ML19261A752

Minutes of 221st ACRS Meeting Held in Washington,Dc,On 780907-09.Minutes from Various Subcommittee Groups

ML19261A752
Person / Time
Issue date:	09/07/1978
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-1580, NUDOCS 7902080129
Download: ML19261A752 (500)

Text

{{#Wiki_filter:- Q6 KS -/.5% TABLE OF C0liTEliTS h. h t i MillUTES OF TiiE i; i '3 [ b h H l3 221ST ACRS MEETIl4G ,t h ;,,J Q Q i ; J Ll-yl } 4 y SEPTE!EER 7-9, 1978 WASHIliGTON, DC l9ll 2l*/7* I. Chairman's Report................................................ I A. Reviewers................................................... 1 B. Changes of Basis for Acquisi tion of ACRS Fellows............ 1 C. Meeting wi th Reaktor Sicherhei ts Kommission................. 2 D. Meeti ng wi tn Groupe Permane n t............................... 2 II. Meeti ng on Fa s t Fl ux Tes t Fa cili ty.............................. 2 A. S u b commi t t e e Re po r t......................................... 2 B. I n t rod u c ti o n................................................3 f C. S ta tus of the liRC S ta ff Review.............................. 3

1. Review..................................................

3 .d 2. Engi nee red Sa fe ty Fe a tu res.............................. 5 3. Containment Margin....................................... S 4. Calculations of Radiological Consequences from Pos tul a te d Ac c i d e n ts.................................. 5 5. Pi p i n g I n te g ri ty........................................ 6 6. Advice to DOE........... ..t........................... 6 D. Appl i ca n t 's Pres e n ta ti ons................................... 6 1. Le a k De te c ti o n.......................................... 6 2. Ul trasonic Tes ting Inspection of Piping................. 8 III. Meeting wi th the Ris k Assessment Review Group................... 8 IV. Meeting with the fiRC Staff and Industry on Anticipated Trans i ents Wi thout Scram...................................... 9 1 79020801M t

TALLE OF COUTEdTS 221ST ACRS MEETIGG A. S u b commi t t e e Re p o r t........................................... 9 B. E PR I P re s e n ta ti o n............................................. 9 C. Navy P re s e n ta ti o n............................................. 10 D. Statement f rom Membe r of the Publ i c........................... 10 E. Atomi c Indus trial Forum's Presentations....................... 10 V. Meeting with the Executive Director for Cperaticns and Mis Staff.. 11 A. Office of i.uclear Reactor Regulation Workload / Resources Situation...................................................11 8. Pipe Cracks Observed in Foreign Sciling Water Reacters......... Il VI. Meeting on Reactor Safety Research for Emergency Core Cocling Systems............................................ 11 Reactor Sa fety Research Su' ccami ttee Report................... 12 A. o B. ECCS S ubccani ttee Rep o rt...................................... 12 C. Office of Nuclear Regulatory Research Report.................. 13 D. Office of Nuclear Reactor Regulation Report................... 13 E. Offi ce of Executi ve Legal Di rector Report..................... 14 F. Office of Executive Director for Cperations Report............ 14 VII. Meeting with the NRC Staff on Recent Operating Experience, Operating Actions, Generic Matters Related to Light-Uater Rea c to rs, a nd Fu tu re Age nda..................................... 15 A. Beaver Valley 1: Failure of Main Transformer................. 15 B. Browns Ferry: Safety Relief Valve and Compressor Malfunction................................................. 16 C. Use of Non-Specification Welding Material in Babcock and Wilcox Reactor Pressure Vessels......................... 16 D. Future Agenda................................................. 16 ii

TABLE OF C0!!TEl.TS 221ST ACRS MEETII;G VIII. Executi ve Sessions (0 pen to Public)............................... 17 A. S ub c omi tte e Re p o rts.......................................... 17 1. Regul a to ry Ac ti vi ti es.................................... 17 a. Reg ul a tory G ui de 1. 7 2 ( Re v. 2)........................ 17 b. Regul a tory Gui de 1.134 ( Rev. 1 )....................... 17 2. Rea cto r Sa fe ty Res ea rch................................... 17 B. Sub ccami ttee Ac ti vi ti es....................................... 18 1. Fu tu re S c h e d u l e........................................... 18 C. Ac ti vi ti e s o f I'embe rs......................................... 16 1. hr. B e n a e r................................................ 1 8 D. Paper: Rol e of ACRS i n !;ucl ear Sa fe ty........................ 19 E. Advanced Gas-Cool ed Reac ters.................................. 19 F. ACRS Repor ts a nd Le tters...................................... 19 1. Regulatory Guides......................................... 19 2. f'emorandum to Comissioner Bradford....................... 19 3. Le tter to Rcb e rt Grcy..................................... 19 iii

TADLE OF C0llTEf4TC 221ST ACRS MEETING A p p e n d i x I - A t t e n d e e s................................................. A-1 Appendix II - ACRS Future Agenda....................................... A-7 Appendix III - FFTF: P roj ec t S ta tus Re p o r t............................ A-8 Appendix IV - FFTF: Background Material............................... A-37 Appendix V - FFTF: CDA Ene r g e ti cs..................................... A-6 9 Appendix VI - FFTF: Co n tai nme nt l'a rgi ns............................... A-7 3 Appendix VII - FFTF: Calculations of Raciological Consequences from Pos tul ated Acci de nts............................. A-85 Appenoix VIII - FFTF: NRC Ad vi c e to DC E............................... A-9 4 Appendix IX - FFTF: Le a k De te c ti on Sys tem............................. A-104 Appendix X - FFTF: UT Ins pecti on of Pi pi ng............................ A-l l 4 Appendix XI - Surmary of Report of Risk Assessment Review Group........ A-142 Appendix XII - ATUS: S ta tus Re po r t.................................... A-15 3 Appendix XIII - EPRI Appraisal of ATWS................................. A-220 Appendix XIV - ATilS: A I F Po s i ti o n..................................... A-2 7 9 Appendix XV - Office of Nuclear Reactor Regulation Workloac, Priori ties, and Manpcwer Resources..................... A-336 Appendix XVI - History of Pipe Cracks in U.S. Nuclear Power Plants..... A-382 Appendix XVII - Members' Assignments for RSYs Me e ti n g................... A-3 9 2 Appendix XVIII - LOCA/ ECCS Safe ty Research............................. A-394 Appendix XIX - Overview of ECCS/LOCA Research Programs................. A-404 Appendix XX - Classification of Needs for ECCS/LCCA Research........... A-419 Appendix XXI - Beaver Valley 1: Failure of Main Transformer a nd of Di es el to Loa d................................. A-4 27 Appendix XXII - Browns Ferry: Malfuncticns of Relief Valves and Compressor....................................... A-437 iv

TABLE OF C0liTE!;TS 221ST ACRS 14EETIriG Appendix XXIII - Babcock and Wilccx Resctor Pressure Vessels: Use of fion-Specif' cation Weldi ng fiaterial........... A-440 Appendix XXIV Schedule of ACRS Subconr.ittee Meetings and Tours....... A-458 Appendix XXV - Request for Participation on Panel Discussing Design Loads and Loading Combinations for !!uclear Pow e r P l a n ts.......................................... A-4 6 0 Appendix XXVI - GCRA Reques t for fleeti ng wi th ACRS..................... A-462 Appe ndi x XXV I I - Re g ul a to ry Gui de s..................................... A-46 3 Appendix XXVIII - Memo., Chairran Lawreski to Comissioner Bradford on ACRS ' Approach to Safety l'a tters................ A-464 Appendix XXIX - Ltr, ACRS Exec. Director to Robert Gray, on the Appoi ntment of ACRS !! embers.......................... A-493 Appendix XXX - AJci tional Documents Provided for ACRS ' Use............. A-496 y

nuclear powerplanti The future October 31,1977, page 56912. In ae-kposedUsNu ~ ~ schedule for ACRS activities will also cordance with these procedures, oral Dated. August t$.1978. be disnJaed or sTitten statements may be present-JssTin M. Nienotson 8 612 p. 6 30 p.m.: Ezeentti'e ses-ed by members of the pubile, record. Na A. Q ston (opent-The Committee will hear ings will be permitted only durma and discuss the reports of ACRS sub. those portions of the meeting when a (FR Doe. 78-23568 PNed 8-22-78. 8.45 am] committees on proposed revisions to transcript is being kept. and quet,tions NRC regulatory guides. may be asked only be members of the The Committee will also discuss its Committee, its consultants, and staff. [7590-N] proposed report to the NRC regarding Persons desiring to make oral state-NUCLEAR REGULATORY the fast flux test fact 11ty, ments should notify the ACRS Execu. COMMISSION tive Director as far in advance as prac. ra! DAY. st:PTEMBER 8,1978 t! cable so that appropriate arrange-ments can be made to allow the neces-ADVISCtY COMMffitt CN atACTOR

1. 130 a.n-9:l$ c.m.: Meeting tetth the stry time during the meeting for such 5AftGUARD5

. NRC Executive Director for Oper-s t ug attons (opent -The Committee will cg nd information concerning hear and discuss information regard-b d d d I a ordance th th rposes ing the Office of Nuclear Reactor Reg-heNin can be ound in docu ents ulation workload and resources. file and available for public inspectten Energy Act (42 U.S C

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9:15 cJ t.-9.4S c.m. Executive sessten as appropriate in the NRC's Pubite the Advisory Comm'ittee on Reactor' (opent -The Comnuttee wtll hear and Document Room 1717 H Street NW.. Safeguards will hold a meeting on Sep-discuss the report of its subcommittee Washington, D.C. 20555. tember 7-9, 1978, in Room 1046, 1717 and consultants who may be present Further information regard!ng H Street NW., Wa:hington. D.C. regarding proposed implementation of topics to be discussed, whether the The agenda for the subject meeting pmvisions to mitigate the conse. meeting has been canceled or resche-will be as follows. quences of anticipated transtents with. duled, the Chairman's ruling on re-THURSDAY. stTTEMBER 7, 3 31g out scram in light water reactors, quests for the opportunity to present 9:45 a.n-1:15

p. m.

