ML20133K390

Summary of ACRS 303rd Meeting on 850711-13 in Washington,Dc

ML20133K390
Person / Time
Issue date:	10/07/1985
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2336, NUDOCS 8510210246
Download: ML20133K390 (232)

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, TABLE OF CONTENTS MINUTES OF THE tm ,i 3 y) 303rd ACRS MEETING  :

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JULY 11-13, 1985 P *

• WASHINGTON, DC I. Chai man 's Report (0 pen) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 II. Watts Bar Nuclear Station (0 pen) ........................ 1 III. Recent Events at Operating Nuclear Power Plants (0 pen) .. 4 A. Davis-Besse Loss of Main Feedwater and Auxiliary Feedwater ................................ 5 B. Hatch I Stuck Open Safety Relief Valve ............. 9 C.- Oyster Creek Scram Discharge Volume Isolation Valves Failure ..................................... 10 D. Rancho Seco Reactor Coolant System High Point Vent Leak .......................................... 11 E. Sequoyah 2 Reactor Trip - Improper Use of Test Instrument ......................................... 11 IV. Diablo Canyon Nuclear Plants Units 1 and 2 (0 pen) ....... 12 V. General Electric Standard Safety Analysis Report (GESSAR-II) (0 pen) ..................................... 15 VI. Quantitative Safety Goals (0 pen) ........................ 20 VII. Meeting with the Commissioners (0 pen) ................... 23 A. Consideration of Earthquakes in Emergency Planning . 23 l B. Safety Goal Implementation ......................... 26 VIII. EPA Standards for High Level Waste Repository (0 pen) .... 28 IX. ANL - West Survey of Control Room Habitability (0 pen) ... 30 X. Long Range Plan for NRC (0 pen) .......................... 31 XI. Human Factors and Maintenance Subcomittees on Natural Aptitude Selection Procedures (0 pen) .................... 32 XII. Briefing Regarding Steam Line Failure in Non-Nuclear Power Plant (Closed) ................................... 32 XIII. Executive Sessions (0 pen) .............................. 33 F510210246 851007 P

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TABLE OF CONTENTS (Cont.) '

. MINUTES OF THE 303rd ACRS MEETING A. Subcomittee Assignments

, 1. Meeting with the NRC Commissioners ............ 33 4 2. SALP Evaluation of Licensees .................. 33

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3. Incident Investigation Program ................ 33 B. ACRS Reports, Letters, and Memoranda

1. Reports, Letters, and. Memoranda  !

ACRS Comments on Proposed NRC Safety Goal Evaluation Report ............................. 34

2. Provisions for Protection Against Sabotage ... 34  ;

3. Long Tern Seismic Program for the Diablo Canyon Power Plant ............................ 34 ,

4. EPA Standards for High-Level Radioactive Waste '

Disposal ...................................... 34

5. Emergency Preparedness for Fuel Cycle ard Other Radioactive Material Licensees ................ 34

6. Investigation of Recent Incident at the Davis- '

Besse Nuclear Power Plant ..................... 35  ;

7. Control Room Habitability...................... 35

8. Materials Research ............................ 35 C. Future Schedule  ;

1. Future Agenda ................................. 35

2. Future Subcomi ttee Activi ties . . . . . . . . . . . . . . . . 35 D. Sustained Meritorious Service ...................... 35 i E. Joint Meeting of Nuclear Safety Comittees ......... 35 ,

F. Proposed Amendments to ACRS By-Laws ................ 36 G. Watts Bar Nuclear Plant ............................ 36 H. Indian Point Nuclear Plant ......................... 36 I. Licensing Process, Considerations of a National Academy of Nuclear Power Safety .................... 36 Proprietary Supplement 1-2 i I

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, r NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFECUARDS y~ I W ASHINGTON, D. C. 20066 g

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Revised: July 9,1895 SCHEDULE AND OUTLINE FOR DISCUSSION 303RD ACRS MEETING JULY 11-13,1985 WASHINGTON, D. C.

Thursday, July 11, 1985, Room 1046, 1717 H Street, NW, Washington, D.C.

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

2) S:45 A.M. - 9:30 A.M. Discuss topics for meeting with Com-missioners (0 pen) 2.1) Consideration of seismic events (ACRS re in emergency plannin10,1985)g(DWM/OSM dated June 2.2) Quantitative Safety Goals (status report)(D0/RPS)

3) 9:30 A.M. - 11:30 A.M. Meeting with Connissioners (0 pen) 3.1) Discuss topics inted above

4) 11:30 A.M. - 12:00 Noon Briefing Regarding Steam Line Failure in Nonnuclear Power Plant (Closed)

SEE HANDOUT 8 -

(Note: This session will be closed to discuss Proprietary Infonnation.)

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

5) 1:00 P.M. - 3:00 P.M. Quantitative Safety Goals (0 pen) 5.1) Discuss proposed ACR5 report regard-ing proposed NRC Quantitative Safety Goals (D0/RPS) 5.2) Meeting with NRC Staff and invited experts, as appropriate

6) 3:00 P.M. - 4:30 P.M. Diablo Canyon Nuclear Plant, Units 1 and 2 (0 pen) 6.1) Subcommittee report regarding ten-year seismic review (CPS /EGI) 6.2) Meeting with NRC Staff and the Licensee, as appropriate

7) 4:30 P.M. - 6:00 P.M. Recent events at operating reactors (0 pen) 7.1) Report of ACRS Subcommittee on Reactor Operations (JCE/HA) i

303rd ACRS Meeting Agenda .

7.2) Briefing by members of NRC Staff

8) 6:00 P.M. - 6:30 P.M. Future Activities (0 pen)

See Tab 4.1 4.1) Discuss anticipated ACRS meetings (MWL)

SEE HAND 0UT 4.2 4.2) Discuss proposed ACRS activities (RFF) i i

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303rd ACRS Meeting Agenda .

Friday, July 12,1985, Room 1046,1717 H Street, NW, Washington, D.C.

Watts Bar Nuclear Station (0 pen)

9) 8:30 A.M. - 9:30 A.M.

9.1) Report of ACR5 Subcommittee on quality assurance activities at the Watts Bar Nuclear Station (GAR /RKM) 9.2) Meeting with representatives of NRC Staff and the Applicant, as appropriate Quantitative Safety Goals (0 pen)

10) 9:30 A.M. - 11:00 A.M. 10.1) Discuss proposed ACRS report re-garding proposed NRC Quantitative Safety Goals (D0/RPS) .

General Electric Standard Safety

11) 11:00 A.M. - 1:00 P.M. Analysis Report (GE55AR II)(Open) 11.1) Report of ACR5 Subcommittee regard-ing the containment integrity for this type facility (D0/RKM) 11.2) Meeting with representatives of the NRC Staff and the Applicant, as appropriate 1:00 P.M. - 2:00 P.M. LUNCH EPA Standards for HLW Resository (0 pen)

12) 2:00 P.M. - 4:00 P.M. 12.1) Report of ACR5 Su bcomittee (DWM/OSM) 12.2) Meeting with representatives of the NRC Staff and the EPA, as appropriate ACRS Subcommittee Activity (0 pen)

13) 4:00 P.M. - 5:45 P.M. 13.1) Report of ACRS Subcomittees regarding:

13.1-1) 4:00 P.M.-4:45 P.M.: Air Systems - NRC Staff /

ANL-West Survey Control Room Habitability Practices (DWM/J05 )

13.1-2) 4:45 P.M.-5:15 P.M:

Long-Range Plan for NRC (MWC/JCM) l

- 303rd ACRS Meeting Agenda - 4-l l

13.1-3) 5:15 P.M.-5:45 P.M.: Human Factors and Maintenance Subconnittees on Natural Aptitude Selection Pro-cedures (GAR / DAW /JOS) 1 t

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- 303rd ACRS Meeting Agenda ,

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l Saturday, July 13, 1985, Room 1046, 1717 H Street, NW, Washington, D.C. l Preparation of ACRS Reports to NRC ,

14) 8:30 A.M. - 12:30 P.M.

(Open)

EPA standards for HLW repository (DWM/OSM)

GESSAR II containment integrity (D0/RKM)

. Quantitative Safety Goals (00/RS)

. Diablo Canyon Nuclear Plant -

Ten-year seismic review (CPS /EGI)

Indian Point - Implementaticn of PRA (D0/RPS)

Provisions to preclude sabotage ,

at nuclear power plants ,

(JCM/J05) l 12:30 P.M. - 1:30 P.M. LUNCH

15) 1:30 P.M. - 3:30 P.M. ACRS Subcomittee Activities TUpen/ciosed) 15.1) Report of ECCS Subcommittee on the MIST facility (DAW /PAB)(Open) 15.2) ACRS Procedures - Proposed changes in ACRS Bylaws ,

regarding conduct of members ,

and procedures for revision of the ACRS Bylaws (DAW /TGM)

(0 pen) 15.3) Reports by ACRS members regarding meeting with the RSK and the GPR (DAW et al/

TGM)(Closed)

(Note: Portions of this session will be held in closed session as required to discuss infomation pro-vided in confidence by a foreign source.)

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TABLE OF CONTENTS  ;

APPENDIXES TO MINUTES OF THE i 303RD ACRS MEETING  ;

JULY 11-13, 1985 l

- f Appendix I -

List of Attendees ........................ A-1 i Appendix II -

Future Agenda .............................. A-8  !

Appendix III -

Schedule of ACRS Subcommittee Meetings ..... A-10 '

Appendix IV ' -

Staff Presentation of Watts Bar IDVP &

Allegations ................................ A-43 ,

Appendix V -

Concerns RE TVA Construction Sites ......... A-52 l I

Appendix VI -

Davis-Besse Feedwater Systems .............. A-70 Appendix VII -

Description of Davis-Besse Loss of MFW & AFW A-73 Appendix VIII .- Recent Significant Events .................. A-80 Appendix IX -

Rancho Seco Isometric Diagram RCS High Point Vent ................................. A-101 -

Appendix X -

.GE Presentation on GESSAR-II Containment Capability ................................. A-102 i

Appendix XI -

BNL Containment Structural Verification  !

