SFRCS)

PriorTo Restart (Cont'd) e Relocate reset buttons for startup feedwater valves to the control board.

e impnwe temperature control of SFRCS steam generator pressure switch sensing lines.

m Revise labels on manual resets for indicating Ilghts associated with steam generator level instrumentation.

m Provide separate manual reset for "SFRCS Full THp" alarm.

O m improve SFRCS power supply operation.

Installforced cooling to cabinets.

Measure power supply loading and estimate useable servicelife.

m Perform engineering evaluation of and measure response time for replacement ampilfier/ calibration boards for steam generator level transmitters.

l 1

D Steam and Feedwater Rupture ControlSystem (SFRCSD(Cont'd)

Long Term a Remove automatic close signals to AF599 and AF608 and leave valves open to improve overall reliabilityof AFWSystem.

m Modify SFRCS to preclude isolating both steam generators if coincident low pressure signals are  ;

received for both steam generators. t a Evaluate modifying control circuitry for main steam isolation valves to improve reliability.

m Evaluate several additional changes to SFRCS to l

improve system reilability and improve decay heat removalcapability.,

m Evaluate additional improvements of Control Room annunciator indication of SFRCS actuation.

m Evaluate removing SFRCS close signals to atmospheric vent valves.

m Establish improved record keeping for SFRCS power supplies and trend to better determine expected servicelife.

m Monitor, periodically, the 125 Volt DC bus to ermure noiseis at acceptablylowlevel.

• Modify steam g;.;;..' wr level instrument monitors.

b09

Safety Features Actuation System PriorTo Restart a Modify SFAS to avoid ungrounded power supply common problems by 1.Wilng separate sensor channel power supplies. Accomplish by completely rewiring power supply wiring to sensor logic.

m Confirm AC and DC contact cunent in SFAS output relaysis within design capability.

m Repair / replace surveillance light cards with damaged components or electrical connections.

m Perform complete logic, calibration and response O "- ta as a' S'^S ca aa ' -

Long Term a Evaluate altamate surveillance card design.

m Evaluatt ?!imination of SFAS closure of main steam isolation valves.

e investigate spurious SFAS incident level 1 trips due to spildng of radiation monitorstrings.

e improve human engineering considerations associated with relative location of SFAS manual trips and associated reset pushbuttons (2) location of reactor coolant pump seal injection and seal retum valve controlswitches.

t

O t Integrated ControlSystem Non-Nuclearinstrumentation '

PriorTo Restart s Perform additional neview of ICS and Non-Nuclear instrumentation to evaluate plant and operator response on loss of powersupplies.

m Remove and replace improper wire wrap terminations.

m improve main feedwater pump runback control upon unit trip (rapid feedwater reduction circuitry).

m Replaceinoperable fuse holders.

a Calibrate selected control modules.

m Perform action plan 16 requirements.

a Modify load balance control for turbine bypass valves, t

O a Upgrade selected controlmodules.

Long Term a Evaluate altematives to rapid feedwater reduction controlscheme.

m improve preventive maintenance system.

m improve proportional and integrating module response.

m Provide additional cooling for ICS cabinets.

m Provide monitoring capability for selected ICS parameters (diagnostic).

m Replace powerselectorswitches.

u lmprove transfer capability for pressure inputs to ICS forturbine bypass valves.

A-/DI

O Security System PriorTo Restart a Evaluate / modify security requirements to improve i op;..^wraccess.

• Review electrical loads on uninterruptible powersupply.

• Revise procedures in event of loss of ventilation to Central Alarm Station.

Long Term a Perform evaluation of several operational / reliability impnwements forsecurity system.

O

1

~

O Control Room Normai and Emergency Ventilation Systems PriorTo Restart a Modify overall control scheme for water <:ooled and air-cooled condensing modes of Emergency Ventilation System (EVS).

• Manually adjust Service Water System cooling flow ,

to water-cooled EVS condenser to accommodate seasonal changes.

m Replace refrigerant solenoid control valves and install additional stop and check valves to facilitate l

0 'ria r atcoat i'aravs-aincrease cooling capacity of EVS.

e install / repair gaskets on Cc6uvi Room door and security room door to limit air leakage.

m Calibrate control and indicating instrumentation.

l 0 A-to3

O .

Control Room Normal and Emergency Ventilation Systems i.ong Term a Install flow modulating contml valve in service l watercooling path for EVS.

m Further increase cooling ceifs. city of EVS. <

a Limit use of manual switches on local controi l panele associated with EVS.

• Perform air flow balance of normal ventilation O system.

8 Review overall adequacy of normal ventilation l

system.

m Develop improved preventive maintenance ,

procedures for entire system including dampers and actuators.

e inspect and refurbish air handling duct work and associated insulation, filters and different!al pressure units to improve cleanliness conditions in Control Room.

e improve operator Indications and control regarding normaland EVS operation.

• Clean and repair humidification system. Evaluate altomato designs.

m Evaluate causes of failures of chlorine detectors and station vent air particulate monitors.

O f -/M

O Station Air and Instrument Air Systems PriorTo Restart e isolate and repair leakage in Station Air and {

instrument Air systems to maximum practical i extent.

m Perform required station air compressor preventive maintenance. -

a Test setpoints of control valves used to maintain E instrument air header pressure in the event of k station airsystem failure.

a Revise procedures and testing to identify the O i - p er ry di supply.

i ir c o - a or ta a c ku p ir m Provide improved reliability of diesel air compressor during inclement weather. .

m Procedurally biw down drains from SA28 to remove accumulated moisture in station air system.

m Provide filtering of air supply through SA2010 to minimize debris accumulation.

O Hsc

O

~

i i

Station Air and Instrument Air Systems (Cont'd) '

Long Term  ;

a Perform engineering study of Station Air and instrument Air systems to evaluate several areas for improving overall reliability (e.g., required emergency air compressor size; isolable vs. non- '

isolable loads; need for 100/100 peig regulators; leakisolation flexibility). '

m install moisture trap on SA28 to improve draining.

a Segregate air intake and diesel exhaust for diesel air compressor.

Q m Modify Station Air and Instrument Air systems to prevent dumping system air to atmosphere when the dryers are byM.

m increase frequency of dowpoint check on instrument air receiver to detect unacceptable moisture leakage from station air headers to instrument air headers; install permanent air dryer around IA408.

m Evaluate providing capability to start station air compressor 1-2 from Control Room.

m install flow meter to permit trending of air system degradation.

O A-to6 L--- - -- . . _ _ _ _ _ _ _ _ _ _

O Station Fire Protection  ;

PriorTo Restart l e Provide audible fire alarm in Control Room.

i a Provide fire watches for areas affected by improperty installed fire dampers.

Long Term s Maintain fire alarm location backup on  !

security computers, a Revise circuitry for ionization smoke detectors to avoid spurious alarms.

m Evaluate improving testing and maintenance accessibility of several smoke detectors and temperature switches.

m Resolve operational problem with ionization smoke

detector installed above control rod drive breakers.

l 8 Complete fire protection enhancement program on schedule consistent with Appendix R commitments.

a Provide time delay for diesel fire pump start to minimias unnemry starts while the electric fire pump brings up system pressure.

m Resolve power supply problems with Viking fire panels.

8 Modify fire alarm display in Control Room so that both panel number and zone number are provided.

m Replace / modify improperfy installed ventilation duct fire dampers.

1 l

0 A-k?

O Component Cooling Water System  !

PriorTo Restart '

s Functionally test CCW pump room ventilation fans.

m Repair or replace nitivi;;a regulator for CCW surge tank.

Long Term a Add CCW room ventilation fan periodic test.

e implement altamate pressure control for CCW surge tank.

m Fix smalloilleaks on CCW pumps.

l f

6 O A -/o S

I I

l i

l Service Water System '

PriorTo Restart a Modify ventilation for service water pump rooms.

m Perform ultrasonic test of selected portions of piping and fittings.

m Resolve service water flow through containment air coolers.

Long Term a Periodically ultrasonically test selected portions of

!O service watersystem.

a Review service water pump shaft p.kini ance post-modiacation.

m Evaluate apparent increased head and flow on service water pump 3-1.

m Evaluate tube corrosion / performance in ECCS room coolers.

a Modify containment air cooler valves to improve

! stroke time.

! a lnspect tubes in component cooling water heat l +Ei.i.6:3.

e improve temperature control for Component

Cooling WaterSystem.

l 0 _

A -/09

l O

Auxiliary Feedwater System PriorTo Restart a Evaluate applicable portions of November 21,1985 transient at San Onofre Unit 1 regarding check valve inoperability and water hammer.

a install air p..J d steam admission valve near i

each auxiliary feedwater pump turt>ine (AFPT);

resolve associated high energy line break items.

a Ensure operability of steam traps on AFPT steam

! supply lines and periodically reconfirm operability.

a Revise AFW pump automatic suction transfer l

O setpoints for switching from condensate storage tankto service water.

m Resolve overpressurization problem for discharge 4

piping from AFW pumps.

e install PGG governor on AFFT #1, as currently on AFPT #2; revise low speed stops to accommodate changesin govemor operation time.

a Evaluate coincident AFW and MFP feed to steam generators.

• Confirm AFW pump discharge valves will open for maximum credible differential pressure.

l l [

,e - -- - - - - .-- w-. m_

i O

I Auxiliary FeedwaterSystem PriorTo Restart (Cont'd) e Remove individual suction strainers tt .AFW pumps and enlarge mesh on common strainer from condensate storage tank.

m Remove control power from suction valves FW786 and FW790 to preclude spurious closure, e increase time delay on suction pressure switch actuation which isolates steam supply to AFPTs.

e implement impiu.;.T.; aim to AFPT overspeed trip mechanism.

O m Calibrate AFW pump flow instrumentation.

e implement controls to preclude steam binding of AFW pumps (NRC IE Bulletin 8501).

m Install pressure switches to detect main steam line breaks upstream of MS106, MS106A, MS107, MS107A.

i 0 4-m 1

O

^

Auxiliary FeedwaterSystem (Cont'd)

Long Tenn e investigate means of monitoring for condensate in i

AFPT steam supplylines.

a impnwe steam generator level control capability

when using AFW pumps.

m Evaluate leaving AF3870 and AF3872 open during plant operation to further improve AFW System l reliability.

! e Evaluate potential for contamination from Service Water System to AFW pump suction.

j u Improve operability of manual mechanical overspeed trip on AFFTs.

almprove access to AFW pump #1 room.

e inspect internals of AFW pump and turbine during next refueling outage. ,

a Provide accurate means of monitoring AFW pump recirculation flow.

e improve AFW pump rooms ventilation co.ei.

e improve Control Room board control, Indication and resetlayout.

I

/-//h 8

O Main Steam System  :

PriorTo Restart s Refurbish and check setpoints for all main steam safety valves (MSSV's).

a install lift stop collars on "R" orffice MSSV's.

i e Install lateral snubber on "A" main steam header.

, a Recalibrate atmc=g-tic vent valve controls.

l e Administratively require plant shutdown if both 1050 poig setpoint MSSV's on a header are inoperable.

m Test smaller inlet bore size for MSSV's.

O e Establish leak rate for testing of main steam non-retum valves.

Long Term ainstall MSSV monitodng system.

a lmprove mechanical design of main steam isolation valve position switches.

m Investigate increased noise and general vibration with full arc admission to main turbine.

O h -//3

O l Steam Generator System PriorTo Restart

! u Provide engineering analysis regarding acceptability of having exceeded procedural pressure / temperature limits during testing.

Long Term a Improve reliability and accuracy of steam generator shell thermocouple temperature readings.

s Evaluate modifications to condenser steam Jet air ejector radiation monitors.

O I

l I

l l

bl

- . - - . , - . - ~ _ . ___,._ - - _. . . . . - - - . - ~ - - - - - - - - - - -

1 0

1 Main FeedwaterSystem i

PdorTo Restart a Evaluate applicable portions of November 21,1985 l transient at San Onofra Unit 1 regarding check valve inoperability and water hammer.

j a lnvestigate and correct cause of power supply

failures for main feedwater pump turtaine (MFPT) controlsystem.

a lmprove reliability of MFPT low pressure drainage i heeder pumps.

e Evaluate via testing ability for automatic vs. manual O control of main feedwater pumpe between 15% and 45% plant power.

j u Minimize potential for overfeeding steam generators after reactor trip by modifying intadocks between startup valve and main feedwater blockvalve.

e improve reliability of MFPT hydraulic oil system.

• Resolve apparent vibration problem on MFR' 1-1 pump end bearing (instrument problem).

i a lnwestigate high pressure differential on MFPT thrust bearing wear monitor.

)

l l

I k-//f

O .

Main Feedwater System (Cont'd)

Long Term a Perform overall reliability study of main feedwater system and associated support systems.

m Evaluate modifications to high pressure drain system to minimize flashing and vibration.

m Repair or replace level indicating sight glasses on high pressure feedwater heaters.

a Conect variance among MFPT turbine speed indicators.

O mi-arove sii aiiiivorarerto-ia r vsiem.

e improve flow control during feedwater cleanup operations performed prior to each reactor plant startup.

e O A -H6

. O .

Post Accident Sampling System PriorTo Restart a Perform overall operability testing to confirm adequacy of several previous design changes. ,

a Correct several component maintenance deficiencies (e.g., sample pump diaphragm replacement; flowmeter i=p45:+T.;at).

m Conduct additional training of Chemistry and Health Physics personnel.

m Confirm time and motion study for taking samples adequately addresses system design changes.

O 'oao rer-a Evaluate altomate PASS isolation valve designs to ensure adequateisolation.

A-//7

O l

Gaseous Radwaste System PriorTo Restart a Revise Station procedures to require periodic sampling of Clean Waste Receiver Tanks for oxygen i

and hydrogen upon failure of monitons.

Long Term l m Eliminate air inleakage to Gaseous Radweste t

System.

! u Evaluate redundant oxygen and hydrogen monitors on Weste Gas Surge Tank and Clean Weste ReceiverThnks.

O m Evaluate relocation of oxygen / hydrogen monitors for the Clean Weste Receiver Tanks to minimize l maintenance personnel radiation exposure.

! a impnwe nitrogen overpressure control on waste gas tanks.

m Obtain reliable operation of degasifier.

A -nB

. . - - ~ ~ ~

SYSTEM REVEW & TEST PROGRAM @)

Two Phase Program Evaluation i

The staff hos estabished on eight-point program to review the SRTP.

1. Evoluote the proposed SRTP to determine the degree to which it con achieve stated program objectives.

2. Assess whether the list of systems important to safe plant operation is sufficiently complete to provide reasonable assurance of safe plant operation. This would include evaluation O of specific justifications for excluding any safety-related system.

3. Review the lists of system functions important to safe plant opera-tion to determine whether they are complete with respect to both specific system functions and plantwide system safety functions.

4. Review selected test outlines to ensure that they encompass oil system functions required for safe plant operation and that the systems are tested under anticipated operating conditions. This

' would include review of proposed justifications for not testing any system function deemed important to safe plant operation or not testing systems at anticipated system operating conditions.

Items 1 - 4 discussed in Safety Evaluation Report.

A-l19 1

__..__._____._.,___.___,ll.~_ET__._____..__ _ .__________________ _

SYSTDi REVEW & TEST PROGRAli (SRTP)

Two Phone Program Evoluotion (cont)

5. Review, witness, and evaluate the results of selected i

system tests.

06. Verify that the licensee has developed, performed, and evaluated the results of remaining test procedures.

I

7. Audit maintenance record reviews, personnel interview result, test leader qualification, etc.

8. Observe SRTP meetings such as JTG meetings, IPRC meetings, etc.

O items 5 - 8 discussed in Rlli inspection Reports.

, A -ta o

- - - - - - - - - - - - - - - - - - - - - , - - - - ~ , - - - - - - - - , - , _ , , --.,,,,---v-n,- - - - - - - , - - - - - - -- - - ----

g -

_- ~ .

Rill REVIEW l

1 l

• Review the SRTP System Review Reports for 10 of the 34 l

Systems, to ensure the review is accomplished within SRTP quidelines.

O Review the licensee's test program similar to the methodology utilized in IE WC-2513 for preoperational test programs [i.e., three-tiered approach).

- Review, witness, and evaluate 4 of 6 integrated tests.

- Review, witness, and evaluate the testing for 10 of 34

- systems deemed "Important to safe plant operations".

Verity the licensee has developed, performed, and evaluated the results of the remaining test procedures.

A -M/

ON-SITE REMEW ACTMTIES

• Since 9/23/85,4 NRC and 7 contractor inspectors have participated in the program evaluation.

O

• Weekly coverage provided from 9/23/85 to 5/23/86 w1en test activi<:ies slowed due to WCP Shaft investigation.

• Weekh coverage planned until restart.

O A -/ n .

NRC REVIEW STATUS

• Reviewed 16 of the 34 System Review Reports.

• Reviewed 1 of 6 integrated tests.

3 tests approved.

None performed.

• Licensee has developed test program consisting ,

I of 112 test procedures and 171 surveillance tests.

TED has performed 72 test procedures.

NRC has reviewed 64 and witnessed 45 test procedures.

TED has approved the results of 28 test procedures.

NRC has reviewed 18 of these packages.

TED has performed 39 surveillance tests.

NRC has reviewed 13 and witnessed 12 O '""'"''4"''""

available for review.)

A -/R3

1

- - - ~ ~ -

1 5f5155 MPORTANT 10 S&ft NBA!W

I Group i 1

Esseter Coelant System I

• Righ Pressure hjection Care fleeding System

• Decay Beat Removal & 14v Pressure lajection Centainment Spray System Centainment Emergency Yentilation Centainment Air Coeling and Rydrogen Centrol l

• Makeup and Purifiation System Group 2

• Electrical 125/250 TDC (lacludes Battery RM H&T)

[ Eeetrical 4.18 KY 3ys.(11.8/4.1EY Transformers)

• Rectrical 480V Distribution (lacludes hverters and Beguired Transformers) f Electrical 13.8 KY Systana (Inchdas Startup and Andliary Transformers) i

• Esmergency Diesel Generators (lacludes "Q" Fuel Oil Tanks and Diesel Room Ventilation) hatrument AC power (includes laverters and Required Transformers)

• = nac neview.4. A -/W

~U 'x ~

SYSTEMS NPORTANT TO SAFE OPERATION [ cont)

Group 3 Anticipatory Reactor Trip System Control Rod Drive System Incore Monitoring [ Includes Core Exit TC)

Reactor Protection System

• Steam Feedwater Rupture Control System

• Safety Features Actuation System Integrated Control System

• Security System Group 4

• Control Room Normal and Emergency H&V System

• Station and Instrument Air

• Station Fire Protection O Component Cooient Water System Service Water System A -M8 l

x 1

! SYSTEMS IMPORTANT TO SAFE OPERAfl0N (

l f Group 5 l

Auxiliary Feedwater System l

Main Steam Steam Generator System O

• Yain Fee 6:ater System ADDITIONAL SYSTEMS FOR SUPERVISORY RE ,

• Gaseous Radwaste

• Post Accident Sampling  ;

Fuel Handling Ventilation O + Miscellaneous containment 1 solation Va:yes

• = NRC Reviewed.

A~*

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( . ( ( ( l r r

. e . . e e e e t e . S S S S l

e l

e m e P P n u t

s P T T v v e n i

e T S P S E E i

t e o t

s e T ( ( R s t t e v T . .

. e s s E

S e T e e T o S T n P T f i s (T (

d u T S r e e 'e S S t F . .

V e P T a s T S r .

n g o t

e n

S y l

\ N i L

e c

I

[dW ,

Q Oo wU go O)g -

a tQ

' :1 1 :ll llf  : !I - i. .  !

,l t tjll

SIGNIFICANT RESULTS OF SRTP System Review

• Inodequate cooling water flow to the Containment Air Cooling Unit.

• Inadequate ventilation in the Service Water Building.

O Test Review

• Inadequate periodic testing of High Pressure Injection Valves.

• Inadequate periodic testing of Decay Heat Removal Valves.

• Inadequate period testing of Auxiliary .

Feed Pump Turbine [AFPT) Steam Generotor (SG) Level Control System.

• Inadequate venting of High Pressure O Injection System downstream high j point vents. A'M8 l

S GN lCAT" RESJ_"S 0 SR"3 ' con':}  !

Sys'em "es':inC

• maro3ery wireC Con':Oinmen; Air Coo er I

O n O n C; ess ':10n CesiCneC ow c030ci':y.

O

• DOC eC UO':e 0W OnC Cis"ri3u': ion for Service WO:er 3umas.

• maro3er ower Oir 3ressure imi; "or .

mu':ia e s:Or:s o' EmerCency Jiese Gener0:or ')G).

O

• no3ero)e Ven':iO': ion Sys:em 20maers.

A-iar

SIGNIFICANT RESULTS OF SRTP [ cont) l l

l Program implementation Review

• IPRC Meeting Winutes had not been provided from 11/15/85 to 4/86.

RESOLUTION:

All ba;k minute have been provided for re,vic .

• A Significant Events Log was not required to be maintained for the performance of STs.

RESOLUTION: A niini-rhonolr9101 leg it noA require,d.

O A method did not exist to ensure required changes to S1s identified in the review process were accomplished prior to test performance.

RESOLUTION: A tracking and audit system has been developed.

• A 10 CFR 50.59 Safety Evaluation was not required to be l performed for each of the new tests developed in the SRTP. -

An audit has been performed to ensure o safety

! RESOLUTION:

i evaluation has been performed for each TP.

s Problems identified in the Test Review were not reported within the time period required.

An audit to verify that oii identified problems O RcS0tVTiON:

have been reviewed for reportabili is ongoing.

-/30

_ . = _ _ - __,

! l i .

APPENDIX V -

NRC STAFF BRIEFING ON DAVIS-BESSE i

i I

l l

i i

l l

k i

ACRS BRIEFING j DAVIS-BESSE i j RESTART FROM 6/9/85 EVENT , ,

. JULY 11, 1986 ,

i

.l i

CONTACT-CONRAD

MCCRACKEN X28124 l

l l

I I

SLIDE 1 I

/ -/31  :

i

s' O i i

BACKGROUND o MULTIPLE BRIEFINGS TO ACRS AND DAVIS-BESSE SUB-COMMITTEE SINCE IIT REPORT ISSUED o 1/10/86 STATUS REPORT TO FULL ACRS COMMITTEE l o 1/29/86 DRAFT SAFETY EVALUATION REPORT ISSUED, INCLUDED OPEN ITEMS o 2/6/86 PRESENTATION OF DRAFT SAFETY EVALUATION TO SUB-COMMITTEE o 6/10/86 FINAL SAFETY EVALUATION REPORT ISSUED o 6/27/86 PRESENTATION OF FINAL SAFETY EVALUATION TO SUB-COMMITTEE l

SLIDE 2 l

l

) -/3h

O FINDINGS o THE PLANT HAS BEEN IMPROVED BEYOND OR RESTORED TO ITS LICENSING BASIS o THE HEALTH AND SAFETY OF THE PUBLIC WILL NOT BE ENDANGERED BY RESUMPTION OF POWER GENERATION o CONTINGENT UPON IMPLEMENTATION OF LICENSEE COMMITMENTS IDENTIFIED IN NUREG-1177, THE STAFF CONCLUDES THAT THE DAVIS-BESSE NUCLEAR POWER STATION MAY RESUME OPERATION ,

6 SLIDE 3 O A-/33 1

l l

\

MAJOR CHANGES TO SER SINCE 2/6/86 MEETING o MOTOR OPERATED VALVES Pg. 3-28 o MAIN STEAM SAFETY VALVES AND ATMOSPHERIC VENT VALVES Pg. 3-33 o SAFETY FEATURES ACTUATION SYSTEM Pg. 3-52  :

o SAFETY SIGNIFICANT HED'S Pg. 3-63 o SINGLE FAILURE CONSIDERATIONS Pg. 5-1 o SIGNIFICANT COMMITMENTS o MAINTENANCE Pg. 3-4 ,

ho o

SYSTEMS REVIEW AND TEST PROGRAM Pg. 3-64 i

SLIDE 4 O A -/3+

's

. O MOTOR OPERATED VALVES PROBLEM DEFINITION:

- FAILURE OF AFW VALVES TO REOPEN DURING JUNE 9,1985 EVENT LED TO ROOT CAUSE DETERMINATION THAT CONTROLS WERE IMPROPERLY SET CONCERN EXISTED THAT OTHER MOV'S MAY NOT FUNCTION RESOLUTION:

ALL VALVES (167) IN SAFETY SYSTEMS AND SYSTEMS IMPORTANT TO SAFETY REWORKED USING MOVATS REPRESENTATIVE VALVES TESTED AT DESIGN BASIS CONDITIONS O

SLIDE 5 O A-/35

l l

O  !

MAIN STEAM SAFETY VALVES AND ATMOSPHERIC VENT VALVES PROBLEM DEFINITION:

- MAIN STEAM PRESSURE FLUCTUATION ANOMALIES DURING THE JUNE 9, 1985 EVENT CAUSED BY IMPROPER OPERATION OF MSSV'S AND AVV'S i RESOLUTION:

AVV CONTROL SYSTEM REPAIRED

- MSSV'S REMOVED, REWORKED AND RETESTED LICENSEE WILL SUBMIT LONG TERM MAINTENANCE PLAN SLIDE 6 2

O 3 -/s t

O SAFETY FEATURES ACTUATION SYSTEM PROBLEM DEFINITION:

l

- PRIOR TO THE JUNE 9, 1985 EVENT, CONCERN EXISTED THAT THE SFAS DID NOT l HAVE SUFFICIENT INDEPENDENCE TO MEET SINGLE FAILURE RESOLUTION:

- MODIFICATIONS INCORPORATED TO MEET CHANNEL INDEPENDENCE CRITERIA 0F IEEE STANDARD 279-1971

?

6 SLIDE 7 0 mer

4 O

SAFETY SIGNIFICANT HED'S PROBLEM DEFINITION:

- PRIOR TO THE JUNE 9, 1985 EVENT THE DCRDR IDENTIFIED 29 SAFETY SIGNIFICANT HED'S. DUE TO THE EVENT RELATED HED'S RESOLUTION OF THESE ISSUES WAS NEEDED FOR RESTART RESOLUTION: -

PERMANENT OR INTERIM FIXES INSTITUTED FOR RESTART INTERIM FIXES TO BE PERMANENTLY RESOLVED PRIOR TO COMPLETION OF STH REFUELING OUTAGE O -

ALL IDENTIFIED HED'S TO BE CORRECTED BY COMPLETION OF 6TH REFUELING OUTAGE i

SLIDE 8 i

Q A -/ 3 8 l

O SINGLE FAILURE CONSIDERATIONS PROBLEM DEFINITION:

SINGLE FAILURE CONCERNS WERE IDENTIFIED BY THE IIT FOR THE SFRCS AND AFWS.

BASED ON THIS CONCERN A SINGLE FAILURE REVIEW OF THE RPS WAS PERFORMED.

RESOLUTION:

LICENSEE'SEVkLUATIONDEMONSTRATEDACCEPTABLESINGLEFAILUREPROTECTION 6

SLIDE 9 Q A-t39

SIGNIFICANT COMMITMENTS o PRIOR TO OR DURING RESTART REWORK AND TESTING OF ALL SAFETY RELATED MOTOR OPERATED VALVES COMPLETION OF SYSTEMS TEST PROGRAM o AFTER RESTART

- RESOLUTION OF CONTROL ROOM HED'S DURING STH REFUELING OUTAGE, CORRECTIO!!S IMPLEMENTED BY COMPLETION OF 6TH REFUELING OUTAGE PLAN FOR ROUTINE INSPECTION OF MSSV'S (90 DAYS FOLLOWING RESTART)

TURBINE STOP VALVE CLOSURE TIME TECHNICAL SPECIFICATION (90 DAYS FOLLOWING RESTART)

?

COMPREHENSIVE AFW RELIABILITY STUDY (90 DAYS FOLLOWING RESTART)

TECHNICAL SPECIFICATIONS FOR MOTOR-DRIVEN FEED PUMP (60 DAYS FOLLOWING RESTART)

COMPLETE DESIGN AND INSTALL SAFETY GRADE DEPRESSURIZATION CAPABILITY PRIOR

TO RESTART FROM STH REFUELING OUTAGE SLIDE 10 l

l D

(J. A-Mo

-_ v - _ . - - - - _ .,-_--.,,.,.,.-.._.__._.,..-__m..m.___- - _ . _ , ,, _ - - _ _ _.-.4_._.--- ._., __ _ _ . . . , . , - . , _ ,

a. - + . _ _ - . , _ _

l I

\

APPENDIX VI

. MAINTENANCE REVIEW -

AT DAVIS-BESSE 0 );-/4l

b U

• TWO MAINTENANCE SURVEYS CONDUCTED BY THE STAFF SEPTEMBER 16-20, 1985 MARCH 2f4-27,1986 TEAM COMPOSITION 4

SEPTEMBER SURVEY: 2 NRR, 2 REGION III, 1 OIE, 2 CONTRACTORS (PNL)

MARCH SURVEY: 2 NRR, 2 REGION III, 2 OIE FIVE OF SIX MEMBERS ON MARCH SURVEY HAD PARTICIPATED IN SEPTEMBER SURVEY

. O

• BOTH SURVEYS CONSISTED OF INTERVIEWS WITH PLANT PERSONNEL SEPTEMBER SURVEY UTILIZED PROTOCOL DEVELOPED IN MSPP MARCH SURVEY l

REVIEW OUTSTANDING MW0s l'

ACCOMPANYING SITE PERSONNEL PLAN OF DAY, WEEKLY STATUS MEETINGS FINDINGS ARE CONSENSUS JUDGEMENT OF TEAM A -Mu

O

• SEPTEMBER SURVEY FOUND WEAKNESSES IN:

CORPORATE COMMITMENT SPARE PARTS / MATERIAL READINESS SUPERVISION PREVENTIVE MAINTENANCE MAINTENANCE WORK ORDER BACKLOG MAINTENANCE PROCEDURES COMMUNICATIONS DEFINED RESPONSIBILITIES TRAINING i

i

• CONCLUSIONS MODIFICATIONS TO MAINTENANCE PROGRAM ADDRESSING WEAKNESSES MANY CHANGES BEING IMPLEMENTED j T00 EARLY TO JUDGE EFFECTIVENESS i ANOTHER SURVEY BE CONDUCTED 4

A -/+3

r 4

MARCH SURVEY REVIEWED SAME NINE AREAS OF WEAKNESS ALSO REVIEWED MAINTENANCE WORK ACTIVITY PLANNING AND SCHEDULING: ADMINISTRATIVE PROCEDURES CONCLUSIONS CONSIDERABLE PROGRESS MADE IN IMPLEMENTING NEW MAINTENANCE PROGRAM CONSISTENTLY COMPLYING WITH STATED BASES FOR DETERMINING MW0s REQUIRED FOR RESTART SOME PROBLEMS NOTED BY TEAM - NOT CONSIDERED MAJOR PROGRAMMATIC WEAKNESSES FUNCTIONING AS INTENDED WITH NO MAJOR IDENTIFIABLE WEAKNESSES REGION III WILL MONITOR OUTSTANDING MW0s TO ASSURE CONTINUED CONTROL AND PROGRESS REGION III WILL CONTINUE TO MONITOR MAINTENANCE AT DAVIS-BESSE A -/4 4

APPENDIX VII NRC REASSESSMENT OF B&W PLANT DESIGNS NRC REASSESSMENT OF B&W PLANT DESIGNS NRC ENC 0URAGED THE B&W OWNERS GROUP (BWOG) TO TAKE THE ~ .