Meeting tettA oral statements and the time allotted

1. J0 a.m.-9:JO a.m. E ecutite session NRC staff (opent-The Committee therefor can be obtained by a prepata telephone call to the.tCRS Executise (opent-The Committee will hear and will hear presentations from and hold Director. Mr. Raymon; F. Fraley, te:e-discuss the report of the ACRS Chair-discussions with representatives of the phone 202-634-3265 t ' ween 8:15 a.m.

man regarding miscellaneous inatters NRC staff the nuclear and utility in-and $ p.m. e.d.t. relating to ACRS activities and mat-destry regarding proposed implemen-ters for discussion with the NRC Com-tation of provtsions to mitigate the Dated: August 17,1 missioners regarding ACRS activities consequences of anticipated transients y." H r e A I d"[.h"g (open L-The t'ce f, wi,thout scram in light water reactors. Adrt r; s:!S p.m.-2.~45 p.m.: E2ecutste sesston na rna y n.mt Offtcer. NRC Commtsstoners (opent-The Committee will hear and tTR Doc. 78-23: ' ~ Jed 8-7 945aml Cor-mittee sill meet with the NRC discuss the report of its subcommittee Conu mstoners to discuss ACRS activi-and consultants who may be present tic', t nng the period of May-August regardmg reactor safety research for 19 a.m.-!!:15 a.m.: Executtre ses. emettency core cooling systems. w} sic .en L-The Committee will hear 2:45 p.ms6:15 p.m.: Meettnq tcith ane acuss the report of the ACRS NRC staff (opent-The Committee APPilCArIONS FCE UCINSES TO HPotT Sutcommittee and consultants who will hear and discuss reports of repre-NUCt1AA Fact! TIES OR MATEtlAt$ may be present regarding proposed op, sentatives of the NRC staff regarding eration of the fast flux test facility. reactor safety research applicable to Pursuant to 10 CFR 110.70. "public 1115 a.m.-12:45 p.-n. and I 45 p.m.. emergency core cooling systems. notice of receipt of an appiteation." 4:J0

p. m.: Fast flur test factlity 6:15 p.m.-7 p.m.: E:ecutive sesston pleast takt notice that the Nuclear (opent-The Committee will hear and (opent-The Committee will discuss Regulatory Commission has received discuss presentations by representa. its proposed reports to NRC regarding the following applications for export Lives of the Nuclear Regulatory Com.

the fast flux test facility and antici. licenses during the period of August 7-misaon WRC) staff, the Department pated transients without scram. 11,19'l8. A copy of each application is on file in the Nuclear Regulatory of Energy and their contractors re-sATmLDAY, sIrrtusta e,1sts Commission's Public Document Room garding proposed operation of the fast I cated at 1717 H Street NW., Wash. flux test factitty. 8:30 a.m.-12 m.: Exceutit'e se'sfon 4:J0 p.m.-5:15 p.m.: Meeting tetth (openL-The Committee vtll complete NRC staff (opent-The Committee its reports to the NRC regarding mat. Dated: August 16,1978, at Bethesda, will hear presentations and hold dis-ters discussed during this meetmg and M d. cussions with representatives of the proposed ACRS comments regarding For the Nuclear Regulatory Com-NRC staff regarding recent licensmg the source term used in accident anal-mission. ysis and anticipated doses m adjacent Grut.D O. Oruncrn. nuclear reactors from postulated acci. Asststant Director. Export / ' Antitrust Division. U.S. Department of Justtee. suite 4:0, 1776 Peachtree Street dents in one of the facilities. Import and Internationat NW., Atlam.a. Ca. 30309 telephone 404-881-Procedures for the conduct of and Safeguards. Office of Interna. 3:20. FTS 237-3820. partJcipation in ACRS meettngs were tiona1 Programs. IEDERAt itG15T11. YOL. 43. NO.164-WEDN!5 DAY, AUGUST 23, 1978 j I

i.f a -wvy\\ UNITED STATES f y 'l ) s. / [,; NUCLEAR REGULATORY COMMISSION . 3 0.Q./ / j ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

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/ WASHINGTON, D. C. 20555 g Revised: September 6, 1978 s SCHEDULE AND CUI'LINE Krs DISCCSSICU 221ST ACRS MEETI!.G SEPIDiBER 7-9, 1978 3 717 H St., !M WASHI!SION, DC 'Ihursday, Snptember 7,1978, Ecom 1046,1717 H Street, !M, Washington, CC 1) 8:30 A.M. - 9:15 A.M. Executive Session (C7en) 1.1) Cnairm n's report 2) 9:15 A.M. - 10:30 A.M. Meeting with NEC Staff (Cb?n) 2.1) Report on recent cp~ ating ex-perience and licene actions 2.1-1) Beaver Vali hoclear Station - f lure of min transforrnr 2.1-2) Browns Ferry Nuclear Station - pri:r.3ry sys-tem blowdown anc failure of control air system 2.1-3) Reactor pressure vessels - use of nonspecification welding material 2.2) Future Schedule 2.2-1) Anticipated ACRS Sub-committee activity 2.2-2) Anticipated ACES acti-vity 3) 10:30 A.M. - 11:00 A.M. Executive Session (Cben) 3.1) Report of ACES Subco=ittee on the Fast Flux Test Facility

a Schedule - 221st Mtg. - 2.- Pcvised: September 6, 1978

4) 11:00 A.M. - 1:00 P.M.

Fast Flux' Test Facility (Ocen) 5) 1:00 P.M. - 2:00 P.M. LUNCH 6) 2:00 P.M. - 5:00 P.M. Fast. Flux 'Ibst Facility (Omn) 7) 5:00 P.M. - 7:00 P.M. Report of Risk Assessment Review Group (Coen) Friday, September 8, 1978, Room 1046, 1717 H Street, W, Washincten, DC 8)

8. 'O A.M. - 9:45 A.M.

Meeting with Executive Director for Cperations (Co:n) 8.1) Report on tam workload /re-sources situation 8.2) Prirary system pipe cracks - pipe cracks in foreign BWR's 9) 9:45 A.M. - 10:15 A.M. Executive Session (Ocen) 9.1) Report of ACRS Subccmmittee on Anticipated Transients Without Scram (A portion of this session will be closed as required to discuss Classi-fied Information related to this rat-ter.)

10) 10:15 A.M. - 1:45 P.M.

Anticipated Transients Without Scram (Ocen) (A portion of this session will be closed as required to discuss Classi-fled Information related to this rat-ter.) 11) 1:45 P.M. - 2:45 P.M. LUNCH 12) 2:45 P.M. - 3:15 P.M. Executive Session (Ocen) 12.1) Report of ACES Subcommittee on Reactor Safety Research for Dnergency Core Cooling Systems 13) 3:15 P.M. - 6:45 P.M. Reactor Safety Research for Dneraency Core Cooling Systems (Coun) s

Schedule - 221st Mtg. Revised: September 6, 1978 14) 6:45 P.'M. - 7:15 P.M. Executive Session (Ocen) 14.1) Discuss proposed ACES report to IEC on the Fast Flux Test t Facility L Saturday, Senterter 9, 1978, Room 1046, 1717 H Street, IM, Washington, DC 15) 8:30 A.M. - 2:30 P.M. Executive Sessicn (Ocen) 15.1) Discuss proposed ACRS: Report to tac on Fast Flux 'Ibst Facility Position /co=ents Ie. An-ticipated Transients Without Scram (A portion of this ses-sion will b3 closed as required to discuss Clas-sified Information related to this matter.) 15.2) Discuss proposed ACRS position / comments on the PSR program for ECCS 15.3) Reports of ACRS Subco=it-tees on NRC Pegulatory Guides o0o S e O vf e

Issue Date: MINLTTES OF THE NOV 2 81978 221ST ACRS MEETItX3 am. e' 9 f 3 p*i piU'l c {lJ

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• l SEPTEMBER 7-9, 1978 l

,j !J e 1 NASHIt& ION, DC y E The 221st neeting of the Advisory Ccr=tittee on Reactor Safeguards, held at 1717 H St. N.W., Washington, DC, was convened at 8:30 a.m., Thursday, September 7, 1978. The Chairman noted the existence of the published agenda for this meeting, and the items to be discussed. He noted that the meeting was being held in conformance with the Federal Advisory Committee Act (FACA) and the Government in the Sunshine Act (GISA), Public Laws 92-463 and 94-409, respectively. He noted that a request had been received frcrn a member of the public to present a statement with regard to the Committee's discussion of Anticipated Transients Without Scrs (ADS), and that time would be made available for this statement when appropriate. He also noted that copies of the transcript of some of the public portions of the meeting would be available in the NRC' 3 Public Document Roan at 1717 H St. N.W., Washington, DC, within approximately 24 hours. (Note: Copies of the transcript taken at this meeting are also available for purchase fran Ace Federal Reporters, Inc., 444 North Capitol St. N.W., Washington, DC, 20001.] I. Chairman's Reoort (Open to Public) [ Note: Raymond F. Fraley was the Designated Federal Employee for this portion of the meeting.] A. Reviewers The Chairman named Messrs. Moeller and Siess as reviewers for the 221st ACRS Meeting. B. Changes of Basis for Accuisition of ACRS Fellows The Chairman informed the Ca=tittee that the Ccs=nission has interpreted the wording of Public Law 95-209, the Act that estab-lishes the ACRS Fellowship Program, to indicate that one objective of the Fellowship Program is to attract new technical people into the nuclear energy field. On this basis, the Camiission believes that it is inappropriate for the Cannittee to request the transfer of any NRC personnel into the Fellowship Program. 1

MINUTES OF 'IEE 221ST ACRS MEETIN3 SEPIDGER 7-9, 1978 C. Meetina with Reakter Sicherheits Kommission (RSK) The Chairman noted the existence of a proposed draft for the agenda for a meeting to be held between the Ccxnmittee and the RSK in November. He requested that the ACRS Executive Director assign subjects from this agenda to appropriate Members for devel-opment for the meting (see Appendix XVII). D. Meetino with Grouce Pe ganent The Chairman reminded the Committee that a meeting was being held with the French Groupe Permanent Reacteur to begin at 1:30 p.m., Monday, September 18, and to continue through Wednesday, September 20, 1978. II. Meetino on Fast Flux Test Facility (FFTF) (Special) (Open to Public) [ Note: Andrew L. Bates was the Designated Federal Employee for this portion of the meeting.] A. Subcommittee Reoort Mr. Kerr, Subcommittee Chairman, discussed the long history of the FFTF project, and noted that this plant, being a Department of Energy test reactor, does not require licensing. This current special review is in the nature of the NRC providing advice to the Department of Energy. He noted some of the following differ-ences between this current review and one for an operating license for a nuclear power plant: e The reviews of the seisnology, geology, and hydrology, which were carried out in an earlier review of the project, are not considered again in the current safety evaluation report (SER). e The meteorological data considered for this site was devel-oped not for FFTF, but for an adjacent site on which the WPPSS Plant Units 1 and 4 are being constructed. e LesJ Stringent requirements for tornado protection are being providad for FFTF than would be required for a commer-cial plant. e The requirements for FPIF for emergency power are less than those for a commercial plant. t 2

MINUTES OF THE 221ST ACRS MEETIIG SEPTEMBER 7-9, 1978 Mr. Kerr enumerated the areas on which the NRC Staff focused in its review (see Appendix III). He noted that the review covers nest of the standard abnormal events that the NRC Staff considers for commercial plants. [ Note: D. E. Simpson, Hanford Engineering Developrnent Laboratory (HEDL), coordinated presentations for the Applicant; T. P. Speis, for the NRC Staff.] B. Introduction D. E. Simpson provided background information for the FPIF Project, including the location of the plant, the site layout, the operating organization, the plant arrangement, the key design parameters, a core map, the design of the driver asserbly, the reactor internals arrangement, the heat transport system, the chronological history of the project, and the current status of the project (see Appendix IV). In answer to a question, D. E. Simpson said that a detailed startup and test program is planned for FFIF. Upon comoletion of the test program, it is planned to operate the plant on a full-power, steady-state, test reactor basis. C. Status of the NRC Staff Review

1. Review T. P. Speis noted that the Staff issued in August an SER (NUREG-0358), summarizing the NRC Staff's safety review of FFIF.