Studies .................................... A-114 l

Appendix XII -

NRC Presentation GESSAR Severe Accident  !

Threat to Containment ...................... A-145 Appendix XIII -

NRR Comments on Safety Goal Evaluation ..... A-164 ,

i Appendix XIV -

F. Rowsome Comments on Safety Goal  !

Evaluation Report .......................... A-169 Appendix XV -

Consideration of Potential Complicating Effects of Earthquakes on Emergency Plan ... A-171 Appendix XVI -

Assessment on Field Applications of Control i Room Habitability Practices ................ A-174 r'

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Appendix XVII -

Briefing Regarding Steam Line Failure in Nonnuclear Power Plants .................... A-185  !

Appendix XVIII - Additional Documents Provided ACRS ......... A-198 1

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F0IA EXEMPTION (c)(4)

PROPOSED MINUTES OF THE 303rd ACRS MEETING JULY 11-13, 1985 WASHINGTON, D.C.

The 303rd meeting of the Advisory Comittee on Reactor Safeguards, held at~1717 H Street, N.W., Washington, D.C. was convened by Chairman D. A.

Ward at 8:30 a.m., Thursday, July 11, 1985.

[ Note: For a list of attendees, see Appendix I. C. Michelson did not attendthemeeting.]

Chairman D. A. Ward noted the existence of the published agenda for this meeting, and identified the items to be discussed. He noted that the meeting was being held in conformance with the Federal Advisory Comittee 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 Ann Riley & Associates, Ltd., 1615 I Street, N.W., Suite 921, Washington, D.C. 20006.]

I. Chairman's Report (0 pen)

[ Note: R. F. Fraley was the Designated Federal Official for this portionofthemeeting.]

D. A. Ward mentioned tnat ACRS Member D. Okrent was awarded the NRC Distinguished Service Award for 1985. The NRC Meritorious Service Award was given to Carol Ann Rcwe, Secretary to the ACRS Executive Director and a 30 Year Pin was awarded to J. C. McKinley, Branch Chief for Branch No. 1. Brief mention was made of the reorganization of upper management at TVA and problems at the Watts Bar, Sequoyah, and other TVA nuclear plants. D. A. Ward also mentioned the Idaho Loft Reactor partial meltdown experiment, which attempted to replicate the 1979 Three Mile Island accident.

II. Watts Bar Nuclear Station (0 pen)

[R. K. Major was the Designated Federal Official for this portion ofthemeeting.]

G. A. Reed indicated that a joint meeting of the Watts Bar and the Quality and Quality Assurance in Design and Construction Subcomittees was held on June 26th to discuss the quality assurance breakdown in the construction program at Watts Bar.

Quality assurance problems were mentioned in the August 16, 1982 ACRS OL Report. The Comittee requested that- it be kept informed of the situation. He noted that TVA has invoked a major quality l

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PROPOSED MINUTES OF THE 303rd ACRS MEETING assurance program including an independent contractor review of the design and construction of a typical vertical section of the plant to confirm the adequacy and safety of the as-completed plant. The Black & Veatch Engineering Company did an independent vertical slice QA review on the auxiliary feedwater system at Watts Bar Unit

1. Some 428 inconsistencies were catalogued of which 165 were due tc the fact that the auxiliary feedwater system was still under construction at the time of the review. TVA's inhouse Independent Review Policy Comittee and Black & Veatch decided that 115 of the inconsistencies or findings were not really deviations. There remained 148 findings which required further generic consideration.

G. A. Reed indicated that the Subcomittee had difficulty at the outset deciding whether the Watts Bar issue was one of quality assurance documentation or the quality of equipment installed in the workplace. As a result, there was a focus on the Black &

Veatch presentation. The Black & Veatch representative indicated that Watts Bar Unit I had been designed and had been installed in accordance with TVA's licensing commitments with the possible exception of three identified unresolved issues. The first of <

these three issues involves problems with cable tray loadings and fire retardant coatings that were put on the cables. The second issue involves loadings on embeddment plates with respect to

, restraints and other attachments. The third issue deals with 3 seismic design and the way in which the peaks were broadened on the '

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seismic spectra.

G. A. Reed explained that TVA's Board of Directors appointed a Nuclear Safety Review Staff (NSRS). This group decided on its own to review the Black & Veatch report. The NSRS group assessment concluded that from the work done by Black & Veatch there were no issues that would preclude imediate approval for operation of Watts Bar Unit 1. He indicated that the NRC Staff at the Subcomittee meeting was not prepared to agree or disagree with TVA or its NSRS group or with the Black & Veatch report. The NRC Staff indicated they had just received a draft of an INP0 Design and Construction Evaluation Report on Watts Bar but were not prepared to coment on the report or the status of release of the report. Also mentioned was TVA's solicitation of allegations from l construction workers and other regular employees to be evaluated by an outside firm named Quality Technology Company. G. A. Reed cautioned that the Watts Bar QA issue be kept separate from generic problems regarding TVA management in order that a manageable review  ;

could be conducted on the quality assurance issue. l G. A. Reed called for discussion of the INP0 Construction Evaluation Report. E. Adensam, NRC, indicated that the Staff has only a set of INP0 field notes which are part of an exit interview dated June 20, 1985. She indicated that the Staff does not have a copy of the draft INP0 Report and is not prepared to discuss the INP0 field notes at this meeting. G. A. Reed noted that the INP0 Report appears harsh as expected since INP0 is attempting to encourage licensees to seek excellence in the workplace. TVA must -

now respond to this report or these criticisms will stand.

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PROPOSED MINUTES OF THE 303rd ACRS MEETING i C. J. Wylie agreed that the Comittee must separate the quality [

assurance issues from other TVA problems in order to reach some closure on the quality assurance issue. He noted that the issue involving the cable tray loading is a common problem on all l construction sites where one does not give attention to tha l installation of cables on a continuing basis. He thought that the ,

overfilling of the cable trays by the field installers is more of i an issue of sloppy installation than a safety issue.

L W. Kerr asked with which one of the issues the Comittee should be l concerned. G. A. Reed stated that the Comittee should be L concerned whether TVA met the requirements the ACRS laid down in its April 16, 1982 letter and whether the QA activity at Watts Bar L is satisfactory to the ACRS. The Black & Veatch vertical slice on '

the auxiliary feedwater system indicated that a big problem did not i exist for TVA from a quality and quality assurance standpoint.

However, the new INP0 report, which is a general review of quality assurance in design and construction, does seem to add further confusion with regard to the QA issue. C. P. Siess pointed out it ,

was his belief that Black & Veatch focused on the quality of the  !

product as well as the quality assurance paperwork problem. He i indicated that the fact that they did not find any deficiencies in i j the_ design is very positive. Despite the possible paper problems, i J he stated that he was confident that significant deficiencies in  :

the design are not present. l t

i E. Adensam, NRC Licensing Branch Chief, in charge of licensing on Watts Bar, contrasted the Black & Veatch report with the detailed Staff review of it. She indicated that the Staff is satisfied that i Black & Veatch fulfilled its requirements in the study, and that the findings are reasonable. This included one of three findings still under review bearing on Bulletin 79-02 which has to do with base- plate flexibility and anchor bolts. j.

E. Adensam noted that half of the Staff's review effort involved the TVA Independent Review Policy Comittee generic review of the Black & Veatch findings to consider the implications for the rest of the plant and other safety systems that might be impacted by the Black & Veatch findings. As a result of some of the concerns that have been raised recently with regard to the report, the Staff has  :

decided that further work is necessary. A dedicated review group [

composed of representatives from Region II and NRR has been i 4

established to determine if TVA did a good job in addressing the >

findings of the Black & Veatch Report regarding their generic  !

applicability and corrective actions on the plant design and ,

construction. This dedicated review group will not only close out j the IDVP review, but will also address allegations received -

regarding the Black & Veatch review (see Appendix IV). [

E. Adensam indicated that the TVA Policy Comittee established a  !

TVA Task Force to~ do the actual work regarding review and generic application of the Black & Veatch work. Because of allegations

regarding the Black & Veatch IDVP, TVA's' Nuclear Safety Review f l Staff looked at the Black & Veatch documents independently, the ,

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PROPOSED MINUTES OF THE 303rd ACRS MEETING Policy Comittee report, and the work of the TVA Task Force and made recommendations to the TVA management regarding corrective actions to be taken. The Staff intends also to look at all reports and related documents that address the allegations.

E. Adensam indicated that the Staff had requested that the licensee provide additional documentation and the answer to certain questions to resolve Staff concerns prior to preparation of an SER.

She mentioned a May 16, 1985 Staff letter to TVA, with two enclosures which outlined the allegations the Staff has received (see Appendix V).

G. A. Reed noted that the proliferation of comittees reviewing the Black & Veatch study is making it difficult to detennine who is  !

responsible for resolving tM issues and allegations. He suggested that authority for the entire review effort ought to be placed in the hands of one qualified individual. He noted that the ACRS, because of the status of the current situation, is not in a position to make any judgments at this time. F. J. Remick agreed that the ACRS could not profitably contribute to the process at this point. He pointed out that the 1982 ACRS letter asked that the Committee be kept informed and the Staff is doing just that.

Certainly the ACRS is not the one to investigate allegations. He suggested that further action ought to be deferred to further developments. D. A. Ward indicated that the next step appeared to be review of the Staff conclusions after the Staff has completed its review. He asked for the Staff's schedule for completion of its review. E. Adensam indicated that the Staff has had some difficulty with scheduling because TVA has not submitted a fuel load date. In answer to a question by C. P. Siess, E. Adensam indicated that the Staff will definitely supplement the SER concerning the IDVP.