LEADERSHIP ROLE IN THIS MATTER ,,

l BWOG PROGRAM DOCUMENTED IN MAY 15, 1986 SUBMITTAL NRC STAFF COMMENTS ON BWOG PROGRAM DISCUSSED AT JUNE I4,1986 l I

MEETING NRC-BWOG WORKING MEETINGS ICS/NNI - MAY 21, 1986

- SENSITIVITY STUDY - JUNE 19, 1986 TAP - TBD SYSTEMS REVIEW - TBD RECOMMENDATION TRACKING SYSTEM - TBD ,,'

e e

. O } ~/W

, l l

NRC PROGRAM PLAN PRIMARILY REVIEW B&W OG EFFORT ,

WILL INDEPENDENTLY REVIEW TAP REPORTS -

WILL REREVIEW DISPOSITION OF PREVIOUS STAFF

. RECOMMENDATIONS WILL PERFORM COMPARIS0N TO OTHER PWRs -

- OPERATOR ACTIONS I

- PRA

- THERMAL-HYDRAULIC RESPONSE B8WOG PROGRAM

- INITIAL ASSESSMENT INDICATES PROGRAM IS GENERALLY ON TARGET ,

- DOCUMENTATION OF RESULTS OF STAFF REVIEW PLA,NNED FOR DECEMBER 1986 (DEPENDENT ON BWOG SCHEDULE)

O

~

-,.-,-n-, , . , . , . . - -- , . - , - _ . . , , _ - . - , - , . - - . - , , - , , , _ , , _ , , . - - - ---------,-------,--.--.+e. r -,

APPENDIX VIII NRC REASSESSMENT PROGRAM SCOPE b

O

' ~

E C REASSESSE NT PROGRAM SCOEE

1. REVIEW OPERATING EVENTS (*) -

2. IDENTIFY PROBLEMS, ROOT CAUSES, AND SENSITIVE SYSTEMS (*)

3. OBTAIN'MW PLANT SPECIFIC ESIGN IFORMATION (*)

L1. ASSESS PREVIOUS STAFF REVIEWS OF ANTICIPATED OPERATIONAL OCCURANES AND STATUS OF IMPLEENTATION (*)

5. REVIEW EXISTING PLANT PRA'S (*) ~

6. ASSESS MW PRA'S

7. COWARE B&W AND OTER PRA'S O

V

8. PERFORM SYSTEM ANALYSES AND SENSITIVITY CK0(S AGAINST CURRENT CRITERIA (*)

9. EVALUATE MARGINS AND SENSITIVITIES EYOND REGULATORY ORECTIVES

10. II NTIFY ADDITIONAL SAFETY CONCERNS

11. REGIONAL OPERATIONS EXPERIENE FEEDBA0(

12. IEEGRATED ASSESSENT TO IENTIFY ALTERNATIW IWROWENTS b

i

• ThSKS WHERE OWER's groto is likely to play a neJor role in cartpleting.

A-/+7

_ NRC REASSESSENT PROGRAM ORECTIVES ,,

o REASSESS BASIC ESIGN AND CHARACTERISTICS OF TE ,

MW REACTORS o COWARE OVERALL SAFETY OF TE EW REACTORS TO OTER PWRS o REDUE FREQUENCY AND COTLEXITY OF ANTICIPATED OPERATIONAL -

TRANSI,ENTS o IMPROVE OVERALL SAFETY OF TE B&W REACTORS 1

Oo IEEEIFY POTERIAL REVISED LIENSING CRITERIA i

l 3

l -

k $0

? i

.. , APPENDIX IX

• KEY DATES RELATED TO THE D&WOG PROGRAM KEY DATES RELATED TO THE B&W OG PROGRAM o DAVIS BESSE LOFW TRANSIENT 6/85 -

1 o B&WOG DISCUSSION OF DAVIS BESSE TRANSIENT IMPLICATIONS 7-8/85 o B&WOG DECISION TO FORM ',,

SPECIAL TASK FORCE 9/85 o RANCHO SECO OVERCOOLING TRANSIENT 12/85 o B&WOG DECISION TO RESTRUCTURE -

& REFOCUS OG ACTIVITIES 1/86 o NRC DECISION TO REASSESS B&W PLANT DESIGN 1/86

l o B&WOG DECISION TO LEAD REASSESSMENT 2/86 o B&WOG REASSESSMENT .

PLAN COMPLETED 5/86 F wi ,

1 l

O STELLO 1/24/86 LETTER HIGHLIGHTS ,

SENSITIVITY OF B&W PLANTS TO OPERATING

~

TRANSIENTS. j NEED TO REEXAMINE B&W PLANT DESIGN ,,

REQUIREMENTS AND NEEDTO SUPPLEMENT.

SAFE TO OPERATE AT THE CURRENT TIME.

- RECOGNITION OF PAST, CURRENT, AND FUTURE MODS. ..

NEED FOR B, ROAD EVALUATION OF B&W DESIGN.

INTEREST IN B&WOG PARTICIPATION & LEAD ROLE.

- COMPLETE REEXAMINATION EFFORT THIS YEAR.

CHANGES TO BE PER BACKFIT RULE.

l MAIN POINT ,,

THE FREQUENCY OF COMPLEX TRANSIENTS IS TOO HiGH l

o O

use ,

=-- - -

- -a s- .a - --_ ,. z__y_ _,,, ,__g _m,,, _ _

/

O TUCKER 2/13/86 LETTER HIGHLIGHTS -

B&WOG'WILL LEAD EFFORT TO REDUCE TRIP FREQUENCY AND IMPROVE TRANSIENT RESPONSE. ~

B&WOG WILL DEVELOP A PLAN AND WORK WITH THE STAFF IN FORMULATING THAT PLAN.

NRC ABLETO MONITOR PROGRAM PROGRESS. .

B&WOG COMMITED TO RESOLVING CONCERNS RELATED TO TRIPS AND TRANSIENT RESPONSE.

O MAIN POINT -

PROGRAM WILL REDUCE COMPLEX TRANSIENT FREQUENCIES.

M

=

N L_._- . - - _-- - . ---

f ACRS MEETING WITH THE B&W OWNERS GROUP JUNE 25,1986 ,

i l

AGENDA l I. INTRODUCTION W. S. Wil US '

G. R. SKILLMAN

11. INFORMATION GATHERING KEY PROJECTS TRANSIENT ASSESSMENT PROGRAM S. T. ROS5 (0 d' ROOT CAU5E S.T. ROSE OTHER DATA S.T. ROSE INTERVIEWS S. E. MAYS SENSITIVITY STUDY S.T. ROSE H. ESTRADA regn ll1 NTEGRATION PHASE KEY PROJECTS ICS/NNI SYSTEM REVIEW C. B. DOYEL l -

MFW SYSTEM REVIEW C. W. TALLY RISK ASSESSMENT S. E. MAYS IV. INTEGRATION AND IMPLEMENTATION L.A. REED l

V.

SUMMARY

OF ACTIVITIES G. R. SKILLMAN -

CONCLUDING REMARKS W. S. WILGUS-/S L .:

e 5 5 m 5 17, a s B

    $      LOSS OF INTEGRATED CONTROL SYSTEM POWER o
    "r e                      -

AND W g 5 5* wm OVERC00 LING TRANSIENT N

 &8 e a 1

AT t RANCHO SECO ON DECEMBER 26, 1985 O O O

                                                                                         .  /.

i i i T F A M _M _E _M B E R S l FREDERICK J. HEBDON. LEADER

                                                ,               HENRY A. BAILEY s%

J.T. BEARD h 5/' g RONALD B. EATON i LGORDON E. EDISON r < f f

              >                                                                                                                                    r 9                                                                                                                       G 9

i, i s- o

  .             -                                            -           C --     4            - _ _ _ _             __           _ __   _ . _ _ _

l RANCll0 SECO NilClEAR POWER STATION t OPERATED BY TliE SACRAMENTO MUNICIPAL llTILITY DISTRICT (SMllD) LOCATED 25 MILES SOUTHEAST OF SACRAMENTO, CALIFORNIA N'. BABC0CK & WILCOX DESIGNED REACTOR LICENSED IN AllGilST 1974 9 O 9

I i ks l SF 20rT1 a AFW ISFAS) l A OTSG FLOW CONTROL VALVE l I AFW PUMP P-318 CONDENSATE IDU AL-O RIVEl 3,g g,g gg,g l TANK

                                                                    ~

f h( FV 20527 l l C AFW llCS) l FLOW CONTROL VALVE l HV 31826 g

=, -, ><

SFV 20578 i g  ! RESERVOIR g2 g HV 31 (SFAS) 77 B OTSG I O FWS-054 FWS 119 FWS 064 l r  %  % >< FV 20528 N  %;= e llCSI l i AFW PUMP P-319 a l V IMOTOR DRIVEN) CONTAINMENT BUILDING BOUNDARY AUXILIARY FEEDWATER SYSTEtt (S!!1PLIFIED) O O O

I FROM MOISTURE , SEPARATOR REHEATORS A OTSG mal FROM STEAM AUXILIARY SYSTEM STEAM STARTUP MFW FLOW CONTROL VALVE l , 4

                           ,  r ,, ,                  ,
                                                                       ,                                                                       Q, FV 20575 I

SECOND POINT FIRST POINT l HP EXHAUST TO

.P CONDENSER

                                                                                \                 HHP
                                                                                                                                                               ,      ,           j FROM LP                                       ['                                         M             M                   py,g            py,g i        i                 N FW HEATERS _                                                                                                                                                  g MAIN MFW       MFW STOP           3 P-317A                         1 r                                    FLOW CONTROL        VALVE           g      BOTSG      N RECIRCULATION TO VALVE                           I       O          3 LP CONDENSER                                                                        I I
                                                                                                                                                                           +        +

I s V FROM LP FV 20526 FV 20530 FW HEATERS P-317B MFW PUMPS CONTAINMENT FV 20576 ' BUILDING DOUNDARY e _ _ _ __ HAIN FEEDWATER S aM(SIl1PLIFIED) e

e i O MOISTU SEPARATOR peg ygp ATMOSPHERIC DUMP ! V ALVE S 13/HE ADE R) I e m l SAFETY VALVES 19/ H E AD E RI e I ' I "' TURBINE BYPASS

                                                                                                             : VALVES TO CONDENSER l                  HV 20505,,                                                               12/HE ADE R)
                                                          *                                      ~

l J L PEGGING STEAM

               ->       e                                             TO rW HE ATERS                                                               -

M HV.20569 l

                  .v.

T g i X X X Xe=UR8 inert eTS l ___ ,0 M ,. . M .S n rv aosos

                                                                                                               = T e _ .. M . . 2,.

yg 1 CO NT AIN M E NT V.20596 l SUILOING TUR8tNE COUNDARY CONTROL HV.20560 PEGGING" STE AM VALVES l TO FW HEATERS , i g u 7 ' ' s-HP TUR8INE I I n I l. gar u l l'n"AinV# lREHEATORS V 8 0750 (1AIN STEAi-1 SYSTEt1 (Sit 1PLIFIED) O O O

t s 4- D. ~

                   .~              ^o ~                                      e, s.

9 i S, c i M C R,c m o. r .~ c. a I, == --

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AlddnS B3M0d 30 W31SAS 70 BIN 03 031VB931NI 1 l r---------- 7 l F----] l 1I Sne s3i ll Sne So, 3PA 9P 3PAtl+ 1I 32 52 2s 2E .. _. . l1 , S30010 ONIW33NO113ny g i li F r__ y _-,_ _ 4_ lI lI _ .. Il i

                                  ;lr--

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                                                                       -  -~'
                                                                                   *-                     ~

lI ll lIliiIi _ m. ll I lil l1I ll iti ill I 1 lill II l ill o-  ! II ii liil i lllut

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               %       3Wn11V A      -
               #         0 531
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66 SQVO13VS3001 4 ~> W3dSNYW1 Sne 010W (L Sne 1VilA NON WOW 10 3L SAG 1VilA WOW A O J ~

                                                         /, -M O

l t DESCRIPTION OF Tile EVENT INITIAL CONDITIONS i 76% POWER RCS TEMPERATilPE 582*F N

                                                              $I t

RCS PRESSURE 2150 PSIG y - ICS IN FULL AllT0MATIC RANCil0 SECO DOES tt0T liAVE MSIVs O O O

b i DESCRIPTION OF THE F. VENT (CONTINilED) LOSS OF ICS DC POWER CAUSED BY SPUPI0llS ACTION OF THE POWER SilPPLY MONITOP WHICil OPEt!NED SWITCHES S1 AND S? AFW(ICS) FLOW CONTROL VALVES TO 50% h MAIN AND STARTllP MFW FLOW CONTROL VALVES TO 50% $ l TURBINE BYPASS VALVES TO 50% g ATMOSPHERIC DUMP VALVES TO 505 MFW STOP VALVES SHUT MFW PUMPS TO MINIMllM SPEED (N0 FLOW) LOST REMOTE (HAND) CONTROL FROM THE CONTROL ROOM

     #                                            9 e

DESCRIPTION OF THF EVFtlT (CONTINUED)

 -'  AFW PUMPS STARTED ON LOW MFW PUMP DISCHARGE PRESSilRE PEACTOR TRIP ON HIGH RCS PRESSilRE (16 SECONDS AFTER LOSS OF ICS DC POWEP)          g M

PLANT C00LDOWN BEGAN

 -   NONLICENSED OPERATORS SENT TO SHUT OR ISOLATE THE TBVs, ADVs    AND AFW(ICS) FLOW CONTROL VALVES 9

O O O

i DESCRIPTION OF THE EVENT (CONTINilED) OPERATORS STOPPED THE MFW PilMPS SFAS ACTIJATED (2 MINUTES 54 SECONDS AFTER THE REACTOR TPIP) INITIATED FUl_L llPI OPENED AFW(SFAS) FLOW CONTROL VALVES SHilT SUCTION VALVE FROM MAKEUP TANK TO MAKEUP PUMP AND "A" HPl PllMP N(s ' DR I OPERATORS SHUT THE AFS(SFAS) FLOW CONTROL VALVES '9C PRESSURIZER EMPTIED (5 MINIJTES AFTER REACTOR TRIP) REMAINED EMPTY FOR AB01.iT 3 MINUTES BUBBLE FORMED IN THE UPPER REACTOR VESSEL HEAD ADEQUATE RC3 SilBC00 LING MARGIN MAINTAINED t 9 9 9

1 iDFSCRIPTION OF THE EVEtlT (CONTINilED) l l - OPERATORS TRIED TO DETERMINE CAUSE OF LOSS OF ICS POWER 1 PLANT ENTERED THE PTS REGION OPERATORS ISOLATED TBVs AND ADVs 4i OPERATORS ATTEMPTED TO SHilT Tile AFW(ICS) FLOW ('ONTROL VALVES T PARTI ALLY CLOSED THE "B" VALVE (TH0llGilT TO DE FULLY CLOSED) CLOSED THE "A" VALVE (THOUGHT TO BE PARTIALLY OPEN) COMPLETED CLOSING THE "B" VALVE BROKE THE "A" VALVE (REOPENED) OPERATOPS DECLARED AN " UNUSUAL EVENT"

          #                                                                               9                     e

l i DESCRIPTION OF THE EVFFT (CONTINilFD)

       "A" 0TSG OVERFLOWED FOR ABOUT 9 MINUTES OPERATORS FOUND THE  "A" AFW MANUAL ISOLATION VALVE SEIZED ND OPERATORS SHilT SWITCHES S1 AND S2 AND RESTORED DC POWER WITilIN THE ICS
                                                                                         $i A

CONTROLS REENERGIZED IN HAND AND VA!VES OPENED FilLLY

              -  OPERATOPS SHUT VALVES, INCLUDING "A" AFW(ICS) FLOW CONTROL VALVE AFW FLOW TO OTSGs STOPPED RCS TEMPERATURE 386*F PLANT HAD COOLED DOWN 180*F IN 26 MINUTES t

O O O

 ~~

I i DESCRIPTION OF Ti1E EVENT (CONTINUED) BACKUP SHIFT SUPERVISOR COLLAPSED A MAKEUP PUMP DESTROYED BECAUSE SUCTION VALVES SHUT BY SFAS AND OPERATORS w REACTOR BUILDING PADI ATION MONITOR PllMP OVERilEATED BECAllSE SilCTION VALVE SHUT BY SFAS NONLICENSED OPERATORS ENTERED MAKEUP PUMP ROOM LOST ICS DC POWER AGAIN RESTORED IMMEDIATELY O O O

I GENERAL FINDINGS AND CONCLUSIONS THE INCIDENT WAS SIGNIFICANT BECAllSE: A SINGLE FAILURE IN THE INTEGRATED CONTPOL SYSTEM, WHICH IS A NONSAFETY-

                                                                                                            \

RELATED SYSTEM, SUBJECTED THE PLANT TO AN UNDESIRABLE OVERC00 LING TRANSIENT. a

          .                                                                                                  '9:

DURING THE TRANSIENT, i

                         -   THE RCS COOLED DOWN 180*F IN 26 MINUTES,
                         -   THE PRESSURIZER EMPTIED,                                           ,
                         -   A BUBBLE FORMED IN THE REACTOR VESSEL HEAD,
                         -   THE PLANT ENTERED THE PRESSURIZED THERMAL SHOCK REGION,
                                                                                  ~
                          -  THE SAFETY FEATURES ACTilATION SYSTEM (SFAS) ACTUATED, WATER OVERFLOWED FROM A STEAM GENERATOR IflTO THE MAIN STEAM LINES.

O O O

4 GENERAL FINDINGS AND CONCLUSIONS (CONTINilED) THE FUNDAMENTAL CAUSES FOR THIS TPANSIENT WERE DESIGN WEAKNESSES AND VilLNERABillTIES IN THE ICS AND IN Tile E0lllPMENT CONTROLLED BY THAT SYSTEM. THESE WEAKNESSES AND VllLNERABILITIES WERE NOT ADE0llATELY COMPENSATED BY OTHFR DESIGN FEATilPES, Pl.AtlT y PPOCEDURES OR OPERATOP TRAINING. THESE WEAKNESSES AND VllLNERABILITIES WERE LARGELY KNOWt! TO SACRAMENTO MUNICIPAL UTILITY DISTRICT (SMllD) AND THE NRC STAFF BY VIRTilE OF A NUMBER OF PRECljRSOR EVENTS AND THROUGH PEl.ATED ANALYSES AND STilDIES. YET, ADE0VATE PLANT MODIFICATIONS WERE NOT MADE S0 THAT THIS EVENT WOULD BE IMPP0BABLE, OR S0 THAT ITS COURSE OR CONSEQUENCES WOULD BE SIGNIFICANTLY ALTERED. IN

SUMMARY

THE INFORMATION WAS AVAILABLE AND KNOWN WHICH C0l1LD HAVE PREVENTED THIS OVEPC001.ING TRANSIENT; BUT IN THE ABSENCE OF ADEQUATE PLANT MODIFICATIONS, THF IFCIDENT SHollLD HAVE BEEN EXPECTED. P O O O

PRINCIPAL FINDit!GS AND CONCLllSIONS I

1. Tile DECEMBER 26,1985 OVERC00 LING TRANSIENT WAS It'!TI ATED BY THE FAILURE OF A SINGLE MODULE IN THE NONSAFETY-RELATED ICS (1.E., THE SPURIOUS TRIPPING OF THE POWER SUPPLY MODULE THAT INTERRUPTED ALL +/-24 VDC POWER).

k THF MOST PROBABLE CAllSE OF THIS FAILURE N I WAS A DESIGN WEAKNESS THAT APPARENTLY MADE THE CIRClllT SilSCEPTIBLE TO ERRATIC OPERA-TION IF " CONTACT RESISTANCE" BETWEEN THE 24 VDC BilS AND THE POWER SilPPLY MONITOR WERE TO DEVELOP, AND THE DEVELOPPENT OF A HIGH PESISTANCE CONNECTION (1.E., A BAD CRIMP CONNECTION) IN THE WIRING BETWEEN THE +24 VDC BUS AND THE POWER SUPPL.Y MONITOP WHICH EXPOSED THE DESIGN WEAKNESS AND CAUSED THE MODULE TO TRIP. (SMllD HAS AGREED TO FURTHER EXPLORE THE CAUSE OF THE FAILURE OF THE POWER SUPPLY MONITOR BY HAVING AN INDEPENDENT LABORATORY CONDUCT ADDITIONAL ANALYSES). G . G S

i i PRINCIPAL FINDINGS AND CONCLUSIONS (CONTINilED)

2. UPON 1.0SS OF ICS DC POWER AND THE SilBSEQUENT AUTOMATIC REPOSITIONING OF A NilMBER OF VALVES !N THE PIANT, THE DESIGN OF THE ICS ALSO CAUSED THE l.0SS OF REMOTE CONTROI. OF THE AFFECTFD VALVES FROM TiiE CONTPOL POOM WHICH NECESSITATES MANUAL ACTIONS L0 RALLY AT THE VALVES.

k'N. 8

3. AN AFW MANilAL ISCLATION VALVE C0llLD NOT BE SHilT BY THE OPEPATORS AFTER THE Fall.llRE OF.

THE AFW(ICS) FLOW CONTROL VALVE. THE FAILURE OF TliE AFW MANUAL ISOLATION VALVE WAS

    ,THE RESULT OF A LACK OF ANY MAINTENANCE OF THIS VALVE DLIRING THE OPERATIONAL LIFE OF THE PLANT. THE LACK OF A MAINTENANCE PROGRAM RESULTED IN THE VALVF BEING IMADEQUATELY LilBPICATED, WHICH CAllSED THE VALVE TO SElZE. IT APPEARS THAT THE LACK OF A MAINTENANCE PROGRAM COULD AFFECT THE OPERABILITY OF OTHER MANUAL VALVES AT PANCII0 SECO.

i PRINCIPAL FINDINGS AND CONCLilSIONS (CONTINIJED)

4. RANCHO SECO EMERGENCY OPERATING PROCEDURES (EOP) DO NOT ADDRESS THE LOSS OF ICS POWER, THE LACK OF SPECIFIC GUIDANCE SEEMS TO BE A WEAKNESS IN THE PLANT-SPECIFIC E0Ps AVAILABLE TO THE OPERATORS ON DECEMBER 26, 1985. THE PANCHO SECO ANTICIPATED A TRANSIENT OPERATING GUIDELINES (ATOG) SUPPLIED BY THE BF.W OWNERS GROUP INCLlIDE AN EXPLICIT PROCEDURE FOR A LOSS OF ICS POWER AND THE ATOG DIRECTS OPERATORS TO THAT PROCEDURE. HOWEVER, THIS PROCEDURE WAS NOT INCLUDED IN THE RANCHO SECO E0Ps.

i O O O

i PRINCIPAL FINDINGS AND CONCLllS10NS (CONTINilED)

5. THE E0PS AT RANCHO SECO DIRECT THE OPEPATORS TO TRIP THE APP THIS WAS NOT DONE DilRING THE TERMINATE FLOW IF THE FEEDWATER FLOW CANNOT BE ISOLATED.

DECEMBER 26, 1985 INCIDENT. THE OPERATOPS WERE RELUCTANT TO STOP THE AFW PUMPS WHEN THEY HAD DIFFICULTY STOPPING FLOW TO THE ONC VALVE OPERATION. THE OPERATORS HAD DECIDED THAT THEY WO!!LD STO HOWEVER, THE OPERATORS FAILED TO WATER STARTED TO FLOW INTO THE MAIN STEAM LINES. ' ADEQUATELY MONITOR OTSG WATER LEVEL AND, AS A RESULT, WATER WAS INTRODUC STEAM LINES. THEIR RELUCTANCE APPEARS TO BE THE PESULT OF Tile SUBSTA PLACED ON THE AFW SYSTEM BY NRC AND OTHERS, AND A LACK 0F CONFIDENCE IN OF THE AFW PUMPS (1.E., FEAR THAT THE PUMPS W0llLD NOT RESTART IF STOPPED.)

• O e

i PRINCIPAL FINDINGS AND CONCLllSIONS (CONTINilED)

6. THE OPERATORS HAD CONSIDERABLE DIFFICULTY RECONCILING THE DICl10TOMY BETWEEN AVOIDING THE PRESSURIZED THERMAL SHOCK (PTS) REGION (E.G., REDUCING HPl FLOW) AND REGAINING i PRESSURIZER LEVEL (E.G., INCREASit!G HPI FLOW IN ACCORDANCE WITH THEIR E0Ps). TRAL t'ING AND PROCEDURES WERE NOT ADEQUATE TO RESOLVE THIS CONFLICT AND TO SOME EXTENT TENDED TO PROVIDE CONFLICTING INDICATIONS OF THE APPROPRIATE PRIORITIES.

7. THE OPERATORS RECEIVED NEITHER CLASSROOM NOR SIMUI.ATOR TRAINING ON Tile OVERALL PLANT RESPONSE TO EITHER THE TOTAL LOSS OF ICS DC POWER OR THE RESTORATION OF ICS DC POWER.

i i PRINCIPAL FINDINGS AND CONCLilSIONS (CONTINUED)

8. THE OPERATORS WHO INVESTIGATED THE LOSS OF ICS POWER DID NOT ADEOUATELY UNDERSTAND THE ICS POWER SYSTEM CONFIGURATION. WITH 120 VAC POWER STILL AVAILABLE FROM Tile 1C BUS AND THE ICS DC POWER SilPPLIES DE-ENERGIZED, Tile MOST CREDIBLE CAUSE FOR THE LOSS OF ICS DC POWER WAS THE OPENING 0F SWITCHES S1 AND S2. '!0 WEVER, THE OPFPATORS DID NOT j({

RECOGNIZE THIS FACT AND, AS A RESllLT, DID NOT SHUT THE SWITCHES UNTIL 26 MINUTES INTO 'd

                                                                                                    '9 THE TRANSIENT. THE FACT THAT SEVERAL OPERATORS DID NOT RECOGNIZE THAT SWITCHES SI AND S2 WERE OFF SUGGESTS THAT THEIR TRAINING ON THIS CRilCIAL SYSTEM WAS NOT ADEQUATE.       IN ADDITION, ALTHOUGH SIMPLIFIED DRAWINGS OF THE NON-NilCLEAR INSTRUMENTATION (NNI) POWER SUPPLIES WERE POSTED ON THE NNI CABINETS, COMPARABLE DRAWINGS F0P THE ICS POWER SUPPLY HAD NOT BEEN PROVIDED.

t O O O

PRINCIPAL FINDINGS AND CONCLUSIONS (CONTINUED) 9.. IT DOES NOT APPEAR THAT NONLICENSED OPERATORS PROPERLY OPERATED THE AFW(ICS) FLOW CONTROL VALVES. AN OPERATOR APPLIED EXCESSIVE FORCE WITH A VALVE WRENCH TO CLOSE AN AFW(ICS) FLOW CONTROL VALVE. HE DID SO BECAUSE HE DID NOT ACCURATELY DETERMINED THE POSITION OF THE VALVE WHILE ATTEMPTING TO SHUT IT COMPLETELY. AS A RESULT OF HIS g - ACTIONS, THE VALVE WAS DAMAGED, REOPENED, AND THE MANUAL (LOCAL) CAPABILITY TO OPERATE > THE VALVE WAS LOST. THESE CONSEQUENCES SilGGEST TPAINING WEAKNESSES IN TH USE OF VALVE WRENCHES, THE PROPER METHODS FOR MANUALLY OPERATING AND OVERRIDING AIR-0PERATED VALVES, AND Tile llSE OF AVAll.ABLE AND BACKUP INDICATIONS TO DETERP.INE VALVE POSITIONS. THESE WEAKNESSES SUGGEST AREAS WHERE HANDS-ON TRAINING RATHER THAN WALK-THROUGH OR TALK-THROUGH TRAINING MAY BE NECESSARY.

  'O                                             O                                          e

PRINCIPAL FINDINGS AND CONClllSIONS (CONTINilED)

10. WHILE THE DEFICIENCIES IN SMUD'S RADIOLOGICAL CONTROL AND EMERGENCY PREP PROGRAMS DilRING THE DECEMBER 26, 1985 INCIDENT DID NOT JFOPARDIZE THE PilBLIC HEALTil AND SAFETY DilE TO THE RELATIVELY MINOR RADIOLOGICAL CONSE0tlENCES OF THIS INCIDENT, THEY D0 INDICATE WEAKNESSES IN SMUD's PROGRAM AND THE TRAINING OF RAN I

T

11. THE PRC STAFF WAS LED TO BEllEVE THAT THE EMERGENCY FEEDWATER INITIATION (EFIC) SYSTEM W0llLD BE IllSTALLED IN 1984 IN RESPONSE TO A NUMBER OF NRC REQUIREMENTS, INCLUDING TM1 ACTION ITEM ll .E.1.2. APPARENTLY SMUD DECIDED T0 INSTALL AN ALTERNATE TO SATISFY ll.E.1.2 WITH THIS ALTERNATE SYSTEM IN RESPONSE TO II.E.1.2. SMUD's INTENT DESIGN WAS NOT MADE CLEAR TO THE NRC STAFF, WAS NOT APPROVED BY Tile STAFF, AND MAY NOT AS A RESULT, THE EFIC SYSTEM, SOME HAVE COMPLIED WITH THE RE0lilREMENTS OF ll.E.1.2.