He discussed the matters considered at the various ACRS Subcom-mittee meetings (see Appendix III). He said that the object of the NRC Staff's review was to evaluate the adequacy of the FFIF design, and to insure its safe operation. He noted however, that DOE has the responsibility for deciding how to respond to NRC's advice and has the final responsibility for the design and safe operation of the plant. NRC has not provideo s..-site inspection, nor will NRC be involved during FPIF operation under the terms of the existing NRC-DCE interagency agreement. The NRC Staff does expect to receive certain acceptance test information in several areas identified in the SER. Som of the limitations of the review include o the mixed Pu-U-oxide fuel was reviewed for low burnups and low operating temperatures only, 3

MINUTES OF THE 221ST ACRS MEETING SEPTDGER 7-9, 1978 e the effect of the closed loops on the core was not re-

viewed, e the possible use of advanced fuel in the future was not

reviewed, e at DOE's request, safeguards and security provisions were not reviewed.

T. P. Speis said that the NRC Staff had been asked by the subcommittee chairman to address the following unresolved areas with respect to FETF: e piping integrity and containment margins 'or core-melt events; and e core destructive accident energetics. With respect to core accident energetics, the NRC Staff has concluded that the residual risk which arises frcra uncertain-ties in in-vessel post-accident heat removal makes it prudent to provide for the consequences of this low likelihood event in the best possible manner. The NRC Staff has further concluded that the accident energetics are not likely to exceed the containment capability of the primary system. Thus, assuming vessel melt-through, the NRC Staff has determined that over a wide variety of postulated accident conditions, containment pressurization and the generation of hydrcgen resulting from sodium and core debris interactions constitute the principal challenge to containment. This is an issue between the NRC Staff and DOE, the differences of which lie in the interpret:- tion and utilization of data on concrete attack rates and hydrogen production and disposition. The NRC Staff belie'. that the data on sodium-concrete and debris reaction rans currently available are not adequate to justify their applica-tion to FETF. In a discussion of operator actions with regard to fuel failures, T. P. Speis said that it is the intent that, although the reactor is designed to be able to accommodate at least 1% fuel failure, when failure is first detected, the plant will be shut down, and the cause of the failure determined. Following an evaluation, a decision can then be made whether to cperate with this failed fuel condition, or whether to remove and replace the damaged elements. 4

MINUTES OF THE 221ST ACRS MEETUG SEPIDGER 7-9, 1978

2. Encineered Safety Features J.

F. Meyer, NRC Staff, discussed the calculated FPIT core-disruptive accident energetics and the engineered safety features provided to withstand the energetics of such events (see Appendix V). He said that the NRC Staff takes the posi-tion that the FFTF vessel, vessel components, and primary loop are adequately designed to withstand the nechanical loading from a conservative spectrum of core disruptive accident events.

3. Cont <;inment Margin

'.. R. Marchese, NRC Staff, discussed the Staff evaluation of FFT containment margins (see Appendix VI). He noted that the coaclusions reached were based on best estimate calcula-tions. The NRC Staff is supporting a confirmatory research program irected toward understanding the materials interaction phenome: associated with sodium-concrete reactions and inter-actions mtween nelten fuel-steel and concrete. In the near future, both small-and large-scale tests will be perfc.med utilizing the FFI'F types of concrete; namely, basalt and magne-tite.

4. Calculationf of Radiological Consecuences from Postulated Accidents J. A. Long, NRC Staff, discussed the calculations of radiological consequences from various accidents postulated for the FFI'F (see Appendix VII).

He said that the NRC Staff has examined many calculations that have been subnitted by DOE, and that recognizing that the NRC Staff does not always agree with the constants used, that when the same constants are used, the results obtained fram the calculations are similar. He therefore concluded that the NRC Staff and DOE are in agreement with regard to the mechanics of the calculations. Mr. Moeller suggested that agitation, by such simple means as a fan, of the vapors in the containment following an acci-dent can help facilitate agglomeration of droplets and reduce doses from purged releases. He concluded that in all the cases examined, dose limits were below the 10 CFR 100 limits. Mr. Siess noted that while these doses might be acceptable under 10 CFR 100, they would not be acceptable to the Staff on a licensed plant under 10 CFR 50. Under 10 CFR 50 rules, doses are acceptable only if due to leakage, but are not acceptable as a result of purging. 5

MINUTES OF THE 221ST ACRS MEETIIG SEPTEMBER 7-9, 1978

5. Piping Intecrity H.

B. Holz, NRC Staff, discussed the following points relative to piping integrity: e The NRC Staff believes that there should be continuous verification of the primary piping integrity. He said that, of all the piping, the nest sensitive was the inlet downcomer to the reactor, between the check valve and the reactor inlet. Elsewhere in the piping system, a double-ended pipe rupture could be withstood, with scram, and this event would not be a core-disruptive initiating event. This matter is covered in Appendix B to the SER. Appendix B also shows that a leak would be detected before a break, and that it was extremely unlikely that gross failure would occur without some prior warning, and therefore leak detectors would provide sufficient tine for warning to shut the reactor down. e A base line ultrasonic test (UT) of the piping should be performed where possible to evaluate the adequacy of construction practices. Requirements for the examination of high stress welds in the pump discharge over the life of the plant represent approximately 4% of the pri:rary welds, or six welds. Four inspections of each weld per year represent a total of only 24 inspections. The NRC Staff does not believe it is unreasonable to require such a small number of inspections.

6. Advice to DOE H. B. Holz su:=arized the NRC presentation, and discussed the advice given by the "RC to DOE (see Appendix VIII).

D. Apolicant's Presentations

1. Leak Detection R. D. Warrick, HEDL, identified the disagreement between the Applicant and the Staff on this item as one regarding the seismic qualification of the leak detection system.

The Applicant fully agrees that leak detection is a safety related item. The Applicant believes that the leak detection systen of the FETF meets all of the regulatory requirements (see Appendix IX). He noted that the reactor will be shut down in the event of an earthquake. The only problem that a nonqualified leak detection system would present in this event is that it might 6

MINUTES OF THE 221ST ACRS MEETING SEPTmBER 7-9, 1978 not be capable of detecting leaks post shutdown. Such leaks, should they cccur, will not interfere with the decay-heat renoval system. He noted further that the requirements of the plant are such, that in the event of a failure of the leak detection system, the plant will be shut down. In answer to a question, R. D. Warrick said that the calculated survival time of the reactor following loss of all a-c power is approximately one week. This long time is the result of a very slow heat-up rate. He noted also that anal-ysis of the ability to renove heat from the reactor shows that heat removal with no secondary loop available is effective. Even without the primary loop, considerable decay heat can be stored initially and, at higher temperatures, renoved from the reactor vessel without disruption of the core geometry. The analyses have been conservative. R. D. Warrick said that an extensive test program has been carried out on this leak detection system and, in particu-lar, on aerosol generation tests. Many variables have been examined in these tests: e the leak rate, e the temperature of the sodium, e the oxygen content of the at::osphere, e the noisture content of the at osphere, and e the gecnetry of the piping installation. In addition, approximately 50 tests have been made on the sensitivity of the system. He said that the tests have shown that, contrary to the NRC Staff position, the insulation around the piping does not interfere with the release of sodium vapor frcrn leaks and the subsequent detection of vapors by the leak detection system. He discussed the various leak detection systems which are provided in the FFTF. H. B. Holz said that the NRC Staff will settle for just one leak detection system, provided it is seismically qualified. 7

MINUTES OF THE 221ST ACRS MEETING SEPTEMBER 7-9, 1978

2. Ultrasonic Testina Insoection of Piping D. D. Stepnewski, HEDL, discussed the NRC Staff's proposed requirements for UT inspection capability for the primary piping in the FFTF, noting that the Applicant does not believe that a UT surveillance system is essential to safety.

He discussed the requirements for, and the wock that has been done to develop, a UT inspection system (see Appendix X). The Applicant believes that the leak detection system is the best available method to detect piping cracks. [ Note: Because of time limitations, the Comittee was unable to complete its review of the FFTF during the 221st ACRS Meeting. It was proposed that this matter be continued for further consideration at the 222nd ACRS Meeting.] Members suggested that the following additional items relating to FFTF might te discussed at the 222nd ACRS Meeting. e method for handling tritium, e studies of sand gravel filters including allowance for sodium aerosols, e consideration of both conventional and sodium fire pro-

tection, e calculated consequences of a postulated loss of the heat

sink, e effects of containment ventilation on 10 CFR 100 dose

limits, e evaluation of the use of s: Toke detectors for early indica-tion of sodium leaks, and e a postulated scenario for loss of primary piping.