D. A. Ward then indicated that the next step appeared to be review of the Staff's SER when it is in a final draft. He thought that the decision about another Subcomittee meeting or - just full Committee discussion could be determined at a later date. The Comittee briefly discussed the method the Staff will use to evaluate and resolve the allegations on Watts Bar. E. Adensam indicated there it an action office in NRR that will have responsibility for resolving allegations and reporting back to the alleger if that individual can be identified. C. P. Siess noted that a more formal organized approach was taken to allegations on Diablo Canyon and Waterford 3. E. Adensam in: plied that if the number of allegations reached a very large, hard ta :ranage number, a different procedure than the action office would probauly be used.

III. Recent Events at Operating Nuclear Power Plants (0 pen)

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

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PROPOSEDMINUTES0FlHE303rdACRSMEETING A. Davis-Besse Loss of Main Feedwater and Auxiliary Feedwater E. Jordan, Division Director, Emergency Preparedness and Incident Response, I&E, indicated that the NRC Staff has drafted a Commission Paper, SECY-85-208, entitled " Incident Investigation

. Program" which proposed to the Commission a manner for dealing with 4

operating reactor events. It is a response to the Brookhaven study on the need for an NTSB-type board for investigating nuclear plant accidents and ACRS findings. The Staff advised the Commission of its intention to proceed, for the interim, to use this method as a a response to the Davis Besse event on June 9, 1985. A multi-disciplinary team was established made up of technical experts from various NRC offices. A memorandum from the ED0 identified E. Rossi as team leader, (reporting directly to the ED0)

J. T. Beard of NRR, L. Bell, I&E Training Center, and W. Lanning, AE00, as experts in the various areas that seemed appropriate for this particular investigation. This special investigative team will prepare a single report which will focus on fact finding, identify the root causes, and provide findings and conclusions.

Recommendations would not be part of this report but would be developed subsequently by the program office responsible for the varicus areas. He indicated several steps in the investigation at Da~vis-Besse including the fact that statements were taken from plant personnel involved on shift at the time of the event: l Strip chart records were reviewed Procedures, logs and manuals were reviewed Equipment involved was inspected He noted that the program is being administered by AE00 despite the fact that E. Rossi reports directly to the ED0 in tenns of findings.

A. DeAgazio, NRR Project Manager for Davis-Besse, described the feedwater system at the Davis-Besse nuclear plant. He explained that the Davis-Besse plant has two once-through steam generators, two steam-driven main feedwater pumps which provide flow to either steam generator, and two turbine-driven feedwater pumps that provide auxiliary feedwater flow to either one or both of the steam generators depending on valve alignment (see Appendix VI). He noted that this plant also has a small capacity electric motor-driven startup feedwater pump normally used just during startup and latter stages of plant shutdown (the capacity of the electric motor pump).is pump less than

[Since the that of one turbine-driven characteristic auxiliary curve of the electric feedwater startup feedwater pump is somewhat flat, it is difficult to say whether it has half the capacity of an auxiliary feedwater pump, but it can be assumed that it has no more than half that capacity depending upon thepressureofthesteamgenerator.] J. C. Ebersole noted that at the time of the incident, plant management . contemplated installation of additional pumps at some points in the future. A.

DeAgazio indicated that there is a license condition in effect that would require that Davis-Besse provide a new startup feedwater pump 4

in a new location with a capacity equal to one of the auxiliary

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PROPOSED MINOTES'0F TSE 30 E ACRS MEETING  ;

feedwater pumps. This new startup feedwater pump is not in the

, plant yet. It is scheduled for installation during the next  ;

refueling outage which would nominally be the spring of 1986. j A. DeAgazio indicated that just prior to the event the plant was  !

operating at 90 percent power with one main feedwater pump on ,

automatic operation and one main feedwater pump in manual control. t The reason for this was that they were experiencing difficulties with the speed governors on the main feedwater pumps. Both of the l feedwater speed governors had been replaced with new models at the  !

last. refueling outage. The initiating event was the tripping of ,

the main feedwater pump in autcmatic on overspeed. The reduction  ;

in feedwater flow initiated a power run back and approximately 30 l seconds later the rower level was down to, 78 percent, i Nevertheless, the run back wasn't fast enough and a high pressure reactor trip occurred (see Appendix VII). The main steam isolation  ;

valves then tripped with the effect of stopping all steam flow from i the steam generators since one main feedwater pump running in  ;

manual was deprived of steam. It coasted down on steam stored in  !

the system beyond the isolation valves and, at approximately four l minutes, it tripped. He noted that there should not have been a >

signal from the steam and feedwater rupture control system at that j time which closed the main steam isolation valves since steam  !

generator levels were normal. This signal provides for starting uf '

the auxiliary feedwater pumps in the event of low steam generator l water level and provides actuation of the auxiliary feedwater ,

i system in the event that all four reactor coolant pumps are lost [

(to promote natural circulation).  !

H. W. Lewis asked why the main steam isolation valves closed at '

this point. A. DeAgazio indicated that the Staff does not know i' exactly why this occurred. He noted that if there is a steam line break, both main steam lines are isolated. If there is a low steam  ;

pressure trip on the steam and feedwater rupture control system, ,

the main steam isolation valves close on both steam generators. It  :

does not matter which steam generator has the low pressure.  !

1 A. DeAgazio discussed the feedwater flow paths on a low steam  !

generator water level or on a high feedwater to steam pressure  !

i differential indicative of a feedwater line break (see Appendix '

l VI). 4 i A. DeAgazio indicated that six minutes into the event there was an i actual low level steam and feedwater actuation signal generated for r steam generator number one. The operator, recognizing the steam generator water level was dropping, attempted to initiate auxiliary '

feedwater and not depend upon the steam and feedwater control system. He activated the Steam Feedwater Rupture Control System (SFRCS) on low pressure instead of low water level. By activating '

the SFRCS on low pressure, the SFRCS was signaled that both >

generators had experienced a steamline break or leak and the system i responded, as designed, to isolate both steam generators. About one  ;

minute later, the operator recognized the error he had made and ,

attempted to correct it but two auxiliary feedwater isolation >

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PROPOSED MINUTES OF THE 303rd ACRS MEETING valves failed to open. The operators were then dispatched throughout the plant to attempt to restart these pumps. An attempt was made to get the electric motor-driven startup feedwater pump in operation and to replace some control fuses which were pulled to prevent starting of this pump against a closed suction valve. The operators were successful in restarting the startup feedwater pump and also successful in resetting the turbine-driven feedwater pumps. Thus, at 16 minutes into the event, the Davis-Besse operators had begun restoring feedwater flow to the steam generator.

A. DeAgazio indicated that, just prior to restoring feedwater flow to the steam generators, the pressure in the reactor coolant system rose to the point where the PORVs were actuated. They actuated three times, and there was indication that the PORY failed to reseat after the third actuation. There is some question whether '

this was an actual failure of the valve. The latest informt. tion from the site is that the valve has been disassembled and nothing abnormal found. The reason why the PORV stuck open, if it did stick open, is not known. Once the feedwater had been restored the plant entered a normal cooldown.

A. DeAgazio indicated that there were at least 13 or 14 different malfunctions or failures or unexpected occurrences. The initiatin event was the main feedwater trip on overspeed (see Appendix VIII)g.

H. W. Lewis suggested that the fact that there were nine independent failures of components in the sequence that followed the tripping of one main feedwater pump suggests that there are serious implications from this event. J. C. Ebersole suggested that this is a milestone occurrence in the context th!Lt one may have to be somewhat less optimistic about the usefulness of PRAs regarding the probability of this combination of events. H. W.

Lewis noted that there is zero probability that there would be ten independent failures. He suggested waiting until all of the findings are out. H. W. Lewis hoped that the Staff's incident investigation team would be able to get at the root cause of this event.

J. C. Ebersole asked how many minutes were left before core damage would have occurred from core uncovery. B. Sheron, NRC, indicated that hand calculations were done and more sophisticated calculations with RELAP 5 will be reported to the ECCS Subcommittee on July 31st. He suggested that if the operator had taken no action whatever to turn on the makeup pumps or startup feed pumps or open a PORV, core uncovery would have . occurred in about 50 minutes. The steam generators will normally dry out in a few minutes under these conditions. This scenario involves the displacement of water by steam collecting in the high points until the vent path is uncovered which is a surge line to the pressurizer. It is then just a case of boil off. It was determined that if both makeup pumps were started, the PORV opened, and the startup feedwater pump started, there was sufficient capacity to remove decay heat and the core would have stayed covered. Starting just the two makeup pumps and opening the PORV

PROPOSED MINUTES OF THE 303rd ACRS MEETING would still have kept the core covered. Startup of one makeup pump would have uncovered the core, but would have extended the time considerably. He pointed out that the event occurred at 90 percent power and the calculational results would have been different at 100 percent power. If the plant had been operating at 100 percent power and just two makeup pumps and no startup feedwater pump had been initiated at the time of steam generator dry up (two or three minutes into the event) there would not have been core uncovery, but had the operators waited 20 minutes to initiate the makeup pumps core uncovery would have occurred.

G. A. Reed indicated that he was surprised at the estimate of 50 minutes until core uncovery. He suggestsd that this was because this particular B&W plant has high set steam generators versus other B&W plants with low set steam generators. One gets the drain down advantage of the loop piping on the steam generator primary side. B. Sheron indicated that the volume of the core is about the same in both cases, such that the raised loop plant does not behave much differently then the lowered loop plant in terms of time to core uncovery. Also a factor is the relative location of the surge line on the. hot leg. G. A. Reed noted only one PORV on the plant had an in-series block valve. B. Sheron noted that all B&W plants have only one PORV. He indicated that the PORV, HPI or makeup system was not originally designed for feed and bleed, and the PORVs were installed in these plants to protect the lifting of safety valves. Obviously, a plant that has a higher PORY capability with more PORVs will have a much higher capability of feed and bleed.

J. C. Mark took note of the fact that except for the one operator error during the succeeding hectic 15 minutes in the control room.

the operators did everything about right. and very quickly. A.