FEATURES OF WHICH W0llLD HAVE REDUCED THE SEVFRITY OF THE DECEMBER ?6, 1985 INCIDENT, HAS NOT YET BEEN INSTALLED AT RANCHO SECO. 8 G e

e i i PRINCIPAL FitIDINGS AND CONCLUSIONS (CONTINUED)

12. ALTH0llGH THE RCS TEMPERATURE DROPPED 180*F IN 26 MitlVTES, IT W0llLD HAVE HAD TO PAPIDLY DROP ANOTHER 215*F (1.E., TO AN RCS TEMPERATllRE OF ABOUT 170*F), WHILE PRESSURE WAS MAINTAINED AT APPROXIMATELY 1400 PSIG, IN ORDER TO SER10llSt.Y THPEATEN REACTOP VESSEL h

INTEGRITY, (N

                                                                                                }

13. THE DECEMBER 26,1985 OVEPC00 LING INCIDEllT DOFS NOT APPEAR TO HAVE SEPIOUSLY TilREATEtlED HOWEVER, THE PLANT HAS HAD A NilMBER THE INTEGRITY OF THE RANCHO SECO REACTOR VESSEL.

OF OVERC00 LING INCIDEtlTS IN ITS 12-YEAR OPERATING HISTORY. EACH TIME THIS OCCURS THE POTENTIAL EXISTS FOR ADDITIONAL DPERATOR ERRORS AND EQUIPMENT FAILURES TIIAT MiGPT EXACERBATED THE EVENT AND SERIOUSLY THREATEN PEACTOR INTEGPITY. THilS, THE SIGNIFI-CANCE OF THIS INCIDENT LIES IN THE FACT TilAT UNDER ALTERNATE SCENARIOS MORE SERI0llS CONSEQUENCES C0llLD OCCUR.

    #                                           9                                           O

PRINCIPAL FINDINGS AND CONCLUSIONS (CONTINilED) lb. IT IS NOT CLEAR THAT THE OVEPC00 LING TRANSIENT WAS WITHIN THE FIN ALTHOUG,il PTS HAS BEEN ADDRESSED REPORT (FSAR) ANALYSIS OF THE RANC110 SECO PLAtlT. GENERICALLY, THE FSAR ACCIDENT ANALYSIS FOR RANCHO SECO DOES NOT ADDRESS } THIS T THE MOST COMPARABLE AMALYSIS IN THE FSAP IS FOR T11E C00LDOW IN BREAK, HOWEVEP, THIS ANALYSIS INCLUDES ONLY 100 SECONDS OF THE TRANSIENT. ADDITION, THE RANCHO SEC0 FSAR ANALYSIS OF MAIN STEAM LINE BREAKS APPEARS TO AND NONCONSERVATIVE IN TilAT IT ASSUMES TilAT THE NONSAFETY RELAT FULLY TO MITIGATE THE CONSEQUENCES OF THE ACCIDENT.

  #                                      9                                           e

i I PRINCIPAL FINDINGS AND CONCLllSIONS (CONTINUED) i

15. THERE WERE A NUMBER OF PRECUPS0RS TO THE DECEMBER 26, 1985 INCIDENT AT RANCHO SECO.

THESE PRECURSORS INDICATE TilAT IMPROVEMENTS IN THE REllABILITY OF T TO EFFICIENTLY MIT! GATE A LOSS OF ICS POWER llAVE NOT BEEN DEVELOPE g} RANCHO SECO DESPITE NllMER0llS EFFORTS Otl THE PART OF THE NRC STAFF T REllABILITY OF THE ICS AND TO ENSURE THAT THE NECESSARY PROCEDURES T N WHILE Tile STAFF HAD MITIGATE SUCH AN EVENT WOULD BE AVAILABl.E TO THE OPERATORS. RAISED THESE ISSUES ON A NUMBER OF OCCASIONS OVER THE PAST 6 TO 8 YEARS, SMUD PERSO HAD NOT IMPLEMENTED THE ACTIONS, AND THE NRC STAFF liAD t!0T TAKEN EFFECTIVE ACTION T0 t ENSURE THAT THE IMPROVEMENTS IN RELIABILITY AND THE PROCEDilRES WERE DE IMPLEMENTED AT RANCHO SECO. THE SPECIFIC FINDS ASSOCIATED WITH THESE PRECURSORS INCLUDE:

• O e

PRINCIPAL FINDINGS AND CONCLllS10NS (CONTINilEDI A. ALTHOUGH THE ICS POWER StlPPLY IS SIMILAR TO THE NNI POWER SilPPLY, PAPTICllLAPLY WITH RESPECT TO THE ROLE OF THE POWER SilPPLY MONITOR, SMUD's PRINCIPAL EMPHASIS FOLt.0 WING THIS  : THE LIGHTBilLB INCIDENT IN MARCH 1978 WAS ON THE NNI RATHER THAN ON THE ICS. EMPHASIS SEEMS TO HAVE BIASED SMUD's SUBSEQUENT REVIF_WS OF ISSUES ASSOCIATED WITH i THE NNI AND ICS. T B. THE LOSS OF ICS POWER TRANSIENT AT RANCHO SECO ON JANilARY 5, 1979 WAS SIMILAR TO THE DECEMBER 26, 1985 INCIDENT. HOWEVER, IT WAS M0T AS SEVERE AS THE "LIGHTBULB INCIDENT" AS A PESULT, CHANGES IN THE DESIGN AND DID NOT RECEIVE THE SAME LEVEL OF ATTENTION. OF THE ICS WERE NOT MADE AND PROCEDURES FOR LOSS OF ICS WERE NOT DEVELOPED,

                      #                                         9                                             9

i PRINCIPAL FINDINGS AND CONCI.IISIONS (CONTINilED) C C. IN MARCH 1979, B&W ISStJED A REPORT (BAV-1564) IN WHICH THEY ANAL.YZED THE RELIABILITY OF THE ICS. ALTHOUGH THE B8W ANALYSIS NOTED A NUMBER OF CHANGES THAT APPEARED h TO B WARRANTED IN THE ICS, SMllD CONCLUDED THAT NO CHANGES WERE NECESSARY. A SUBSEQUENT k I . i ANALYSIS OF THE ICS BY THE 0AK RIDGE NATIONAL LABORATORY CRITIClZED THE B8W AND NOTED TiiAT IT WAS OF LIMITED SCOPE AND DID NOT APPEAR TO MEET THE REQUIRE THE ORIGINAL ORDER. THE STAFF CONCLUDED THAT N0 IMMEDI ATE CliANGES WERE REQUIPED AT PANCHO SECO AS A PESllLT OF THE B8W ANALYSIS. THE LONG TERM ISS11ES ASSOCl ATED WITH TiiE B&W REPORT WERE TO BE CONSIDERED IN UNRESOLVED SAFETY ISSllE (USI) A-47, " SAFETY IMPLICATIONS OF CONTROL SYSTEMS."

    ;#                                         9                                             9

PRINCIPAL FINDINGS AND CONCLllS10NS (CONTINilED) D. AS A PESULT OF THE LOSS OF POWER TO NNI AND ICS AT OCONEE IN NOVEMBER 1979, NPC ISSllED BULLETIN 79-27 DESCRIBING A NUMBEP OF ACTIONS TO BE CARRIED OllT BY LICENSEES. ALTHOUGH THE BULLETIN RAISED SIGNIFICANT CONCERNS ABOUT THE CONSEQUENCES OF A LOSS OF POWER TO INSTRUMENTATION AND CONTROL SYSTEMS, SMUD CONCLUDED THAT NO ADDITIONAL DESIGN MODIFICATIONS WERE NECESSARY AND THAT EVENT-ORIENTED PROCEDURES TO DEAL WITH SilCH EVENTS WERE NOT NECESSARY. IT WOULD APPEAR THAT BULLETIN 79-27 WAS INITIALLY INTENDED TO SOLICIT DETAILED INFORMATION FROM LICENSEES THAT COULD FORM BASIS FOR AN IN-DEPTH } REVIEW 0F THE ISSUES ASSOCIATED WITH CONTROL SYSTEMS COMPARABLE TO THE REVIEW 0F SAFETY-RELATED SYSTEMS CONDUCTED AS PART OF AN OPERATING LICENSE REVIEW. BASED ON THE INITIAL SCOPE OF THE REVIEW, THE CONCLUSION WAS REACHED THAT SMUD's RESPONSE DID NOT CONTAIN SUFFICIENT INFORMATION AND DID NOT ADE0llATELY ADDRESS THE CONCERNS IN THE BilLLETIN . AFTER THE PP0GRESSIVE NARROWING OF TiiE SCOPE OF THE REVIEW, IT WAS DECIDED THAT THE SMllD RESPONSE WAS ADEQUATE, DESPITE WHAT APPEAR TO BE A NilMBER OF WEAKNESSES IN THE SMllD RF.SPONSE. THUS, THE CONCLilSION WAS FINALLY REACHED THAT SMllD HAD PROVIPF.D REASONABLE ASSilRANCE THAT THEY HAD ADDRESSED THE CONCERNS IN RilLLETIN 79-27, AND THAT , THE LONG TERM IMPLICATIONS OF BilLLETIN 79-27 WOULD ADDRESSED AS PART OF IISI A-47. O O O

PRINCIPAL FINDINGS AND CONCLl!SIONS (CONTINUED) I E. FOLLOWING THE FEBRUARY 1980 LOSS OF NN1 POWER AT CRYSTAL PlVER, Tile NRC IDENTIFIED All ISSUE ABOUT THE FAILURE MODE OF ATMOSPilERIC DUMP VALVES ( ADVs) ON LOSS OF ICS 't-POWER. SMUD's RESPONSETOTHISISSUEDIDNOTINCLUDETHEOTilERVALVESATRANCHOSECOSg I THAT REPOSITIONED ON LOSS OF ICS POWER (1.E., THEY CONFINED IT TO THE NARROW ISSUE ASSOCIATED WITH ADVs). IN ADDITION, SMllD DEFERRED THIS NARROW ISSUE TO INSTALLATION OF THE EFIC SYSTEM, WHICil TO DATE HAS NOT BEEtl illSTALLED AT RANCHO SECO. THE NRC FOUND THIS RESPONSE TO BE ACCEPTABLE.

  #                                            9                                          e

I i ( PRINCIPAL FINDINGS AND CONCLilSIONS (CONTIN!!ED) i F. BECAllSE OF CONCERNS ABOUT THE TRANSIENT RESPONSE OF B&W-DESIGNED REA ROLE OF ICS AS AN INITIATOR OF SUCH TRANSIENTS, NRC CONDl)CTED AN EXTENSIVE STljDY I HOWEVER, IT DOES NOT APPEAR THAT THESE AND MADE 22 RECOMMENDATIONS IN NUREG-0667. RECOMMENDATIONS WERE SENT TO SMUD FOR ACTION OR THAT THE RECOMMENDATION RELEVANT TO THE DECEMBER 26, 1985 INCIDENT WERE IMPLEMENTED AT RANCHO SECO. i , l l 1 f 8 9 e

PRINCIPAL FINDINGS AND CONCLUSIONS (CONTINUED) G. THE MARCH 19, 1984 PARTIAL LOSS OF NNI POWER AT RANCHO SECO AGAIN DEMONSTRATED THAT FAILURE OF NONSAFETY-RELATED EQUIPMENT AT B&W-DESIGNED PLANTS HAS THE POTENTIA CAUSE PLANT TRANSIENTS AND TO CHALLENGE Tile OPERATOR'S CAPABILITY TO MITIGATE THE DESPITE THE FACT TRANSIENT WITHOUT OVERC00 LING AND UNDERC00 LING THE PRIMARY SYSTEM. } THAT THIS EVENT OCCllRRED NEARLY 2 YEARS AG0, THE DECEMBER 26, 1985 INCIDENT DEMON- 3 STRATES THAT NEITHER SMUD NOR THE NRC STAFF HAS IMPLEMENTED EFFECTIVE ACTIONS RESOLVE TlilS SITilATION, IN QUESTIONS ASKED BY THE STAFF AND RESPONSES PROVIDED BY THE B&W OWNER'S GROUP FOLLOWING THE MARCH 1984 LOSS OF NNI POWER AT RANCHO SECO, THE TEAM AGAIN SEES STRONG EVIDENCE OF A NARROW FOCUS ON THE INCIDENTS INITIATED BY INA THE QUESTIONS PRIATE CONTROL SYSTEM ACTIONS IN RESPONSE TO FALSF. INPilTS FROM THE NNI. AS A PESULT, THE FULL SIGNIFICANCE OF IN GENERAL DO NOT REFER DIRECTLY TO THE ICS. THE LOSS OF POWER TO THE ICS WAS NOT ADDRESSED.

• O O

PRI!NCIPAI. FINDINGS AND CONClllSIONS (C0tlTINUED) H. WHILE THE SCOPE OF THE ANALYSIS PERF0PMED UNDER USl A u7 IS RPOAD, IT APPEARS THAT TO DATE THE ACTtlAl. STUDY INCLUDES ONLY THOSE EVENTS WITH THE POTENTIAL TO PRODilCE SllCil EVENTS ARE RAPE S0 SEQUENCES OUTSIDE THE DESIGN BASIS OF THE REFERENCE PLANT. THE STUDY DOES NOT APPEAR TO ADDRESS SUBSTANTIVE ISSUES OF THE FREQUENT CHAL TO PROTECTION SYSTEMS AND FRE0llENT ABNORMAL OPERATING OCCURRENCES, SUCH AS-THOSE IN IDENTIFIED IN BAW-1564, BULLETIN 79-27, AND NilREG-0667. IN ADDITION, THE ANALYSIS DOES NOT CONSIDER THE EVENTS THAT ARE SIGNIFICANT AT OTHER THAN THE REFER DIFFERENCES IN PLANT DESIGN THAT C0llLD CAUSE AN EVENT TO BE SIGNIFICANT AT A THEREFORE, IT APPEARS THAT THE ANALYSIS PER-PLANT ARE NOT ADEOUATELY CONSIDERED. FORMED TO DATE UNDER llSI A-47 DOES t40T ADDRESS THE LONG-TEPM ISSilES RAISED I 79-27, BAW-1564, OR NUREG-0667 THAT ARE RELEVAtlT OT THE DECEMBER 26, 1985 IPCIDENT. THUS, RESULTS OF THE RESOLUTION OF USl A-47 ARE OF OtilTE LIMITED APPLICABillTY TO THE RESULTS ARE NOT B&W-DESIGNED PLANTS BEYOND T11E REFERENCE PLANT TIIAT WAS STUDIED. 3 DIRECTLY APPLICABLE TO MOST OTHER B8W-PLANTS SUCli AS PAtICHO SECO BECAllSE OF TH DIFFERENCE IN THE DESIGN OF THE ICS.

    #                                           9                                               e

OTHER FINDINGS AND CONCLUSIONS

1. IT APPEARS THAT THE TRANSIENT INITIATOR (1.E., THE LOSS OF ICS DC POWER) WAS NOT FULI.Y RECOGNIZED BY CONTROL ROOM OPERATOPS tlNTIL 2 MINilTES AFTER THE POWER WAS LOS ALTH0tlGH THE "lCS AND FAN POWER FAllllRE" ALARM ALERTS OPERATORS ABOUT ICS POW FAILURES, IT APPEARS THAT ITS IMPORTANCE WAS SOMEWHAT OBSCllPED BECAUSE IT ALSO ACTS AS A TROUBLE ALARM FOR FAN FAILURE OR FOR LOSS OF ONE OF THE REDUNDANT ICS DC P SUPPLIES, NEITHER OF WHICH REQUIRES IMMEDIATE OPERATOR ACTIONS OR ltilTIATES A TRANSIENT.

2 '. THE ANNUNCIATOR PROCEDURES MANllAL WAS NOT USED BY THE OPERATORS FOLLOW EVEtt IF THE ANNilNCIATOR PROCEDURES MAtlllAL HAD BEEN USED, IT FAN POWER FAILURE" ALARM. CONTAINED VERY LIMITED GUIDANCE CONCERNil1G THE lFiLICATIONS OF THIS ALARM AN HAVE BEEN OF NO VAlllE TO THE OPERATORS IN PFCOGN171NG OR RESTOPIt!G THE LOSS POWER.

      #                                           9                                           e

i OTHER FINDINGS AND CONCLUSIONS (COMTINtJED)

3. THE ICS PERFORMANCE IIPON RESTORATION OF POWER ARE STILL NOT FULLY llNDERST00D, ESPECI ALLY BECAUSE PERF0PMANCE MAY DEPEND ON THE DilRATION OF THE POWER INTERPUPTION. HOWEVER, WHEN ICS DC POWER IS RESTORED, REACTOR OPERATORS REGAIN REMOTE CONTROL OF PLANT E0tilPMENT FROM g

THE CONTROL ROOM. (IT IS THE TEAM'S UNDERSTAFDING THAT THE BP.W OWNER'S GR0llP IS PLANNING TO CONDUCT AN INVESTIGATIVE PROGRAM THAT WILL INCLl1DE THIS MATTER.)

4. MOST OF THE INDICATORS IN THE CONTROL POOM (BOTH METERS AND RECOPDERS) ARE PART OF THE NNI SYSTEM; HENCE, THEY ARE GENERALLY INDEPENDENT OF THE ICS. HOWEVER, THERE ARE EXCEPTIONS THAT HAD NOT BEEN RECOGNIZED PRIOR TO THE DECEMBER 76. 1985 INCIDENT. FOR EXAMPLE, THE MFW FLOW RECORDERS APE AFFECTED BY THE ICS. DURING THE DECEMBER 26, 1985 INCIDENT, THE RECORDER FAILED TO A VAlllE NEAR MID-SCALE WHEN MFW FLOW WAS ACTUALLY ZERO.

O O O

e OTHER FINDINGS AND CONCLUSIONS (CONTINUED)

5. BECAllSE OF A PERCEIVED SENSE OF llRGENCY, TWO NONLICENSED OPERATORS MADE AN EMERGENCY ENTRY INTO THE MAKEUP PilMP WITH0llT RESPIRATORY PROTECTION OR ADEQUATE PPOTECTIVE CLOTHING, NEITHER OF WHICil WAS READILY AVAILABLE. AS A RESULT, THEIR CLOTHING WAS CONTAMINATED AND THEY WERE EXPOSED TO A!RBORNE RADIDACTIVITY. k

                                                                                                }

6. THE OPERATORS DID NOT REhEMBER A RECENT MODIFICATION HAD BEEN MADE TO PERMIT THE TBVs AND ADVs TO BE CLOSED FROM THE REMOTE SHilTDOWN PANEL (OUTSIDE THE CONTROL ROOM)

INDEPENDENT OF THE AVAILABILITY OF ICS POWER. THIS CHANGE WAS MADE TO ACCOMMODATE A FIRE IN THE CONTROL ROOM. ALTHOUGH THIS MODIFICATION HAD BEEN INCORPORATED IN THE CONTROL ROOM FIRE PROCEDllRES, SMUD DID NOT REVIEW OTHER PROCEDURES TO DETERMINE THE APPLICABILITY OF THIS MODIFICATION. 9 9 9

i OTHER FINDINGS AND CONClllSIONS (CONTINUED)

7. ADDITIONAL STAFFING ABOVE THAT REQUIRED BY PLANT TECHNICAL SPECIFICATIONS AND OTHER SMUD REGULATORY COMMITMENTS ALLOWED OPERATORS TO PERFORM CERTAIN TASKS SIMULTANEOUSLY. h);

ss WITH STAFFING AT THE MINIMUM REQUIRED LEVEL, THE ACTIONS PERFORMED WOULD HAVE HAD TO I

    -    BE PERFORMED SEQUENTI ALLY, WOULD HAVE TAKEN LONGER, At'D COULD HAVE EXACERBATED THE        C OVEPC00 LING TRANSIENT.

8. NEITHER THE OPERATORS NOR THE SHIFT TECHNICAL ADVISOR (STA) COULD IDENTIFY AN INSTANCE OF WHEN THE STA PROVIDED ENGINEERING EXPERTISE DURif1G THE ItCIDEllT. HOWEVER, THE OPERATORS FOUND THE STA VALUABLE AS AN EXTRA PERSON ON SilIFT TO HELP OUT DURING TliE i

INCIDENT, 9 O O

O si OTHER FINDINGS AND CONCLUSIONS (CONTIN 11ED)

9. IT APPEARED TO THE TEAM THAT SMUD PERS0liNEL FOUND THE PROCESS OF TROUBLESHOOTING IN lilGHLY CONTROLLED, SYSTEMATIC, AND WELL-DOCUMENTED MAf4NER, AS PROPOSED BY THE TEAM, TO THIS DIFFERENCE CONTRIBUTED BE QUITE DIFFERENT FROM THEiR USUAL MAINTENANCE PRACTICES.

, TO THE DIFFICULTY THAT THL TEAM EXPERIENCE IN PEVIEWING THE TROUBLESHOOTING PROGRAM.

10. THROUGHOUT THE TEAM'S REVIEW OF THE DECEMDER 26, 1985 INCIDENT, SMUD PFPSONNEL HAD CONSIDERABLE DIFFICULTY PROVIDIllG INFORMATION IN THE DETAIL THAT Tile TEAM REQUESTED.

THUS, SMUD PERSONNEL REPEATEDLY SUMMARIZED DATA, ANALYSES, AND PLANS WITHOUT INCLUDING THE ACTUAL DATA AND ANALYSES. AS A RESULT, Tile TEAM HAD TO REQUEST THE DETAILED UNDERLYING DATA AND ANALYSES, WitICH SilBSE0llENTLY WERE PROVIDED. THIS ITERATIVE PROCESS DELAYED THE TEAM'S ONSITE INVESTIGATION. G 9 e

OTHER FINDINGS AND CONCLUSIONS (CONTINilED)

11. IN JUNE 1983, THE B&W OWNER'S GROUP REPORTED (BAW-1791) THE RESULTS OF AN ANALYSIS WHICH PREDICTED AN OVERC00 LING TRANSIENT CAUSED BY A LOSS OF ICS POWER COULD OCCUR or AT IF b(y ?

B&W-DESIGNED REACTORS WITH A HIGH PROBABILITY (ABOUT 4X10-2 PER REACTOR i Y THIS PROBABILITY WERE APPLICABLE TO ALL EIGHT B&W-DESIGNED OPERATING REACTORS, SUCH A sa; THUS, TRANSIENT COULD OCCUR AT SOME B8W-DFSIGNED PLANT APPR0XIMATELY EVERY 3 YEARS. IT WOULD APPEAR THAT THIS ANALYSIS PREDICTS THAT EVENTS COMPARABLE TO THE DECEMBER 26, 1985 INCIDENT WOULD OCCUR APPR0XIMATELY ONCE EVERY THIRD YEAR EVEN IF THE EFIC SYSTE IN ADDITION, THE REPORT NOTES THAT ONE WERE INSTALLED AT ALL B8W-DESIGNED PLANTS. BF.W-DESIGNED PLANT HAD A COMBINATION OF COMPONENTS THAT CAUSE THE TRANSIENT FREQUE FINAll.Y, THE TO BE EVEN HIGHER. THE TEAM DEDUCED THAT THE PLANT WAS RANCHO SECO. GENERIC B8W PTS ANALYSIS (BAW-1791) IS NOT DIRFCTLY APPLICABLE TO RANCHO SECO BECAUSE IT ASSUMES THAT THE EFIC SYSTEM IS INSTALLED. O- O O

0

     '             NRR STAFF PRESENTATION TO THE O                                    ACRS                  APPENDIX XI NRC STATUS REPORT ON TVA REVIEW

SUBJECT:

TVA - STATUS REPORT ON THE NRC STAFF P.EVIEU

       ~

DATE: JULY 10, 1986 PRESENTER: R. H. WESS!!AN, TVA PROJECT STAFF B. J. YOUNGBLOOD, DIRECTOR, PWR PD#4 L. P. CROCKER, FACILITIES OPERATIONS BRANCH C. R. STAHLE, SR. PROJECT !!ANAGER - SE000YAH PRESENTER'S TITLE / BRANCH /DIV: PWR LICENSING-A PRESENTER'S NRC TEL. NO.: 492-7761

         .                                        492-8060 SUBCOMMITTEE: A0 H0C SUBC0!!!!!TTEE ON TVA C. WYLIE, CHAIR!!AN l

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STATUS OF STAFF ACTIONS ON TVA FACILITIES [v'} 0 NO SCHEDULES AVAILABLE FROM TVA, BUT STAFF ESTIMATES OF TVA SUBMITTAL DATES CONTINUE TO SLIP 0 CORPORATE PLAN - REVISED IN MARCH 1986

                                     - STAFF COMMENTS - MAY 1986
                                     - PRELIMINARY EVALUATION - JUNE 1986 l                                     - TVA PLANNING A SECOND REVISION 0    COMMISSION EVALUATION OF INTIMIDATION AND HARASSMENT FORWARDED TO TVA'ON JUNE 2, 1986 0    SEQU0YAH RESTART EXPECTED IN JANUARY 1987 EQ INSPECTIONS LARGELY COMPLETE TVA DEVELOPING DESIGN CONTROL VERIFICATION PROGRAM PRELIMINARY INSPECTION IN JUNE 1986 WELDING INSPECTIONS LARGELY COMPLETE REVIEW 0F EMPLOYEE CONCERNS IS ONG0ING 0 WATTS BAR NOT EXPECTED TO BE READY FOR LICENSING BEFORE MAY 1987 l

l 0 BROWNS FERRY RESTART DATES REMAIN UNCERTAIN , O h/W

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NRC 50,5f1(F) LETTER REQUIREMENTS IN CORPORATE AREA O SPECIFY ACTIONS BY TVA BOARD TO REMAIN INFORMED AND INVOLVED IN NUCLEAR PLANT PERFORMANCE 0 DESCRIBE MANAGEMENT CHANGES TO STRENGTHEN REGULATORY PERFORMANCE 0 ESTABLISH CORPORATE CONTROLS TO ASSURE AN INTEGRATED COMMITMENT TRACKING SYSTEM O IMPROVE PROGRAM FOR ESCALATING OA AUDIT FINDINGS I G 9

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REVISED CORPORATE PLAN (MARCH 10, 1986) O . BOARD ACTIONS 0 BOARD TO REMAIN INFORMED OF NUCLEAR PROGRAM ACTIVITIES 0 IMPROVED CORPORATE OVERSIGHT MANAGEMENT CHANGES 0 ESTABLISH MANAGER OF NUCLEAR POWER AS RESPONSIBLE FO NUCLEAR PROGRAM 0 REORGANIZE NUCLEAR PROGRAM TO ESTABLISH STRONG CENTRAL ORGANIZATION WITH COMPLETE AUTHORITY OVER ALL NUCLEAR MATTERS 0 REDUCE LOCAL AUT0NOMY OF INDIVIDUAL SITES 0 ' MAJORINFUSIONOFOUTSIDEMANAGERIALANDTECHNICALTALENT IN LINE POSITIONS OF NEW ORGANIZATION i 0 DEVELOP IN-HOUSE MANAGERIAL TALENT TO EVENTUALLY ASSUM CONTROL OF PROGRAM ) 0 NUCLEAR SAFETY REVIEW STAFF (NSRS) CHANGED TO NUCLEAR MANAGER'S REVIEW GROUP (NMRG) - TO BE USED TO SUPPOR ) 0F NUCLEAR POWER COMMITMENT TRACKING AND DA - 0 INTEGRATED COMMITMENT TRACKING SYSTEM AND PROGRAM FOR ESCAL A NG OL AUDIT FINPINGS PEMAIN CONSISTENT Wl'P OPyr.p " w.:. . g_ 1

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STAFF EVAlllATION OF REVISED CORPORATE PLAN 0 TVA SUBMITTAL PROVIDES'A CONCEPTUAL DESCRIPTION OF CHAN TVA INTENT, BUT LACKS IMPLEMENTATION DETAILS 0 PLAN'S CONCEPT ACCEPTABLE BUT STAFF COULD NOT MAKE FINAL FINDING WITHOUT THE DETAILS OF IMPLEMENTATION 0 RESULTS OF PRELIMINARY STAFF REVIEW FORWARDED TO COMMIS MAY 9, 1986 0 ADDITIONAL INFORMATION FROM TVA REQUESTED MAY 1, 1986 O STAFF CONCERNS AND NEEDS INCLUDED: DETAILS OF STAFFING, INTERFACE PROCEDLIRES, AND OPERABILITY OF 0FFICE OF NilCLEAR POWER REVISED OA TOPICAL REPORT DETAILS OF ENGINEERING ORGANIZATIONAL CHANGES 0 TVA StlBMITTED REVISED QA TOPICAL REPORT ON MAY 1, 1986 AND

                            .- IS        PREPARING RESPONSE TO OTHER STAFF CONCERNS i

0A TOPICAL REPORT UNDER STAFF REVIEW . O STAFF WILL INSPECT IMPLEMENTATION OF PLAN O A -/ 9 8 5

STAlllS OF STAFF EVI&l (CORPORAE PIRD 0 AWAITitE WA stb 11TTAL OF ADDIT 10W. ItFORt%T10tl TO TK CORPORATEPLMI O FWMCBEIT C0flSULTNIT IS EING OBTAllED 0 PRESBIT CONCERNS

         - SPNI 0F IWMGDEff C0ffiROL
         - TPAlfiltE OF fWlAGERS
         - TPRISITION TO WA BPLOYEES O

A-/99

APPENDIX XII TVA BRIEFING TO ACRS, WASHINGTON, DC a JULY 10, 1986 , i O i TVA BRIEFING 1 Tu-l ACRS - i Washington, DC !O JULY 10,1986 . l l l C. C. Vlason (O paco 1 . _ _

TVA'S REVISED CORPORATE NUCLEAR PERFORMANCE PLAN l - l ! o HIRING, DEVELOPMENT AND RETENTION OF EXPERIENCED i NUCLEAR MANAGERS E o RESTRUCTURING OF TVA'S NUCLEAR ORGANIZATION Y o t j IMPROVEMENT IN MANAGEMENT SYSTEMS AND PROGRAMS

              - INCREASING UPPER-MANAGEMENT AWARENESS
              - MANAGEMENT SYSTEMS AND CONTROLS.                         '
              - TIMELY CORRECTIVE ACTIONS o

RESTORING EMPLOYEE CONFIDENCE IN TVA NUCLEAR MANAGEMENT o PROGRAMMATIC IMPROVEMENTS o IMPLEMENTATION OF iHE REVISED PLAN \

i 1 , _0RGANIZATION l j PROBLEM i ! e LACK OF COMMUNICATION AND COORDINATION AMOUNG TVA'S NUCLEAR DEPARTMENTS SOLUTION o

BRING ALL NUCLEAR MATTERS UNDER ONE CONTROL i

e REMOVE NON-NUCLEAR MATTERS l l e CLEAR, SIMPLE LINES OF AUTHORITY / RESPONSIBILITY 6 j e CONSISTENCY ACROSS ENTIRE TVA NUCLEAR EFFORT l l e STRENGTHENING WEAK AREAS SUCH AS QA, ENGINEERING, LICENSING O O O

S O L N F A AI OR E N E V C l CE G NS N ANT L I I R A A O 8 TV P . E R N R 0 1 & EA T CRL C U U R N T 8 - N N N O O D O MIC N U 0 RT l S S T 7C N E N O PI I C R A E N L E G C C S U O N E N A O - R N

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                ' uS                                                       N          N "CSA                                                              GO NT I

R C I E U E R N ST l A QN NE R NO E E C AA G A ER A  ! R N N IllS l8IE l lIlEl18 A E A Q l GM I

   -                                                                            N G
                                                                 .      .GIN O

I T N K C V I A T F N U

         .*      E F I      N           M     D F A                                          A           E     O A T                                          L           T     R S S                                          P           A     P RW                                                       R A E E i                      R GN L v                     A   I C E                      E S U R                     L N N                        C E U C N   I L

R A S E L N C O I I U T N- A R

              .                                                N      E O       PO N

E R C E I I F e F O O W F O P S R N I A O E I T L C R A U E N P O

I I MANATER ' OFFICE OF NUCLEAR POCER S. A. WHITE , DEPUTY MANA3ER C.C. MASON l

                       ,   HUCIEAR S AFETY 81EVIEW S TAf P af.W. Wlili T g

SUPPOni f ur8CilOt43: NUCIEA883CEliS34Q

                                                                                   .tw. ssurHAM MANAGER AS$1SI ANT M AllAGER QUAtIIT AS$UnA!4CE                                       JJt. Dun Al L                                           .