The Chairman suggested that it would be helpful if the handouts to a used during the follcv-up presentation" were provided to the Members prior to the meeting. III. Meetina With the Risk Assessment Review Grouc (Open to Public) (Note: John C. McKinley was the Designated Federal Employee for this portion of the meetirs.] 8

MINUTES OF THE 221ST ACRS MEETING SEPTEMBER 7-9, 1978 Dr. H. W. Lewis, Chairman of the Risk Assessment Review Group, provided the Cmmittee with a summary of the report the group had prepared for the Commission and which was scheduled for publication on September 13, 1978 (see Appendix XI). Dr. F. von Hippel said that the report is a negotiated docu-ment, involving many compromises, and must be read in its e,tirety to obtain the full flavor and importance of the findings of tne Review Group. Dr. Lewis said that the Review Group has concluded that KASH-1400 contains a number of discrepancies, and should not be used uncritically in the regulatory process. However, the Group believes that the fault-tree methodology has merit, and that probabilistic analysis should be used in the regulatory process. The Review Group also believes that the process of peer reviea went astray in the review of WASH-1400. The Review Group also found that the allegation that WASH-1400 encompassed intellectual disbonesty is without merit. IV. Meeting with the NRC Staf f and Industry on Anticioated Transients Without Scram (A'IWS) (Open to Public) [ Note: Thomas G. McCreless was the Designated Federal Employee for this portion of the meeting.] A. Subcommittee Report Mr. Kerr, Subcommittee Chairman, discussed the current posi-tions of the NRC Staff and the nuclear industry with respect to A'IHS and reviewed the discussions held at recent A'IKS subcommittee meetings. (For project status report, subcommittee meeting minutes, and general background material, see Appendix XII.) In answer to a question, R. Mattson, NRC Staff, said that the NRC Staff would provide a response to the industry response to the Staff's position prior to the next ACBS meeting. B. EPRI Presentation G. S. Lellouche, EPRI, presented an EPRI appraisal of the A'1HS matter, based on statistical analysis (see Appendix XIII). He noted that the probability figures derived by EPRI differ significantly from those developed by the NRC Staff. 9

SEPTDSER 7-9, 1978 MINUTES OF THE 221ST ACRS MEETING C. Navy Presentation (Closed to Public) r Wis presentation was classified Confidential Restricted The minutes of the classified portion of the meeting will [ Note: Data. be kept on file in the ACRS office.] D. Statement from Member of the Public (Oper. to Public) Birchall, Cmbustion Engineering Co., requested time at this meeting to correct what he believed to be a misinterpre-W. E. respect to the Cmbustion Ergineering position with He said that Cmbustion Engineering does not row, nor has tation of it ever, considered A%5 to be a significant threat to the A%S. and safety of the public. concern to ATE has become a significantATE has had a dest come to believe that industry as a licensing problem. When C mbustion influence commercially on the nuclear industry. f the included a supplementary protection system in the de i proposes to make this modification based solely on the licen CESSAR-80 plant, requirements established by the NRC,that A%S does n h and safety of the public." E. Atomic Industrial Forum's (AIF) Preser :ations_ J. Ward, Chairman of the AIF Ccmnittee on P' actor Licensing AIF and industry and Safety, discussed the A%S problem from theHe concluded that A% point of view (see Appendix XIV).present a hazard to t discussed the Burstein, Wisconsin Electric Company, importance to the nuclear industry of an equitable S. look at A%5 as an example of philosophy, a the A%S ratter. to the implications of infinite inquiry. S. Burstein suggested that, while many believe that the regu-latory agencies have the ultimate responsib The indus-derives from the efforts of plant owners and operators. try perceives that ATE was invented around the ACRS examined, neither the NRC Staff nor the industry has been ab table. He said that, dettonstrate that ATE is in fact an issue.a point whe 10 N=

MINUTES OF 'nIE 221ST ACRS MEETING SEPTDGER 7-9, 1978 liabilities must be capped. The fixes that are currently being proposed by the NRC Staff will cost the nuclear industry billions of dollars, and the increase in safety that is expected to result from such modifications does not justify such an outlandish expenditure. He said that industry has accepted fixes in the past because it was cheaper to accept and modify than it was to fight the Ccmission. Mr. Mark requested the NRC Staff, when it responds to the comments made by the nuclear industry, to make it clear in its presentation just when 10 CFR 100 limits are exceeded. [ Note: The Ccmittee did not complete its consideration of the A'IMS issue at this meeting. Additional presentations from both Industry and the NRC Staff are planned for the 222nd ACPS Meeting. Mr. Kerr believed that the Committee would not be ready to write a report until sometime af ter the 222nd ACRS Meeting.] V. Meeting with the Executive Director for Ooerations and His Staff (Open to Public) [ Note: Thomas G. McCreless was the Designated Federal Employee for this portion of the meeting.] A. Office of Nuclear Reactor Reculation Workload / Resources Situation H. Denton, NRC Staff, discussed the current and projected workload for the Office of Nuclear Reactor Regulation, the prior-ities set for accomplishing the workload, and the resources available to the Office (see Appendix XV). B. Pipe Cracks Observed in Foreign Boilina Water Reactors D. Eisenhut, NRC Staff, reviewed the history of pipe cracks and studies o f intergranular stress corrosion cracking of stain-less steel in the U.S. (see Appendix XVI). He discussed new data recently received from the Federal Republic of Germany regarding instances of detected cracks in BKR piping in Germany. He noted that these data are currently incomplete, ard the German Govern-ment has not authori ~1 release of it. VI. Meetino on Reactor Safety Researci. for Emeroency Core Coolina Systems (Open to Public) [ Note: Andrew L. Bates was the Designated Federal Employee for this portion of the meeting.] 11

MINUTES OF THE 221ST ACRS MEETItG SEPIDEER 7-9, 1978 A. Reactor Safety Research Subconnittee Report Mr. Siess, Subcommittee Chairman, noted that in the prepara-tion of NUREG-0392, REVIEW AND EVALUATION OF THE NUCLEAR REGULA-TORY COMMISSION SAFETY RESEARCH PROGRAM, A Reoort to the Concress of the United States of America, the Ccmittee tried to provide comprehensive coverage of the entire NRC Safety Research Program, and also provide some depth in each area of research. The Cccmit-tee commented in the report that it did not have time to consider the assignment of relative priorities among the various areas of research, or the long range aspects of the research program, but that it would try to do so in the future. The Committee did mention, that in connection with the LEA /ECCS research, systems engineering is strongly directed toward LOCA/ECCS research, and that there should be a more significant research effort on inte-grated system design in such areas as shutdown heat removal systems. At this time LOCA/ECCS research dominates the entire program in terms of costs. In order to give the Ccmittee an opportunity to reach a collegial position regarding the LOCA/ECCS research, knowledgeable people from the NRC Staff had been requested to discuss the objectives and the future plans for the LOCA/ECCS program. At its 220th meeting, the Committee heard frcm them that a change in direction in the program is planned for the future, and that the NRC Staff proposes to phase out this program over the next decade. The budget for this research is scheduled to increase for a few years, and then decrease. There is much work still to be done, so that while the budget may decrease, it will still be a major program. The Committee has endorsed these proposed objectives in its activities report to the Ccm ission. B. ECCS Subcommittee Reoort Mr. Isbin, Subcommittee Chairman, discussed the current status of the ECCS/IDCA research programs (see Appendix XVIII). In answer to a question, Mr. Isbin said that the German PWR ECCS/LOCA test results are not limited to the design they are testing. The test results will be applicable to U.S. designs. In answer to a question, S. Levine, NRC Staff, said that at this time the U.S.-German research program is not cpen to other participants, but that there are negotiations to include other countries. 12

MINUTES OF THE 221ST ACRS MEETIIG SEPIE4BER 7-9, 1978 Mr. Siess discussed the Committee's postion regarding addi-tional work to resolve ECCS problems by citirg the Ccmmittee's Report on Acceptance Criteria for Cmergency Coolino Systems for Light-Water-Cooled Nuclear Power Reactors, dated Septe ber 10, 1973. He noted that the Committee's Recort on Evaluation Models for Commission Criteria for Emergency Core Coolina Svstems for Licht-Water-Cooled Nuclear Power Reactors reaffirms this position. C. Office of Nuclear Reculatory Research Reoort S. Levine discussed the history, goals, approach to model developnent and testing, the future scope of planned LOCA/ECCS research, and the program schedule for the LOCA/ECCS research (see Appendix XIX). In answer to a question regarding We adequacy of the RES

budget, S. Levine said that the budget is not really limited, and that RES is spending reney in many areas.

Although it is difficult to obtain the needed expertise to do the research at a useful rate, RES is carrying out research in those areas for which a need has been identified. D. Office of Nuclear Reactor Reculation (NRR) Recort R. Mattson, NRC Staff, said that the objective of safety research should be confirmation of the conservatism of LOCA/ ECCS licensing models. He identified the following needs: e confirmation of specific judg:ents, o quantification of the margin of safety, e verification of performance, e verification and quantification of reliability, and o ECC system improvements (see Appendix XX). In answer to a question regarding the methods of quantifica-tion of margins of safety, S. H. Hanauer, NRC Staff, said that he believes that the methodology used in WASH-1400 should be applied to this calculation. The quantification of margins of safety must include both the probability of an event and its consequences. R. Budnitz, NRC Staff, noted that the analyzability of reactor systems is a design criterion. 13

MINUTES OF THE 221ST ACRS MELTING SEPTEMBER 7-9, 1978 R. Mattson said that NRR believes in ECG improvements, but not at the expense of other licensing requirements. He noted that currently there is no pressure to reduce conservatism in areas of reactor design or operation and that LOCA/ECCS criteria are not limiting in the operation of plants. At this time, DNB, not Appendix K, is limiting reactor operation. E. Office of Executive Leoal Director (ELD) Report J. Scinto, ELD, discussed the legal impact of LOCA/ECCS research. He said that Appendix K does not propose fixed require-ments, but refers to aporopriate data, and some fixed mandatory assumptions and parameters that must be used in the evaluation model. With respect to new research results, there is no dif fi-culty in nost of Appendix K, (except for the fixed requirements) to incorporate the results of new research in the new evaluation model. If research demonstrates that the evaluation models previously used were conservative, there is no requirement to modify the evaluation nodel. At his option, the applicant or licensee may modify the nodel and take advantage of any improved results from the research, but he is not required to do so. If the research demonstrates nonconservatism, the applicant or licen-see is required to nodify the evaluation nodel, and recalculate the operating parameters. In this case, the NRC might require shutdown or reduction of power until the new calculations could be completed. In answer to a question, J. Scinto said that there is no legal requirement for the NRC to sponsor research of a particular type. Sponsoring of such research is the agency's obligation to itself, to the public, and to the industry. He said also that there is no mandated legal requirement that future rulemaking with regard to Appendix K be carried out in a full adjudicatory-type rulemaking procedure. F. Office of Executive Director for Ooerations (EDO) Recort S. H. Hanauer said that there is no position regarding the applications of ILCA/ECCS research in the Office of the EDO. He set forth the following percanal views: o NRC research in LOCA/ECCS is essential for the continued licensing of LWRs. The current level of research is appro-priate. o The aim of this research should be to obtain adequate assur-ance that the ECCS will perform adequately when called upon. 14