DeAgazio indicated that from a control room design review of the incident, the one operator error was attributed to a human engineering defect.

W. Kerr indicated that he thought the auxiliary feedwater system i was an automatic start system. A. DeAgazio indicated that the auxiliary feedwater system is automatically started by the steam and feedwater rupture control system. W. Kerr asked why the operator who punched the wrong buttons had to punch any buttons at all. A. DeAgazio indicated that this is a manual action followup in most plant procedures to confirm that an automatic action has taken place. G. A. Reed agreed that it appeared that the operators did a good job except for the one error. However, he expressed i concerns regarding several design errors, such as the exclusive use of steam-driven pumps as auxiliary feed pumps and a very complex i valving arrangement to introduce water into one or the other of the

! steam generators. He also noted his concern regarding the CESSAR

! system-80 plants and the large number of closed valves in the

! valving arrangement at Davis-Besse. He noted that valves have a l

habit of not functioning and are often quite unreliable. The systems would be better off without closed valves from a reliability point of vi-ew, J. C. Ebersole added that the j l

PROPOSED MINUTES OF THE 303rd ACRS MEETING turbine-driven pumps are cooled by ac-driven environmental cooling fans which makes them interdependent with the ac system. An electric power failure or a steam failure, or any other kind of failure, can remove these pumps from service.

E. Jordan complained that it was 36 minutes after the turbine trip that the NRC was advised with cursory information on this event.

W. Kerr thought it more important that the operators look after the accident then inform the NRC Information Center. E. Jordan indicated that he had to insist on both of these items since the NRC would not have been in a position to initiate emergency response capability and make determinations as far as offsite measures, had the event gone awry. In answer to the question by G.

A. Reed, E. Jordan indicated that the Davis-Besse operators met the regulatory requirements which called for a response within one hour.

B. Hatch I Stuck Open Safety Relief Valve G. Rivenbark, NRR, indicated that on May 15, 1985, a crane passing overhead ruptured a line that provides water pressure to the charcoal filter deluge valve (the fire protection valve). This pressure which normally kept the valve closed caused the valve to open and spray water into the charcoal filters. The operator discovered the problem 15 or 20 minutes later when water began dripping into the control room through the air conditioning ducts.

J. C. Ebersole asked how the crane which is out in the turbine hall got . involved in this incident. G. Rivenbark indicated that the crane was passing from one turbine building to the other turbine building and passed over the control room. The floor immediately above the control room is the location of the air conditioning equipment.

G. Rivenbark indicated that as a result of the water dripping into an instrument panel in the control room, one of the SRVs opened several times and closed and finally stayed open. The operator tripped the reactor and started feedwater pumps and quickly recovered the reactor water level that had begun to drop. While the operators spent about 30 minutes attempting to close the SRV by pulling incorrectly labeled fuses, the SRV closed by itself. J. C.

Ebersole noted that the placement of the deluge system with a potential for straight drop down into the control room from a water pipe is almost akin to the classical scenario of having a toilet on

the ceiling which overflows after you flush it. G. Rivenbark agreed that if the drains plug up, the water will flood into the charcoal filter box and will run over and leak into the control room. J. C. Ebersole thought that the licensee should get the water ingress potential completely away from the panel board where the safety relief valve was electrically connected, not just to clean the plugged drains. G. Rivenbark indicated that the utility i

is looking at the possibility of removing the water altogether from the area.

l l

PROPOSED MINUTES OF THE 303rd ACRS MEETING C. Oyster Creek Scram Discharge Volume Isolation Valves Failure D. Powell, I&E, indicated that the event at Oyster Creek which occurred on June 12, 1985, is. of interest because it mimics an event that occurred at Hatch in August 1982 where uncontrolled l leakage occurred out of the scram discharge volume drain valves.

l Following a scram signal on high drywell pressure while the operators tried for 38 minutes to clear the scram signal, two particular valves failed by different failure mechanisms. One valve failed to go fully shut, or was hypothesized to have been originally fully shut and 'to have been forced open due to the pressure build-up from the leaking valve. This valve had an improperly sized spring in the valve actuator. Reactor coolant was discharged to the reactor building drain tank. Release of steam from floor drains and blistering paint on a hot pipe caused a portion of the reactor building deluge system to activate. There was no damage to the equipment inside the reactor building due to the actuation of the deluge system or due to the steam. There was some radioactive contamination, mostly short-lived radionuclides, at the 23-foot level. The potentialfsignificance of this event was the uncontrolled leakage of radioactive coolant outside of the containment. D. Powell noted that the licensee had CRD seal high temperature alarms which respond intermittently. The problem of CRD high temperature alarms is not significant for this particular event but for later operation of the plant. Should those seals degrade due to the high temperatures, there is the possibility-that they would be unable to operate properly. He noted that the failed valves were categorized as B valves under ASME Section 11 and were never leak-rate tested after installation of the backfit modifications.

J. C. Ebersole pointed out that there is controversy about the potential leakage from the scram dump volume even due to metallurgical failure of the volume proper, not to mention valve failure as being a potentially serious event. These valves should be recognized as safety-related valves. He asked if this was a standard problem for all BWRs. D. Powell indicated that probability studies done on the system show that the probability is not high for failure of the piping compared with other types of events that would lead to core damage. J. C. Ebersole indicated that the probability of piping failure was virtually zero, but the analysts ignored the presence of the valves which made an aperture through the pipe. J. C. Ebersole asked if the Staff intends to upgrade these valves. D. Powell indicated that NRR was not pursuing the matter at this point. E. Jordan indicated that the Staff will probably issue an information notice and then make a subsecuent decision on whether action is needed.

J. C. Ebersole asked how the utility stopped the leakage. D.

Powell indicated that they basically blew down the System by starting up the reactor water cleanup system and bled the letdown portion of that to the condenser.

PROPOSED MINUTES OF THE 303rd ACRS MEETING D. Rancho Seco Reactor Coolant System High Point V e t Leak H. Wong, I&E, indicated that on June 23, 1985, while in hot shutdown, the Rancho Seco plant had a 20 gallon per minute nonisolable primary coolant system leak on the high point vent system on the B steam generator. He described the RCS loop top portion of the candy cane and the additional piping which was the original design. The cause of the leak appeared to be missing supports, and fatigue failure as a result of an RCS vent line addition. Two additional pipe supports had to be modified and the addition of one cross-brace member was required. Investigation

, reveals that these support changes had not been performed although records stated that work had been done and inspected (see Appendix IX).

H. Wong indicated that what was important was not so much the additional pipe that was added as a TMI modification, but the removal of the spool piece in the middle which was to provide that the nitroger system would not be ;ontaminated during operation.

The spool piece was planned for removal during operation and would be put in place for nitrogen blanket purposes during outages. The purpose of this was to transmit loads across both sides of the pipe. The dummy spool piece was designcd so that it would go back during operations basically giving rigidity to the pipe so that loads would be transmitted properly. Without the spool piece, the pipe was hanging as a cantilever. In 1983 the stainless to stainless weld made at the time of the TMI 2 change cracked. P. G. ,

Shewmon asked what kind of administrative action is likely to be taken against the person who signed off that these pipes were in

place and the hangers were in proper order. H. Wong indicated that '

he was not aware of any licensee actions.

E. Sequoyah 2 Reactor Trip - Improper Use of Test Instrument ,

E. Weiss, I&E, indicated that on May 22, 1985, Sequoyah Unit 2 tripped from 100 percent power on overpower Delta T. This event

' demonstrates how a plant can trip following an approved procedure despite all the precautions in place to prevent maintenance or surveillance activity from causing this sort of event. An instrument technician had to take temperature readings from four protection cabinets located approximately 15 feet apart using a digital voltmeter. The voltmeter leads were incorrectly connected to the ammeter sockets in the voltmeter with the result that the internal resistance of the voltmeter was much lower than it should t have been. The technician had to go to all fcur protection sets within three minutes. He tripped one channel and although the reactor operator in the control room noticed the trip it all happened too quickly for him to respond. The plant went down on two out of four coincidence.

1 E. Weiss mentioned a similar event where a digital voltmeter caused a shorting of the output transistors on a reactcr protection sys te.n. An information notice was issued on that subject. The' corrective actions taken for this particular event include a precaution about procedures. One must look for the proper expected ,t value of voltage before proceeding to the next piece of equipment.

- 11 ,

PROPOSED MINUTES OF THE 303rd ACRS MEETING

' J. C. Ebersole suggested that TVA ought to outlaw multi-purpose meters for its personnel. W. Kerr thought it might be best to do this sort of testing when the plant is not operating. J. C.

Ebersole noted that this was a full power thermal measurement, a thermal heat balance value which has to perfonn during operation.

IV. Diablo Canyon Nuclear Plants Unit I and 2 (0 pen)

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

C. P. Siess explained that a requirement for a long term program to reevaluate the seismic design basis for Diablo Canyon was included as a licensed condition for Unit 1 primarily as a result of concerns ' expressed in a July 14, 1978, ACRS letter. The letter recomended that the seismic design of Diablo Canyon be reevaluated in about 10 years, taking into account applicable new information.

The license condition required reevaluation of the geology, seismo-tectonics, earthquake magnitude, the ground motion of the site, and earthquake engineering. The Licensee, Pacific Gas &

Electric Ccmpany (PG&E), issued a plan in January 1985 which the Staff has urder review. The ACRS discussed the plan with the Licensee and their consultants at a .subcomittee meeting near Los Angeles on March 21. The ACRS consultants viewed the plan, in general, with favor. There were no significant concerns expressed.

A joint meeting of the Extreme External Phenomena and Diablo Canyon Subcommittees was held on July 10, to hear the Staff evaluation of the PG&E plan. In the several months between the time the plan was submitted and the time the Staff finished its review, the Staff submitted a list of comments, questions and suggestions to PG&E.

PG&E responded to all of them. The Staff concluded that the program was acceptable and issued a draft of its evaluation for ACRS coment. ACRS consultants were particularly pleased that Dr.