II J. Etts I El llUfJAtt RESOt*1CE OEVELOPMENT M E. T AYLOR 14UCLEA4 $ATEIT staff h AS$tSTANT TO TIIE F.A. SZCZEPANSK8 Q EnaPLOYEE CONCEANS LLPUfYMANAGES E K. SLIGER {  ! R P. DElllSE ' t I MAf(AGER MANAGER PROJECT M ANAGEMENT 840C8EAn ENGN4EEntflG & OttG AlalZAllOtt OPEst A TIOtis CONSIRUCitON seNelonie C.C. MASON [V A C Alli) J. l'. D ARLIHQ Watte ses 18 alt 3 5 W. a. an O wN I I MANAOEA MANAGES ENGuiEERING ColisinUCitosa

m. W. C ANIRELI C. OOt1HJE I I ASSISTAlif S TO gg A r4 AGE R '

g's g$ NUCLEAll SERVICES L. L. J A Cit SON ( A stlag) y i L. E. M ARilla i t J.18UtION . I l 1 i

                           !                                                        I                                                        i saOwHS FERRY                           SEQUOVAH              WATIS SAS S                  ( Ac ting) att sana ILt AsERCa                W. T. CO T TLE

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MANAGEMENT PERSONNEL '

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PROBLEM i i e LACK OF LEADERSHIP AND DIRECTION > r SOLUTION l i o . e  ! O GO "OUTSIDE" FOR THESE PEOPLE (CONTRACTOR) !9i i i e 't USE THEM ON A TEi.1PORARY BASIS (TWO YEARS) I

• CONTINUE SEARCH FOR TALENT ,

e CONTINUE SEARCH OUTSIDE FOR TALENT WILLING TO HIRE IN AS A TVA EMPLOYEE , o CONTRACTOR MANAGERS WILL BE TVA LINE MANAGERS, WITH A l  ! PERMANENT TVA " DEPUTY" TO TRAIN  ; i 9 O 9

NUCLEAR OFFICE i i QUALITY ASSURANCE oF NUCLEAR POWER NUCLEAR OFFICE OF QU ALIT Y ENGINEERING I ASSURANCE l DEPUTY ENGINEERING ASSURANCE l PROCUREMENT Q U ALIT Y TEC HNIC AL AND g y ALgr y SUPPORT SYSTEMS s EVALUATION ASSURANCE SITE SITE SITE SITE Q U ALIT Y Q U ALITY Q U ALIT Y Q U ALIT Y ASSURANCE ASSURANCE ASSURANCE ASSURANCE SEQ UOYAH OffR Y WATTS S AR SELLEFONTE 9 . G O

C _ S O . E t P N - _ t R S _ E O . _ N o F JE UR _ t O s C C e

e. T T v\

T B e E O o se T e _ m p _ er e tN y C r W O p e A P ce e s or T RS A T OT N t S J R A I s E U E 0, G l S CC C E e A TT T R e S R a M o es n , a g e N t U of N l C u s L cl e W e O E a r SID A C T RF T AC 6 e s R e SNCTA A m e S S E N A C t r u A R HIO N R V E G I O cI S I C M N ol n E S E N S T T R U e s C S O T I E S D Q R F I O U A C ON I N T C A T A HI O M N S f C O ec N N D S L PU O E T C P R OC FF U U E L WLE'C I T E S t V C A RA E e T R R 1 4 O I 0 Se D O N W RF D N AC S es A F CT A ER MIO c P D e e R s S E s s Y E R ee A RS S S Vas OST R I 1 S Cet A T Es E S L T I A V N E T T S P P O

• T R R N O .

A N A C s C O CU t DN E aE L _ AD DU r D E R y U A D R R R S E E - S S S E F _ R I V E I L I C D E . S M C A M NO N A _ A T G TR E - E R s A a - e C A E T L N T S S Nk Q@

OFFICE NUCLEAR NUCLEAR POWER SAFETY AND LICENSING NUCLEAR SAFETY & LICENSING DEPUTY k ASSISTANT TO g LICENSING  % SUPPORT GENERIC SI M NUCLEAR SAFETY LIC ENSING MANAGERS (4) REGULATORY COMPLI ANC E REVIEW

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OFFIC E NUCLEAR ENGINEERING NUCLE AR POWER NUCLEAR ENGINEERikG t h N NUCLEAR Q U ALITY ASSISTANT TO k l ASSURANCE T I I I MANAGEMENT ENGINEERING ENGINEERING & PROJECT OPER ATIONS SYSTEMS TEC HNIC AL ASSURANCE ENGINEERING SUPPORT SERVICES SERVIC ES 9 9 9

MANAGEMENT CONTROL SYSTEM IMPROVEMENTS e INTEGRATED SYSTEM OF CORPORATE PROCEDURES

              -ALL NUCLEAR PROCEDURES REVIEWED AND REVISED AS NECESSARY TO BE CONSISTENT

• PREPARATION OF NEW POSITION DESCRIPTIONS FOR MANAGERS

              -STRESS ACCOUNTABILITY AND RESPONSIBILITY                       N
              -USED TO EVALUATE MANAGERS PERFORMANCE e IMPROVEMENTS IN PLANNING AND INTEGRATION OF NUCLEAR ACTIVITIES e INTEGRATED COMMITMENT TRACKING SYSTEM e REVISED SYSTEM TO ENSURE TIMELY AND EFFECTIVE CORRECTIVE ACTION FOR CONDITIONS ADVERSE TO QUALITY e EXPANDED OPERATING EXPERIENCE REVIEW PROGRAM e INCREASED UPPER MANAGEMENT AWARENESS OF NUCLEAR ACTIVITIES e                                 O 9

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NRR STAFF PRESENTATION TO THE ACRS C) APPENDIX XIII STAFF PRESENTATION ON STANDARDIZATIO!

SUBJECT:

STANDARDIZATION DATE: JULY 9, 1986 PRESENTER: DIN 0 C. SCALETTI PRESENTER'S TITLE /BR ANCH/DIV: PROJECT MANAGER / SSPD / DPWRL-B PRESENTER'S NRC TEL. NO.: 492- 8208 SUBCOMMITTEE: IMPROVED LWR DESIGNS k-A/6~

O STANDARDIZATION -- DISCUSSION OUTLI-NE BACKGROUND NEED FOR REVISION 1978 STANDARDIZATION f POLICY STATEMENT ,

I

      '                 CHRONOLOGY OF RECENT POLICY DEVELOPMENT i

GOALS OF PROPOSED STANDARDIZATION POLICY PRELIMINARY NUREG OUTLINE CONFORMING CHANGES TO REGULATIONS PROPOSED LEGISLATION k -Alb

O 1 l BACKGROUND APRIL 1972 INITIAL POLICY STATEMENT ISSUED !-f

• MARCH 1973 REFERENCE SYSTEM, DUPLICATE PLANT AND MANUFACTURING LICENSE CONCEPTS ANNOUNCED O* AUGUST 1974 REPLICATE PLANT CONCEPT ANNOUNCED JULY 1977 STATEMENT REAFFIRMING SUPPORT OF STANDARDI-ZATION, AND REQUESTING COMMENTS AND SUGGESTIONS ON CHANGES ISSUED AUGUST 1978 MOST RECENT POLICY STATEMENT ISSUED A - 2 /f

;                                            NEED FOR REVISING 1978 STANDARDIZATION               ;

POLICY STATEMENT. t I l STAFF'S EXPERIENCE IN IMPLEMENTING STANDARDIZATION PROGRAM . .- t.

d PROVISIONS OF SEVERE ACCIDENT POLICY STATEMENT i

i PROVISIONS OF DRAFT NUCLEAR POWER PLANT LICENSING AND STANDARDIZATION ACT l l k -d?/f

         --ww-e w m,

O CHRONOLOGY OF RECENT POLICY DEVELOPMENT FEBRUARY / APRIL 1985 - SRM REQUESTED REVISIONS TO 1978 POLICY BE PREPARED FOR COMMISSION CONSIDERATION NOVEMBER 1985 - AIF PROPOSAL TO STAFF

    .(

DECEMBER 1985 - COMMISSION BRIEFING MARCH 1985 - INDUSTRY VIEWS SENT TO COMMISSION NUMARC EEI EPRI APRIL 1986 - SRM WITH POLICY GUIDANCE l MAY 1986 - REVISED' DRAFT POLICY TO COMMISSION l l l l

1 l G0ALS OF PROPOSED STANDARDIZATION POLICY ESSENTIALLY COMPLETE PLANT 1 ESSENTIALLY FINAL DESIGN DETAIL

I i

REFERENCE SYSTEM DESIGN CERTIFICATION i l l t l t I k o1o? O L

       - - _ _ _ _   _     _ _ _ . - - - . . . _ , . - . _ . .                       , _ , . . _ _ _ , . - , _ .                  _ . . .   ..w m _-_    ,,-_--w_w,,.,mm

PRELIMINARY NUREG OUTLINE ' BACKGROUND AND PURPOSE OF NUREG STANDARDIZATION POLICY STATEMENT DESIGN CERTIFICATION CONCEPT

    . TRANSITION OPTIONS k

.f COMPLETENESS OF DESIGN SCOPE AND DETAIL CHANGES TO APPROVED STANDARD DESIGNS LICENSE FEES RELATED POLICIES AND REGULATIONS DESIGN CERTIFICATION RULEMAKING OPTIONS RENEWALS 0 A -AA /

f CONFORMING CHANGES TO REGULATIONS. a s 10 CFR PART 50 f APPENDIX M - MANUFACTURING LICENSE CONCEPT II

                                                                                       ~

APPENDIX N - DUPLICATE PLANT CONCEPT l APPENDIX 0 - REFERENCE SYSTEM CONCEPT 10 CFR 50.34(F) - CP/ML RULE 1 i SCHEDULE - TO COMMISSION APPROXIMATELY 90 DAYS FOLLOWING APPROVED STANDARDIZATION POLICY l l l O ~-'

O HIGHLIGHTS OF PROPOSED LEGISLATION CERTIFIED STANDARDIZED DESIGNS (10 YEAR APPROVALS) PREAPPROVED SITES (10 YEAR APPROVALS) , f i f

• ONE-STEP LICENSING

                                                                                                                                                                                 ~

()

• STANDARDS FOR BACKFITTING TO LICENSES, DESIGN

) APPROVALS AND SITE APPROVALS ALLOCATES FEES TO USERS I I

• PROVIDES CONGRESSIONAL ENDORSEMENT FOR STABILITY OF THE LICENSING PROCESS i

A-Da3 i O f { i

APPENDIX XIV STATEMENT OF F. SEARS, ATOMIC INDUSTRIAL FORUM STUDY GROUP V STATEMENT OF C. FREDERICK SEARS ON NUCLEAR POWER PLANT STANDARDIZATION BEFORE THE ADVISORY COMMITTEE ON REACTOR SAFETY ON j JULY 10, 1986 i

    $ l. INTRODUCTION G

MY NAME IS C. FREDERICK SEARS, I AM VICE PRESIDENT, NUCLEAR AND ENVIRONMENTAL ENGINEERING AT NORTHEAST UTILITIES. I APPRECIATE YOUR INVITATION TO APPEAR BEFORE THIS COMMITTEE. I AM SPEAKING TODAY ON BEHALF OF THE ATOMIC INDUSTRIAL FORUM'S (AIF) STUDY GROUP ON THE PRACTICAL APPLICATION OF ] STANDARDIZED NUCLEAR POWER PLANTS IN THE UNITED STATES. WITH THE U.S. ECONOMY EXPANDING AND DEMAND FOR ELECTRICITY l ON THE RISE, MANY UTILITIES ARE BEGINNING TO PLAN FOR NEW i GENERATING CAPACITY THAT WILL BE NEEDED IN THE 1990'S AND LATER. UNFORTUNATELY, NUCLEAR POWER IS CURRENTLY PERCEIVED BY THE UTILITY INDUSTRY AND THE INVESTMENT COMMUNITY AS A RISKY INVESTMENT. IT IS VERY EXPENSIVE AND REQUIRES TOO LONG TO LICENSE AND CONSTRUCT A PLANT. IN THIS COUNTRY, IT IS GENERALLY NO LONGER CONSIDERED AN ECONOMIC GENERATION ALTERNATIVE BECAUSE OF THESE PROBLEMS. O k'kdY l

MY STATEMENT TODAY WILL FOCUS ON PROPOSED REVISIONS TO THE CURRENT LICENSING PROCESS WHICH WE BELIEVE WILL AID IN CORRECTING THIS SITUATION. IN ADDITION, THESE PROPOSED REVISIONS SHOULD ENHANCE flUCLEAR PLANT SAFETY. SPECIFICALLY, I WILL ADDRESS A NUCLEAR PLANT DESIGN CERTIFICATION (DC) PROCESS AND A COMBINED CONSTRUCTION AND OPERATING LICENSE (COL). IN DOING THIS, I WILL ALSO DESCRIBE BRIEFLY THE INFORMATION THAT THE INDUSTRY BELIEVES WOULD ADEQUATELY SUPPORT AN APPLICATION FOR A DESIGN CERTIFICATION. THIS IS A FRESH APPROACH TO AN ISSUE THAT ALL OF US, -- THE UTILITY INDUSTRY, VENDORS, THE NRC, THE PUBLIC AND THE CONGRESS -- HAVE BEEN CONCERNED WITH FOR YEARS. 11 THE DESIGN CERTIFICATION q THE FIRST ELEMENT OF OUR PROPOSAL IS THE DESIGN

   ,          CERTIFICATION (DC) PROCESS. IN OUR VIEW, DESIGNS FOR         ~

STANDARD PLANTS WOULD BE DEVELOPED BY A CONSORTIUM MADE UP OF UTILITIES, NSSS VENDORS, TURBINE GENERATOR VENDORS, A/ES, I CONSTRUCTORS, ETC. NECESSARY INFORMATION TO CERTIFY THE DESIGN WOULD BE DOCUMENTED IN A PLANT SAFETY REPORT (PSR). THE REPORT WOULD BE SUBMITTED TO THE NRC FOR REVIEW AND APPROVAL. APPROVAL OF THE DESIGN WOULD CULMINATE WITH THE ISSUANCE OF A DESIGN CERTIFICATION WHICH WOULD BE VALID FOR l 10 YEARS WITH OPTIONS FOR RENEWAL. THE OBJECTIVE OF THE DESIGN CERTIFICATION IS TO DEVELOP A DETAILED PLANT DESIGN WHICH COMPLETELY SATISFIES REGULATORY REQUIREMENTS BEFORE CONSTRUCTION BEGINS, THEN LICENSE IT AND FREEZE IT! A -a?a f 1_ _- _ . . _ - _ _ -_ -.

l IN ORDER FOR A DESIGN CERTIFICATION (DC) TO BE ISSUED, THE . DESIGN MUST BE SUFFICIENTLY FINALIZED SO THAT THERE IS A O' CLEAR DEFINITION OF ALL RELEVANT SAFETY ASPECTS. THE NRC REVIEW OF THE PLANT SAFETY REPORT (PSR) WOULD CONFIRM THAT SUBSEQUENT CONSTRUCTION, INSPECTION, TESTING AND OPERATION CAN ALL BE PERFORMED IN ACCORDANCE WITH PRE-APPROVED METHODOLOGIES. IHE PLANT SAFETY REPORT (PSR) WOULD CONTAIN THE FOLLOWING INFORMATION: DESIGN BASIS CRITERIA ANALYSIS AND DESIGN METHODS FUNCTIONAL DESIGN OF SYSTEMS PHYSICAL ARRANGEMENT OF AUXILIARY, BALANCE OF PLANTS AND NUCLEAR STEAM SUPPLY SYSTEMS AND COMPONENTS ACCEPTANCE AND STARTUP IESTING REQUIREMENTS PROBABILISTIC RISK ASSESSMENT METHODOLOGY

    ,- THE NRC COMMISSION AND THE APPLICANT MUST AGREE IN ADVANCE

( ON THE DEPTH OF DESIGN DETAILS NECESSARY TO LATER CONFIRM THAT THE PLANT MEETS SPECIFICATIONS. THIS CONFIRMATION WOULD BE ACHIEVED THROUGH INSPECTIONS, TESTS, ANALYSES AND ACCEPTANCE CRITERIA WHICH WOULD BE AGREED TO BEFORE HAND. THE DESIGN DETAIL MUST ALSO SUPPORT THE PREPARATION OF EQUIPMENT PROCUREMENT SPECIFICATIONS AND QUALIFICATION, INSTALLATION AND TESTING PROCEDURES. THE DESIGN CERTIFICATION (DC) PROCESS IS DEFINITELY NOT

         " BUSINESS AS USUAL".            IT REPRESENTS A SIGNIFICANTLY DIFFERENT PHILOSOPHY OF DESIGNING, CONSTRUCTING AND REGULATING NUCLEAR POWER PLANTS.                IN THE PAST, PRELIMINARY DESIGN INFORMATION WAS FURNISHED TO THE NRC AT THE CONSTRUCTION PERMIT STAGE. LATER, DETAILED DESIGNS AND REGULATORY REQUIREMENTS WERE FINALIZED IN PARALLEL WITH PLANT CONSTRUCTION.             THIS TWO-STEP PROCESS PERMITTED BOTH THE PLANT OWNER AND THE NRC TO MAKE SIGNIFICANT AND OFTEN COSTLY PLANT CHANGES DURING THE OPERATING llCENSE STAGE.                  IN THE PROPOSED DESIGN CERTIFICATION (DC) PROCESS, THE DESIGN A-M4                                       ,

i i  ! l

_q_ f-'s ENGINEERING IS ESSENTIALLY COMPLETE, THE REGULATORY k) REQUIREMENTS HAVE BEEN THOROUGHLY DEFINED AND INCORPORATED AND PUBLIC INPUT HAS BEEN CONSIDERED BEFORE CONSTRUCTION OF THE PLANT BEGINS. THE " ESSENTIALLY COMPLETE" DESIGN DEVELOPED TO SUPPORT THE DESIGN CERTIFICATION PROCESS WOULD INCLUDE THE FOLLOWING DOCUMENTS, DRAWINGS AND INFORMATION AS APPROPRIATE: DESIGN BASIS CRITERIA PLANT GENERAL ARRANGEMENTS OF STRUCTURES AND COMPONENTS PROCESS AND INSTRUMENTATION DIAGRAMS CONTROL LOGIC DIAGRAMS SYSTEM FUNCTIONAL DESCRIPTIONS COMPONENT AND PROCUREMENT SPECIFICATIONS INCLUDING ACCEPTANCE IEST REQUIREMENTS CONSTRUCTION AND INSTALLATION SPECIFICATIONS DA PROGRAM (s

   \_,)                 -

EMERGENCY PLAN SECURITY PLAN ALARA/ RADIATION PROTECTION PLAN ACCIDENT ANALYSES DRAFT TECHNICAL SPECIFICATIONS RISK ANALYSIS IT IS IMPORTANT TO NOTE THAT THE ABOVE DOCUMENTS ARE THOSE WHICH WILL CONTROL THE ACTUAL CONSTRUCTION AND OPERATION OF THE PLANT. ANY ADDITIONAL DOCUMENTATION REQUIRED TO SUPPORT THE NRC SAFETY EVALUATION WOULD BE PROVIDED WITH THE PLANT SAFETY REPORT (PSR) AS PART OF THE DESIGN CERTIFICATION (DC) APPLICATION. AS I HAVE SAID, THE PLANT DESIGN WOULD BE ESSENTIALLY COMPLETE BEFORE SUBMITTAL OF THE PLANT SAFETY REPORT (PSR) TO THE NRC. THE ONLY PLANT DESIGN ACTIVITY REMAINING WOULD O BE ASSOCIATED WITH SITE SPECIFIC FEATURES AND INCORPORATING EQUIPMENT NAMEPLATE INFORMATION. A - any? i

()

   ,           THE EFFORT TO SUPPORT A DESIGN CERTIFICATION (DC)

APPLICATION WILL RESULT IN A LARGE EXPENDITURE OF MANPOWER AND FUNDS. IT IS ESTIMATED THAT THIS EFFORT WOULD COST BETWEEN $150 AND $200 MILLION. Ill. COMBINED CONSTRUCTION AND OPERATING LICENSE (COL) THE KEYSTONE TO THE EFFECTIVENESS OF THE DESIGN CERTIFICATION PROCESS IS ITS LINKAGE TO A COMBINED CONSTRUCTION AND OPERATING llCENSE. A DESIGN CERTIFICATION ALONE WILL NOT REDUCE THE INVEETHENT RISK SUFFICIENTLY TO PERMIT NUCLEAR POWER TO BE COMPETITIVE WITH OTHER AVAILABLE GENERATION ALTERNATIVES. IN FACT, IT IS DOUBTFUL IF ANY UTILITY WILL ACCEPT THE RISK OF COMMITTING TO A NUCLEAR PLANT WITHOUT A HIGH DEGREE OF ASSURANCE THAT IT CAN BE CONSTRUCTED ON SCHEDULE AND WITHIN BUDGET AND THAT IT WILL BE PERMITTED TO OPERATE WHEN COMPLETED. IN THE CONSTRUCTION AND OPERATING LICENSE (COL) PROCESS, THERE WILL BE AN OPPORTUNITY FOR FURTHER PUBLIC PARTICIPATION ON SITE SPECIFIC CONSIDERATIONS CONSISTENT WITH THE INTENT OF THE CODE OF FEDERAL REGULATIONS. ANY DISCUSSIONS OF THE ADEQUACY OF THE PLANT DESIGN WILL HAVE ALREADY OCCURRED AT THE TIME OF THE DESIGN CERTIFICATION APPLICATION REVIEW AND APPROVAL. ONLY SITE SPECIFIC ISSUES WILL BE CONSIDERED AT THE CONSTRUCTION AND OPERATING LICENSE (COL) STAGE. 1 I AFTER THE CONSTRUCTION AND OPERATING l! CENSE (COL) HAS BEEN l ISSUED, AND CONSTRUCTION BEGUN, CONFIRMATORY AUDITS WILL BE CONDUCTED BY BOTH THE NRC AND THE OWNER. THESE AUDITS WILL i ENSURE THAT THE PLANT IS BEING CONSTRUCTED IN ACCORDANCE WITH PRE-APPROVED REQUIREMENTS AND THAT THE OWNER IS SATISFACTORILY PERFORMING THE AGREED UPON TESTS AND

   '"             INSPECTIONS. UPON COMPLETION OF CONSTRUCTION, THE NRC WOULD FINALIZE THE AUDIT PROCESS AND THE OWNER WOULD BEGIN PLANT OPERATION.

Md

THE DESIGN CERTIFICATION / CONSTRUCTION AND OPERATING llCENSE (s) (DC/ COL) PROCESS WILL RESULT IN SIGNIFICANT PROGRESS TOWARDS CONSISTENT PERFORMANCE REGULATION AND AWAY FROM VARIABLE REGULATION WHICH CHARACTERIZES THE EXISTING PROCESS. IV. BENEFITS THE PRIMARY BENEFIT OF THE DESIGN CERTIFICATION / CONSTRUCTION AND OPERATING LICENSE (DC/ COL) PROCESS, IF IT IS SUCCESSFUL, IS THAT NUCLEAR POWER WILL AGAIN BE COMPETITIVE WITH OTHER GENERATION ALTERNATIVES. WE BELIEVE THE PROCESS WILL BE SUCCESSFUL BECAUSE A STABLE AND EFFICIENT LICENSING PROCESS HAS BEEN SHOWN TO PERMIT A SIGNIFICANT REDUCTION IN THE REAL COST OF NUCLEAR POWER. THE REDUCTIONS IN COST OCCUR FROM TWO SIGNIFICANT FACTORS:

1) IHE COST OF BORROWED FUNDS WILL BE REDUCED BECAUSE INVESTORS AND LENDERS WILL PERCEIVE THAT ONCE A DESIGN 7~

( ,, CERTIFICATION (DC) AND A CONSTRUCTION AND OPERATING LICENSE (COL) HAVE BEEN OBTAINED, PROJECT PROGRESS WILL BE PREDICTABLE AND INVESTMENT RISK SUBSTANTIALLY REDUCED. j

2) THE TIME FROM START OF CONSTRUCTION TO PLANT OPERATION WILL BE SIGNIFICANTLY REDUCED. THIS WILL OCCUR BECAUSE THE DESIGN WILL BE ESSENTIALLY COMPLETE AND REGULATORY I REQUIREMENTS DEFINED AND STABILIZED PRIOR TO INITIATION OF CONSTRUCTION. IHE RESULT WILL BE REDUCED REWORK, OPTIMlZED DESIGNS, AND IMPROVED CONSTRUCTION MANAGEMENT.

A REDUCTION IN CONSTRUCTION TIME WILL SIGNIFICANTLY REDUCE ESCALATION COSTS AND FINANCE CHARGES FOR THE PROJECT. IHE PROPOSED DESIGN CERTIFICATION / CONSTRUCTION AND OPERATING LICENSE (DC/ COL) PROCESS WILL ALSO ENHANCE PUBLIC PARTICIPATION IN THE LICENSING PROCESS. DURING THE DESIGN CERTIFICATION (DC) PROCEEDINGS, THE PUBLIC WILL HAVE THE

                                                       ~A2f

                                      -7 OPPORTUNITY TO EXPRESS THEIR CONCERNS ABOUT THE ADEQUACY OF

[ THE PLANT DESIGN. THIS WILL PERMIT THEM TO EXPLORE TECHNICAL ISSUES TO A GREATER DEPTH THAN IS NOW POSSIBLE AT THE CONSTRUCTION PERMIT STAGE BECAUSE THE DESIGN WILL BE ESSENTIALLY COMPLETE. AT THE CONSTRUCTION AND' OPERATING LICENSE (COL) STAGE, THE PUBLIC CAN PARTICIPATE IN SITE RELATED ISSUES. THE OVERALL PROCESS ALLOWS THE PUBLIC TO PROVIDE INPUT WHEN IT IS MOST MEANINGFUL, THAT IS, BEFORE CONSTRUCTION IS AUTHORIZED. THE DESIGN CERTIFICATION / CONSTRUCTION AND OPERATING LICENSE (DC/ COL) PROCESS ALSO ENHANCES SAFETY AND BETTER UTILIZES INDUSTRY AND REGULATORY PERSONNEL. IT WILL ALLOW THE NRC, NSSS VENDORS, ARCHITECT-ENGINEERS AND UTILITIES TO CONCENTRATE ON KEY ISSUES AND FEWER DESIGNS. IT WILL REDUCE THE DEMANDS ON THESE TECHNICAL RESOURCES CAUSED BY EVER CHANGING CRITERIA AND DESIGNS. 1 IHE ADDED STABILITY OF THE ENGINEERING, CONSTRUCTION, AND REGULATORY PROCESSES AND THE RESULTING SHORTENED CONSTRUCTION PERIOD WILL ENABLE UTILITIES TO PLAN NEW PLANTS WITH MORE CONFIDENCE. UTILITIES WILL BE ABLE TO MORE ACCURATELY PREDICT FINAL COSTS AND TO EFFECTIVELY MATCH CAPACITY ADDITIONS WITH GROWING CUSTOMER DEMAND. l V. 10MMARY IN

SUMMARY

, WE WOULD EMPHASIZE THAT NUCLEAR POWER IS GENERALLY NOT A COMPETITIVE GENERATION ALTERNATIVE FOR FUTURE PLANTS UNDER THE CURRENT LICENSING PROCESS. IHE INDUSTRY HAS PROPOSED LICENSING REFORM WHICH UTILIZES THE CONCEPTS OF A DESIGN CERTIFICATION (DC) AND A CONSTRUCTION AND OPERATING LICENSE (COL). WE HAVE ELABORATED ON THE EXTENSIVE DESIGN INFORMATION WHICH WILL BE COMPLETED AND AVAILABLE TO THE NRC AT THE TIME OF THE DESIGN CERTIFICATION (DC) APPLICATION AND WE HAVE DESCRIBED HOW THE PROCESS WOULD WORK. g l . - . _ _ _ _ - _ _. - _

THE DESIGN CERTIFICATION / CONSTRUCTION AND OPERATING LICENSE (DC/ COL) PROCESS IS INTENDED TO PROVIDE A MAJOR STEP TOWARD REVIVING NUCLEAR POWER AS A COMPETITIVE GENERATION ALTERNATIVE FOR THIS COUNTRY. THIS WILL OCCUR BY REDUCING THE COSTS ASSOCIATED WITH NUCLEAR POWER AND RESTORING INVESTOR CONFIDENCE IN OUR ABILITY TO DESIGN, REGULATE, CONSTRUCT AND OPERATE THESE FACILITIES SAFELY, RESPONSIBLY, AND PREDICTABLY. WE URGE YOUR COMMITTEE TO TAKE AN ACTIVE AND AGGRESSIVE ROLE IN BRINGING ABOUT MEANINGFUL AND TIMELY LICENSING REFORM SUCH AS DISCUSSED HERE TODAY. IHANK YOU FOR THE OPPORTUNITY TO APPEAR BEFORE YOU. 1 i 1 A .R3/ i

l O O O i Nuclear As a Generation Alternative i

• Utilities are planning new capacity 1
• Nuclear Power not attractive alternative i - Risky investment l  % - Expensive b -Time to license and construct a plant is too long g

e

• Contributory to this situation is current licensing process

          - Unstable
          - Unpredictable                                             -

l

• Revisions to licensing process required

1 O O O The Design Certification Process 4 i

• Consortium Designs Standard Plant L .

l

• Submits Plant Safety Report (PSR) to NRC i

l

• NRC issues a Design Certification

                                      - 10 years + renewal options k              - Design " frozen" l                       b g

• PSR will address l

                                      - Design Basis Criteria l
                                      - Analysis and Design Methods l
                                      - Functional Design and Physical Arrangements of Systems '
                                      - Plant Physical Arrangement
                                       - Acceptance Test / Requirements
                                       - PRA Methodology

• Name Plate information not available

o o o l l Design Engineering to Support l Design Certification

• Design esentially complete on front end

! A * " Essentially" complete design will encompass

!        i k              - Design Basis Criteria w

4.