MINUTES OF THE 221ST ACRS MEETING SEPTEMBER 7-9, 1978 e Research should be carried out in areas where knowledge is incomplete, and where conservative assumptions have been made in the interim, to assure that the assumptions have in fact been conservative, e Licensing is tied to ECG problems, o Research to determine the reliability of the ECG is neces-sary. When the current gap in ECCS information is filled, the level of WCA/ECCS research probably can be reduced. e Much useful information will be obtained from the LOFT nuclear blowdown testing, which should give insight regarding the adequacy of the safety rargins. e Code verification is necessary, although -experimentation cannot be done to cover all conditions. Verification is not an all or nothing program. e The verification function is central to the final phases of the research program as it matures. The current program seems to be about right for this purpose. He noted that NASH-1400 calculi.ces the probability of pipe breaks not to be a significant contributor to the overall risk. Therefore it would seem that refining the calculations for ECCS in the direction of making them more realistic and less conservative accomplishes little, since it will not change the overall risk. In this context, the only function of research is to determine that major errors have not been made, or that non-conservatisms have not been assumed. If, after the research, the calculations still show negligible contribution to risk, further refinement is unnecessary. VII. Meeting with the NRC Staff on Recent Ooerating Exoerience, Ooeratino Actions, Generic Matters Related to Light-Water Reactors, and Future Acenda (Open to Public) (Note: John C. McKinley was the Designated Federal Employee for this portion of the meeting.] A. Beaver Vallev 1: Failure of Main Transformer V. Thomas, NRC Staff, discussed an incident in which the rain transformer failed and a diesel generator failed to load at Beaver Valley, Unit 1. (The content of Thomas' presentation is contained in the documents included in Appendix XX'.) 15

MINUTES OF THE 221ST ACRS MEETING SEPTD4BER 7-9,1978 Mr. Okrent requested that the NRC Staff discuss with the Ccrnmittee the generic implication of these events based upon a risk analysis. Mr. J. J. Ray, acting as ACRS consultant, suggested that a relay testing program should be established to test the sequence of relay operation. He questioned a system design at Beaver Valley 1 that would permit the sequence of events that occurred. B. Browns Ferry: Safety Relief Valve and Comoressor Malfunction C. DeBevec, NRC Staff, discussed events at Brown c Ferry 3 including safety relief valve malfunctions and the massive failure of an air compressor providing controlled air pressure to all units at Browns Ferry (see Appendix XXII). He noted that the compressor failure, while causing malfunction of the reactors, does not affect safe shutdown. C. Use of Non-Soecification Weldino Material in Babcock and Wilcox Reactor Pressure Vessels W. Hazleton, NRC Staff, noted that i9 has been discovered that some non-specification welding materin. has been used in some Babcock and Wilcox reactor pressure vessels (see Appendix XXIII). Mr. Okrent questioned the quality control procedures used by Babcock and Wilcox. He asked how adequate confidence was be be achieved that the proper weld material had been used in the past. G. Emmanuel, Babcock and Wilcox, said that in 1967 and 1968, when the procedures were set up for testing of welding material coils, tests were required only on samples made from one .d of each coil. Since 1973, samples from both ends of the c are tested. D. Future Acenda The Ccrnmittee agreed on a future agenda for the review of cases (see Appendix II). No new cases were scheduled for the 222nd ACPS Meeting. O. Vassallo, NRC Staff, noted that the Staff has changed its position regarding its time schedule for review of the nea Westinghouse Integrated Protection System proposed for RESAR-414. The NRC Staff will now complete its review of the system in aprox-imately six weeks, prior to the issuance of a preliminary design authorization, and will issue a supplement to the SER on this matter. 16

MINUTES OF THE 221ST ACRS MEETING SEPTEMBER 7-9, 1978 Mr. Kerr suggested that the Ccmittee may wish to review this supplement. VIII. Executive Sessions (Open to Public) [ Note: James M. Jacens was the Designated Federal Employee for this portion of the meetiig.] A. Subcommittee Reoorts

1. Regulatory Activities

a. Reculatorv Guide 1.72 (Rev. 2)

The Ccmittee concurred with the NRC Staff position regarding proposed Regulatory Guide 1.72 (Rev. 2, Draf t 2), Sprav Pond Picing Made f rom Fiberglas-Reinforced Thermo-setting Resin.

b. Regulatorv Guide 1.134 (Rev.1)

The Ccmittee deferred action on proposed Regulatory Guide 1.134 (Rev.1, Draf t 1), Medical Evaluation of Person-nel Recuiring Ooerator Licenses, and remanded the matter back to the Regulatory Activities Subcommittee for further work with the NRC Staff.

2. Reactor Safety Research Mr. Siess requested, and the Ccmittee agreed, that time should be allotted at future meetings for consideration of the following items relative to their inclusion in the Ccmittee's second annual report on the Commission's safety research program:

e advanced reactor research, e overall priorities, o waste management, and e the role of risk assessment. The Ccmittee agreed to reserve one full day at the 222nd ACRS Meeting for these discussions. 17

MINUTES OF THE 221ST ACRS MEETING SEPTDiBER 7-9, 1978 The Car:nittee agreed to hold additional discussions regard-ing the II)CA/ECCS research program. Mr. Isbin stressed the Unportance of LOCA/ECCS research, suggesting that the credibility of the NRC is dependent upon the outcome of this research. Independent code assessment is essential to the successful completion of the LOCA/ECCS proc' Jam. He suggested that it would be inadvisable to reduce the lev,1 of research for II)CA/ECCS too much at this time. Mr. Siess suggested that the Ccmmittee should advise the Commission on an appropriate level of safety research for advanced reactors. Mr. Ray suggested that the Canmission should learn from past experience, and that it should be a leader in the develop-ment of safety research programs for advanced reactors. The Canmittee agreed that the follcwing subjects would be discussed at the 222nd ACRS Meeting: e advanced reactor safer; research, o waste management safety research, and e the rr.lative priorities for the safety research programs. Discussions on risk assessment also will be held at the 223rd ACRS Meeting. B. Subcommittee Activities

1. Future Schedule Schedule of future ACRS subcommittee meetings and tours was distributed (see Appendix XXIV).

C. Activities of Members

1. Mr. Bender Mr. Okrent noted that he was been requested to serve on a panel discussing " Design Loads and Loading Canbinations for Nuclear Power Plants" to be presented under the auspices of the Pressure Vessel and Piping Division of the American Society of Mechanical Engineers at its 1978 Annual Winter Meeting (see Appendix XXV).

18

MINUTES OF THE 221ST ACRS MEETING SEPTEMBER 7-9, 1978 Mr. Okrent noted his inability to participate on this panel. It was the consensus of the Ccruittee that partici-pation on this panel is appropriate. Mr. Bender noted his willingness to participate in place of Mr. Okrent. D. Pacer: Role of ACRS in Nuclear Safety The Committee authorized publication of the paper, The Advisory Committee on Reactor Safeguards - Its Rola in Nuclear Safety, by Messrs. Lawroski and Moeller. It is planned that this paper will be offered for publication in a technical journal. It was suggested that any further comments be forwarded to the ACTS Executive Director. E. Advanced Gas-Cooled Reactors The Ccmittee agreed that it could not review, during 1978, the standard reference gas-cooled reactor plant proposed by Gas-Cooled Reactor Associates (see Appendix XXVI). This review will be scheduled for early 1979. F. ACRS Recorts and Letters

1. Regulatorv Guides The Committee prepared a memorandum to the Executive Director for Operations informing him that the Ccri:mittee con-curred in the regulatory position of Regulatory Guide 1.72 (Rev

2),

Sorav Pond Picino from Fiberolas-Reinforced 'Ihermo-setting Resin.

2. Memorandum to Commissioner Bradford The Ccmittee endorsed a response from the Chairman to Commissioner Bradford regarding a letter frcrn Professor Mingle, University of Kansas, and a speech by Mr. A. Philip Bray, regarding the Committee's approach to safety matters (see Appendix XXVIII).

3. Letter to Robert Gray The Ccmittee authorized a reply by the ACRS Executive Director to Mr. Robert Gray, of Save the Valley, regarding the appointment of ACRS members (see Appendix XXIX).

The 221st ACRS Meeting was adjourned at 11:35 a.m., Saturday, September 9, 1978. 19

221st ACRS Meeting Meeting Dates: September 7-9, 1978 APPENDLX I ATTENCEES ADVISCRY CCMiTITES CN REACICR SAFTM Stephen Lawroski, Chairman Myer Bender Max W. Carten Harold Etherington Eerbert S. Isbin William Kerr J. Carson Mark William M. Mathis Dade W. Moeller David Ckrent Milton S. Plesset Paul G. Shewman Ches+er P. Siess .TRS STAFF Raynend F. Fraley, Executive Director Marvin C. Gaske, Assistant Executive Director James M. Jacobs, Technical Secretary Herman Alderman Andrew L. Bates Paul A. Boehnert Sam Duraiswamy Elpidio G. Igne Morten W. Libarkin Richard K. Major temas G. McCreless John C. McKinley Pobert E. McKinney Ragnwald Muller Gary R. Cuittschreiter Jean A. Ecbinette Richard P. Savio Hugh E. Voress Ecbart L. Wright CCNSULTANTS Arden L. Bement Ivan Catton Jeremiah J. Ray William R. Stratten /7- /

~ NRC ATTE?lDEES 221ST ACRS MEETING Thursday, Sect. 7 Div. of Project Mot. Nuclear Reactor Reculation L. P. Crocker

3. Grimes D. V. Vassallo V. Noor;.

W. Hazelton P. O'Reilly Div. of Inspection & T. Speis Enforcement J. K. Tong R. P. Denise W. J. Raymond C. P. Tan R. Woodruff P. F. Riehm V. Thomas

0. P. Chapa C. J. DeBevec R. W. Houston G. W.Reinmuth Div. of Operatins Reactors M.

Ceriance Div. of Systems Safety NRM J. J. Burns F. 3. LItton LI'FB H. 3. Holz MPA R. H. Moore R-9

APPLICANT ATTENDEES 221ST ACRS MEETING Seotember 7,1978 SANDIA LABORATORIES WESTINGHOUSE HANFORD R. Acton R. E. Peterson L. D. Muhlesteirs BNL D. E. Simpson R. G. Gasser D. D. Stepnewski J. E. Werner S. L. Additon ANL R. Warrick W. Lehte A. R. Schide J. Matenatene SAI Pa'u'l Ford