Slemons of the U. S. Geological Survey would be working with the Staff. S. Brocoum, NRC GeoSciences Branch, indicated that the Staff will hear progress reports from PG&E quarterly and meet with them at a minimum every six months.

C. P. Siess indicated that the Subcommittee on July 10, found the program acceptable with one exception. The NRC has requested that PG&E assess the significance of the conclusions drawn from the seismic reevaluation studies with a probabilistic risk analysis as well as deterministic studies as necessary to assure the adequacy of seismic margins. D. Okrent has taken exception to the Licensee's proposal to do a Level 1 PRA (level of damage states only). He indicated that the Staff is satisfied with a Level 1 PRA as a basis for assessing the significance of whatever conclusions may arise regarding ground motion. He proposed that the ACRS accept the recomendations of the Subcommittee regarding the program plan and leave the question of the PRA level to a discussion with the Staff and the Licensee.

S. Israel, NRC, discussed the PRA analysis process. He indicated that the PRA analysis " front end" starts off with plant systems PROPOSED MINUTES OF THE 303rd ACRS MEETING analyzed in event-tree, faul t-tree fashion to detennine those 4

sequences that will lead to core melt. This analysis yields various plant damage states with information about the core

~ condition, core melt, and the condition of the containment. Of particular interest in a seismic analysis would be containment failure prior to core melt which would occur if the earthquake failed the containment. Other plant damage states of interest I

would be core melt without any core cooling, which usually occurs as a result of loss of all electric power, both offsite and onsite.

There are other damage states dealing with transients and LOCAs which basically are early core melts (large LOCA without core injection) or later core melts due to small pipe breaks or transient situations where one loses core cooling and ultimately i the core fails. Of concern is the availability of containment

) cooling which would be determined by an analysis of plant systems in the fault-tree process. This is basically a level 1 PRA, a plant damage state analysis. A Level 1 PRA does not include containment failure modes. A Level 2 PRA tries to characterize what happens to the containment, given one of the plant damage

, states. One is concerned about where the potential containment J failures occur and how they occur. Potential containment failure modes would include steam explosions (the vessel explodes and penetrates the containment), early hydrogen burns, late hydrogen

{ burns, overpressurization failures of the containment as well as basemat melt through. The split fractions are usually characterized by a release category. A source term is then

calculated for each one the release fractions. The Level 3 i

analysis, a consequence analysis, uses the source terms for the contairment failure modes for determining offsite consequences. S.

Israel explained that the' Staff has found, in terms of offsite consequences, that early fatalities are dominated by containment failure prior to core melt. Those types of sequences which include containment failure prior to core melt are basically the ones

, associated with a seismic event where the earthquake rocks the containment away from the auxiliary building or fails the containment. One could potentially cause early fatalities from long term containment failure usually through a station blackout

sequence (core melt without containment cooling) where containment

pressure builds up over a period of time and the containment fails i releasing radioactive materials to the environment.

S. Israel indicated that one of several reasons why the Staff did  :

not require the Licensee to extend his analysis beyond Level 1 is ,

the conflict between the old source term used in WASH 1400 and the i new source tenn being developed. Another reason that the Staff did  ;

not require more than a Level 1 PRA is the fact that such analyses can become very expensive. Such analyses will require j sophisticated computer calculations, phenomenological calculations in terms of containment loadings, sensitivity studies dealing with core debris coolability, as well as core-concrete interactions. He stressed that the purpose of the Diablo Canyon Study is to i investigate seismic characteristics of the site as opposed to other PRA work done to determine containment potentials. A third reason the Staff does not wish to require PG&E to do more than a Level 1 i

l

PROPOSED MINUTES OF THE 303rd ACRS MEETING PRA is the fact that traditionally the Staff has performed back-end analyses for PRAs and even though the licensees or applicants have submitted back-end analyses, the Staff does audit calculations anyway. It is like paying twice for the same work. Another reason for only requiring a Level 1 PRA is the fact that the Diablo Canyon site is a low population site, and the potential societal risks are not expected to be limiting. The thrust of the Diablo Canyon study is to deal with seismicity and geology and not resolve the ongoing source term severe accident work.

D. Okrent explained that the Diablo Canyon study would calculate accident states but would not evaluate containment failure modes except those that resulted from the earthquake itself. They would not evaluate the containment capability to withstand pressure and temperature nor would they evaluate whether the design basis, or as designed capability to withstand pressure and temperature in any way, is weakened by a severe earthquake. There'would therefore be a gap in their ability to assess the likelihood of different release categories. Obviously, the containment capability is vital to this analysis.

D. Okrent pointed out that the Diablo Canyon plant has a different configuration from other plants for which PRAs have been done and unless the Diablo Canyon study proceeds beyond Level 1, the Staff will be in a poor position to make estimates of releases, of the mode of containment failure, and the frequency of containment failure release categories. He suggested that the Licensee with a better knowledge of the plant is best qualified to do the Level 2 work. D. Okrent suggested that his estimate for doing the additional work for Level 2, despite the confusion about the source term, as well as estimates for the increased analysis costs is more conservative than that calculated by the Staff. He suggested that the - Staff's position is wrong and he thought that the Comittee ought to recommend a Level 2 PRA for the Diablo Canyon Plant.

D. Okrent made the point that a core melt guideline is not adequate without looking at risks. It is an incomplete lo~ok even though it is relevant to know the core melt frequency. He indicated that if one could get an estimate of containment release frequency, one wculd have the essential information for estimating the safety of the reactor.

C. P. Siess indicated that he was in a position to draft a letter which indicated that the Comittee approved the PG&E program plan and that the Comittee agrees with the proposed Level 1 PRA that the Staff has accepted. Comittee members could append additional coments . The letter could also recomend going beyond Level 1 to Level 2 or Level 3 for the PRA depending upon the sense of the Comittee. P. G. Shewmon pointed out that the only thing missing from the Diablo analysis is containment performance. D. Okrent indicated that if you do a PRA you want to estimate the releases from the containment. You have to evaluate the capability of the containment to withstand pressure and temperature. You do that in a Level 2 and not in a Level 1 PRA.

PROPOSED MINUTES OF THE 303rd ACRS MEETING

, V. General Electric Standard Safety Analysis Report (GESSAR II) (0 pen)

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

D. Okrent indicated that last month's full Committee meeting on GESSAR II discussed the subjects of containment capability and hydrogen but was not able to explore these subject in sufficient ,

depth.

Rom Vij, GE, presented a comparison of containment design pressures, ultimate pressure capability (calculated by different architect engineers), dome configuration, containment diameter, and containment fill at the base for GESSAR-II and several other BWR plants. He indicated that GESSAR-II is a free standing steel containment backed by a concrete shield building. The design i pressure is 15 psig (the ultimate pressure capability calculated to be 85 psig). The dome configuration is torispherical with a containment diameter of 120 feet. He pointed out that GE developed a methodology (given an energy dump after a break inside the drywell) to calculate pressure and temperature transients and the design . pressures and temperatures on the drywell in the containment. GE has suggested an equipment layout for the containment. The actual size dimensions set for GESSAR-II are not variable. GE has not provided the detailed structural design for every containment and this accounts for some of the differences in the diameters of the Grand Gulf, Perry, River Bend, and Clinton plants. R. Vij indicated that the detailed structural design is provided for the GESSAR-II containment and therefore the architect engineer is not allowed any flexibility.

R. Vij addressed the issue of fabrication flaws in the containment.

He indicated that GE's fracture mechanics group performed an analysis at the point of maximum stress in the containment for the ultimate pressure capability. It turned out to be in the knuckle region of the dome. The lower bound fracture toughness properties of the . material were input to the analysis and then the plate material of the containment dome, the weldment material and the heat effected zone around the welds were followed for resultant crack fomation. They picked out the most conservative value.which will give the smallest flaw and it came out to be the plate material. The fracture analysis results show that a potential crack, which is three inches long and a half inch deep, can be  ;

tolerated without unstable propagation to failure. This half inch depth is more than 25 percent of the wall thickness. He contended that with controls on the welding procedures, including preheating l and multi-pass to improve the impact properties of the weldings, GE does not expect to miss flaws more than about 10 percent of the wall thickness during inspactions. All the welds are 100 percent l radiographed on the outside where a flaw of up to two percent depth of the wall can be detected. Therefore, GE believes that undetected flaws should not compromise the calculated ultimate s' capability of the containment.

PROPOSED MINUTES OF THE 303rd ACRS MEETING J. C. Ebersole inquired regarding the viability of containment penetrations under high temperatures and pressures. R. Vij indicated that this sort of detail was not considered in the fracture analysis. D. Okrent asked what the possibilities are of missing a larger flaw and tne ' likelihood of the vessel failing at lower pressure from this flaw. R. Vij indicated that the carbon steel in the welds is considered stronger than the base metal by GE's metallurgists. He assured the Committee that there was no way that flaws of a large size could be missed, either in the inspection of base metals which are hard-rolled metal plates or during welding where there are at least 10 or 12 passes in one and

. three-quarter inch thick plate. The 100 percent radiography will pick up two percent angular flaws. D. Okrent asked the Staff whether, after radiographic analysis, flaws have been found larger than 10 percent of the wall thickness. J. Knight, NRC, indicated that the Staff really does not have the experience base to answer than question. But, he assumed that it is quite possible that one could initially have had flaws larger than 10 percent of the wall thickness ground out and corrected. D. Okrent indicated that he was trying to determine whether there is a significant risk of '

vessel failure due to the presence of larger flaws perhaps at a lower pressure. J. C. Ebersole suggested that perhaps the x-raying of welds should be a confirmatory process rather than an initial survey for finding defects. R. Vij indicated that radiography is a confirmatory process since the test results are used by insurance corrpanies and the architect engineers, as well as the NRC. R.