                        - Plant General Arrangements of Structures and Components
                        - Process and instrumentation Diagrams                                                    .
                        - Control Logic Diagrams
                        - System Functional Descriptions
                        - Component and Procurement Specifications including Acceptance Test Requirements

!    O                                O                               O Design Engineering to Support 1     ,

l ! Design Certification (contw.) j * " Essentially complete design will encompass (cont'd.) ! - Construction and Installation Specifications g - QA Program f i i t - Emergency Plans tu q - Security Plan l

           - ALARA/ Radiation Protection Plan
           - Accident Analysis
           - Draft Technical Specifications
           - Risk Analysis

• Will Not encompass

              " Name Plate"information incorporation
            - Site specific verification

l O O O The Construction and Operating i License Process ' 1  ! !l

  ~

• Combined construction permit and operating license (COL)

I l a Will make Design Certification (DC) most effective ! 4 - DC alone will not sufficiently reduce risks i

b - COL will provide high degree of assurance completed plant g will be allowed to operate l

I

• Public participation on site specific issues at COL stage

                                                                         ~

• After COL, NRC will perform confirmatory audits against pre-approved critera
• Move toward performance based regulations

O O O Benefits of DC/ COL

• Cost of nuclear power will be lowered i

! - Cost of borrowed funds lower

• Less risk i
• More stable process i A l i - Total finance charges lower

l. b '

i g

• Essentially complete design before construction l
• Stable regulatory requirements before construction j
• Shorter construction time
• Rework minimized 1
• Design optimized 4
• Construction management enhanced 1020-6 on 330

o O O t l Benefits of DC/ COL (cont'd.) j

• Public Participation enhanced

               - Will occur before construction 4
                - Better information available A

• Safety Enhanced and Industry / Regulatory '

i ! b Resources Better Utilized

to N - Concentration on fewer designs / key issues Frozen" criteria

• Better Utility Planning

                   - More predictable costs
                    - Shorter construction time l

                                                                                                                                     = ........... ......

APPENDIX XV EPRI ALWR PROGRAM PRESENTATION 1 , EPRI ALWR PROGRAM PRESENTATION i i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS i i O JULY 10, 1986 i < l l K. E. STAHLK0PF I D. M. NOBLE I , I l O naar  ; k- - - - - - - _ - - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _

e EPRI ALWR PROGRAM PRESENTATION i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS JULY 10, 1986 INTRODUCTORY REMARKS UTILITY /EPRI OBJECTIVES FOR ALWR DEVELOPMENT EPRI ALWR PROGRAM TIES WITH FOREIGN ENTITIES 4 SIMPLIFICATION / STANDARDIZATION / CERTIFICATION ALWR REQUIREMENTS DOCUMENT MAJOR REQUIREMENTS AND PRIORITIES STATUS SAFETY AND LICENSING ISSUE RESOLUTION i i ALWR SMALL PLANT CONCEPTUAL DESIGN EVOLUTION ECONOMY OF SCALE / UTILITY INTEREST MAJOR GOALS AND PRIORITIES STATUS

                                      ~

0 3896SM6-3 i

O

                .EPRI ADVANCED LIGHT WATER REACTOR PROGRAM THE EPRI PROGRAM STRUCTURE THE PROGRAM CONSISTS OF THREE SEPARATE BUT RELATED MAJOR ELEMENTS:

UTILITY REQUIREMENTS DOCUMENT THE DEVELOPMENT OF A COMPREHENSIVE SET OF DESIGN REQUIREMENTS FOR THE ALWR REGULATORY STABILIZATION COOPERATIVE EFFORT WITH NRC TO IDENTIFY AND RESOLVE ALL OUTSTANDING ISSUES OF NUCLEAR PLANT SAFETY SMALL PLANT DESIGN DEVELOPMENT THE INVESTIGATION AND DEVELOPMENT OF SMALL (<600 MWE) NUCLEAR PLANT OPTIONS O 39915M6-5

O EPRI ADVANCED LIGHT WATER REACTOR PROGRAM PARTICIPANTS... THE EPRI PROGRAM DRAWS UPON A BROAD BASE OF EXPERT PARTICIPANTS, INCLUDING: UTILITY INDUSTRY EPRI MEMBER UTILITIES REPRESENTED BY A UTILITY STEERING COMMITTEE OF SENIOR UTILITY PROFESSIONALS CONTRACTORS O - TEAMS OF NSSS SUPPLIERS, ARCHITECT ENGINEERING FORMS, AND MAJOR UTILITY COMPANIES, INCLUDING: COMBUSTION ENGINEERING / DUKE POWER GENERAL ELECTRIC /BECHTEL WESTINGHOUSE /SARGENT & LUNDY/ COMMONWEALTH EDISON STONE & WEBSTER / YANKEE ATOMIC DIRECT ASSISTANCE TO EPRI FROM MPR ASSOCIATES S. LEVY, INC. i O ^*" 3991SM6-6

O EPRI ALWR PROGRAM CREDIBLE SAFE STANDARD PLANT EPRI ALWR PROGRAM IS FULLY SUPPORTIVE OF INDUSTRY EFFORTS TO DEVELOP A STANDARDIZED NUCLEAR PLANT DESIGN THE EPRI ALWR PROGRAM, LINKED TO PLANT STANDARDIZATION, CAN PLAN A SIGNIFICANT PART IN ITS SUCCESS. IN PARTICULAR: THE RESOLUTION OF REGULATORY ISSUES, VIA THE EPRI REGULATORY STABILIZATION PROGRAM, CREATES A FAVORABLE CLIMATE FOR STANDARDIZED PLANT LICENSING THE ALWR REQUIREMENTS DOCUMENT SERVES AS A COMPREHENSIVE TECHNICAL BASIS FOR STANDARD PLANT DESIGN THE EPRI ALWR PROGRAM IS COORDINATED WITH DOE'S PROGRAM TO SECURE NRC CERTIFICATION OF AN ADVANCED BWR AND PWR DESIGN. 1 O l 3991SM6-8

aaJh-_ A m.4 4 m 4h -au4 A Je.J-MAm-...-Es w# -s-- aum--+-aL*E.J4-Eus - wh.a.4- M-4 -....e,+.4-u- *m-M.eim4 2 +-. -aC---e-4-.- m ---------+- - - - sA --Aa--- - A-ha l 4 4 i t i i 4 I l i r i - 4 i 1 l l i I l l 1 i 1 5 i .i I 4 p p i I 6

FOREIGN PARTICIPATION h JAPAN MOU IN FINAL STAGES OF NEGOTIATION WITH JAPAN ATOMIC POWER COMPANY (JAPC) FOR COOPERATION ON DEVELOPMENT OF (1600MWE) BWR, EPRI WILL RECEIVE RESULTS OF $1 MILLION 1986 WORK FUNDED BY JAPC AND PERFORMED BY GE. JAPC WILL RECEIVE RESULTS OF EPRI PHASE I STUDIES ON SMALL BWR. RESULTS OF EPRI PHASE Il STUDIES THROUGH THE END OF 1986. RESULTS OF RELATED DOE INFORMATION ON SMALL 9 BWR DEVELOPMENT THROUGH THE END OF 1986. DEPENDING ON THE STATUS OF THE WORK AT THE END OF 1986, BOTH PARTIES WILL CONSIDER EXTENSION OF THE AGREEMENT TO 1987 AND BEYOND. 3898SM6-4

JAPAN EPRI HAS RECEIVED A LETTER FROM KANSAI POWER EXPRESSING INTEREST IN PARTICIPATING IN EPRI'S ALWR PROGRAM. EANSAI IS MOST INTERESTED IN DESIGN SIMPLIFICATION OF PWR'S. EPRI HAS RESPONDED POSITIVELY TO KANSAI'S INTEREST IN THE PROGRAM DISCUSSIONS ARE UNDERWAY TO SOLIDIFY THEIR PARTICIPATION O l O """ "*+'

FOREIGN PARTICIPATION TAIWAN MOU SIGNED WITH IAIWAN POWER COMPANY SUBJECT TO APPROVAi. OF REPUBLIC 0F CHINA GOVERNMENT.

            $225,000 PER YEAR FOR FOUR YEARS PAID TO EPRI.

PARTICIPATION IN UTILITY REQUIREMENTS DOCUMENT ONLY. STATION TWO ENGINEERS AT PALO ALTO. OPPORTUNITY TO PARTICIPATE ON STEERING COMMITTEE. 38985M6-3

4 b FOREIGN PARTICIPATION t KOREA i - A DRAFT MOV HAS BEEN EXCHANGED WITH KOREA ELECTRIC POWER COMPANY. l 1 -

                                   $150,000 PER YEAR FOR FOUR YEARS PAID TO EPRI.

PARTICIPATION IN PWR PORTION OF UTILITY REQUIREMENTS DOCUMENT ONLY. i 4 STATION TWO ENGINEERS AT PALO ALTO. J O OPPORTUNITY TO PARTICIPATE ON STEERING COMMITTEE. i i i 4 1 l 4 4 I i 3898SM6-7 k ~ A Yf i 1

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

INDUSTRY /EPRI ADVANCED LWR PROGRAM g SYSTEMS 8 MATERIALS DEPARTMENT v K. E. STAHLKOPF DIRECTOR INDUSTRY /EPRI ALWR PROGRAM D. M. NOBLE SR. PROGRAM MANAGER O UTILITY  ! SMALL REGULATORY REQUIREMENTS PLANT STABILIZATION DOCUMENT CONCEPTUAL DESIGN D. M. NOBLE

• J. C. DEVINE. JR.

, l l W. R. SUGNET a l

• ACTING O

l l

i . a i

ALWR UTILITY REQUIREMENTS DOCUMENT i

OVERVIEW... 1 THE REQUIREMENTS DOCUMENT IS AN EXTENSIVE COMPILATION OF THE

DESIGN. CONSTRUCTION AND PERFORMANCE REQUIREMENTS OF THE ADVANCED i

l LWR; AS ESTABLISHED BY THE U.S. UTILITY INDUSTRY. l THE DOCUMENT REFLECTS UTILITY /NRC ON ADVANCED LWR LICENSING ' , REQUIREMENTS, AND WHEN COMPLETE WILL BE ENDORSED VIA AN NRC-i ISSUED SAFETY EVALUATION REPORT. ] O THE DOCUMENT IS THE STARTING POINT FOR SUBSEQUENT DETAILED 1 ' ENGINEERING FOR AN ADVANCED LWR AND WILL BE A BASIS FOR j DEVELOPMENT OF STANDARD PLANT DESIGNS. l i i l l A -A+.r j 3991SM6-7

ALWR UTILITY REQUIREMENTS DOCUMENT STRUCTURE AND CONTENT CHAPTER SCOPE (BWR AND PWR) 1 OVERALL REQUIREMENTS INCLUDING DESIGN BASES, MATERIALS, CONSTRUCTIBILITY, MAINTAINABILITY, OPERABILITY, AVAILABILITY, ETC. 2 POWER GENERATION SYSTEMS 3 PRIMARY COOLANT SYSTEM AND NON-SAFETY AUXILIARY SYSTEMS 4 REACTOR SYSTEMS 5 SAFETY SYSTEMS 6 ARRANGEMENTS 7 FUELING AND REFUELING 8 PLANT COOLING WATER SYSTEMS 9 SITE SUPPORT SYSTEMS 10 PLANT CONTROLS 11 ELECTRIC POWER SYSTEMS 12 RADIDACTIVE WASTE PROCESSING SYSTEMS 13 TURBINE GENERATOR SYSTEMS h JC03980SM6-2

ALWR REQUIREMENTS DOCUMENT PROGRESS EXECUTIVE

SUMMARY

FOR ALWR REQUIREMENTS STATES ALWR PROGRAM OBJECTIVES, POLICY, PHILOSOPHY, TOP-TIER REQUIREMENTS AND IMPLEMENTATION SCENARIO ISSUED TO NRC ON JULY 8, 1986 CHAPTER 1 - OVERALL REQUIREMENTS CHAPTER HAS BEEN PREPARED, REVIEWED EXTENSIVELY BY UTILITIES AND INDUSTRY, AND APPROVED BY THE ALWR UTILITY STEERING COMMITTEE O - ISSUED 10 NRC auLv 8, 1988 i O &&n 3971SM6-1

O ALWR REQUIREMENTS DOCUMENT EXECUTIVE

SUMMARY

OBJECTIVES PROVIDE CONCISE AND VISIBLE STATEMENT OF ALWR OBJECTIVES POLICY / PHILOSOPHY TOP-TIER REQUIREMENTS h O 39725M6-4

EXECUTIVE

SUMMARY

KEY POINTS ALWR FUNDAMENTAL OBJECTIVES TECHNICAL EXCELLENCE - THE ALWR MUST MEET THE NEEDS OF THE UTILITY INDUSTRY IN ALL RESPECTS, INCLUDING SAFETY, ENVIRONMENTAL COMPATIBILITY AND TECHNICAL PERFORMANCE ECONOMIC ADVANTAGE - THE ALWR MUST BE ECONOMICALLY ATTRACTIVE TO THE UTILITY INVESTOR, CONSIDERING BOTH FIRST COST AND LIFE CYCLE COST, INVESTMENT PROTECTION - THE ALWR MUST PROVIDE VERY HIGH PRO-y TECTION OF THE UTILITY INVESTMENT, PARTICULARLY IN TERMS OF: EXTREMELY LOW RISK OF SEVERE ACCIDENT ASSURED LICENSABILITY PREDICTABLE (AND CONTROLLABLE) CONSTRUCTION COST / SCHEDULE PREDICTABLE OPERATING COST AND PLANT AVAILABILITY kdV 3991SM6-3

O EXECUTIVE

SUMMARY

KEY POINTS ALWR PHILOSOPHY THE FOLLOWING WORKING PRINCIPLES SHALL GUIDE THE DEVELOPMENT OF ALWR REQUIREMENTS AND CONCEPTUAL DESIGN: SAFETY SIMPLICITY DESIGN MARGIN RELIANCE ON EXPERIENCE OWNER /0PERATOR FOCUS 1 l l 9 l 3972SM6-6

i l lo SAFETY: NUCLEAR SAFETY IS OF PARAMOUNT IMPORTANCE AND MUST PLAY THE DOMINANT ROLE IN ALWR REQUIREMENTS ALWR WILL IMPROVE ON EXISTING U.S. NUCLEAR PLANT SAFETY IMPROVED SAFETY WILL BE ACHIEVED THROUGH: SIMPLIFICATION O - IMPROVED DESIGN MARGINS IMPROVED OPERABILITY / HUMAN FACTORS CONSIDERATIONS QUANTITATIVE RIS< ASSESSMENT WILL BE EMPLOYED A-A'?8 3972SM6-7

O SIMPLICITY PLANT SIMPLIFICATION IS A PRIMARY MEANS OF ACHIEVING ALWR PROGRAM REQUIREMENTS, IN TERMS OF SAFETY, AVAILABILITY, OPERABILITY, MAINTAINABILITY, AND INITIAL CAPITAL AND LIFE-CYCLE COSTS. MAIN FOCUS IS ON SIMPLICITY IN OPERATIONS STRAIGHTFORWARD, PREDICTABLE AND UNDERSTANDABLE SYSTEMS RELIANCE ON FIRST PRINCIPLES RELIANCE ON INHERENT SAFETY MECHANISMS

   -     IMPROVED ACCESSIBILITY, FOR CONSTRUCTION, OPERATIONS AND MAINTENANCE REDUCED NUMBERS OF COMPONENTS PLANT SIMPLIFICATION OPPORTUNITIES WILL BE IDENTIFIED AND EVALUATED ON A CASE-BY-CASE BASIS, DURING PREPARATION OF EACH REQUIREMENTS DOCUMENT CHAPTER O

3972SM6-8

J l i - DESIGN MARGIN j THE ALWR IS TO BE A " ROBUST" PLANT WITH SUBSTANTIAL ! BUILT-IN MARGINS TO PROVIDE INHERENT CAPABILITY TO DEAL l WITH ADVERSE SITUATIONS. I i - RELATED TO SIMPLIFICATION INCLUDES SUCH DESIGN FEATURES AS DVERSIZED PRESSURIZER IMPROVED CORE THERMAL MARGINS INVOLVES SOME TRADE-0FFS, TO BE EVALUATED ON A CASE-BY-CASE BASIS I 3972SM6-9

O' RELIANCE ON EXPERIENCE ALWR IS TO BE AN EVOLUTIONARY DESIGN, BASED ON PROVEN TECHNOLOGY

    -   WHERE PRACTICAL, FULLY MATURE DESIGNS, FEATURES, EQUIPMENT, ETC, WILL BE EMPLOYED (WITH LITTLE OR NO EXTRAPOLATION)
   -    TECHNOLOGICAL IMPROVEMENTS CAN BE EMPLOYED, IN CASES WHERE THERE IS CLEAR BENEFIT THERE IS HIGH CONFIDENCE IN SUCCESS, BASED ON EXPERIENCE IN COMPARABLE APPLICATIONS, EXTENSIVE TESTING, ETC.

ALTERNATIVES ARE UNACCEPTABLE OR INCONSISTENT WITH ALWR REQUIREMENTS. O 3972SM6-10

i' j i I  ! 1 4 i i 1 l ! OWNER /0PERATOR FOCUS 4 l

THE ALWR REQUIREMENTS DOCUMENT IS WRITTEN FOR AND REVIEWED BY THE ALWR UTILITY STEERING COMMITTEE -

i 1-I L 1 i l t f l l

                                                                                                                                   \

i i i l k ~858 )  ! ! i i 3972SM6-11 i

EXECUTIVE

SUMMARY

KEY POINTS ALWR TOP-TIER REQUIREMENTS h SAFETY TARGET FOR MEAN ANNUAL CORE DAMAGE FREQUENCY 10-5 EVENTS /YR OR LESS SEVERE ACCIDENT (EVENTS OF FREQUENCY GREATER THAN 10-6/YR) WOULD CAUSE LESS THAN 25 R DOSE WITHIN ONE-HALF MILE AVAILABILITY 87% AVAILABILITY (AVERAGE OVER PLANT LIFE) DESIGNED F0R 92% ANNUAL AVAILABILITY, WITH PROVISION FOR ADDITIONAL OUTAGES TOTALING 6 M0. EVERY 10 YEARS h FOR MAINTENANCE (THEREFORE, NET 87% AVAILABILITY OVER PLANT LIFE) TWO-YEAR REFUELING INTERVAL, 18-DAY (POWER-TO-POWER) DURATION INADVERTENT TRIPS FEWER THAN ONE PER YEAR PLANT LIFE 60 YEAR DESIGN LIFETIME, WITHOUT A NEED FOR EXTENDED (MULTI-YEAR) OUTAGE O T370044,1A

O ALWR TOP-TIER REQUIREMENTS (CONT'D) RADWASTE

                                                         -                                                                                                             3 ALWR DESIGNED TO PRODUCE NO MORE THAN 2500 FT /YR RADI0 ACTIVE WASTES FOR OFF-SITE DISPOSAL a

PERSONNEL EXPOSURE l ALWR DESIGNED TO PERMIT OCCUPATIONAL EXPOSURE LESS THAN 100 MAN-REM /YR, AVERAGE OVER PLANT LIFETIME. COST ALWR COST TARGETS: CAPITAL 4.5c/KWH FUEL 1.24/KWH OgM 0.84/KWH LIFE CYCLE 6.54/KWH , (1985 COSTS LEVELIZED FOR 30 YRS) 1 CONSTRUCTIBILITY ALWR DESIGNED TO SUPPORT FIRST PLANT CONSTRUCTION SCHEDULE OF 54 MONTHS, FROM START OF STRUCTURAL CONCRETE THROUGH INITIAL OPERATION

                                                                                                                            )-AS/

3972SM6-15 i j

V

ADVANCED LWR PROGRAM REGULATORY STABILIZATION l

OBJECTIVE: DETERMINE THE LIST OF STABLE REGULATORY REQUIREMENTS WHICH MUST BE MET BY ANY NEW LWR DESIGN b

                                 'h 9

I O 9

                       -3991SM6-9

? b O ADVANCED LWR PROGRAM REGULATORY STABILIZATION STATUS NRC IS PARTICIPATING JOINTLY WITH EPRI IN THE REGULATORY STABILIZATION EFFORT ACRS BRIEFED SEPTEMBER 5, 1984 COMMISSIONERS BRIEFED FEBRUARY 7, 1985 0F APPR0XIMATELY 700 ISSUES IDENTIFIED, 46 REMAIN TO BE 4

         ~

ADDRESSED BY THE ALWR PROGRAM ) ISSUE RESOLUTION METHODOLOGY REPORT TRANSMITTED TO NRC j DECEMBER 3, 1985 NRC RESPONSE TO ESTABLISH PROCESS FOR RESOLVING CURRENT ISSUES, EVALUATING FUTURE ISSUES AND REVIEW 0F REQUIREMENTS FOR THE ALWR PLANT l.' l O

                                              +=

3991SM6-10

O 800 - TREND OF ISSUES (700) 700 - , (649) (703) g (670) 600 - (***) b' (606) g (595) [ [(549)

            ~

W . g 500 - (516) (320)

 ?          -

u. .

O 400 - ~ C:' w ~ cc h 300 f , Z 200 -

          -3        .

l (137) 100[ (133) 1978 1979 1980 1981 1982 1983 1984 1985 1986 YEAR O

O 80

                             }         PROGRESS ON REr1AINING ISSUES
                        < 00 - ,

fgg) fg3)

                               ',            TOT AL ISSUES (670)

(629) (638) (649) 600 - . (606) (588) CO - w . U 500 - Lo - 3 .- u., - O 400 - d ! W - (345) C l E 300 - - D . Z - 200 -

                             ;                  . (140)

I 100 - -

                             -                                                  REM AINING ISSUES (94) (90)       (101)

( - (58) (56) (49) (46) 0 i + i i 1983 1984 1985 1986 YEAR A-253

SUMMARY

OF REMAINING ISSUES g ISSUE NUMBER TITLE A-29 NUCLEAR POWER PLANT DESIGN FOR REDUCTION OF VULNERABILITY TO INDUSTRIAL SABOTAGE A-44 STATION BLACK 0UT A-45 SHUTDOWN DECAY HEAT REMOVAL REQUIREMENTS A-47 SAFETY IMPLICATIONS OF CONTROL SYSTEMS A-48 HYDROGEN CONTROL MEASURES AND EFFECT OF HYDROGEN BURNS ON SAFETY EQUIPMENT B-22 LWR FUEL h B-29 EFFECTIVENESS OF ULTIMATE HEAT SINKS D-2 EMERGENCY CORE COOLING SYSTEM CAPABILITY FOR FUTURE PLANTS ll.E.4.3 CONTAINMENT DESIGN -- INTEGRITY CHECK 2 FAILURE OF PROTECTIVE DEVICES ON ESSENTIAL EQUIPMENT 23 REACTOR COOLANT PUMP SEALS 29 BOLTING DEGRADATION OR FAILURE IN NUCLEAR POWER PLANTS O 3896SM6-12

l

SUMMARY

OF REMAINING ISSUES (CONTINUED)

   ~q            ISSUE b           NUMBER                                TITLE 51                    PROPOSED REQUIREMENTS FOR IMPROVING RELIABILITY OF OPEN CYCLE SERVICE WATER SYSTEMS 67.7.0                STEAM GENERATOR STAFF ACTIONS -- IMPROVED EDDY CURRENT TESTS 75                    GENERIC IMPLICATIONS OF ATWS EVENTS AT SALEM NUCLEAR PLANT 76                    INSTRUMENTATION AND CONTROL POWER INTERACTIONS 79                    UNANALYZED REACTOR VESSEL THERMAL STRESS DURING NATURAL CONVECTION C00LDOWN 82 O                                 BEYOND DESIGN BASIS ACCIDENTS IN SPENT FUEL POOLS 84                    CE PROVs

, 91 MAIN CRANKSHAFT FAILURE IN TRANSAMERICA DELAVAL EMERGENCY DIESEL GENERATOR 93 STEAM BINDING 0F AUX 1LIARY FEEDWATER PUMPS 94 ADDITIONAL LOW TEMPERATURE PROTECTION ISSUES FOR LIGHT WATER REACTORS 96 RHR SUCTION VALVE TESTING l 99 RCS/RHR SUCTION LINE VALVE INTERLOCK ON PWRs 101 BWR WATER LEVEL REDUNDANCY

                                                         #asy 3896SM6-13

SUMMARY

OF REMAINING ISSUES (CONTINUED) ISSUE NUMBER TITLE 103 DESIGN.FOR MAXIMUM PROBABLE PRECIPITATION 105 INTERFACING SYSTEM LOCA AT BWRs 107 GENERIC IMPLICATIONS OF MAIN TRANSFORMER FAILURE 110 EQUIPMENT PROTECTIVE DEVICES ON ENGINEERING SAFETY FEATURES 114 SEISMIC INDUCED RELAY CHATTER 116 ACCIDENT MANAGEMENT 117 ALLOWABLE OUTAGE TIMES FOR DIVERSE SIMULTANEOUS EQUIPMENT GUTAGES h 118 TENDON ANCHORAGE FAILURES 120 ON-LINE TESTABILITY OF PROTECTION SYSTEMS 121 HYDROGEN CONTROL FOR LARGE DRY CONTAINMENTS i 122.1A DAVIS-BESSE LOSS OF ALL FEEDWATER EVENT--COMMON MODE. FAILURE OF AUXILIARY FEEDWATER PUMP DISCHARGE ISOLATION VALVES IN CLOSED POSITION 122.18 DAVIS-BESSE LOSS OF ALL FEEDWATER EVENT-- EXCESSIVE DELAY IN RECOVERY OF AUXILIARY FEEDWATER O 3896SM6-14

l

SUMMARY

OF REMAINING ISSUES (CONTINUED) i I ISSUE O NuM8eR TITLE 122.lC DAVIS-BESSE LOSS OF ALL FEEDWATER EVENT-- INTERRUPTION OF AUXILIARY FEEDWATER FLOW 122.2 DAVIS-BESSE LOSS OF ALL FEEDWATER EVENT-- ADEQUACY OF EMERGENCY PROCEDURES, OPERATOR TRAINING AND AVAILABLE PLANT MONITORING SYSTEMS 124 AUXILIARY FEEDWATER SYSTEM RELIABILITY 125 LONG-TERM GENERIC ACTIONS AS A RESULT OF THE DAVIS-BESSE EVENT OF JUNE 9, 1985 i HF-01.4.5 APPLICATIONS OF ARTIFICIAL INTELLIGENCE i O HF-01.5.1 LOCAL CONTROL STATIONS j HF-01.5.2 ANNUNCIATORS HF=01.5.3 EVALUATE OPERATIONAL AID SYSTEMS HF-01.5.4 COMPUTERS AND COMPUTER DISPLAYS l 38965M6-15 l -

O TREATMENT OF FUTURE ISSUES ISSUE m IDENTIFICATION Definition of issue V ISSUE EVALUATION Prioritization of issue V - ISSUE m RESOLUTION Screening Criteria for Application to ALWR II APPLICATION

                >   Incorporate into Requirements Document TO THE ALWR O

I O TREATMENT OF FUTURE ISSUES ISSUE l enuficadon of issue IDENTIFICATION Definition of issue V ISSUE Screening Criteria for Applicability EVALUATION . Prioritization of issue V ' ISSUE ene c Resoludon RESOLUTION Screening Criteria for Application to ALWR 1 P APPLICATION

                  >   Incorporate into Requirernents Document TO THE ALWR l

l 9

O e'a"r oerisizario" su8aec's PLANT OPTIMIZATION SUBJECTS ARE PROPOSED CHANGES TO CURRENT REGULATORY REQUIREMENTS THAT RESULT IN OPTIMIZATION OF PLANT DESIGN, CONSTRUCTION, AND OPERATION. NECESSARY TO ACHIEVE THE ALWR OBJECTIVES OF: IMPROVED SAFETY IMPROVEMENTS IN OPERATIONAL PERFORMANCE IMPROVED ECONOMICS REVIEW 0F CURRENT TECHNOLOGY INDICATES THAT SOME i HIGHLY PRESCRIPTIVE EXISTING REGULATORY REQUIREMENTS MAY BE OVERLY CONSERVATIVE. UTILITY STEERING COMMITTEE DIRECTION: LIMIT NUMBER CONCENTRATE ON SUBSTANTIAL BENEFIT WITH AFFECTING SAFETY HIGH PROBABILITY OF NRC ACCEPTANCE BEING PURSUED INDEPENDENT OF ALWR PROGRAM A -ASE 3896SM6-9

PLANT OPTIMlZATION SUBJECTS STATUS FIVE SUBJECTS BEING PURSUED FOR CHAPTER 1 LEAK BEFORE BREAK METHODOLOGY OPERATING BASIS EARTHOUAKE AND DYNAMIC ANALYSIS METHODS SOURCE TERMS EQUIPMENT SEISMIC QUALIFICATION BY EXPERIENCE TORNADO DESIGN NO SUBJECTS FOR CHAPTER 2 ONE SUBJECT LIKELY FOR CHAPTER 3 BWR MAIN STREAMLINE VALVE AND ISOLATION CONTROL SUBJECTS FOR CHAPTERS 4 AND 5 BEING IDENTIFIED O l 3896SM6-10

CONCLUSIONS ISSUES CONTINUE BEING IDENTIFIED BY THE NRC AT A l . RATE OF GREATER THAN 30 ISSUES PER YEAR i AN EFFECTIVE PROCESS FOR TREATING CURRENT ISSUES HAS BEEN DEVELOPED TREATMENT OF FUTURE ISSUES XEY TO REGULATORY REQUIREMENTS STABILIZATION NEW PROCESS BEING IMPLEMENTED PROCESS EXPECTED TO REASONABLY CONTROL BACKFITS 1 O OPTIMlZATION SUBJECTS ARE IMPORTANT TO VIABILITY OF ALWR PROGRAM i l O h.w? 3896SM6-11

__e- -, . m. .e- a m.m e Am e-ea-a.J.--e4 -e. e- e--a AA- ___ *.M.e= = .m24 -.,h.hh-*--AhW *A h44+-Mh he4 *te=-m+--o*e---.m.-m.em+h.h aJ4nwA4 .--- m ._J. - __a_ 4 4 0 l l l f i 1 i e i i l I f i l i i I s 9 h l l 4 1 h I l k l i I 1 l J. t .h l l i I l 4

                                                                                                        - 2
    .. .                                          _ _ .._... . . -                                   _a                            .