1. - 2

PUBLIC ATTENDEES 221ST ACRS MEETING Thursday, Sect. 7 Robert A. Bari, ANL, Upton, NY R. Borsum, B&W, Derwood, MD H. C. Huang, Westinghouse, Pittsburgh, PA Hans K. Fauske, Argonne Nat'l Lab., Hinsdale, IL Larry D. Kenworthy, Int'l Energy Assor.., Ltd., Gaithersburg, MD Angus Kimmins, WPPSS, Richland, WA Donald F. Knuth, KMC, Washington, DC John Lacante, Westinghouse, Monroeville, PA William Trevor Pratt, BNL, Shoreham, NY Donald Robinson, Clinch River Breeder Reactor, Oak Ridge, TN Noel Shirley, GE, Gaithersburg, MD James H. Taylor, B&W, 707 Old Trents Ferry Road D. -Schaffstall, KMC, Inc.Reston, VA S. Eisenberg,.Self G. Emmanuel, S'abcock & Wilcox K. E. Moore, Babcock & Wilcox P. A. Morris, Scandpower, Inc. J. J Swift, EPA /Y-

NRC ATTENDEES 221ST ACRS MEETI.'G Friday, Seotember 8,1978 Div. of Project Management Div. of Reactor Regulatory Research L. P. Crocker R. A. Benedict R. M. Scraggins G. Edison R. R. Landry L. S. Tong Div. of Systems Safety S. Fabic R. Mattson E. Davidson F. Schroeder A. W. Serkiz A. Thadani T. M. Novak D. F. Ross Nuclear Reacter Reculation J. Knight G. M. Holahan W. Hazelton J. Burns Div. of Ooerating Reactors D. Eisenhut H. W. Woods V. S. Noonan B. Grimes G. Chipman Executive Director for Ocerations H. L. Ornstein Div. of Site Safety & Systems Evaluation International Procrams H. Faulkner S. M. Coplan /?- b

PUBLIC ATTENDEES 221ST ACRS MEETING Friday, Sect. 8, - A.M. James W. Ashkar, Boston Edison Co., Boston, MA Charles S. Behanan, Carolina Power & Light Co., Raleigh, NC Marcos L. Boothby, Stone & Webster Engr. Corp., Cherry Hill, NJ C. S. Benwart, B&W, Lynchburg, VA Robert J. Breen, EPRI, Palo Alto, CA Charles B. Brinkman, CE, Bethesda, MD William E. Burchall, Combustion Engr., Bloomfield, CT William E. Burns, GA Power Co., Atlanta, GA R. G. Cockrell, Wash. Public Power Supply System, Richland, WA James F. Davis, Power Authority of the State of NY, Massapeyva, NY Louis 0. Delgeorge, Commonwealth Edison, Chicago, IL E. D. Fuller, Oeneral Electric, San Jose, CA Abel A. Garcia, EPRI, Copertino, CA Paul V. Holton, Bechtel, Springfield, VA Michael P. Harrell, Ebasco Services, Inc. 2 Rector St., NY, NY Douglas R. Jaquette, Stone & Webster, Cambridge, MA Angue Kimmins, WPPSS, Richland, WA Joseph W. Leavines, Gulf States Utilities, Port Neches, TX Lcuts B. Long, Southern Company Services, Birmingham, AL Samuel Miranda, Westinghouse, Pittsburgh, PA John P. Morin, Long Island Lighting Co., Hicksville, NY R. C. L. Olson, Baltimore Gas & Electric, Ludterville, MD Earl Page, Detroit Edison, Troy, MI H. C. Pfefferten, General Electric, California Rose Marie Randall, Offshore Power Systems, Jacksonville, FL William H. Rasin, AIF/ Duke Power Co., Charlotte, NC D. E. Schaffstall, KMC, Reston, VA Fred Stetson, AIF, Reckvillc, MD Terry G. Tyler, TVA, Knoxville, TN Friday, Sect. 8 - P.M. Achilles G. Adamantiades, EPRI, Chevy Chase, MD Stan Benjamin, Assoc. Pr. Garrett Park, MD Richard A. Guide, Naval Reactors, Apt. 511 4500 S., Arlington, VA T. Scull, Philadelphia Elect. Co., Haddcw Clifts, NJ J. C. Zink, Public Service Co. of Oklahcma, Tulsa, OK Saturday, Sect. 9 - M1cnael Horrell, Ebasco Services, Inc. 2 Rector St., NY, Ni 10006

APPENDIX II ACRS FUTURE AGENDA 9/5/78 ACRS MEETING TYPE OF REACTOR SER ISSUE PROJECT REVIEW VENCOR DATE OCTOBER NONE NOVEMBER ZIMMER OL GE 10/2/78 DECEMPER E0PPSAR/ESAR-205 PDA B&W 11/1/78 JANUARY SALEM 2 OL W 12/1/78 FEBRUARY SAN ONOFRE 2&3 OL CE 1/1/79 SHOREHAM OL GE 1/1/79 SEQUOYAH 182 OL W 1/1/79

1. - 7

APPENDIX III FFTF: Project Status Report ppa 7EC:' STNIUS PEPCP5' CN 'ISE FAST FLUX TEST FACILITY (FPI'F) AUGUST 25, 1978 Purrose: te Project is requesting ACRS advice to assure that the Current Plant Design is Satisfactory for Cperation without undue risk to the public. In addition,14PS advice is sought in specific areas where technical differences may exist between the prcject and the NRC Staff. Acolicant: CCE and its centractors. Reactor Cesicn: Power - 400 Mit Primary Ccolant - Scdium/3 loops Seccedary Coolant - Sodium /3 1ceps Secondary heat dumped to air blast heat exchangers. Fuel: U/Pu Cxide 74 fuel assemblies 6 centrol assemblies 3 safety assemblies 8 test assembly gesiticns 92 reflector assemblies 0 0 Primary Temperatures: Inlet-858 F/ cutlet - 600 F te Fast Flux Reactor is intended to-provide a high temperature, high flux (7 x 10"n/cm'-sec) envirenment for the testing of U1FER materials and fuel. te reacter is a 400 Mit lecp type Um with heat being rejected to the at=csphere thrcugh air blast heat excnangers en each of the 3 secondary sedit=. lecps. Se core is fueled with a 22-26% Plutemium oxida in CO2 *i#E2f** Prior ACPS Letters: (attached) July 15, 1975 July 13, 1971 May IS, 1973 June 14, 1971 January 13, 1972 May 14, 1971 i ACRS Review Histerv: Early ACES review of the FF:'? fccused en the ability of the reactor building and containment to centain a larger than postulated core disruptive accident. Se Cc=mittee's initial report (7/14/71) indicated the need to retain flexability in the design for a core retentien system be-Icw the reactor. In January 1972, the Cm~=:uttee recc== ended that "an intensive prcgram be started to develcp an ex-vessel, post accidenc ccre retention and ccoling system H-7

Status Report FFI? suitable for installation telcw the reacter, so that the required information will be available in time to enable installation prior to reac*ar startup, should the system be needed." In May 1973, the Ccmmittee made additional reca=enda-tiens that the search for significant accident event sequences be centinued and that develeg-ment work en an ex-vessel core retention system be centinued. In July 1975, the Committee agreed with the NRC Staff that sealing the reactor head cavity wculd not centribute significantly to safety and that the existing space beneath the reactor guard vessel should be retained so as "a not make im;:cssible the future installation of an ex-vessel core retention system. Ccnstruction Historv: Excavation cc:"plete - January 1971 First Concrete pcured - October 1971 Containcent in place - August 1972 Reactor Vessel in place - Ceced er 1974 Peactor Internals installed - July 1975 First Core Assembly Ccmplete - Ceced er 1976 Heat Transport Compenents Installed - Cctober 1977 Contair=ent Leak Test - June 1978 Sodium introduced in plant - June 1978 Secondary I.cep Scdium fill - July 1, 1978 Secpe of NRC, ACES Paview for Ceeratiens: CCE requested that the NRC review and: (1) Provide advice regarding the adecuacy of the FF"? design to ensure safe cperaticn. (2) Previde advice en the adequacy of the FFIT Technical Specifications. CCE has indicated that the NRC need not include assessments of the *as built" configuration, cen-struction audits, or evaluation of acceptance test results to verify that the plant is cen-structed in acccrdance with design criteria and documentation. Provision for safeguards and security were also excluded frem review, f-

Status F.e;x:rt Fr? Precent AC M S t.+c _,4.. Sutcc.nt ttee teetings were held en July 12, 1978 ~i Feview: and August 10, 1973. .v.e. =.. 4.~. - H.i.." '.' *.. *w a*- lo. 'i => " ~.-...-.d *.e.a. July 1'.. 2 - (1) ?.e EnF pers:=e1 reviewed the status of cen-structien and exterirental plans for de facility. Constructicn is in cc: plete vi= all 138 systems turned cver to da crerati g project by de cen-struction crganizaticns. ?.e,:rrarv system was filled wie =e sediun in early Juli 1973. Ccre loading will take place in 1979 and initial criti-cality is pla=ed for de su: er of 1979. 444... 3 a. ,,4,r....s, (2) %.e. /.,. .,t.,a o.= w.. . 4 ~ .~. by a ECE rec :est =at de 2C previde advice rega: ding :ne adecuacy of te design and tecnni-cal specificati:= f:r safe cperat:.cn. ?.e review did not include an asses =ent of de "as built" ccnfiguratien, c nstructicn audits er evaluaticn of acceptance test results. Frc-versicns for sadeguards and security were al - excluded frc:n :ne EC review. (3) .n.... 4,,. .e s3.u.. .,/. ~.ed a.e re.,ct. 4 ay e...n _s..A. . a..#" a.l.i.., ac,".d...e..a.* '.e ra.=c.~. e, a - is a fast breeder syste:n designed f:r 4C0 fueled witn U/?u cxide rd c cled by sedic. S.ree c.ri arv. c clant lec;s re. eve heat fr. the core to sec ndary scdiu lecps wnica c t.a.e1 e_ .,.4 x.i _e ,x.._.. era. ..a. am.-. -. handling is icne wi:n invessel and exvess. fueling racnines wnica remove fuel frc:n na core and reacter ressel to an =tarun dec.y storage regicn. ?.e core nas 74 fuel asce.:: lies, 6 cent:01 ::ds, 3 safety reds, and 3 incere test

csitiens (4 closed-lccp tests and 4 cpen-lecp test

pcsiticns). ?.ere are 92 reflector assextlies.