Klecker, NRR Division of Engineering, indicated that the Staff has made similar flaw calculations using a different analytical model and has found that one could tolerate roughly a two inch long through-wall crack or its equivalent crack area if you allow it to go a little longer. He stressed that any crack that may pre-exist should be found during an inspection even if one puts a margin of two or three on the crack size for uncertaint'es. J. C. Ebersole asked what would happen if the crack were to start later after fabrication. R. Klecker indicated that the Staff does not know of any mechanism that would cause a crack to grow at a later time because the containment is essentially only supporting its own weight. D. A. Ward asked if there is a high probability that a crack would be found in normal inspections. R. Klecker thought that the probability would be quite high because of the-requirements of the ASME code. The allowable crack sizes are considerably smaller than what GE, BNL, and th? Staff are calculating to be close to the critical crack size. He thought there was a sufficient margin so that the chances of missing a crack would be very small. The Committee discussed the detection of through-wall cracks. J. Rosenthal, NRC, indicated that the Staff calculates late containment failure (assuming no flaws) of the order of 11 to 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> into the event. Presuming that flaws did exist and hence the containment would fail earlier in time, one can look at the difference in consequences between a r. early containment failure and a late containment failure. The Staff calculates about a factor of three in persons-rems betwecn early and late failures. He indicated that the GESSAR-II containment will have venting procedures. The containment is vented in a l

l l

PROPOSED MINUTES OF THE 303rd ACRS MEETING

. period of 11 to 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> which allows ample time for the operators

, to take action.

M. Reisch, Brookhaven National Laboratory (BNL), indicated that BNL was asked by the NRC to verify structural failure analysis results presented in Appendix G of the GESSAR report, and to perform independent analyses for these structural members. The BNL deterministic studies concentrated on the torispherical steel containment, the drywell steel head, and the concrete roof slab. A reliability evaluation was then done on the torispherical containment (see Appendix XI). D. Okrent asked if extensive plastic deformation occurs, does one know the behavior of flaws that were subcritical before the plastic deformation? Are they changed in size along with the plastic deformation, or~do they hold their original size? R. Klecker indicated that as one approaches plasticity, in generally tough materials the small cracks will tend to blunt first before they begin to run. In the Staff's analysis one takes the cracks just to the point where they blunt but do not extend by crack extension or unstable tearing. Therefore, if one has a small crack it would tend to open up but the sharpness of the crack itself would tend to decrease. C. p. Siess asked if the large plastic deformations are confined to the knuckle region of th'e torispherical shell. M. Reisch indicated affirmatively. C. P.

Siess asked if BNL has looked at any way in which the plastic deformations might affect the integrity of large penetrations such as the equipment hatch. M. Reisch indicated that BNL did not check the capacity of large openings in the containment building, such as the equipment or personnel hatches, since it was assumed that they were reinforced and should have a higher capacity. J. Rosenthal indicated that the Staff thinks that the hatches are not structurally stronger than the containment.

J. Rosenthal stated that the Staff is concerned about thermal effects on seals and whether the times to containment failure are very long. If the wetwell fails, a relatively benign release can be expected. There is a lot of margin between the wetwell failure and the drywell and the head of the drywell should a crack exist.

If there is no flaw in the containcent, then the containment may never fail. If there is a flaw, the containment may fail early. If it fails early relieving the pressure, you don't fail the drywell or the pool. The Staff is rather confident that one will maintain the drywell and pool integrity. W.~ Kerr asked if, from these analyses, it is possible to make a statement _that, if containment failure occurs, it will occur as predicted by this analysis or is it likely to occur at a penetration or some other location. J.

Rosenthal indicated that the Staff was far more concerned about deflagrations causing failures of penetration seals due to the thermal environment rather than containment failure occurring at penetrations due to pressures or forces on the penetrations caused by distortions in the structure.

R. Vij explained that the GESSAR-II drywell ,is a . pressure boundary for larger pressures only. This volume is' in direct comunication ,

through 120 vents below the water surface c contained within the

" ~~

g '7-PROPO5ED MINUTES OF THE 303rd ACRS MEETING ,

s1  ;

i drywell wall. When one . has a very sharp transient the .only L resistance you have to this free communication is about 5 psi or  :

[. so. The drywell boundary is not challenged at all in a low '

<.s pressure phenomenon. During any slow pressure build up within the ,

N drywell or within the containment, the drywell head, ceiling, I i, walls, or containment penetrations are not challenged at all beyond  :

l 5 or 6 psi. R. Vij indicated that the drywell has a capability  :

only for handling some ~ duty for ,large pressurizations such as a i

steam line break or a hydrogen detonation and combustion. J. C.  !

Ebersole ' asked the GE design pressure for the maximum pipe break in i the drywell. R. Vij indicated that it was 30 psig.

~ \

R. Vij discussed the dominant containment failure modes. These i included hydrogen detonation in the containment, hydrogen  !

combustion in the c'ontainment, hydrogen slow burning, and steam and j or noncombustible ~ gas overpressurization (see Appendix X). With a i hydrogen detonation in the containment, the loads are shock wave i internal pressure on the containment and external pressure on the  !

!,o drywell. For a local. detonation the containment' failure is assumed l to occur'above the water line. Below the water line, there is 1.75 inch thick steel plate backed by 8 feet of concrete. For a global j detonation the containment failure is also assumed above the water- (

line and drywell failure is also -postulated. Therefore, both  !

barriers are postulated to fail. For hydrogen combustion one  !

generates internal pressure on the containment and small external  !

pressure on the drywell. Containment failure is assumed above the l water line with no drywell failure since there are no loads above 5  ;

or 6 psig differential. No containment failure is postulated for  !

the case of slow hydrogen burning since pressure does not exceed  ;

! the capability of the containment and the drywell again only sees 5  ;

, or 6 psig differential. For the case of steam and/or l noncombustible gas overpressurization there is a slow pressure  ;

generation and a small internal pressure on the drywell because of j the free comunication between the drywell-and the containment. GE t assumes containment failure above the water line at the containment i

, ultimate pressure capability (85 psig) and no drywell failure since  !

i there are no significant loads. D. Hankins, GE, indicated that  !

these analyses assumed that there are no ignitors. GE is .

-postulating that hydrogen in most cases will build up to a fairly }

high concentration before combustion and that is why high pressures ,

are assumed in some cases.

s i-

/(

D. W. Moeller asked if there are vacuum breakers between the wet- L well and the drywell. He asked if there are data on the failure fi j

ratess or projected failure rates available. D. Hankins indicated that GE has submitted a study to the Staff on supression pool i bypass which included the failure of the mechanical and motor-operated valves. There is a very low probability for the failure- l of both of those because there'are breakers between the wetwell and  ;

drywell. D. Okrent asked if there is any way water in the j suppression pool can drain out, aside from a failure such as a hole (

in the containment. - R. Vij indicated that the ECCS suction lines  ;

l have one valve on the outboard and if one assumes a break between  !

the - valve and the containment- boundary, which is equivalent to  !

I f i

1 i  !

L .

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

~ ' ' ~

PROPOSED MINUTES OF THE 303rd ACRS MEETING i L

assuming a hole in the containment, such circumstances would lead

, to draining or a loss of a substantial part of the pool water, j R. Vij indicated that GE considered some validation methods with a structural configuration of the structures with the loads that are j imposed. The stresses were calculated assuming that the highest  !

j stress points failed first. No failures were assumed where loads ,

l were significantly less than the design loads. Based on the  ;

analysis, GE concluded that the drywell structure head and i i personnel lock are not' challenged and the suppression pool bypass -

due to drywell boundary failure should not be a concern.  :

~

R. Vij indicated that GE perfonned a very simplified analysis which  !

-assumed that the molten core burns a 6 foot deep hole in the basemat. The temperature of the molten core is assumed to be 4,000 degrees Fahrenheit. A heat conduction analysis and a linear [

elastic stress analysis were performed with those dimensions and the corresponding displacements. The temperature in the drywell

, wall at the basemat was found to be on the order of 150 degrees '

! Fahrenheit. The deflection of the drywell wall was approximately  ;

half an inch at that point. The - calculated stresses in the  !

concrete were .about_ 3,500 psi where the maximum compressive strength of the concrete is about 4,000 psi. In the steel, GE calculated stresses close to 40 ksi.as compared to a yield strength -

of 46 ksi. Therefore, GE does not expect any danger to the drywell i j

walls overall stability because of these assumptions for the molten Core.

• i J. Rosenthal discussed the severe accident threat to the GESSAR-II [

containment (see Appendix XII). He indicated that the Staff's main '

concern is ablation of the concrete from the molten corium. He  ;

' described a core melt scenario which led to the potential for loss  !

of integrity of the pedestal beneath the reactor vessel. In answer 1 to an inquiry by D. Okrent, he confirmed' that drywell and  ;

conceivably wetwell integrity would be lost because there would be  !

the potential for ripping out major piping that penetrates the  !

drywell and wetwell.. E S

J. Rosenthal listed the dominant containment failure modes i considered by the NRC Staff. He indicated that the Staff f considered overpressure failure -due to noncondensible gas i generation as well as hydrogen deflagration and detonations. He l c indicated that the Staff looked at structures, seals, piping penetrations and their fragility. He noted that the Staff

• considered neither steam explosions nor direct' heating since the latter was viewed primarily as a high pressure problem. The  ;

dominant sequences on the GESSAR plant are low pressure because of [

< the perceived reliability of the ADS (automatic depressurization  !

system). He indicated that the Staff looked at a range. of l containment failure times. The -late failure is at 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> for a

wetwell failure due to a potential fabrication flaw and the early j containment failure is about 150 minutes. There is a difference of i about three in consequences between early and late failure. He  :

discussed the Staff's concerns over excessi.ve drywell/wetwell i I

i i

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PROPOSED MINUTES OF THE 303rd ACRS MEETING leakage, where certain sequences model total bypass of the suppression pool via vaporization release. He indicated that the ultimate consequences of these sequences hinga on whether one bypasses the suppression pool or fails the pool. This is the main Staff concern. D. Okrent thought that the Committee ought to focus on the ablation discussion since there is a significant probability of a loss of integrity late, but nevertheless a loss of integrity of containment. There would be release of whatever radionuclides are available at that point. This contrasts with the claims that some containments stay relatively tight for a large family of accidents. H. Etherington suggested that ablation might be a rather destructive process involving chunks of concrete breaking away rather than just washing away.