                                                                                                             '                                C. Ff 4 J EP*t APPENDIX XVI U.S. ADVANCED LWR PROGRAM O

U. S. ADVANCED LWR PROGRAM r SMALL PLANT PROGRAM l

• DEVELOP CONCEPTUAL DESIGNS FOR SMALL (<600 MWE) LWRs l-GE/BECHTEL/MII l

WESTINGHOUSE / BURNS & ROE O i V ( l l l l

                                                                                                                 ~

hOf

    - .- - - - -  # y- - - --e - ---,#   ,-wr     .---------w---.,w.         - - - ------i------+w.,--.m--         __ ---         ,  -.--     -  - - - - - -- - -.----w-

 'O 809 [-

PROGRESS ON REMAINING ISSUES END (,0a> (nu

                                      '                              TOT AL ISSUES                                   '

(670) 600 1 - (629) (*) (649) (606) (588) to .- v> 500 2 Cr) . 4 0 400 - E Lu . (345) O cc h 300 - , Z . 200 -

                                                                          .       (148)

\ l 100 - - ettMAmmo ssuts (94) (98) (101) l l , i '

                                       ;                                                                                                      (58)     (56) (49) (44) 0                                                   i                                             e                                               i                             '

1983 1984 1985 1966 YEAR

                                                       ,.                                                      A -A6b 2
     -,,--m            e
            <   n-         .,a_-
                                   ~ - , , - , _ - , - - - - ,         ---g- - - , - - - - , ,        ,,.----g    ---,----~,---,,,,---,-w,,--                _ - - - - - - - - - -        _ _ _ - _

ADVANCED LE P9GiMM l REGULATORY STABILINTIGI OBJECTIVE: DETERMINE THE LIST OF STABLE REGULATIRY REQUIREMENTS WHICH MUST BE HET BY ANY NEW LWR IIESIGN O 1 t - O A 46 /

  -r-r--wvy----   ,-e-.r7,.7,    --p--w---y,y,--* - - - - - - - . - - . , - - - - . - - - . , - - .                  , , ,_ ,er,,     --%.,m,--,-w-m.----,.--.-.-..--                 ---- --. w.     ---,--- .--

 'O                                                                                                           . ;-

I FOREIGN PARTICIPAT11M KOREA A DRAFT MOU HAS BEEN EXCHAleGES N'IT41 ESREA ELECTRIC POWER COMPANY.

                                               $150,000 PER YEAR FOR FOUR YEARS PAID To EPRI.

PARTICIPATION IN PWR PORTION OF UTILITY REQUIREMENTS DOCUMENT ONLY. STATION TWO ENGINEERS AT PALO ALTO. OPPORTUNITY TO PARTICIPATE ON $TEERING COMMITTEE. A-Ath

  ,----,-.m,----w-        ,   o   e-,~~--m,---          -w m       ,--,--r-  -w-,---     ------w

FOREIGN PARTICIPATIGI TAlWAN M00 SIGNED WITH IA!WAN POWER COMPANY SusJECT TO APPROVAL OF REPUBLIC OF CHINA GOVERNMENT.

                                     $225.000 PER YEAR FOR FOUR YEARS PAID To EPRI.

l PARTICIPATION IN UTILITY REQUIREMENTS D0cuMENT ONLY. STATION TWO ENGINEEkS AT PALO ALTO. OPPORTUNITY TO PARTICIPATE ON STEERING COMMITTEE. i I Y

                                                                      . . e-se               * ' " 'N
 -w-- - -     e       r , w mm

O JAPAN EPRI HAS RECEl'VED A DETmER PROM K48tSAI POWER EXPRESSING l'NTEREST IW PWRT!!CllPATI4tG IN EPRI'S ALWR Pa0 GRAM. KANSAI IS ROST INTERESTED. IN KSIGII $1MPLIFICATION OF PWR'S. EPRI HAS RESPONDED POSITIVELY TO KANSAl'S INTEREST IN THE PROGRAM DISCUSSIONS ARE UNDERWAY TO SOLIDIFY THEIR l O PARTICIPATION l l l i i l

                                                                                        /9'kbh

• Q .

FOREIGN PARTICIPATION

                                   -              JAPAN i

MOU IN FINAL STAGES OF NEGOTIATION WITN JAPAN

        .                                                     ATOMIC POWER COMPANY (JAPC) FOR COOPERATlell GII
        !                                                      DEVELOPMENT OF (<600MWE) BWR.                        _

i EPRI WILL aECEIVE RESULTS OF $1 AILLIGII 1986 WORK FUNDED BY JAPC alsD PEsFORRED BY EE. JAPC WILL mECEIVE RESULTS OF EPRI PHASE I STUDIES ON SMALL BWR. , j RESULTS OF EPRI PHASE Il STUDIES THROUGH TME END OF 1986. R'ESULTS OF RELATED E INFORMATION ON SMALL ! BWR DEVELOPMENT TNIKEMH THE END OF 1986. I DEPENDING ON TNE STATUS OF THE WORK AT THE END OF 1986, BOTH PARTIES WILL CONSIDER EXTENSIOII 0F THE AGREEMENT TO 1987 AND BEYOND. A cet,r

                                                                                                                                       ,_      +  emm e M * * "

> - - - ,,,-w,- --,-__-,. _ .,m- . - -,n__--- . - - , .. , . , . - . - _ - _ , , - _ . . _ ,,,,,,,-_,-,_.,--,.e.

                             .w     .   .c..~,,     .

O EPRI ALWR PROGRAM CREDIBLE SAFE STANDARD PLANT EPRI ALWR PROGRAM IS FULLY SUPPORTIVE OF INDUSTRY EFFORTS TD DEVELGP A STANDARDIZED NUCLEAR PLANT DESIGN THE EPP! ALWR PROGRAM, LINKED TO PLANT STAEARDIZATi(Bl. CAN PLAN /s "2i GNIFICANT PART IN ITS SUCCESS. IN PARTICULAR: THE RESOLUTION OF REGULATORY ISSUES, VIA THE EPRI REGULATORY. STABILIZATION PR.0 GRAM, CREATES A FAVORABLE CLIMATE FOR STANDARDIZED PLANT LICENSING O - THE ALWR REQUIREMENTS DOCUMENT SERVES AS A COMPREHENSIVE TECHNICAL BASIS FOR STANDARD PLANT DESIGN THE EPRI ALWR PROGRAM IS C00RDIMTED WITH DOE'S PROGRAM 10 SECURE NRC CERTIFICATION OF AN ADVANCED BWR AND PWR DESIGN. A MS b

                       .. x     _ - - -

O ADVANCED !!WR RR0 GRAM CONTINUING NRC !!NTERFACE

                           -    FOLLOWUP EFFORT AS REQUIREENTS DOCUMENT CilAPTERS REVIEWED BY NRC
                           -    ENSURE REGULATORY ISSUE RESOLLITIGd IRCBRPGRA
                            -   ACHIEVE FAVORABLE SER FOR EACH CHAPTER 2D POR
,                               DOCUMENT O                                                   -
                                                                                                                               +sa
                                                                                                                                       *6
                                                                                                                               *h MNwe
         -.,,w-----w.-.,-,           , . - _ .   . . . - -                                     . , _ - _ _ _ _ . . - , - . _ .            , __

jf "ll 5

                                                                                                                                                                                      . ?' :

EPRI ADVANCED LIG!!T ETER itEACTOR PROGRAM FARTLCIPANTS... THE EPRI PROGRAM DRAWS UPON A WROAD BASE OF EXPERT PARTICIPANTS INCLUDING: UTILITY INDl1STRY EPRI MEMBER WTIILITilES REPRESENTED BY A WTILITY STEERING COMMITTEE OF SENI:0R tlTIILITY Mt0 FESS 10NALS CONTRACTORS l TEAMS OF NSSS SUPPLIERS, ARCHITECT ENGINEERING FORMS, AND ' MAJOR UTILITY COMPANIES, INCLUDING: COMBUSTION ENGINEERING / DUKE POWER GENERAL ELECTRIC /BECHTEL WESTINGHOUSE /SARGENT a L MDY/CO MONWEALTH EDISON STONE 1 WEBSTER / YANKEE ATOMIC DIRECT ASSISTANCE TO EPRI FROM MPR ASSOCI:ATES S. LEVY, IINC. A-ASV __,m _ , . _ _ . - _ - _ . _ _ _ _ _ _ _ _ _ _ _ _ , _ . _ ,.,,.--.._,,.,.-,_.,___,___,,__-% - - _,,,,,,._.-_m- . __ ,. .___,,__

                                                                                                               'i EPRI ADVANCED LIGHT WATER REACTOR PROGRM T'HE EPR! PROGRAM STRUCTURE THE PROGRAM CONSISTS OF THREE SEPARATE B1JT RELATED MAJOR ELEMENTS:

UTILITY REQUIREMENTS DOCUMENT THE DEVELOPMENT OF A COMPREHE'!S!VE SET OF DESIGN REQUIREMENTS FOR THE ALWR

                    ~

REGJ'LATC5Y STA3}LIZATICM CC0FERATIVE EFFCRT W,iTH NRC 70 llDENTIFY MD RESOLVE ALL 00TSTEDING llSSUES OF NUCLEAR PLAT SAFETY SMALL PLA';T DESIGN DEVELOPMENT THE INVESTlGATION AND DEVELOPMENT OF SMALL (<600 MWE) NUCLEAR PLANT OPTIONS l l 0 f-A49 . s

5 bo G a DOE ADVANCED LIGHT WATER REACTOR PROGRAM

    . G0AL:    FOSTER DEVELOPMENT AND CERTIFICATION OF SIMPLER, SAFER, AND MultE REllABLE LIGHT WATER REACTORS FOR 1HE FUTURE.

O CLOSE C00RDlHAT10N WITH THE ELECTRIC POWER RESEARCil INSTITUTE / INDUSTRY ALWR PROGRAM. h, o PRINCIPAL WORK AREAS D e n N 84 h - LARGE PLANT APAlR AND APWR DESIGN VERIFICATION gg MID-SIZE INNOVATIVE LWR DEVELOPMENT $E AE SPECIAL APPLICATIONS PROGRAMS o :: o STRATEGY: USE ADVAllCED RET.CTOR CONCEPTS AS ORGANIZING FORCE TO TEST SUCCESS OF R&D E M TASKS, DESIGN DEVELOPMENT EFFORTS AND LICENSING. UTilllL USER STATEMENTS OF PRODUCI REQUIREMENTS. s E E

i EPRI/ DOE SCHEDULE COMPATIBILITY i 1986 1987 1988 1989 1 1990 EPRI ACCELERATED SCHEDULE i ) Chapters Y l l 1, 2 Y Chaptors 3' 4' 6 LEGEND A . . ' I t Chapters v Submit SSAR g 6, 10 y Module N Chapters

!  N        7. B. D. 11, 12                                                                v   NRC R view
                                          ==.============,=============================
     = = . , = = , = = , = = , =

_ _ _ SAR Prepor. DOE VERIFICATION PROGRAM ___ l

1. Licensing Dos. is l .

2. Nuclear Island 7 *

                -   Reactor /Sofety Systems               :-----       V (Chpt'.s 3. 4 5)
                -   Bfdg/ Arrangements                    ____________-

3 7 (Chpt 6) p 7

                -   Auxillary Support Sys and 1&C         _______,,______-

(Chpt's 7 thru 12) PRA/FMEA and Technical --===============- Specifications

3. Turbine Island M 5/30/so

e% NRR STAFF PRESENTATION TO THE O ACRS APPENDIX XVIII STAFF REVIEW 0F EPRI-ALWR REQUIREMENTS DOCUMENT

SUBJECT:

STAFF REVIEW 0F EPRI-ALWR REQUIREMENTS DOCUMENT DATE: JULY 10, 1986 i PRESENTER: D. MORAN O , PRESENTER'S TITLE / BRANCH /DIV: PROJECT MANAGER SAFETY PROGRAM EVALUATION BRANCH, DSR0 PrKF _NTER'S NRC TEL. NO.: 49-27422 SUBCOMMITTEE: FULL COMMITTEE k c?fL

        *                     *                            +
                           ,%                          +

e

        %.g
              .3
                    ~                 .

L

                                                                                                 ^ STAFF REYlEW PLAN

(' . {'

                   ],'               ,. CHAPTER TRANSMITTAL SCHEDULE
                 .                         REQUIREMENTS DOCUMENT CHAPTER TITLES

s. ,

CONTENTS 0F CHAPTER 1

                            ;              STAFF DISCIPLINES REQUIRED DISCUSSION OF THE STAFF END PRODUCT

o MATRIX OF CHAPTERS YS. STAFF DISCIPLINES MAN LOADING DETAll b

w 1 ke = [O l

          .                                                                                                       # m is

3

   ~

O

          ;_R_EOUIREMENTS DOCUMENT CH_APTE_R TRANSMllTAL SCHEDULE CHAPTER          DATE TO BE TRANSMITTED 1            JULY 1986 2            SEPTEMBER 1986 3,4 & 5          JUNE 1987 c              6            DECEMBER 1987 l

O 7 THRU 12 MARCH 1988 13 SEPTEMBER 1988 k ( k - Sff

h.--- - m +A h 4 I S REQUIREMENTS DOCUMENT CHAPTERS CHAPTER TITLE 1 1 OVERALL REQUIREMENTS 2 POWER GENERATION SYSTEMS 3 REACTOR COOLANT AND NON-SAFETY , REACTOR AUXILIARY SYSTEMS 4 REACTOR SYSTEMS 10 5 SAFETY SYSTEMS 6 BUILDING DESIGN AND ARRANGEMENTS 7 FUELING AND REFUELING 8 PLANT COOLING WATER SYSTEMS 9 SITE SUPPORT SYSTEMS 10 INSTRUMENTATION AND CONTROL SYSTEMS AND PLANT CONTROL STATIONS 1i ELECTRIC POWER SYSTEMS 12 RADIOACTIVE WASTE PROCESSING SYSTEMS 13 TURBINE GENERATOR SYSTEMS A-MF

   ~O CONTENTS 0F CHAPTER 1 SECTION                                         TITLE 1             INTRODUCTION 2               GENERAL REQUIREMENTS FOR DESIGN 3               DESIGN BASIS EVENTS

4 STRUCTURAL DESIGN BASIS 5 MATERIALS 6 RELIABILITY AND AVAILABILITY 7 CONSTRUCTABILITY 8 OPERABILITY AND MAINTAINABILITY 9 OUALITY ASSURANCE 10 LICENSING A -a?74

O i STAFF DISCIPLINES REQUIRED

SITE AND ENYlRONMENTAL REACTOR SYSTEMS

[ PLANT SYSTEMS MATERIALS O eLANT STRUCTURES ELECTRICAL SYSTEMS / INSTRUMENTATION AND CONTROL COORDINATION FOCAL POINTS l LICENSING DIYlSIONS i

PWR-A

, PWR-B BWR HUMAN TECHNOLOGY l l A-277

O .

STAFF END PRODUCT - ONE SER WITH 13 SECTIONS OR CHAPTERS SECTIONS WILL BE DRAFTED AS CHAPTERS ARE REVIEWED, SIX MONTHS ALLOTTED FOR REVIEW OF EACH EPRI SUBMITTAL. l THE STAFF REVIEW WILL BE DIRECTED AT DETERMINING WHETHER THE REQUIREMENTS DOCUMENT ADEQUATELY REFLECTS NRC PULES l AND REGULATIONS TO PROTECT PUBLIC HEALTH AND SAFETY. \ l um

   . -.         ..._ - - -.                  __x____

O NRC REVIEW RESOURCE ALLOCATION . FISCAL YEAR 1986 PROJECT MANAGER + 2 FULL TIME EQUIVALENTS REQUIREMENTS DOCUMENT REVIEW 1800 MANHOURS PER CHAPTER 600 MANHOURS PER MONTH INTEGRATION (AFTER CHAPTER 9) l 0 A-279 _ _ . __ __ _ =_ _

DSR0 REVIEW RESPONSIBILITY - ALWR REQUIREMENTS DOUMENT CHAPTER SUBMITTAL SITE / MATERIALS STRUCTURAL l&C REACTOR PLANT DATES EVIRON. SYSTEMS SYSTEMS

1. OVERALL 6/86 X X X X X X REQJIREMENTS

2. POWER 9/86 X X X X GENERATION SYST.

3. PRIMARY 6/87 X X X COOLANT AND NON-SAFETY AUX.SYST.

4. REACTOR 6/87 X X X SYST.

S. SAFETY 6/87 X X X X SYST.

6. ARRANGEMENTS 12/87 X

7. FUELING & 3/88 X REFUELING i 8. PLANT 3/88 X X 1 COOLING l'

I WATER SYST.

9. SITE 3/88 X X SUPPORT SYST.

10. INSTRUMENT. 3/88 X

                        & CONTROL

11. ELECTRIC 3/88 X X POWER SYST.

12. RADIOACTIVE 3/88 X X WASTE PROCESSING SYST.

(

13. TURBINE GENERATOR 9/88 -MhO X X SYST.

F

4 FUTURE INTERACTIONS WITH ACRS l

                                      -         STAFF WILL KEEP ACRS INFORMED OF PROGRAM STATUS, INCLUDING FURNISHING COPIES OF REQUIREMENTS DOCUMENT CHAPTERS To ACRS, STAFF WOULD LIKE TO MEET WITH ACRS ON KEY TOPICS ASSOCIATED WITH ALWR REQUIREMENTS DOCUMENT.

l SUGGEST MEETINGS ON GPOUPS OF CHAPTERS AT APPPOPRIATE l TIMES AFTER INITIAL STAFF REVIEW IS COMPLETE AND PRELIMINARY STAFF POSITION ESTABLISHED. ' - FINAL MEETING PRIOR TO ISSUING COMPLETE SER, ( 4 1 0 A-R8l

1

      ,                                      APPENDIX XIX O                                   PROPOSED COMMISSION POLICY STATEMENT ON TECHNICAL SPECIFICATIONS PROPOSED COMMISSION POLICY STATEMENT ON TECHNICAL SPECIFICATIONS PRESENTED TO ACRS FULL COMMITTEE JULY 11, 1986

( O BY TECHNICAL SPECIFICATION COORDINATION BRANC11, DIVISION OF HUMAN FACTORS TECHNOLOGY OFFICE OF NUCLEAR REACTOR REGULATION O A Rrz

O O O I NRR TECH'NICAL SPECIFICATIONS IMPROVEMENT PROGRAM

                                     ~

e PHASEI i l Problem identification and l x Recommendations, TSIP Report L \ El l PHASE II implementation, TSCB m

'O PHASE I PROBLEM IDENTIFICATION AND RECOMMENDATIONS BACKGROUND 1982 PROPOSED RULE CHANGE NUREG-1024 TSIP AND AIF REPORTS (SECY 86-10)

                          -                                  COMMISSION MEETING, FEBRUARY 1986 O                         -                                   ACRS BRIEFING, FEBRUARY 1986 THREE CATEGORIES OF PROBLEMS IDENTIFIED LACK 0F WELL DEFINED CRITERIA FOR TECHNICAL SPECIFICATIONS RELUCTANCE TO USE OTHER TOOLS TO DOCUMENT AND ENFORCE REQUIREMENTS HUMAN FACTORS AND OTHER TECHNICAL WEAKNESSES O                      .                                                          s-arr i

4

     'O RECOMMENDATIONS COMMISSION POLICY STATEMENT TO DEFINE SCOPE AND PURPOSE OF TECHNICAL SPECIFICATION REVISE STS TO CORRECT CURRENT HUMAN FACTORS AND TECHNICAL WEAKNESSES O

CONTINUE DEVELOPMENT AND USE OF PRA FOR TECHNICAL SPECIFICATIONS . UPGRADE THE USE OF OTHER TOOLS INCLUDING 10 CFR 50.59 i:  ! i t i

I ! O O O ' i ! PLAN TO REFORNi NRC REO.UIREMENTS j RELATED TO THE TECHNICAL SPECIFICATIONS OLD SYSTEM _ NEW SYSTEM , , e e I l I I

1. :=,--- =

l l

                                ^
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ntwSt ate 8KAFDRY .t I

] REQiNninetNTS 3 8 4 NRC LKENSEE ' A9M9090SL N0Nin00NTATF0W U1 1 - _ _ _ ._

IMPLEf1 ATION SER FOR PLANT SPECIFIC E LIC. AfiEilD REVISED TECHNICAL SPECIFICATIONS SER FOR REVISED STS A (g POLICY STATEMENT a N ISSUED SUMt1ER SIX MONTHS INDIVIDUAL LICENSE 198 t AMENDMEIIT APPLICATIONS

e CURRENT All, OMlERS TECHNICAL 1937 GROUP SPECIFICATIONS IllDUSTRY REVISED STS m

                                                                               -__     _m-____.__   _ _ _ _ _ _ _ _

f 7 i t_ t L. q

O -

i l POLICY STATEMENT OVERVIEW l-i i t  : HOW WAS THE POLICY STATEMENT DEVELOPED? e E j BASED ON RECOMMENDATIONS OF EARLIER STUDIES 1 l i CRITERIA VAllDATED BY TRIAL USE STUDIES & RISK EVALUATIONS FINE TUNING BY NRC/ INDUSTRY WORKING GROUP I MEETINGS i* i [ f i - i l ' l l I i r O parr t _ re, rem __ _ _ mis-=--

8 O POLICY STATEMENT OVERVIEW (CON'T) WHAT IS IN THE POLICY STATEMENT? SUBJECTIVE STATEMENT OF THE SCOPE AND PURPOSE OF TECHNICAL SPECIFICATIONS OBJECTIVE CRITERIA FOR DETERMINING CONTENT OF REVISED TECHNICAL SPECIFICATIONS O

• 4 ADDITIONAL SYSTEMS IDENTIFIED PLANT SPECIFIC AND GENERIC RISK CONSIDERATIONS GUIDANCE FOR IMPROVING TECHNICAL SPECIFICATIONS BASES CONTROL MECHANISMS ENFORCEMENT O A -279

. 9 e i i, , IO POLICY STATEMENT MILESTONES ( ' TSIP/AIF REPORTS OCTOBER 1985 l i ! COMMISSION MEETING /ACRS BRIEFING FEBRUARY 1986 l t PROGRAM PLAN MARCH 1986 i i . TRIAL USE/ RISK EVALUATION MARCH 1986 i j FIRST DRAFT ISSUED MAY 1986 [ 1 1 ACRS MEETING JULY 1986

CRGR REVIEW AUGUST 1986 l  !

i  : ISSUE FOR PUBLIC COMMENTS OCTOBER 1986 I i ! POLICY STATEMENT ISSUED JANUARY 1987 t I f I b"S?b \ i j I i . --- --

PURPOSE THE PURPOSE OF TECHNICAL SPECIFICATIONS IS T0 IMPOSE RIGID CONDITIONS OR LIMITATIONS UPON REACTOR OPERATION NECESSARY TO OBVIATE THE POSSIBILITY OF AN ABNORMAL SITUATION OR EVENT GIVING RISETOANIMMEDIATEy0PUBLICHEALTHANDSAFETY, L T## l i CRITERIA

1. INSTALLED INSTRUMENTATION THAT IS USED TO DETECT, AND INDICATE IN THE CONTROL ROOM, A SIGNIFICANT ABNORMAL DEGRADATION OF THE REACTOR COOLANT PRESSURE BOUNDARY OR;

2. A PROCESS VARIABLE THAT IS AN INITIAL CONDITION OF A DESIGN

BASIS ACCIDENT OR TRANSIENT ANALYSIS THAT EITHER ASSUMES i THE FAILURE OF OR PRESENTS A CHALLENGE TO THE INTEGRITY OF

! A FISSION PRODUCT BARRIER OR;

3. A STRUCTURE, SYSTEM, OP. COMPONENT THAT IS PART OF THE l 4 PRIMARY SUCCESS PATH ANT WHICH FUNCTIONS OR ACTIV+T9ES TO l MITIGATE A DESIGN BASIS ACCIDENT OR TRANSIENT THAT EITHER

! ASSUMED THE FAILURE OF OR PRESENTS A CHALLENGE TO THE INTEGRITY OF A FISSION PRODUCT BARRIER, l I f G A-M/ , i..___ _ - - _ _ _ - _ - _ _ . _ - _ . _ _ _ _ _ . _ _ - -. _ . _ - __ _ .. _

j- 11 j i i l l 5 1 1 < l l l IMPACT OF POLICY STATEMENT . i I i [ LCOs: 40% REDUCTION

i. .