/~f-/O

Status Report FFTF (4) te Project Staff reviewed the electrical and instrumentatien and centrol systems. The electrical system has two indeperdent offsite supplies as well as redundant ensite diesel generators. S e diesels are connected to emergency power supplies which are not quali-fled as Class 15 (they are not seismic or tornado qualified). A redundant Class 1E system powered by batteries is provided for reacter shutdown, pcst-accident cenitcring, ard engineered safety features. (5) Padicactive waste dispcsal was reviewed. Liquid and solid wastes will be stored in existing facilities en de Hanford site. Gasecus waste (Krypten and Xenen) will be collected and held up in a cyro-genic system. (6) t e Project reviewed the design basis acci-dents which include reactivity addition events, less of cooling events, and various plant transient events. Precautices were taken in be plant design to prevent de initiation of ECCAs, and studies were carried cut en de margins for unexpected er unforeseen events. Se centainment margin for an ECCA is on de crder of 150 W/sec. Sodium voiding of de central fuel elements.cculd provide S3.00 of re-activity insertien, ccmplete ccre voiding w uld previde S1.00 of negative reactivirf. (7) Se Project reported en two items dat had rerained cpen at de ti.me of de c0nstruction permit review. With de agreement of de NBC Staff and the ACES, the Project sealed the centainment icwer cavief ard have

raintained an cpen head cc=partment.

(8) The NRC Staff briefly su:ne-i ad de 12 cpen issues tney have with de and Staff's basis for Seir safety revie, . Se FFr?. te cpen issues will te reviewed in de Staff SER to be issued August 1,1978 and will be discussed at the August 10, 1978 AC2S Sub-committee meeting. /7- /

Status Report FFl?. Aucust 10, 1978 subec:mittee Meeting (High]Ights of this =eeting attached.) Cpen issues between the NBC Staff ari the Project: Se NPC Staff Safety Evaluation Paport (August 1, 1978) lists 11 major items as cutstanding between the Project and the Staff (SER Section 1.9, pg.1-23). Rese include:

1) Natural Circulation

• 2)

Centaiment Adequacy 2cr Core Melt Events

• 3)

Piping Integrity 4) Cell Liner Integrief 5) Lcese Parts Monitcring 6) Maximum Fuel Channel Exit Temperature 7) Tornado Ccnditions 8) Centaiment Isolation valve Position Indicators 9) Centrol Rocm Isolation 10) Fire Eacards Isolation 11) Cicsed T.ceps

• Expected to be unresolved at the time of the full ACES Review.

Subsequent to the issuance of de SER and prior to de August 10, 1978 ACRS Subec=mittee meeting items 1, 4, 5, 6, 7, and 3 were censidered resolved. Fevies of items 9,10, and 11 have not yet bean ccmpleted by the Staff and will be recorted en in supplecents to the SER. At the August 10, 1978 Sutccmmittee meeting unresolved issues en Piping Integrity, Con-tainment Adequacy for Core Melt Events, Electrical Penetration, and Centrol Fccm Habitability wre re-viewed. It is expected that only the Piping Integ-rity and Centaiment Adequacy for Core Melt events will remain as unresolved issues at the full AC"6 review (items 2 and 3 abcve). / / _2

Status Report FPIF Discussion of Issues: Containment Adecuacy: te principal differences between de Project's evaluation and the Staff's position is the treatment of the sodium / concrete and fuel-steel / concrete attack rates and the manner in which radicactivity is released. De Project believes that sodium /cencrete reactions will be limited to a depth of abcut 12", the staff does not believe an upper tcund can te placed en de reaction depth. The Staff advccates censideration of a centro 11ed release rate to prevent centairment overpressurication (SER Secticn 15.3.7). te Staff also urges the project to further censider the problem of hydrogen accumulation and the pcssibilities of explosive mixtures. The project dces not believe that an explesure mixture of hydrogen and oxygen can form in a contairrent atmosphere containing a sodium aerosol. the NRC telieves stratification of hydrogen and sodium at:cspheres may rake explosive mixtures p:ssible. Piping Intecrity: te NRC Staff has agreed that failure of de reacter inlet piping need not be censidered a CA initiater provided dat certain conditiens are met. tese included preservice and inservice inspection, leak detection instrumentatien and apcrepriate materials surveillance. CCE is in de pa ccess of develeping a high temperature (400 F) ultrasonic testing device, tey have ccmmitted to inservice inspection of selected, high stress welds en tne seccndary sedium lecps wnen develeprant of de t.Tr device is ccmplete. te NFC telieves that provisiens for i.spection of de primary system should also be provided. S e project will install aercsol leak deteccer preces in the reacter guard vessel and in the heat transport cells. te NEC has recc= mended that dey be en Class lE pcwer; the Project has not agreed to this. A. Bates Peactor Ergineer / /3

HIGEICf"S eCFS.e.mc,.yvme_e. ,Cm.

s. v. C*.I

.a Fo. r Fw, d r.~ e,3.C., r-, .a ACCCST 10, 1978 @SHI'XT.C:1, D.C. De ACPS Fast Flux Test Facility (Fri'F) Schce=nittee held a meeting on August 10,1978, at 1717 E Street, N.W., Washington, D.C. to re-view the FP*F project. Dr. Ker: (Schce=nittee Chairran), Cr. Carben, Cr. !brk, Dr. Shewmen, and the folicwing ACFS censultants were present; Cr. Stratten, Dr. Seale, and Cr. B eent. PRESE27"A""C'! 3Y "'EE NEC DR The IE Staff indicated dat, suhsequent to de ACFS Schce=nittee meeting en July 12, 1978, several of de cutstanding issues in relation to instru-mentaticn and centrol had been resolved. Cther cutstanding issues remain to be resolved are: 1. Piping integrity 2. Centainment margins to : re melt events 3. Flectrical genetratien 4. Centrol recm hacitahility te 13C Staff and de ??"? project are in de precess of finding a mutually agreed upcn solutien for dese unresolved issues. A st= nary of de tac Staff advice to de repartment of Cnergy en several issues is attached. NW

Fn? Highlights August 10, 1978 w ~ =.~~. w ~n. ~..e.J 3v. n...= .= m.. P .,w .u De FETF project indicated that doy have been providing all the necessary informatica (calculations, evaluaticn of assumptiens, and experimental results) to the 17.C Staff so as to resolve de cutstanding issues. The FCF project recccendatiens to' resolve de differences between dem and the :3.C Staff, and future invcivement of de EC Staff in de safety of the FT F project are attached. In relatien to de ACFS role in de review cf de ???F project, the project is requesting ACFS advice to asure bat de current plant design is satis-factorv. for =.eratien wi2 cut undue risk to de cublic. In additien, ACFS advice is scught in specific areas where technical difference may exist between the FFIF project and de EC Staff. SUECC:0!!""EE CESERW?":CNS Cr. Shew cn requested de EC Staff to provide "ddi-4~"' infer =aticn en hydrogen generation and behavice under ccnditiens of interest. Dr. Seale requested de ::EC Staff to provide cdditicnal infer =atica en de burning before detenation aspects of the sedium-hyd:cgen-crigen mixture. De Subec nittee indicated that it wculd recc=end this project to be full Cc=nittee fer review during the 201st ACFS meeting. 0

ACRJ Suecomn s rTee MEET)HQ ??Yf Avao.sr to,19 7 8 W A.5 miJ cn rog.D C H. Ho L&

SUMMARY

AND ADVICE TO 00E A) NATURAL CONVECTION - PREVIOUSLY ADDRESSED NOT AT ISSUE BUT'ESSEITIAL TO OVEPALL SAFETY CONSIDERATION OF FFTF 1) DEM0tiSTRATE THAT NC IS A SAFE METHOD OF DECAY HEAT REMOVAL 2) CODE VERIFICATION OVER RANGE OF REACTOR OPERATION TO SUBSTANTIATE SAFETY ANALYSIS 3) LIMIT REACTOR OPERATI0tl UNTIL NC CAPABILITY HAS BEEN VERIFIED B) CONTA!NMENT MARGIN FOR CORE MELT EVE'lTS THE STAFF RECOMMEiDS THAT: 1) DOE SHOULD PERFORM LARGE SCALE MATERIALS IllTERACTI0tl EXPERIMENTS TO OBTAIN BETTER DATA ON HYDROGEM GENERATION AND BEHAVIOR UNDER CONDITIONS OF INTEREST. ff- / l

~ SUMtiARY AND ADVICE TO DOE (CONT'D) 2) A NEANS SHOULD BE PROVIDED TO MONITOR AND CONTROL THE HYDROGEN CO.PlCEi!TRATION IN THE C0i1TAlf2ENT BUILDING UNDER TlOSE CONDITIONS IN WHICH HYDROGEN CONCENTRAT'.0il COULD THREATEN CONTAINtEli INTEGRITY, 3) A MEANS SHOULD EE PROVIDED TO CONTROL THE VENT RATE OF THE CONTAINMEHT EUILDING TO PREVENT OVERPRESSURI7_ATIOil 4) STAFF BELIEVES THAT CONSIDERATION SHOULD BE GIVEN TO POSTULATED A',CIDENTS OF LONG DURATION REGARDING CONTRT_ acci: HABITAfILITY C) PIPING INTEGF THE STAFF'S Letter OF JULY 11, 1976 CONDITIONALLY AGREED THAT FAILURE OF THE REACTOR It!LET PIPING NEED NOT BE CONSIDERED A CDA IMITIATOR PROVIDING THAT CERTAIN C0flDITIONS WERE MET (SECTION 1-24). THESE CONDITIONS INCLUDED PRESERVICE AND INSERVICE /7-/7

-31 SUPMARYANDAdVICETODOE (Cont'c) INSPECTION LEAK DETECTIO:1 AND APPROPRIATE MATERIALS SURVEILLANCE. THE PROJECT. HAS UNDER DEVELOPMENT AN ULTRASONIC METHOD EUT IT WILL NOT BE DEVELOPED BY START UP FOR FFTF AND HAS PROPOSED TO USE IT IN THE SECONDARY SODIUM SYSTEM. CONSIDERING ALL THESE FACTORS THE STAFF RECOP.MENDA-TION Iii SECTICII 19.0 MAY BE STATED AS FOLLOWS: 1) DOE SHOULD CONTIlllJE WITH THE PRESENT ULTRASONIC SYSTEM UNDER DEVELOP!!ENT AND IiiPLEMElT IT AS S00N AS PRACTICAL FOR THE !!0T CROSSOVER PIPING SYSTEM WELDS. CONDUCT A COLD PRESERVICE INSPECTION NOW. ~ 2) INSTALL THE REMOVAELE SECTIONS OF INSULATION OVER THESE WELDS AND RETAIlI THE PROVISI0ilS FOR TRACKS AS ORIGINALLY PROPOSED TO THE STAFF. ' 3) INSTALL THE AEROSOL LEAK DETECTOR PROBES IN THE REACTOR GUARD VESSEL AS PREVIOUSLY PROPOSED NEAR THE REACTOR INLET PIPING. TEST THEM WITH AN IONIZING GAS AS ORIGliiALLY PROPOSED R- / P