C. Thomas, Division of Licensing, indicated that on June 27 the Commission approved the severe accident policy statement with one minor modification. That modification does not affect the GESSAR-II review. On July 3 the Commission issued a Staff requirements memorandum asking the Staff to modify the severe accident policy statement as noted and to forward it to the Office of the Secretary for publication in the . Federal Register. In accordance with the provisions of the severe accident policy statement the Staff is prepared to amend the GESSAR-II FDA to pemit it to be referenced in new CP and OL applications. The amendment would allow GESSAR-II to be referenced but the Staff would not issue a new CP or OL for an application that referenced GESSAR until successful completion of the~ severe accident review. The Staff will further amend the GESSAR-II FDA when the review is completed to allow it to be referenced for CP and OL applications issued for a fixed period of time.

C. Thomas explained that the Staff is experiencing resource problems since this is the first standard design to undergo a severe accident review under the provisions of the severe accident policy statement. For this reason,' the review has taken a lot more time and resources then envisioned at the outset. NRR, in t particular, has been extrenely strained regarding resources to support this review. They are having a particularly difficult time regarding technical assistance and travel dollars. The GESSAR-II review is beginning to adversely impact some other NRR reviews.

For that reason the Staff urges that the subcomittee and the ACRS complete its review as soon as reasonably possible. The Staff would like to complete this review with an ACRS letter at the September ACRS meeting. D. Okrent noted that there are a few issues that are still not resolved and the Staff does have concerns in the area of security or sabotage protection. He indicated that the Committee would do its best to cooperate. C. Thomas also noted NRR's pending reorganization that will probably further exacerbate the resource problem.

VI. Quantitative Safety Goals (0 pen) l [ Note: R. P. Savio was the Designated Federal Official for this portionofthemeeting.]

I PROPOSED MINUTES OF THE 303rd ACRS MEETING V. Stello, DEDR0GR, indicated that he was somewhat troubled by 'the Comittee's discussion of the safety goal issue with the Commission earlier that . morning (Thursday, July 11). He indicated that the Comittee continues to remind the Staff that there are issues in which they are particularly interested and on which they wish to be kept informed of the Staff's deliberations. He stated that the Staff has 'not evolved to a point where it is prepared to declare what its final position will be on the quantitative safety goal issue. He wanted the Comittee to be aware that there is a variety of views on a number of contentious issues. He thought the Comittee was unduly severe in casting the Staff's preliminary deliberations in an unfavorable light before the Comission.

V. Stello indicated that there are two areas in which advice from the Comittee would be helpful: Thefirstis]heissueofsetting a performance criterion. He contrasted 10 ' measured by the classical PRA calculation, with a proposed 10-S per reactor-year valge for a melt which would penetrate the primary system. The 10- number predicts the capability of the core to sustain severe core damage. This core damage may lead to a full scale corg melt where the core physically melts through the vessel. The 10' goal refers to the case where the core has already left the reactor vesse1~ and the fission products are inside the containment. He thought it a fundamental philosophical point of whether one must ask PRA technology to take the next step of discriminating between those two kinds of calculations. The latter approach may' delay the use of safety goals in any meaningful way until the Staff has made further advances in PRA technology. He indicated that he preferred to accept PRA technology as it exists today and recognize that there is some degree of conservatism that must be applied.

The second area where ACRS action would be useful would be in the area of cost-benefit analysis where there is a great deal of controversy.

V. Stello s' tressed that the safety goal is not a mechanism to replace the current regulatory process. It is an additional element in the regulatory process and it will not replace the j defense-in-depth concept. He suggested that one ought to know the results of a cost-benefit analysis when evaluating a plant modification so that if one decides to go that way there ought to be a reason for doing it. On the other hand, if the results of a PRA show that traditional methods of regulation overlook something needed for safety, then the PRA ought to supplant those traditional methods. Similarly, if the PRA clearly indicates that one ought not to do something, then the Staff ought to give that observation substantial weight.

T. Speis, NRR, sumarized some of the views provided to the E00 by H. Denton. He indicated that H. Denton is not comfortable with the proposed core mejt frequency guideline number. Calculations show that use of 10 per reactor-year yields a 45 percent chance of occurrence of a serious accident in the next 20 years, and a 10 percent chance of two or more such accidents. He suggested that H.

Denton believes the 10-4 per reactor-year number is too large and PROPOSED MINUTES OF THE 303rd ACRS MEETING that too much reliance has been placed on perceived knowledge of

. fission product behavior and containment performance. H. Denton believes that there ought to be added conservatism in the performance guidelines.

T. Speis took note of some ambiguities in the Steering GroupQ report regaroing the extent of core damage assumed with the 10 value (see Appendix XIII). W. Kerr suggested that the implication of H. Denton's message is that nuclear power plants that comply with a criterion of 10-S for release of larger amounts of fission products would be safer than existing plants. V. Stello suggested that nuclear power plants would be m goal for core melt frequercy of 10 ye per safer by adopting aNot reactor-year. safety only would the likelihood of a melt through of the reactor vessel be lower but the likelihood of severe core damage would also be lower.

F. J. Remick suggested that .when the original safety goal was developed the Commission's responsibility was recognized as protection of the health and safety of the public. Its primary goals wgre public risk guidelines. The historical prospective on the 10- core melt guideline was that it was to be a secondary goal. He noted that the Steering Group recommends that it be elevated to the status of a primary guideline. 5 He asked whether the Staff had given thought to submitting the 10- per reactor year number to public coment. D. A. Ward indicated that he thought what was driving the Staff to lower the core melt guideline number was not a concern about causing more offsite cancers or accidental fatalities but was the re might occur under the 10-}atively high guideline frequency (estimated byof core the NRCmelt which Staff to be a 45 percent chance of a core melt accident in the next 20 years).

D. Okrent indicated that, before hUREG-0880 was written, the ACRS proposed some tri guideline of 10 gl criteria including a This per reactor-year. meanguideline large scale core-melt clearly meant 1

melt through of the reactor vessel. The ACRS proposal was an effort to achieve both defense-in-depth (asking for both a mean core melt frequency and a containment performance guideline given a large scale core melt). He suggested that H. Denton was expressing

, 'a lack of confidence in the current capability to predict containment performance as claimed in PRAs. V. Stello disagreed.

T. Speis indicated that NRR agrees with the Steering Group's recommendation that averted onsite losses be included (see Appendix XIII). T. Speis indicated that it is a matter of principle that all costs should be. displayed.

F. Rowsome, NRC, explained that cost-benefit analysis already has a well established role in the NRC in the backfit policy in the regulatory analysis of new generic reactor safety standards and in

, NEPA. The safety goal area has been chosen as the arena in which the Agency will codify how it implements cost-benefit analysis (see l

Appendix XIV). V. Stello explained that the Staff proposes to 1 impose plant modifications if ever it concludes that a plant is not l l l t l l

-PROPOSED MINUTES OF THE 303rd ACRS MEETING adequately safe. If it is a matter of compliance, the proposed

~

, rule before the Comission says that you do not do a cost-benefit analysis. W. Kerr suggested that there are exceptions to this rule such as Branch Technical Positions to which new reactors are being required to conform. V. Stello indicated that Staff's positions are embodied in the Standard Review Plan. The Staff's instructions to a utility are that if they wish to license a nuclear plant, they must show conformance to the Standard Review Plan. If they wish to deviate from the Standard Review Plan, they must tell the Staff how they expect to meet the basic underlying regulation.

F. Gillespie. Director, Division of Risk Analysis and Reactor Operations, RES, indicated that the Research position is fundamentally in support of the Steering Group's report. The Research staff does not wish that the defense-in-depth concept be lost. He thought that emphasis should be put on consideration of uncertainties in case studies done by the Staff. One does not want to lose the ability to make improvements just because a plant meets the 10-4 per reactor-year guideline. D. A. Ward asked if all that was meant was implementation of ALARA. F. Gillespie stated that if one can show that a fix has a significant safety benefit, even though the plant meets a particular core-melt frequency safety goal, the -policy should have enough flexibility to allow the staff to show that such a significant increment in safety is in fact worth it. C. P. Siess asked if there were some de minimis level where one would not consider modification even though a significant safety benefit can be shown. F. Gillespie indicated that as long as the incremental benefit exceeded the cost it would not be considered de minimis. D. A. Ward indicated that he was not sure whether a case was being made for ALARA or unrestricted ratcheting.

F. Gillespie explained that he was not prepared to set some particular de minimis cost level to cut off potential fixes.

VII. Meeting with the Comissioners (0 pen)

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

Asselstine, F. M. Bernthal and L. W. Zech]

A. Consideration of Earthquakes in Emergency Planning N. Palladino indicated that on December 18, 1984 the Comission issued a proposed rule which stated that earthquakes need not be considered in emergency planning.

D. W. Moeller indicated that the ACRS Subcomittees on Reactor Radiological Effects and Site Evaluation met with the NRC Staff, and representatives of FEMA, and called in a number of consultants to discuss the subject of emergency planning as related to natural phenomena with specific emphasis on earthquakes. As a result of these discussions and deliberations the Subcomittees reached certain conclusions. The Subcomittees saw no technical reason for j the exclusion of earthquakes from the natural phenomena considered in offsite emergency planning for nuclear power plants. The degree to which natural phenomena should be considered has to be 1

PROPOSED MINUTES OF THE 303rd ACRS MEETING plant-specific and site-specific because the frequencies and

, severities of these events vary over a wide range from one geographic area to another. The Subcomittees concluded that only limited consideration of earthquakes is appropriate. The major effort should be to become aware of problems and alternative approaches to their resolution. He stressed that it would be important to evaluate what might disturb normal emergency response in case an earthquake occurred at some time close to the time of an accident. The goal should be to assure that emergency plans as developed contain sufficient flexibility to cope with the potential ,

added impact of such events. Comissioner Bernthal asked what was meant by. the earthquake occurring at some time close to an accident. D. W. Moeller indicated that the ACRS considered two

possible scenarios. Either the earthquake caused the accident (initiator) or the accident occurred and an earthquake happened to occur sometime close to it in time. The probability of the latter, however, is acknowledged to be extremely low.