4 t i LERs: REPORTS ON THE RELOCATED TECHNICAL  ! SPECIFICATION REQUIREMENTS WILL NO l LONGER BE MADE

         @                                                                     \

RISK: NEUTRAL  ! E l

i e

i I s I u h

12 HANDOUT 1 WOLF CREEK - SAFETY SYSTEMS WHICH MEET CRITERIA Captured by Criterion 3.1.1 Reactivity Control Systems 2 3.1.2 Boration Systems 3 3.1.3 Moveable Control Assemblies 2 3.2 Power Distribution Limits 2 3.3 Instrumentation 3.3.1 - Reactor Trip System 3 3.3.2 - Engineered Safety Features 3 3.3.3 - Radiation Monitoring for Plant Ops. 1,3 3.3.3.6 - Accident Monitoring 3 3.4 Reactor Coolant System 3.4.1 Reactor Coolant Loops and Coolant Circulation 2,3 3.4.2.2 Safety Valves (Modes 1, 2, and 3) 3 3.4.3 Pressurizer B/U Heaters and Water Level 2,3 3.4.4 Relief Valves (All PORVs and Block Valves) 3 3.4.5 Steam Generators (Modes 1 through 4) 3 3.4.6 Reactor Coolant System Leakage 1,2 3.4.8 Specific Activity 2 3.4.9 Pressure / Temperature Limits (RCS) 2 3.4.10 Structural Integrity 3 3.5 Emergency Core Cooling Systems 3,2 3.6.1 Containment Systems - Primary Containment 2,3 lV 3.6.2 Depressurization and Cooling Systems Containment Isolation Valves 3 3.6.3 3 3.7.1.1 Safety Valves 3 3.7.1.2 Auxiliary Feedwater System 3 3.7.1.4 Specific Activity 2 3.7.1.5 Main Steam Line Isolation Valves 3 3.7.3 Component Cooling Water System 3 3.7.4 Essential Service Water System 3 3.7.5 Ultimate Heat Sink 3 3.7.6 Control Room Emergency Ventilation System 3 3.7.7 Emergency Exhaust Systems 3 3.7.12 Area Temperature Monitoring 2 3.8 Electrical Power Systems 3 3.9 Refueling Operations 3.9.3 - Decay Time 2 3.9.4 - Containment Building Penetrations 2,3 3.9.7 - Crane Travel 3 3.9.9 - Containment Ventilation System 2,3 3.9.10 - Reactor Vessel and Storage Pool - Water Level 2 3.9.13 - Emergency Exhaust System 3.10 Special Test Exceptions 2 3.11.1.4 Liquid Holdup Tanks 2 3.11.2.6 Gas Storage Tanks 2 O A-M3

13 HANDOUT 2 LIMERICK - SAFETY SYSTEMS WHICH MEET CRITERIA Captured by Criterion LCO 3.1 Reactivity Control Systems LC0 3.1.1 Shutdown Margin 2 LC0 3.1.2 Reactivity Anomalies 2 LCO 3.1.3 Control Rods 2,3 LCO 3.1.4 Control Rod Program Controls 2 LCO 3.2 Power Distribution Limits 2 LC0 3.3 Instrumentation LC0 3.3.1 - Reactor Protection System Instru. 3 LCO 3.3.2 - Isolation Actuation Instrumentation 3 LC0 3.3.3 - Emergency Core Cooling System 3 Actuation Instrumentation LCO 3.3.4 Recirculation Pump Trip Actuation Instru. LCO 3.3.4.2 - End of cycle Recirculation Pump Trip 3 System Instrumentation LC0 3.3.6 Control Rod Block Instrumentation 2 LCO 3.3.7 Monitoring Instrumentation LC0 3.3.7.1 - Radiation Monitoring Instrumentation 1,3 LCO 3.3.7.5 - Accident Monitoring Instrumentation 3 LCO 3.3.7.6 - Source Range Monitors 2 LCO 3.4 Reactor Coolant System LCO 3.4.1 Recirculation System 2 LCO 3.4.2 Safety / Relief Valves 3 LCO 3.4.3 Reactor Coolant System Leakage 1,2

 /9   LCO 3.4.5     Specific Activity                               2             .
 \  / LC0 3.4.6     Pressure /TemperatureLimits(RCS)                2 V

LC0 3.4.7 Main Steam Line Isolation Valves 3 LCO 3.4.8 Structural Integrity 3 LCO 3.5 Emergency Core Cooling Systems 3 LC0 3.6 Containment Systems 3 LC0 3.6.1 Primary Containment 2 LCO 3.6.2 Depressurization Systems 3 LCO 3.6.3 Primary Containment Isolation Valves 3 LC0 3.6.4 Vacuum Relief 2 LCO 3.6.5 Secondary Containment 2 LCO 3.7 Plant Systems LCO 3.7.1 - Service Water Systems LCO 3.7.1.1 - RHR Service 3 LCO 3.7.1.2 - Emergency Service Water System 3 LCO 3.7.1.3 - Ultimate Heat Sink 3 LCO 3.7.2 Control Room Emergency Fresh Air Supply System 3 LC0 3.8 Electrical Power Systems 3 LC0 3.8.4.3 Reactor Protection System Electrical Power 3 Monitoring LCO 3.n Refueling Operations LCO 3.9.1 - Reactor Mode Switch (Mode 5) 2 LCO 3.9.4 - Decay Time 2 LC0 3.9.6 - Refueling Platform 2 LCO 3.9.7 - Crane Travel 2 LCO 3.9.8 - Reactor Vessel Water Level 2 LC0 3.9.9 - Spent Fuel Storage Pool - Water Level 2 V LCO 3.10 Special Test Exceptions 2 LCO 3.11.1.4 Liquid Holdup Tanks 2 LCO 3.11.2.6 Main Condenser 2 A_

14 O WOLF CREEK / LIMERICK - LIMITING CONDITIONS FOR OPERATIONS WHICH DO NOT MEET THE CRITERIA AND CURRENTLY HAVE ACTION STATEMENTS THAT LIMIT REACTOR POWER. WOLF CREEK LC0 3,3,3,5 REMOTE SHUTDOWN INSTRUMENTATION LC0 3,4,7 CHEMISTRY LC0 3,4,9,2 PRESSURIZER - PRESSURE / TEMPERATURE LIMITS LC0 3,4,11 RCS VENTS LC0 3,6,4,1 HYDROGEN ANALYZERS O LC0 3,6,4,2 LC0 3,6,4.3 HYDR 0 GEN CONTROL SYSTEMS HYDROGEN MIXING SYSTEMS LC0 3,7,1,3 CONDENSATE STORAGE TANK (PLANT SPECIFIC) LC0 3,7,2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LC0 3,7,8 SNUBBERS LC0 3,8,4,1 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES O s-aw

15 I v) LIMERICK

  'LC0 3.1.5          STANDBY LIQUID CONTROL SYSTEM
  *LC0 3,3,4,1        ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION
  *LC0 3.3.5/3,7,3 RCIC SYSTEM ACTUATION INSTRUMENTATION LCO 3,3,7,4      REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND       -

CONTROLS LC0 3.3,9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION LC0 3,4,4 CHEMISTRY LC0 3,6,1,5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY LC0 3,6,2,2 SUPPRESSION POOL SPRAY LC0 3,6,6,1 PRIMARY CONTAINMENT HYDROGEN RECOMBINER SYSTEMS LC0 3,6,6,2 DRYWELL HYDR 0 GEN MIXING SYSTEM LC0 3,6,6,3 DRYWELL AND SUPPRESSION CHAMBER 0XYGEN CONCENTRATION

  *LC0 3,7,3           REACTOR CORE ISOLATION COOLING SYSTEM LC0 3,7,4         SNUBBERS LC0 3,8,4,1       ELECTRICAL EQUIPMENT PROTECTIVE DEVICES
   ' RETAINED AS TS REQUIREMENT BASED ON OPERATING EXPERIENCE AND RISK INSIGHTS, o                                    n ,,

e APPENDIX XX REACTIVATION OF NUCLEAR POWER PLANT CONSTRUCTION PROJECTS NUREG-1205 l Reactivation of Nuclear Power Plant

;                       Construction Projects: Plant Status,
 !                      Policy issues, and Regulatory Options i

l Miller B. Spangler 1 Special Assistant for Policy Development Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Washingtono D.C. 20$$$ I i Briefing of the Advisory Committee on Reactor Safeguards i l July 11, 1986 l l r O A -291'

NUftEG.1206 - CONTENT $ P,a21 (mv) AB5 TRACT..............................................................

SUMMARY

iii ix

1. STUDY SCOPE AND OBJECTIVES....................................... I

2. DETERMINANTS OF STAFF REQUIREMENTS AND ORGANIZATIONAL C00R0lNAT10N...................................... 3 2.1 The Number of Reactivations and Their Timing................ 3 2.2 Reactivation Options Chosen and Plant Status Considerations.............................................. 3 2.3 The Number and Difficulty of Technical and Environmental Issues........................................ 3 2.4 NRC's Choice of Regulatory Options for Treating Reactivation issues................................ 3 2.5 Organizational Coordination................................. 5

3. STATUS AND OUTLOOK FACTORS OF CANCELLEO OR DEFERRE0 PLANT 5.................................................. 7 3.1 Status Summary of Units with Cancelled Construction......... 7 3.2 Status Summary of Units with Deferred Construction.......... 14

3.3 Limerick-2

The Case of a Deferred Unit That Has Been Reactivated............................................ 22

4. IDENTIFICATION OF POSSIBLY RELEVANT SAFETY AND ENVIRONMENTAL 155UES ARl51NG FROM 51GNIFICANT NEW INFORMATION.................. 29 4.1 Safety Issues............................................... 29 4.2 Environmental Issues........................................ 33

5. IDENTIFICATION OF REGULATORY OR POLICY ISSUES OF POSSIBLE RELEVANCE 10 THE ADEQUACY OF EXISTING RULES AND POLICIES OR THEIR NEED FOR REV1510N.......................................... 37

6. REGULATORY OPTIONS AND DECISION CRITERIA......................... 43 6.1 Decision Criteria and Regulatory Purposes Guiding LWR Reactivation Policy Development............................. 43 6.2 Regulatory Options for Dealing with Reactivation Issues..... 4B 6.3 Potential Impact of Research Programs on the Choice of Regulatory 0ptions....................................... 54 6.4 The Image Problem of Identifying and Choosing between Alternatives to Deal More Effectively with a Complexity of Regulatory issues........................................ 58

7. REFERENCES....................................................... 63 APPEN0lX A - FORECASIS OF ELECTRICAL ENERGY DEMAND AND NEE 0 FOR A00E0 PLANT CAPAC11Y,........................... ..................... 67 1, Explanatory Note.... .. ......... ............ ......... 69

           !!. Ten Year Projections and Analyses by the North Afrerican Electricity Reliability Council of U.S. Electric Power Supply and Dernand Plus Need f or Added Plant Capacity. . . . . . . . . . . . . . . 71 III. Ten Year Projections by the U.S. Departrnent of Energy /EIA of Domestic Energy Consumption and Supply by Sector and Fuel Type........................       . ............................              91 IV. Excerpts from the Report of the Edison Electrical Institute on the Outlook f or Reopening the Nuclear Option. . . . . . . . . . . . . 101 APPEN0lX B     ORDERS OF THE PENNSYLVANIA PUBLIC UTilliY COMMISSION BEARihG ON THE DEC1510N PROCESS BY WHICH THE LIMERICK-2 NUCLEAR GENERAllNG STATION WAS REACTIVIATE0 ON DECEMBER 23, 1985......                    . 111 A-AW

O O O SCOPE OF PRFL1MINARY STUDY OF LWR REACTIVATION ISSilES TASK 1: PREPARE A TABULATION OF CURRENT STATUS OF CANCELLED OR DEFERRED PLANTS TASK 2: IDENTIFY SAFETY AND ENVIRONMENTAL ISSilES ARISING FROM SIGPIFCANT NEW INFORMATION TASK 3: IDENTIFY REGULATORY OR POLICY ISSUES AND INFORMATION USEFUL TO A DETERMINATION OF THE ADEQUACY OF EXISTING RULES OR POLICIES AND w THE DESIRABILITY OF IMPROVEMENTS g3 TASK 4: IDENTIFY REGULATORY OPTIONS AND DECISION CRITERIA FOR DEALING WITH

       '                             THE AB0VE ISSilES 3,

NO

     '4D            TASK 5:           IDENTIFY DECISION CONSIDERATIONS THAT WOULD DETERMINE STAFF REQUIREMENTS

O O O

10. 11. 12.

7. 8. 9. , + ADVANCES IN:

STATUS OF FINANCtAL + ELECTRICITY 4 OTHER + SOctO-4 g C DEMAND C REGULATORY C POLITICAL A, CANCELLED C FACTORS - - # OR DEFFERRED

                         ~                        ~

GROWTH FACTORS FACTORS , r TECHNOLOGY SCIENTIFIC PLANTS INon-N RC) KNOWLEDGE

                                        \*  j o              , o' '            ,o"  r
                                                                                                   ,.+,ge

3. + or - [

UTILITY 5. ~ ADEQUACY e

                                                    -[+ or.

I CHOtCE OF '

• I RE ACTIVATION I OF CURRENT NR'C

                                                                             ~

CHOICE OF l I REGULATORY

                               /S                                                  OPTIONS o
                                                                                                                                     \

I I s. h I 2. NO.& 1 NO. b TYPE OF TECH 1 l l TIMING OF l , l

                          ' LWR REACTI 3
                                                                                                                                    & ENVIR*L                     j C                                                                                                                                 ISSUES ve                                                         ,g, VAT 10NS g                                                   ,f,,,_                                               "

D 1. NRC STAFF REQ'TS TIMING

                                                                                                                  ^
                                                                                     + or -

1 variable _ l l Figure 1 --A conceptual model Of interrelationships between factors and events impacting utility choice of I LWR reactivation options and NRC choice Of regulatory options and staffing requirements. ] 1 1 l

8 . gg

                                       }
                   \                                 Tabla 1. Statn of lists alranted Constrwetion Permits                                            trwction has been canc';Iled a v

(Based on reconnaissance level informatton; data not validited) P't Ordered Sta t us of Onu te f *Pa ssau at or Mt t s Site Disposl*n PUC or State-tewel Issues / Actions Reactor teactor CP Cancel ' ' ' ' " .s 19ul or Disposed va l e t' Reactor 17mit Tyne vendor Issued Date M.eintenan:e ftt Monit- or Futwre Use-P TI I I-Millions Program oris9 At T 6se of Cancel 5tatus Change M555) TG l cMI FWI W' 7 leo equifuneet NA NA NA Awa d able lef P none VogtIe 3 6 74 9-8 2-74 NA NA NA Available MfP L.,ne Vogtle 4 PWR We 6-28-74 9-8 2-74 0' delivered , Sorry 3 rvu usw 87-20-F4 2-38-77 7 No egul Puent NA NA NA Avellable NfP, financial teone Sorry 4 PvR B&W 32-20-74 2-58-77 ff dellwered MA NA NA Available NFP. financial Mae 8 Equipment sold NA NA NA Available NfP, financial scor.e foorth Anna 3 PWR 84W 7-26-74 6-22-83 leorth Anna 4 Fwt g&w P-76-74 13-2s-e0 0 or scrapped NA NA NA Available MfP. f Inancial Mene __ O som equipment NA NA NA Available MfP, financial Mone Jamesport & FwR we 3-4-79 t-89-so Jamesport 2 PWR We 3-4-F9 3-19-80 g dejleeged NA NA NA Available MfP. ftnancial Mone ' 4C** FI Medium Routine No Available,F2 gefr Isone Marris 2 FWR We 3-27-78 32-28-81 NfP Isone Harris 3 PWR ve 3-27-78 12-t8-41 l'# Fl Medium Routine No Available,F2 , . Harris 4 Fwt we 3-27-78 32-18-81 Ic.e r1 Medium Routine No Avallable.F2 MfP None j Cherchee ! Pwn C-E 12-3o-77 6-29-43 18 Ealutpoent sold IsA NA NA Sold lefP MA Cherokee 2 PvR c-E 82-3o-77 33-2-a2 0 or scrapped NA NA NA Sold NfP NA C rence 3 Pva C-E 12-10-77 In-2-e 2 O NA esA NA Sold Mfp NA arN e 64ill ! PWR we 4-4-is I F- 33 -e 3 55c.e 901 of major I.arge F) F3 Undet'd, F4 Financial, tate, and F5 I*arble Mill 2 Pvt we 4-4-78 R2-18-41 35e.' equip't onette t.arte F3 F3 Undet'd. F4 NfP tasues F5 Routine peo DOE Lease per 6 financtat saaves leone W6F-4 PWR B6W 2 78 t-22-42 26' F6 t.arge No FF mir & financtet tas.es geone i if fr-5 PWR C-E 7-II-78 3-83-84 Ig' F6 ta rge Mintoal Irrone I FbR We 12-27-F7 7-24,y, O R R R NA NA NA Available F8 Nir, F9 Mone we O Iso egulp't del. NA NA lea Avallebte MfP FIO Sterling 1 Fwn e-26-FF s-2s-so F-10-73 11-6-80 MA NA NA Available Fil Financial /TMI-2,WIP Isone ! Forted River FUR C-E 5 R R R Callauay 2 PWR We 4-86-76 30-9-88 1 R R R Medium Routine 3eo Avstlable NfP leone m Small SP Iso tfndetered,Fl2 Financial issues FIS Bailly I BWR CE s-2-74 s-at 1 R R R Hope Creek 2 sua CE 8 8 74 82-29-88 18' P14 Samall lea teo Unavail..FIS ran actal & rate f asees teone Small NA Available Nfr Isone Cisnton 2 gwR cE F-y4-y6 to-84-83 BC R R F16 NA River Bend ? OUR CE 3-25-77 t-29-84 tc.e FIF R Fl? Medium Routine Iso Available MfP, financial & rate Isone l IlwR CE 10-27-72 5-21-44 93c FIS FIB FIS Larpe.F19 NA NA Gal converstee MA NA j l i.- -- leone Phipps Send I swa CE t 86-78 2-16-83 29' R R F20 1.arge SP, Routine Iso Up for sale Mir F20 Up for sale ufP leone Phi s Bend 2 SWR CE t-86-78 2-16-83 5' R R larec SP. Routine see 3F 1.arge SP, Routine No Available NfP leoce i I e ow Creek I BWW C-E I4-29-F8 s-29-84 NA NA NA leo Available Mfr leone Q Yellow Creet 2 BWit C-E 33-29-7s s-29-84 3' 44' lea NA NA NA NA NA lAree Large SP. Routin, Routine too Available Mir peone Q llartsville Al SWR BUR CE CE 5-9-77 8-2944

                                                                                            %-9-77      S-29-44      34'      NA     NA ISA    Large              Routine        geo      Available        ufP                             Isone N    llartsville Al Routine                 Available        NfP                             None flertsville 81                       BWR     CE                         5-9-77        3-22-81    17'      NA     NA   NA   larPe                             les 3-22-83       7'            NA   NA   tJrge              Routine        Iso      Available        Mir                             Mone Itartsville B2                       BWR     CE                         S-9-77                            NA
   *

• tegend MS$$ - leuclear Steam SePyly System; TC - Turbine-Cenerator: OIE - Construction Meteriale and Equipment; IEA - Isot Applicable; SP - Site. etab -

t!! sttee Plang R - Resold, cancelled before delivery, or screPped; PUC - Public tft111ty Commiselong MfP - leeed for Plant toeve; F1 - Footnote 1. b Purchase price: escludes installation or maintenance costs (large - over $100 million; medium - $10 to 100 mit!!on; small - less than $10 million).

• OL application recetwed.

Available means- there are no plans to sell the site and no lincun constraf ats to its ultimate use for a coal or nuclear plant addition. e No CP cancellation by MRC has yet been issued for this unit. f See nest pate for add tt fonal footnoths.

Additirnal Nat*s far Trble 1 TI- Tor Estris-2. 981 cf NS$$ d211vsrsd to site, but cuch Isss for units 3 & 4 (reactor vessel and internals, piping, pumps, etc.); yard sale yielded little sales; TGs purchased for all' units and stored of f site in Penna, and some piece of rotors

 .                sold.

T2- Site use for a possible coal plant addition may be feasible if plant would not exceed ( regional air pollution limits (already pressed) under EPA bubble concept; su'.tability of site for ash ponds not studied. T3- About 5-102 of onsite equipment has been sold by PSI from both units 1 & 2 under Investeent Recovery Program; protective maintenance of remaining equipment monitored by onsite 16E inspector until a year ago. T4- While a consortium has considered completion of Marble Mill Unit 1 no offer to purchase has been made. FSI states that there are no current or contemplated negotiations for such a sale. The PSI'has requested cancellation of the construction permit and is currently awaiting NRC's decision. PSI has reviewed converting Marble Mill to coal and while there are difficult problems with conversion of a nuclear unit. PSI has not judged any of these problems to be insurmounta,ble from an engineering standpoint. PSI has no plans to convert Marble Hill to coal-fired units. T5- Investment recovery by PSI threatened by adverse Indiana Suprese Court ruling against rate-base decision by the PUC to permit investment recovery of cancelled Bailly plant, a ruling which may be challenged at the U.S. Supreme Court level. T6- The combined acquisition value of delivered equipment (including some sof tware services) for WNP-4 and WNP-5 is $558 million, of which $95 million (acquisition value) was sold at a recovery price of $20 million; the remaining equipment is onsite. T7- A future coal-fired plant is not a practical option for the WNP-5 site and is not being considered for the future use of this site; they do not foresee appreciable growth in electricity demand over next 20 years; WNP-5 is a stasese twin to WNP-3 at the Satsop site. T8- Site for Tyrone-1 is regarded ideal for nuclear plant and not so good for coal with inadequate rail service. T9- At time of cancellation, unable to obtain permit free State of Wisconsin because of FUC evaluation of inadequate need; part of Northern States Power Co. service area is in Wisconsin per a wholly owned subsidiary there. Flo- Construction Fernit for Sterlir.g-1 was rstracted by State of New York because of changed evaluation of need by the PUC; it is now forecasted that a baseload plant j addition will be needed by 1995. ! T11- A study in progress for installing either a coal plant or a vaste recovery (refuse combustion) plant at the Torked River site with resolution undetermined arising. in part, from NfF and financial issues. T12- Bailly site is inadequate in size for adding a coal plant, but could accommodate peaking units in addition to the coal unit already onsite; no interest in reviving a nuclear unit for the site. T13- Possible appeal to U.S. Supreme Court for a reversal of Indiana Supreme Court decision denying investment recovery in the utility's rate base for abandoned nuclear construction work. Fl&- Practically all of the equipment delivered to the Hope Creek-2 site was either scrapped. , sold or used for spare parts with less than 85 million (acquisition value) remaining to be sold; the generator was used in the Sales plant and the reactor pressure vessel was cut up and scrapped. F15- The part of the site designated for Hope Creek-2 is now occupied by support service

              . buildings and, therefore, site is unavailable for adding a nuclear plant in the future; a coal plant addition could be accommodated at the site if adjacent land is purchased.

F16- The TG and NSSS for C11nton-2 was ordered but cancelled and never delivered; such of the remaining equipment delivered to site is warehoused for use as spare parts for Unit 1. T17- Major components of the River Bend-2 NSSS being housed at the site (inerted reactor pressure vessel, psamps, motors, etc.) with no plan to sell; TG not fabricated; substantial quantities of CNE warehoused both onsite and offsite. i FIB- Major equipment for Zimmer remains onstle; however, control tuds and nuclear fuel ! removed from site and some equipment sold; the main steamline was cut and capped ! and the reactor building (with substantial equipment inside) was sealed and written off for tax purposes. l T19- Sealed of f equitment in reactor building has large, but unrecoverable value since l ! remainder of plant is being converted to coal use. T20- The reactor vessel for Phipps Bend 1 & 2 was cut up and sold for scrap; other equipment delivered to site was either scrapped or will be scrapped.

                                                                                ~

b 6 l - . - - - - - - .- _ _ - - _ - . - - - -_-- -.

e .

                  \                             Yable 2. Stctuo af LWRJ granted Construc u 'Permito whose cenetruction hei been dettered.                  (

(Based on reconnalemance level information; data not fully validated) Defer- Constr'n Equipment Statue of Onette Equipment or Mtle Site Die- FUC or State-Isvel issues or Actione 1.WR Reactor ven- Issue Progress Ordered Valu,0 Maintenance I&E Monit- posal or Reactor Unit tal 8 oring Future Use At Time of Deferral Status Change Type dor Date Date (1) Or Resold $-M1111one ProRram Large EPP Yes Undeter- Rate leeues Resolution of Grand Culf 2 BWR CE 9-4-74 8-15-84 35 (All major Financing lessnea rate issue for equipment mined (MK Need teoue Unit 1 111) on ette) C large EPP Yes Undeter- Rate leaves Seabrook 2 PWR We 7-7-76 9-25-84 24 (All major No change equipment eined Financing teoues

• on ette)

Need teoue 44d Large EPP Yes NA Rate issues Perry 2 SWR CE 5-3-77 7-17-84 (All major Need feeue No change equipment e (MK 111) on ette) C Large EPP Yes Site Need feeue WNP-1 PWR 84W 12-24-75' 1981 63 (All major No change equipment leased Financing leaues on ette) from DOE (All major Large EFF Yes Undeter- Meed issue y WNP-3 PWR C-E 4-11-78 March 76C Financing leaves No change 1983 equipment mined ' on ette) Large EPP Yes Undeter- Rate issues IGearing in Midland 1 PWR B4W 12-15-72 7-15-84 85C (All major Financing issues Spring 1986 on equipment mined Conversion study rate tooues f on site) I Yes Undeter- Ditto Ditto Midland 2 PWR B4W 12-15-72 7-15-84 85* (All major Large EPP equipment eined ! I on ette) 1

• Legend: N555 - Nuclear Steam Supply System; TG - Turtilne-Generator; CME - Construction Materials and Equipment; N A - Not Appilcable;

! Ord - Ordered; R-Resold; EPP-Equipment Preservation Program; PUC-Public Utility Comunission. b Purchase price; excludes installation or maintenance costs (large - over $100 million; medium - $10 to 100 milifon; small - less than $10 stillon). C OL appitcation received. dPerry-2 constructiort excludes fact 11tles in comanon with Perry-1.

                ' Full security measures in place for Perry-1 and Perry-2.

I CP empired 12-1-84; extension requested on 5-24-84 for completion by 12-1-89. 9CP empired 7-1-84; extension requested on 9-11-84 for completion by 7-1-89 h Construction of unit 2 halted on 4-19-84.

                  ' Resale is an option under consideration.

POTENTIAL SAFETY AND ENVIRONMENTAL ISSUES ARISING FROM SIGNIFICANT NEW INFORMATION A. SAFETY ISSUE EXAMPLES: 0 THERMAL HYDRAULIC BEHAVIOR (IN-VESSEL 8 EX-VESSEL) 0F SEVERE ACCIDENT SEQUENCES . 9 SOURCE TERM AND FISSION PRODUCT BEHAVIOR 4 CONTAINMENT RESPONSE TO COREMELT ACCIDENTS 0 EMERGENCY INSTRUMENTATION AND EQUIPMENT 4 VARIOUS BACKFITS ARISING FROM THE TMI ACTION PLAN (NUREG-0737) 0 HYDR 0 GEN CONTROL ISSUES 0 RESOLUTIONS FOR USIs AND HIGH TO MEDIUM PRIORITY GENERIC SAFETY ISSUES 0 INTERGRANULAR STRESS CORROSION CRACKING AND PIPE BREAK PHENOMENA INCLUDING FRACTURE MECHANICS

,      8  DIVERSITY AND RELIABILITY OF DECAY HEAT REMOVAL SYSTEMS AND STATION BLACK 0UT ISSUES S  SAB0TAGE AND FIRE PROTECTION B. ENVIRONMENTAL ISSUE EXAMPLES:

0 ACCIDENT HAZARDS AFFECTING SITE SUITABILITY 0 DISCOVERY OF ENDANGERED SPECIES, ADDITIONS TO LISTS OF HISTORIC LANDMARKS, SCENIC RIVERS, ETC. 8 NEED-FOR-PLANT ISSUES i e ALTERNATIVE ENERGY ISSUES l 0 CHANGES IN WATER 8 AIR POLLUTION CONTROL REGULATIONS l 0 NEW COURT RULINGS AFFECTING TRADITIONAL REGULATORY OR REVIEW PRACTICES OR INTERPRETATIONS i e SIGNIFICANT CHANGES IN AREA DEMOGRAPHY AND LAND OR i WATER USES AFFECTING IMPACT ANALYSES l O e A -30f

O O O REGl1ATORY QlESTIGIS FOR ACHIEVING COST-EFFECTIE EASlEES OF EQUIPT1HT MINTDIANCE AND QUALITY PRESERVATION OF DEFERRED OR CMFIIFil NUCifAR POER PLANTS (1) Which equipment (e.g., as determined by advances in severe accident risk

assessment) is in greatest need of maintenance preservation measures and which is of low or insignificant priority?

(2) What ranges of environmental stresses (e.g., moisture, temperature, mildew, dust or other air pollutants) that may be encountered in mothballing or warehousing practices are the most threatening to quality assurance of the priority equipment identified in item I? (3) What possible stochastic events (e.g., vagrant or rodent activity, wind-blown missiles, fire, flooding, etc.) could damage equipment or violate protective coverings exposing equipment to deleterious environmental conditions? (4) What does the length of equipment storage period signify for the severity of the QA/QC issues identified in itens 2 and 3? h (5) How effective are presently used QA inspection and auditing techniques and preoperational or start-up testing methods in uncovering inadequate main- { g tenance practices and,the state of corrosion, damage, or other quality

  .          deterioration of the safety prioritized equipment?

(6) What are the most cost-effective methods of dealing with the above issues and what should be the roles of NRC and industry in identifying and encouraging the use of the best available practices? (7) If some of the above issues are relatively unimportant, is there adequate information available to support this judgment?

KEY I a E PRDCEDURES FOR REVIEW 0F QUALITY ASSURANCE FOR E)0' ENDED CONSTRUCTION DElA ( IP-32050, JAN. 1, 1983 )

• QA procedures to meet inspection requirements are based on a current list that identifies the location, storage level, and preventive main-tenance requirements of all safety-related equipment and materials.
• I&E observations of QA/QC work require initial selection of four representative safety-related items that are sorted / retained at the site and, if applicable, an appropriate sample of off-site locations.

(Inspection every 6 months).

• I&E observations include inspection of protective coverings and coat-ings: cleanliness preservation; weather protection; fire protection; rodent protection; protection against unauthorized site intrusion; and preventive maintenance requirements established for safety-related activities.
• Review of QA records for the four selected representative safety-related items.
• Inspection guidance provides for flexibility in scope and frequency of audits /surveillances.

l

• The licensee's QA plan should include provisions for additional mea-s sures if duration of construction delay exceeds the time previously established. (Different preservation requirements may be necessary if storage time is significantly increased.) -
• Much of QA program for extended construction delay is expected to be similar to that for active construction but additional surveillance procedures would be expected. (If construction is over half completed, quarterly inspection may be required).
• QA/QC requirements will vary from site to site depending on construction status and environmental factors such as temperature variations, atmo-spheric pollution, or proximity of site to salt water.

i

• Sample selection for I&E observations should reflect the importance of the " activity" to safety; the detection of unusual conditions would warrant additional observation or evaluation.
• Preventive maintenance should include periodic exercise of valves and rotating machinery; lubricants and dessicants changed as appropriate.
• The listing of equipment "important to safety" requires updating during construction suspension.

s

                                                                                                  ~

10

g. - . . . - - - - -, ,-,,,%- - - , - , - - . , - - -, ,-

PROPOSED OBJECTIVES OF LWR REACTIVATION POLICY

1. TO IMPROVE STABLITY AND PREDICTABILITY OF REGULATION TO ENC 0URA CHOICES OF LWR REACTIVATION OPTIONS,

2. TO CLARIFY REGULATORY PROCEDURES AND REQUIREMENTS FOR COST-EFFE PRESERVATION AND SITE STABILIZATION PROGRAMS FOR LWRs WITH DEFE l

CONSTRUCTION, AS APPROPRIATE,

3. TO DEVELOP AN INTEGRATED PLAN T0 IMPROVE AND IMPLEMENT A REGU EFFECTIVELY WITH THE PRESENT OR ANY EMERGENT SAFETY OR ENVIR

   ~

CP OR OL PHASES OF PROJECT REACTIVATION,

4. TO AVOID IMPOSING ANY UNNECESSARY BURDEN OF INFORMATION OR i ON BEHALF 0F REGULATORY OR OTHER MEASURES OF MARGINAL OR DUB TECTING SAFETY AND ENVIRONMENTAL VALUES,

5. TO ENSURE THAT NRC's REVIEW PROCESS WILL CONTINUE TO PROVIDE ACCESS FOR EXPRESSING SIGNIFICANT CONCERNS REGARDING LWR REACTIVATION, I

N

1 O O d i DECISION CRITERIA FOR EVALUATING REGULATORY OPTIONS IN LWR REACTIVATION POLICY 4 CRITERIA AS PROVIDED IN THE NEW BACKFIT RULE (10 CFR PART 50,109) l APPLICABLE SITUATIONS i 0 CRITERIA AS PROVIDED IN REGULATORY ANALYSIS GUIDELINES (NUREG/BR-0058, REV, 1) L FOR APPLICABLE SITUATIONS, 4 ' S OTHER PERFORMANCE CRITERIA: HOW WELL THE REGULATORY OPTIONS SERVE THE OBJECTIVES OF THE LWR REACTIVATION POLICY AND HOW WELL THESE HARMONIZE WITH THE EXTANT NRC REGULATIONS AND THE INTEGRATED PLAN FOR POLICY h DEVELOPMENT (E.G., THE SEVERE ACCIDENT POLICY, STANDARDIZATION POLICY,

b SAFETY GOAL POLICY, SITING POLICY, POLICY ON EMERGENCY PREPAREDNESS, k AND POLICY FOR ENVIRONMENTAL PROTECTION)

I l i

i OPTIONS FOR IMPROVED REGULATORY REVIEW 0F LWR REACTIVATION PROJECTS ! A. PLANTS WITH CANCELLED CPS: I. CASE-BY-CASE REVIEW H GENERIC RULEMAKING FOR SELECTED SAFETY AND ENVIRONMENTAL ISSUES (E.G., ALTERNATIVE SITES, ALTERNATIVE ENERGY SOURCES, NEED-FOR-PLANT, WATER-RELATED IMPACTS, SITE-RELATED ACCIDENT HAZARDS, ETC.)