- sui"1ARY AND ADVICE TO DOE (CcNT'D) 4) INSTALL THE SODIUfi AEROSOL LEAK DETECTORS (SID) FR0ii THE HEAT TRANSPORT SYSTEM CELLS AND PLACE THEM CN 1E POWER 5) WHEN THE ULTRASONIC SYSTEM INSERVICE INSPECTION SYSTEM HAS LIEN DEVELOPED, ESTABLISH PERIODIC INSPECTION PROGPAM, DRAINING Tile SODIUM LOOP AND FLUFulfiG IT WITH FRESH SODIUM IF NECESSARY TO REDUCE RESIDUAL ACTIVITY LEVELS TO PERFORM THE llECESSARY PERIODIC INSPECTION. 6) IN ADDITION TO THE IN REACTOR MATE. RIALS SURVEILLANCE PROGRAM WHEN COMPONENTS FROM THE MIMARY HEAT TRANSPORT SYSTE". HOT LEG ARE REMOVED THESE SHOULD'3E METALURGICALLY EXAMI? LED TOO FOR DEGRADATION PROCESSES. D) I?lITIALLY THE STAFF CONSIDERED THE CELL LINERS A SAFETY FEATURE NECESSARY TO PREVENT A CHALLENGE TO CONTAINMENT Ili THE EVENT OF LARGE HOT SODIUi1 SPILLS IN Frir, iHE LT-1 AND SMALL-SCALE FAILdD LINER TESTS INDICATED THAT SMALL FAILURES IN THE LINERS R-t 1

_5_ SuiV1ARY Mii' ADVICE TO DGE (CONT'D) APPEARED TO BE SELF-LIMITI:!G Il TERMS OF SODIUM-CONCRETE REACTIOli. THIS IS PARTICULARLY IMPORTANT TO THE REACTOR CAVITY. WE HAVE RECO:iME!!DED C0ilTI:lUATIO:1 0F FAILED LIiiER TESTS (WELDS) AfiD SODIUM-CLNCRETE REACTION TESTS. WE HAVE ALSO RECOMME!!DED DEVELOPMEIII 0F A PERIODIC TEST PLAN-T0 DEMONSTRATE THE CONTI:lUED FUilCTIONAL CAPABILITY ~ OF THE ACCESSISLE CELL LINERS. E) LOOSE PARTS VONITORIt!G ALTHOUGH NO L0iiGER Ail ISSUE, THE STAFF HAS REQUESTED CONTI.'iUED DEVELOPME!iT OF UNDE.' SCDIUM ULTRAScriIC DETECTORS. THE PROJECT HAS Af.iEED TO C0iiTI::UE SUPPORT OF THIS ACTIVITY. G g o

SUFMARY AND ADVICE TO DOE (Cont'o) F) PAXIMUM FU2L CHANNEL EXIT TEMPERATURE ON THE EASIS OF INFOPJ1ATION CURRENTLY AVAILAELE IN UNVERIFIED THERMAL HYDRAULIC CODES AND UNCERTAINTY IN HOT CHANNEL FACTOR VALUES, AND IN ORDER TO ASSURE NO BOILING IN THE CORE WE RECOMMEND INITIALLY LIMITING THE MAXIMUM FUEL CHANNEL EXIT TEiFERATURE TO 16700F RATHER THAN 1705 F. G) FUEL SYSTEM THIS ITEM IS NOT AN ISSUE. 20TH THE STAFF AND THE PROJECT RECOGN:IE THE UNIQUE CAPABILITIES IN FFTF FOR THE STUDY OF IRRADIATED FUEL PIN PERFORMA.1CE. WE STRONGLY SUPPORT A FUELS EVALUATION PROGRAM AND REQUEST SUCH A PLAN WHEN IT BECOMES AVAILABLE. 4 9

g . Suf1 MARY AND ADVICE TO DOE (CONT'D) H) TORNADO C0ilDITIONS THE STAFF HAS RECO:91E?IDED THAT IN THE EVENT THAT TORNADO CONDITIO!iS ARE EMIrlENT THE FACILITY BE SHUTDOWN RATHER THAN ENTER INTO AN UNPLU!l LED SHUTDOWN FOLLOWIi1G WHICH SUCH CERTAIN OPERATOR ACTIONS MUST BE TAKEi! WITHOUT THE STATUS OF THE FACILITY BEliiG KNOWN. WE HAVE ALSO REC 0ftENDED THE FACILITY NOT BE OPERATED Ii1 THE EVEliT OF ERRATIC BEHAVIOR OF THE CHECK VALVE IN THE OfiE MISSILE HARDENED CIRCUIT. THE PROJECT IS IN ACCORD WITH THIS REC 0fiMENDATION. 6 [ hk

RE C ORD copy hcRs Sueconno rree H seriad. i FFTP' AUCnv5T lo I976 s uns,auw. n.c. METHODS i0R RESOLUlWG DlWERENCES 3EMEEN 2 NRC AND PROJECT 1. NRC AND ACRS AGREEMENT IS GENERALLY AClllEVED REGARDING DESIGN AND OPERATION OF FFTF 2. WHERE UNRESOLVED TECHNICAL DIFFERENCES EXIST, p THE PROJECT TECHNICAL ~ BASIS AND NRC POSITION b ARE PRESENTED TO ACRS; ACRS ADVICE IS SOUGHT g. BY DOE IN REGARD TO Tile ISSUES 3. SHOULD UNRESCLVED ISSUES REMAIN, HEDL WOULD PROV1DE A TECHNICAL RECOMMENDATION TO FFTF PROJECT OFFICE; PROJECT OFFICE WOULD RECOMMEND A DOE POSITION, FOR RESOLUTION BY DOE HQ JiEDL 7808-003.76 DES

e WHAT FUTURE ICUOLUEMENT WiLL YHE QRC HAVE in PROJECT SAFETV? NRC WILL PROVIDE ADVICE REGARDING THE FFTF FSAR AND OPERATING AUTHORIZATION. THE PRO. LECT WILL MAKE AVAILABLE TO NRC INFORMATION ON ACCEPTANCE TEST g RESULTS. SPECIFICALLY NATURAL CONVECTION RESULTS. FURTilER ADVICE FROM NRC WOULD DE SOUGHT BY DOE IN Tile EVEIVf 0F MODIFICATION IN PLANT OR OPERATING PLANS SUCH TilAT DOE CONSIDERS THE CHANGE TO INVOLVE AN UNREVIEWED SAFETY QUESIl0N. HEDL 7808-003.78 DES

TABLE 1.2 2 PRINCIPAL DESIGN VALUES FOR MAJO R COMPONENTS ITEM AND NUMBER SUPPLIED VALud EACH REACTOR CORE. Pcwer Excluding C!csad Loops 400 MWt 2 Peak Flux at 400 MWt 7 x 1015 n/cm.3,e Core Volume 1034i Core Length 36 in. Initial Closed Lacps (2) 2.3 MWt Cociant Max Temp 1220up Test Diameter 2.5 in. REACTOR VESSEL (TYPE 304 SS) Overall Height 43 ft 1 in. Inlet Nozzfes (3) Size 16 in. Outlet Nozzles (3) Size 23in. Head Diameter (00) 25 ft Head Thickness (including shielding) 49.31 in. ' CONTAINMENT d Diameter 135 ft i Overall Height 187 ft Depth Belcw Grade 78 ft Pressure IntJExt. 10/0.2 psig Temperature High/Lcw 250/(-) 100F Allowable Laak Rata 0.1Ydday ELECTRICAL SYSTEM 115 kV Supply Feeder (1) 119 MVA i 13.3 kV Transformer (1) 50 MVA l 13.3 kV Standby Feeder (1) 3 MVA 480 V Emergency Generator (2) 1200 kW CLOSED LOOP EX-VESSELMACHINE i Sara Diameter (Naminal) 8-1/8 in. I Heat Removal Capability 10 kWt T-1.2-2

e \\ TABLE 1.2 2 -(Cont'd) ITEM AND NUMBER SUPPLIED VALUE EACH PRIMARY HEAT TRANSPORT SYSTEM Cooling Circuits (3) 133 MW Hot Lag Piping (Reactor to Pump) 00 28 in. -Structural Design Temperature 0 1050 F - Dasicn Pm s Reactor to HLIV - Ocign Pressure HLIV to Pump 15 p:ig - Valves (3) Size 120 psig 28 in. Hot Leg Piping (Pump to IHX) 00 16 in. -Structural Design Temp at 225 psig 1050ap Cold Leg-Piping Size 00 16in. -Struct Design Temp at 225 psig 8300F -Isolation Valves (3) Size 16 in. - Check Valves (3) Size 16 in. Pumps -(3) Flow at 500 ft Na Head Struct Design Temp at 250 psig 14,500 gpm 10500F Intermediate Heat Exchangers (3) 133 MW Struct Design Temp Shell Side at 225 psig 0 1050 F Struct Design Temp Tube Side at 250 psig 0 1050 F SECONDARY HEAT TRANSPORT SYSTEM Main Piping - Size 00 Branch Piping -Size 00 16 in. 8 in. Piping lHX to OHX Struct Design Temp at 250 psig 0 1000 F Piping DHX to INX Struct Design Temp at 250 psi 8300F Valves Struct Design Temp at 250 psig 0 Pumps- (3) Flow at 400 ft Na Head 1000 F Struct Design Temp at 250 psig 14,500 gpm 0 DHX Units (3) (4 Modules / Unit) 830 F 133 MW Modules (12) 33 MW Tube Side, Struct Design Temp at 250 psig 1000ap Mb T-1.2-3/( T-1.2-4 blank)

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FFTF PROJECT STATUS ~ e CONSTRUCTION ESSENil ALLY COMPLETE n e ACCEPTANCE & START-UP TESTING IN PROGRESS N e FSAR REVIEW APPROACillNG COMPLEil0N e FUEL LOADING TARGET MAY 1979 IIEDL 7807-005.74

FFTF CDA ENERGETICS Staff Position: b O The FFTF Vessel, Vessel Components, and Primary Loop Are Adequately Designed to Withstand the Mechanical Loading Resulting from a Conservative Spectrum of CDA Events Uni I?$ '3. <R .}}