2. ONE-STEP CP/0L HEARING AND. LICENSING PROCESS H A TWO-STEP PROCESS

3. POLICIES AND REVIEW PROCEDURES FOR STANDARD PLANT l DESIGNS (REPLICATE PLANTS, DUPLICATE PLANTS, FDAS, OR DESIGN CERTIFICATIONS OF REFERENCE DESIGNS THROUGH RULEMAKING H A CUSTOM PLANT

4. POLICIES AND REVIEW PROCEDURES INVOLVING A RELATIVELY COMPLETE DESIGN WITH A FULL-SCOPE PRA H A DESIGN AND PRA 0F LIMITED SCOPE

5. OPTIONS FOR REVISED RULES, REGULATORY GUIDELINES, POLICIES, TECHNICAL SPECIFICATIONS, STANDARD REVIEW PLANS, HEARING GUIDANCE, LEGISLATIVE OR ADMINISTRATIVE REFORMS, ETC. TO IMPROVE REGULATORY j

EFFICIENCY H THE CURRENT REGULATORY REGIME B. PLANTS WITH DEFERRED CONSTRUCTION l l ITEMS 3, 4, AND 5 ARE ALSO APPLICABLE TO THIS CATEGORY l OF POTENTIAL PROJECT REACTIVATIONS, BUT POSSIBLY WITH A DIFFERENT WEIGHTING OF DECISION CRITERIA pao 7 j 13

d POTENTIAL IMPACT OF RESEARCH PROGRAMS ON THE CH0 ICE OF REGULATORY OPTIONS 8 OPERATING REACTOR EXPERIENCE DATA 9 SEVERE ACCIDENT RESEARCH PROGRAM (NUREG-0900) 0 ORNL 8 INEL CONTRACT STUDIES ON INSPECTION AIDS TO DEVELOP TOOLS BY WHICH IE AND THE REGIONS CAN UTILIZE RISK INFORMATION TO ASSIST IN THE PLANNING OF INSPECTION ACTIVITIES (ANO-1 PRA PLUS OTHER NUREG-1150 PLANTS) 6 BNL CONTRACT STUDY TO PROVIDE A RELIABILITY AND RISK BASIS FOR EVALUATING TECHNICAL SPECIFICATIONS AND EXEMPTION REQUESTS AND TO SUPPORT NRR EFFORTS TO IMPROVE TECHNICAL SPECIFICATIONS, lD ESPECIALLY THE IDENTIFICATION OF TECH SPECS NOT IMPORTANT TO RELIABILITY / RISK AND TO IDENTIFY DOMINANT FAILURE MODES OF 1 SAFETY EQUIPMENT 4 PNL CONTRACT STUDY T0: DEVELOP GENERIC PROCEDURES FOR VALUE/ I IMPACT ASSESSMENT; PROVIDE REGULATORY ANALYSIS SUPPORT FOR SPECIFIC REGULATORY ISSUES; ORGANIZE AND DEVELOP INFORMATION l NEEDED TO SUPPORT RULEMAKING/STDS ACTIVITIES S -RESEARCH PROGRAM PLAN TO REVIEW EFFECTIVENESS OF LWR REGULATORY RE0'TS IN LIMITING RISK AND TO IDENTIFY CURRENT RE0'TS WHICH IF DELETED OR MODIFIED WOULD IMPROVE THE EFFICIENCY OF NRC REGULA-TION FOR NUCLEAR POWER PLANTS WITHOUT ADVERSELY AFFECTING SAFETY (49 FR 39066) O u A- 3/6

NERC intM*n'W"' NERC etecia,ecrutairiOs.vrutt (United States) (United States)

        %. . .                                    % , = im                                                         % .i e a, n im E5:3 HYDRO (III OILGAs C-] NUCLEAR M OTHER IE22 COAL OAS         / oit   ,i g I

eco-wrono l

                            ;,,      lgg, lll   lll lll III III                     UN                               orwen 0.9**

III III CLEAR i,, o NUCLEAA I g Y oTurn P7% y UN 4 $ 4 ost # N*.

                                                                                              .. s y

85 86 87 88 89 90 91 92 93 94 s us sus an a sa ee ow FIGURE 4 FIGURE 6 RETIREMENTS Planned unit retirements over the 1985-1994 period total 12,600 MW, essentially all of which is in the United able 5 NERC Generating Unit States. The planned retirements amount to about one Additions, Retirements year's load growth. Increasing attention is being given

                        & Conversions

• to extending the life of existing units and to improving 1985-1994 their availability. Such programs are influencing decisions on capacity additions and retirements and Thousands unned  % of  % of of MW states Totsi Canada Total account for some of the deferrals and cancellations The capacity mix of additions between 1985 and 1994 is Nuclear 46.3 40.9 6.4 42.7 shown in Figures 4 and 5 for the United States and Coal 42.2 37.3 2.6 17.3 Canada, respectively.

Hydro 3.2 2.8 5.4 36.0 Oil / Gas 6.7 5.9 0.6 4.0 Pumped PROJECTED ENERGY AND FUEL Storage 5.2 4.6 0.0 0.0 USE Cogeneration 6.6 5.8 0.1 0.0 Other* 2.9 2.6 0.0 0.0 North American utilities expect to provide nearly 3.400 million megawatthours of electricity in 1994 This is over Tital 15.0 27% more electricity than was required in 1984 Coal-Additions

• 113 2 fired generation is projected to its over-50% share of Retirements 12.4 0.2 total electricity generation in the United States by 1994 Oil-to-Coal while nuclear will increase its share to 22%2 and oil / gas Conversions 45 0.0 will decrease to 12% as shown in Figure 6
• Totals may not acc due to roonomg
• Compared to 1995 protections of 7s% for France and 33% to 37% for the U K West Germany and Japan,for examples d " A-3//

1 l

i NERC ASSESSMENTOFPOWER. SUPPLY ADEQUACY

        '- ~ Uerd S:stes. Instated penerstang capacity'es               However. 8.300 NtW of coal-fired capacity was droppec

[m} forecast to be near the minimum acceptable ferels an some regions by 1994 and may not be suffecsent unless from the forecast while less than 3.700 MW was seded. 24 units totaling 12.500 MW were delayed an average of certain important act,ons are taken. In Canada. 20 months each Ariother 18.200 MW of coal-fired elec tricity supply as judged to be adequate for the enfore capacity is planned but not yet under constructson 11is 1965-1994 period. questionable rf all of that amount can be licensed and constructed by the time it is needed Construction was Pnontees for assuring ade quacy of electnc suppfy in the begun on only one large coal-fired unit in the past yea' Unured States for the 1990 s nnclude keepang esisfung generafsng units m operarson. maintaining schedules for Recently, some regulators have not allowed costs for unsis under construelson and shortensnp feed femes for coal and nuclear units to be included in rate base in the sileng freensmg and constructing of nen units other cases. extended phase-in periods have been required. These actions are preventing or ciscour ag+g REQUIREMENT FOR FUTURE utilities from committing to further construction ADEQUACY Reliabitity is i.kely to sutter in the process Electric power supp'y in the Unsted States dunng the ACID RAIN 1985-1994 penod will be adequate af utilities can. In the past, the effect of acid rain legislation was Complete and place in service generating units conssdered to be heaviest east of the Mississippi River which are well along en construction. However calls for further restnction in SO2 ano NO. emissions are being voiced in other areas as well in the Use life extension, refurbrshment and maintenance United States, there remains a need to further programs to get the most out of existing generating document cause effect relationships, assure equipment effectiveness of ar y program offered for adoption. and achieve consensus regarding societal costs Further Devetcp capacity s'tematives having shorter lead emission testriction would lower unit availability and times for licensing and construction. increase the power required to operate new emission contaot equipment. Any control equipment added to Resolve the acid rain question in a manner that will existing plants would require extensive unit downtime not result in an undue redJClion of existing coal- while insta!!ation was made. While the outages required fired capacity or an lengthy outages for retrofrtting for retrofittmg units wrth new control equipment can be units scheduled in phases, some replacement capacity woul: still be required Commit to new capacity that has a demonstrated

            "d SHORTEN LEAD TIMES COMPLETE UNITS NOW UNDER CONSTRUCTION                                                   A reduction in the time between commrtment and operation of a generating unit (lead time) would be of in Canada. 6.400 MW of nuclear capacity.1.700 MW of             great benefit in capacity planning in fact, it could be coal-fired capacity, and 2.100 MW of hydro capacity             considered the single most important s'ep needed to presently planned for service over the next ten years is        assure future electnc supply adequacy atready under construction The expectation is very high that att of most of this capacrty will come ento           Smaller units are also being considered as a means of service on schedule                                             shortening the time between commitment and operation and reducing capital requirements The smaller units' The United States' situation for units under construction       greater ratio of shop tabor to field labor should does not appear as favorable At present, there is               minimtze construction time This rs particularly true of 46.300 MW of nuclear capacity under construction                combustion turbines Another type of unit being studied Dunng 1964. 3 units totating 2.383 MW were indefmitely          uses fluidized bed combustion, which can bum a wide delayed and 8 units totating 9.040 MW were canceled in          range of coats or even refuse in an environmentally addition. 26 untts had their service dates delayed an           acceptable manner. The smatler units should make average of 7 months each Of the 16 units scheduled for          siting easier, and also ease financing of projects commercial operation dunng 1984. only 7 (totaling 7.565 MW) were placed m service. Based on continuing             LIFE EXTENSlON AND difficutties m satisfying all regutatory requirements and        REFURBISHMENT obtaining Csperating licenses, the service dates of the re       n ng nuclear units must be considered                   For various reasons. many systems have made studies of extendmg the life of existing units or of domg major maintenance to improve availabilrty Some have entered United States' utilities also have 24.000 MW of coal-fired      into programs to emp6ement the results of the studies capa ,ty unoe construction These units are not Th*5* p'oo'am5 ^*** ta* **ct of 'ncreas'no capacity ammune to proDiems. but ine outiook .s much bngnie, en a short penod of trme They can aho aHow delaying than for nuclear SmCe last year 18 Coal-fired units the decision date on new capacity additions totating atmost 10.000 MW were placed in service 16         k ~ 3lL

U.S. ELECTRIC GENERATING CAPACITY AND PEAK DENAND WITH POSSIBLE CANCELLATIONS AND RETIRENENTS j 1964-2000 1100-

4.0 %

K  : 1 W 900 I  : 2.5 % N  : j N f 2.0 ) L

     "      L   700      INSTALLED                      -                    -
                    ;    CAPACITY _       -     -

N  :  ! S -

• 4

                    ;e                                                    Edison Electrical Institute h        500    ,

1

       $              9 1      1 9

i 9 1 9 1 9 1 9 1 9 1 9 1 9 1 9 1 9 1 9 1 9 1 9 1 9 1 9 2 0

!                     8       8     8     8     8   8   9     9     9        9   9   9   9   9   9   9     0       !

J' 4 5 6 7 8 9 0 1 2 3 4 5 6 7 8 9 0 YEAR l NOTE: PEAK DEMAND AT STATED GROWTH RATES WITH 20% CAPACITY NARGIN HEEDED FOP l l DEPENDABLE SUPPLY. INCLUDES EXISTING CAPACITY LESS RETIREMENTS FEPOFTED TO NERC AND ASSUNED RETIREMENTS AFTER 1993, AND CAPACITY UNDER CONSTRUCTION LESS ASSUMED CANCELLATIONS.

APPENDIX XXI. POLICY STATEMENT ON DEFERRED PLANTS I l I POLICY STATEMENT i

                                                                             )

ON

         @   ~

DEFERRED PLANTS BRIEFING TO ACRS JULY 11, 1986 T. MICHAELS A -3/f

 . . . . . .                                                          _ _ __ J

l 4 l COMMISSION DIRECTIVES ON DEFERRED / CANCELLED PLANTS i

• PORTIONS OF THE 1985 POLICY AND PLANNING GUIDANCE ITEM IV.B, PLANNING GUIDANCE 2 l
• REQUEST FOR PROCEDURES TO REACTIVATE A PROJECT j- -

AFTER CONSTRUCTION AND LICENSING HAVE STOPPED

l

              . , _ .

• COMMISSIONER ZECH'S REQUEST
• SEVERE ACCIDENT AND STANDARDIZATION POLICY STATEMENTS k

l O nw

   ~

O , l 1 POLICY STATEMENT CONSIDERATIONS f

• DOCUMENTATION, MAINTENANCE AND PRESERVATION REQUIREMENTS FOR DEFERRED PLANTS
• APPLICABILITY OF NEW REGULATORY STAFF POSITIONS FOR DEFERRED PLANTS BEING REACTIVATED

            . . .

• PROCEDURES FOR REACTIVATING DEFERRED PLANTS
• IDENTIFICATION OF REGULATORY IMPROVEMENTS AND l

RESEARCH INITIATIVES

• POPULATION AND STATUS OF DEFERRED AND TERMINATED PLANTS O j-3/h

   ~

O 4 I I l DEFINITIONS DEFERRED PLANT b - CP IS IN EFFECT

                                                                               - CONSTRUCTION CEASED OR REDUCED TO MAINTENANCE LEVEL
                                                                               - LICENSEE HAS NOT ANNOUNCED TERMINATION 3

O - TERMINATED PLANT

                                                                                  .CP IS IN EFFECT
                      ...                                                      - LICENSEE HAS ANN 0UNCED THAT CONSTRUCTION HAS BEEN TERMINATED PERMANENTLY
!                                                                            CANCELLED PLANT
                                                                               - CP HAS BEEN WITHDRAWN BY THE NRC O                                                                                                                                                       A- 3/7

l'  : l l 1 4 3 SCOPE 4 I 1 1

• PROCEDURES DEVELOPED FOR DEFERRAL, CANCELLATION AND i i l REACTIVATION OF DEFERRED PLANTS '

O -

• DOES NOT COVER REACTIVATION OF CANCELLED PLANTS l

i N E O A-3/f P _..-m.--n-+,-----~r _yy.----m-~--w-e-3 -w g -

1 O 4 ELEMENTS OF POLICY , i

• DOCUMENTATION, MAINTENANCE AND PRESERVATION REQUIREMENTS IDENTIFIED

                                                          - APPLICABLE REGULATIONS - 10 CFR PARTS-21, 50,55, 50,71, l-                                                            50 APPENDICES A&B
                                                          - APPLICABLE GUIDES - R.G. 1,28, 1,37, 1,38, 1,58, 1,88                                                                 ,

AND 1,116 i

                                                          - INSPECTION PROCEDURE - 92050 l

i lO i A-3!9 i

  ~

O  : 1 l t ! ELEMENTS OF POLICY (CONTINUED) i

• APPLICABILITY OF NEW REGULATIONS DURING DEFERRAL -

l PLANT SPECIFIC BACKFITS OF NEW STAFF POSITIONS WILL BE !- CONSIDERED IN ACCORDANCE WITH BACKFIT RULE - 10 CFR PART 50.109 GENERIC BACKFITS WILL EITHER BE IMPLEMENTED THROUGH RULEMAKING OR GENERIC ISSUE RESOLUTION PROVISIONS OF OTHER POLICY STATEMENTS APPLICABLE TO PLANTS UNDER CONSTRUCTION MUST BE IMPLEMENTED l O n.e

      ---,.-,-.---_,-,,,-,--,c-,,p.             .--    ,v,.-, , - -e     - - - - -   - - , , - - - - - - - - - - ~     ---m,---=,-- - ----~- - - - --           ' '

l O PROPOSED REGULATORY IMPROVEMENTS

• REGULATORY CHANGES MAY BE NEEDED T0:

i - ESTABLISH THAT CP REMAINS IN EFFECT (EVEN IF IT - ! EXPIRES) UNTIL NRC WITHDRAWS CP

- CLARIFY LICENSEE OBLIGATIONS UPON PLANT TERMINATION

• DETAILED GUIDANCE'0N SPECIFIC INFORMATION NEEDED FOR TERMINATION OF CP
• DETAILED GUIDANCE ON PROCEDURES FOR INSPECTION OF DEFERRED PLANTS PRIOR TO REACTIVATION O A-3A/

 - - .      .                                                                       .~

AIF PARTICIPATION

• STAFF MET WITH AIF, FEBRUARY 19, 1986 i
• AIF PROVIDED LETTER, MARCH 31, 1986 ON " REACTIVATION OF CONSTRUCTION PROJECTS"

( -

                         - AGREES THAT POLICY STATEMENT IS NECESSARY j                         - POLICY SHOULD APPLY TO PLANTS WITH cps
                         - PLANTS WITH WITHDRAWN cps SHOULD BE HANDLED ON A CASE-BY-CASE BASIS
                         - 10 CFR PART 50.109, BACKFIT RULE, SHOULD BE USED TO IMPLEMENT NEW REQUIREMENTS IN EFFECT FOR PLANTS WITH cps
                         - CURRENT PRESERVATION REQUIREMENTS ADEQUATE AS LONG AS

, RECORDS MAINTAINED,

• REVIEW AND COMMENTS ON FINAL DRAFT, JULY 2, 1986 O esn i

      ~

.O POLICY OVERVIEW 1

- NO SAFETY IMPLICATIONS I b - ESSENTIALLY CONSOLIDATES EXISTING REQUIREMENTS i

                                            - C0ORDINATED WITH INDUSTRY i    -

O -

                                            - PRESENTLY IN MANAGEMENT CONCURRENCE CYCLE (NRR, IE, ELD)
                                            - SUBMITTAL TO COMMISSION - JULY 21, 1986 1

I O A -Ba3 1

 !--                . , . . _ _ - . - . . .      _            __- _=           _       _     ._ _ _ _ .-           _ _ - __ - _. - ._. - _.

ft.6 + - -e 1 APPENDIX XXII BWR CONTAINMENT PERFORMANCE BRIEFING NRC SEVERI ACCIDENT POLICY i e FUTURE PLANTS STANDARD . CONVENTIONAL REVIEW PRA c j e EXISTING PLANTS THEY ARE SAFE El10VGH IDCOR PROCESS IS NEEDED INDIVIDUAL PLANT EVALUATIONS GENERIC TREATMENT FOR GEi4ERIC MATTERS l l 1 l A-329 l e- _ --. _ . _ _ _ - _ . . - - _ - - _

2 TWO ACTIVITIES e DEVELOPMENT OF EMERGENCY OPERATING PROCEDURES SCOPE: PREVENTION AND MITIGATION EPG > P8'P > E0P w NOW ON EPG REV. 14 e IDCOR I SCOPE: PREVENTION AND MITIGATION fiETHODOLOGY ---> IPE PROGRAM > IPE METHODOLOGY BEFORE NRC l e COMMON FACTORS ARE IMPORTANT SCOPE PLANT / EVENT ANALYSIS O A-3R C

_. -. . . . . . - - . - - . - _ _ . _ = . _. 3 r - CONCERNS 1 e E0P WHAT ARE THE E0P STRATEGIES? WHAT ARE MINIMA TO TRANSFER FROM j EPG > PGP --> E0P? ARE REQUIREMENTS CLEAR? 1 e IDCOR IPE IS IT AN ASSESSMENT OR AN EVALUATION? SHALL EACH OWNER ENTER THE IPE WITH A BLANK SLATE? ARE THERE ANY GENERIC SOLUTIONS OR STRATEGIES EVIDENT? PRIORITIES? . l e i O A-an m----w--, g---, --,--rw_ -nwn,- e--, -- ,- m m 4 _ _ --w------ ----_

4 O RECALL WASH-1400 e BWR VS PWR PROBABILITY OF CORE MELT -@ BWR <. PWR M"

                                                                                                   ~
                                                                                                       -- t '#

• CONSEQUENCES OF CORE MELT BWR > PWR ,

o BWR CHARACTERISTICS

                                                                                                     -5 BWR-3                                                         2 x 10 TC-g                  TW-7 i

BWR-2 6 x 10-6 TW-g' l. l l A- 327 l l I

5 l 1 , t f l i 1 i A 5 ELEMENT POLICY t I I

1. HYDROGEN  :

t l'  ! i  : 1

2. SPRAYS i

i i

3. PRESSURE ,

] f

4. CORE DEBRIS ,

f l 5. TRAINING AND PROCEDURES i i i i ! + l i t I f a f O part  :

                                ^
                                                                                                       ~

6 O ELEMENT 1 - HYDROGEN 0BJECTIVE: PREVENT HYDR 0 GEN COMBUSTION CAUSED FAIL'JRE REQUIREMENTS: A. 0XYGEN CONTROL M 'E INERT TO START CONTROL INGRESS OF OXYGEN B. HYDROGEN CONTROL NM CONTROLLED BURNING EQUIPMENT SURVIVABILITY ISSUES:

1. WHEN AND HOW LONG NOT INERTED

2. INDEPENDENT POWER FOR IGNITERS Ph "b

3. SURVEILLANCE FOR IGNITERS l

l l l-329

  . . _ _     . . . . . _ _ _ . _ _ _ _ _ . . _ . , _ _ _ _ . ~ _       . __            . _ . . . _ _ .   . _ . . _ . _ _ _ . _ _ _   _

          . .                                                             7 1

ELEMENT 2 - SPRAYS i OBJECTIVE: SPRAY WATER T0:

1. LOWER PRESSURE

2. COOL VULNERABLE EQUIPMENT

3. QUENCH DEBRIS 14 . SCRUB AEROSOLS l

REQUIREMENTS

                               . SPRAY IN DRYWELL %          [

2. BACKUP WATER SOURCES AND PUMPS O -

HOSE CONNECTIONS USE OF FIREMAINS ISSUES:

1. RISK OF IMPLOSION

2. RISK OF HYDROGEN CWFISTION AFTER STEAM CONDEt'SI'00N

3. MANUAL ACTIONS AND TIMING l

i l O 8-330

8 O ELEMENT 3 - PRESSURE  ; OBJECTIVES: 1. AVERT UNCONTROLLED OVERPRESSURE FAILURE

2. CONTROL RELEASE PATH (SCRUBBING)

Rt0DIREMENTS:

1. SUBSTANTIAL CAPABILITY TO VENT WETWELL

2. REMOTE / RELIABLE CONTROL OF VENT VALVE

3. ABILITY TO RECLOSE VENT O

ISSUES:

1. DELIBERATE RELEASE OR RADI0 ACTIVITY

2. WHAT IS REMOTE / RELIABLE CONTROL?

3. IS DUCT BURST IN SECONDARY CONTAINMENT ACCEPTABLE?

4. WHAT IS APPROPRIATE ACTION PRESSURE?

O h-33/

        ..                                                                            9 O

ELEMENT 14 - CORE DEBRIS f OBJECTIVE: REDUCE LIKELIHOOD OF FAILURE BY DIRECT ATTACK I REQUIREMENTS:

1. USE PRACTICAL DEBRIS RETARDING BARRIERS

2. CONSERVE SUPPRESSION P0OL WATER AS A

QUENCHING POOL O

. ISSUE: WHAT IS PRACTICAL? l l 0 A-san

   ~              _   -_           -                  .

10 O ELEMENT 5 - TRAINING AND PROCEDUPES OBJECTIVE: ENSURE OPERATORS ARE READY TO USE PLANT FEATURES TO BEST ADVANTAGE IN SEVERE ACCIDENTS REQUIREMENTS:

1. CLEAR SYMPT 0M BASED STRATEGIES (INTEGRATED)

2. REMOVAL OF UNNECESSARY INHIBITIONS

3. TRAINING / PROCEDURES ISSUES:

1. COMPETING SAFETY REQUIREMENTS

2. DEGREE OF TRAINING O A-333

11 O AN INSTITUTIONAL APPROACH 1 e GENERIC LETTER - ADVANCED NOTICE OF PROPOSED REQUIREMENTS - TO SOLICIT COMMENT e COLLECTIVE PUBLIC REVIEW i e GENERIC LETTER OF REQUIREMENTS O e IMPLEMENTATION WITH IPE AND EPG REV. 11 IMPLEMENTATION i 1 O + ser

APPENDIX XXIII RES/DET PRESENTATION TO ACRS JULY 10, 1986 ES/DET PESENTATION l T0 i ADVISORY COPNITTEE ON EACTOR SAFEGUARDS O JULY 10, 1986 BY: D. SULLIVAN A DATTA O nur l

FIRE PROTECTION RESEARCH PROGRAM (DET)

 \(3

() PHASE I PROGRAM (NUREG-1148) ORIGINAL BUDGET: FY86: $1000K; FY87: $700K REVISED BUDGET: FY86: $658K; FY87: 0 NOTE: PROGRAM ELEMENTS IN THE PARENTHESES ARE DROPPED. BECAUSE OF BUDGET CUT, o FIRE SOURCE CHARACTERIZATION ELECTRICAL INITIATION EXPERIMENTS ELECTRICAL CABINET SOURCE

               -       fCABLE TRAYS, FLAMMABLE LIQUID SPILLS)-

o FULL-SCALE ROOM ENVIRONMENT TESTS BASELINE TESTS: EFFECTS OF COMBUSTIBLE TYPES, FIRE LOCATION, VENTILATION, trE!LINC ::::c;;T, accM GEgMETRY) CONTROL ROOM TESTS: EFFECTS OF CABLE TYPES, 4C-AfH1fET-O TYFES, t.0CATIGii, VENTILATICf' CE!LINC HEIGHT-) b - J P00M=TO -ROOM SM0KE4MRAT10N4ESTS+

                        < cum e nFMnval_ TEg.T4) o      VALIDATE FIRE ENVIRONMENT COMPUTER CODES AGAINST FULL-SCALE ROOM TESTS LASALLE CONTROL ROOM FIRE ANALYSIS l
                      +3D "SAFFIRE M tDEt' l                -     JugaRvann" AND "InMpppN" rnDES) o       COMPONENT FAILURE THRESH 0LD TESTS:

CONSTRUCT 2 TEST CHAMBERS CABLE DAMAGE TESTS: THERMAL ENVIRONMENT ONLY JCABLE DAMACE TESTS. TUTm. EnviRGidEiii, INCLUDING l SUPPRESSION) l i COMPONENT DAMAGE TESTS: RELAYS, SWITCHES, JLOGIC-

                      -GIRCUITS, POWER SUPPLIES, RECGRDERS, CONTROLLERC' IN THERMAL ENVIRONMENT ONLY

( - 4G0reCliEliT DAMAGE TESTS. TOTAL EiiViRONMENH A -336

  ' (3 V

FIE PROTECTION ESEARCH PROGRAM (DET)

                                                                       ~

IFPACT o FIRE CHARACTERISTICS DATA REMAIN INCOPPLETE, NOTABLY ON CABLE TRAY SOURCES, O FIRE TEST SERIES TO ESOLVE CONTROL ROOM ISSUES RAISED IN BOARD NOTIFICATION 8L1-033 IS NOT C0FPLETED, o WITHOUT THE "SAFFIRE" CODE THEE IS NO EANS OF EXTRAPOLATING ROOM

   '          FIRE ENVIRONE NT TEST DATA TO OTHER ROOM CONFIGURATIONS, o  COMPONENT FAILURE THRESHOLD DATA IS UNAVAILABLE, o   AS CONSEQUENCE OF THE ABOVE, LWCERTAINTIES IN FIRE PRAs WRSIST, LICENSING DECISIONS CONTINUE TO BE BASED ON JUDGFENT, o  THE SPECIALIST STAFF AT SANDIA AND BROOKHAVEN DISPERSE, f

l 0 FIRE TEST FACILITIES BUILT AT NRC COST AT FACTORY PtflUAL RESEARCH CORPORATION AND AT SANDIA AE DISPANTLED. O INFLATION FURTHER RAISES COST IN CASE OF ESUPPTION OF THE PROGRAM, h n-a39

4 3 4 FIE PROTECTION ESEARCH PROGPAM (DET) , BUDGETS - FY 76-82 FY 83-85 FY 86 EEDED * (REVISED) TO COPPLETE (PE-NUREG 1148 ! NUREG-1148 PROGRAM PROGRAM) ! 4106 1900 560 950 FIN A1010 3 ._ ( MAINLY TESTING) l 45 600 98 450 FIN A3252 - (ANALYSIS, CODE l DEVELOPENT) 1 4151 2500 658 1400 TOTALS: l, CLMJLATIVE 6651 7309 8709 TOTALS: 4151

• PROJECTION AS OF JULY 10, 1986 O f-838

O APPENDIX XXIV ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE 4

1. Position Paper on Standardization prepared by AIF Study Group on the Practical Application of Standardized Nuclear Power Plants in

- the United States, March 1986 .

2. t er, Lando W. Zech, Jr., NRC Chairman to D. A. Ward, July 1,

3. Memorandum, R. Savio, ACRS Senior Staff Engineer to ACRS Members, Safety Goal Policy Statement, July 9,1986 i

O 1 I O n as9 _ .- - _ __ _ _}}