D. W. Moeller explained that the potential impacts of earthquakes have for many years been given detailed consideration by the NRC regulatory process in the design, construction and operation of ,

nuclear plants. Although FEMA does not consider the potential l impact of earthquakes on nuclear power plant emergency planning on t a formal basis, they have for some' time considered the impacts on ,

an informal basis. An item discussed by the Subcomittee and the '

full Committee was a proposal that it might be possible to rule out l consideration of the impact of certain natural events, such as l earthquakes, on the basis of their very low probability of  !

, occurrence. The Comittee did not reach a conclusion, however, i since the wide range of uncertainties in such probabilities compromises this approach. PRAs for several nuclear power plants t indicate that earthquakes despite their low probability may be  :

significant contributors (initiator) to the risk of core melt accidents. He also noted that FEMA is coordinating a National r Earthquakes Hazards Reduction Program.

(

Comissioner Asselstine asked if the ACRS concern is only with the i very large earthquake, larger than the safe shutdown earthquake  !

(SSE), or does it extend to lesser earthquakes. D. Okrent indicated that studies that he had seen to date suggested that at the SSE level one would not expect serious trouble at a plant. It is only for levels somewhat above the SSE that one starts to

question the possibility of consideration. j t

Comissioner Asselstine suggested that one should look at the I complicating effects of earthquakes on the procedural aspects of emergency planning. Chairman Palladino pointed out the site specificity of the issue and that engineering judgment would have to be applied to decide whether to protect against any complicating event or situation regardless of how unlikely the event might be.

Chairman Palladino thought that the frequency of the postulated natural phenomenon under question would be important.

r

. - - , - ,~ , , - .

PROPOSED MINUTES OF THE 303rd ACRS MEETING Commissioner Asselstine suggested that one may be most concerned

' bout a severe core damage accident with a significant release that a ,

would entail evacuation or a situation in the plant that has that  :

potential. That scenario would require preliminary steps in preparation for such an eventuality. He suggested a sliding scale for initiators and probability. Consideration of a relatively ,

frequent initiator would require study in great detail and minute '

detailed planning. The relatively frequent event with virtually no possibility of causing a nuclear accident would require consideration because of the potential simultaneous occurrence of an independent accionnt and the relatively frequent natural event.

A relatively infrequent event with a high risk of causing a very  ;

serious accident should also be considered to a certain extent.

The low probability event with very little likelihood of causing a severe accident would be discounted. He suggested that earthquakes fall in the third category since they are fairly low probability events but have the potential for being an accident initiator.

Comissioner Asselstine mentioned a July 5,1985 memorandum from the ED0 to the Commissioners and certain numbers in that memorandum -

(see Appendix XV). Comissioner Bernthal requested that the ACRS comment on the relative importance of the full range of natural phenomena (earthquakes, tornadoes, floods, etc.) in terms of their

~

potential impacts on emergency planning. Such an evaluation should be based so far a practicable on a probabilistic approach.

Comissioner Asselstine suggested that the assumption that underlies the proposed rule that is out for comment is that the NRC knows enough both about the absolute probabilities of earthquakes and tornadoes and the relative probabilities of those two natural phenomena as compared with other phenomena to say that these two should be excluded from any consideration as to their impact on emergency planning. C. P. Siess suggested that the Comission is <

overestimating the effectiveness of its policies and regulations on the actual effectiveness of offsite emergency preparedness. He suggested that, regardless of what is in the emergency plan, the local officials will make a judgment as to what they consider natural hazards and prepare appropriately. Local officials are  :

going to prepare most appropriately for what they consider are their natural hazards and not so much what is written into the regulations.

Comissioner Zech noted that the Comission is dealing with uncertainties, probabilities and statistics. He thought it most important for the Comission to develop the most reasonably prudent and comon sense rule possible. To this end, he requested that the Comittee develop a strong ACRS consensus view on this rulemaking.

D. Okrent indicated that hurricanes and earthquakes are the two natural phenomena having the most potential for disrupting large areas from the evacuation point of view and having the potential for being an initiator of a nuclear accident, rather than tornadoes.

PROPOSED MINUTES OF THE 303rd ACRS MEETING 4

B. Safety Goal Implementation D. Okrent indicated that the Comittee came very close to a Conunittee position on the issue of safety goals at the last ACRS meeting. The ACRS has not yet, however, seen a copy of a recommendation from the EDO to the Comission on this subject. The Comittee has had the benefit of memoranda written by senior members of the NRC Staff such as H. Denton and R. B. Minogue and has had the benefit of the Safety Goal Steering Group report. At its Subcomittee meeting on Wednesday, July 10, the ACRS also heard the views of Lester Lave, well known economist who consults for both the NRC and the ACRS, and Philosophy Professor Douglas McLean from the University of Maryland. Both were participants in the first and second panels that reviewed NUREG-0880 in its early formulative years.

D. Okrent speculated that the ACRS will state that the NRC is not

, now ready to reaffirm and implement the 1983 Safety Goal Policy in r its original or. some slightly modified form. Progress had been made and additional effort is needed, but the form of the goals and the plan for implementation are not yet well enough developed.

Note was taken of the considerable differences with the original proposal that exist among members of senior Staff. He anticipated that the Comittee will emphasize in its report that greater attention be devoted to working toward an adequate core melt objective and toward the identification and use of a containment performance objective. He suggested that the Comittee is concerned that the Safety Goal Policy Statement does not give sufficient emphasis to defense-in-depth. There is also concern that inappropriate reliance may be placed on benefit-cost analysis.

Chairman Palladino asked why the ACRS did not choose. to wait for l the Staff document on the safety goal. D. A. Ward suggested that ,

it might be useful for the ACRS to try to influence the Staff document through a letter this month. He noted that that opinion  !

is not unanimous among the Comittee members.

F. J. .Remick welcomed the opportunity to present the Comittee's views but thought that the Staff ought to have the opportunity to evaluate and provide their judgment regarding the just completed two year evaluation program. He thought it appropriate for the Comittee to coment on the Steering Group report but his own personal preference would be against a Comittee letter at this ,

time. '

D. Okrent indicated that ACRS coments will center on the Safety Goal Steering Group report. He indicated that the Comittee agrees with many of its findings and conclusions. The Comittee agrees that PRAs have limitations that must be understood when the results are used and the results of a PRA should normally be used in ,

conjunction with traditional safety review methods in making regulatory decisions. The Comittee agrees tentatively that the statement of the qualitative goals in the 1983 Safety Goal Policy l Statement appears to be satisfactory. The Comittee agrees that,  !

- - - , . - - - -- - - - . , - - ---. - -, ,m. . _ - - -

PROPOSED MINUTES OF THE 303rd ACRS MEETING for sites where no individuals reside within a mile from the plant, an individual should be assumed to reside one mile from the site boundary (for purposes of calculation). In applying the latent cancer fatality safety goal, the Comittee agrees that it is better to consider the population within ten miles rather than 50 miies as proposed in the 1983 policy statement. He noted that while the Steering Group proposed that averted onsite costs in a core melt accident be included in benefit-cost analysis, several senior Staff members have expressed concern that not to include such costs would lead to jeopardizing defense-in-depth. He recognized that inclusion of averted onsite costs and benefits is controversial and he speculated on the possible ACRS position: Benefit-cost calculations ought not necessarily be the most important criteria in decision making concerning safety and the accomplishment of defense-in-depth. He noted that L. Lave and D. McLean stated very positively that the only way to do cost-benefit analysis is to include all significant costs and all significant benefits and the 1983 policy statement was deficient in that regard. Comissioner Bernthal expressed an interest in seeing L. Lave and D. McLean's written comments. D. Okrent agreed to forward them.

Commissioner Bernthal suggested that the Comittee note how the current approach in the Safety Goal -Policy squares with the NRC's mandate under the Atomic Energy Act. Chairman Palladino suggested that the Comittee take account of the nuclear industry's point of view. D. Okrent acknowledged that there are many differences of opinion about the safety goal policy. He expected that the Comittee will recomend that, when the Comission develops a new statement of the safety goal policy, it should state that near compliance with a mean core melt frequency of 10-4 per reactor-year is an NRC objective for all but a few small existing nuclear power plants and that prudence will tend to take -priority over benefit-cost analysis in working toward this goal.

D. Okrent stressed that he expected the Comittee to recomend development of a containment performance guideline, as well as the use of mean values instead of median when assessing core melt frequencies or doing cost-benefit calculations. He noted that the Steering Group proposed a number of detailed implementation procedures. Operating limits proposed are in some instances not sufficiently conservative and in some cases not clear. The Committee has questions regarding these proposed operating limits and wishes to discuss them in more detail with -he Staff. The ACRS recomends that they not be adopted at this time.

D. Okrent indicated that the Comittee will most likely give emphasis to the Comission's policy that future nuclear power plants be safer. He suggested that the ACRS may well recomend a target core melt frequency mean for future nuclear power plants of 10- per reactor year.

F. J. Remick stated that he oersonally did not think that the defense-in-depth concept would be compromised if used in conjunction with prudent normal judgment. He indicated that he did PROPOSED MINUTES OF THE 303rd ACRS MEETING not disagree with the two ACRS consultants who thought that all benefits and costs ought to be considered in a cost-benefit analysis. He questioned the Commission's authority or mandate to get into averted costs to the licensee. Comissioner' Asselstine i

noted that the Atomic Energy Act also talks about minimizing the danger to life and property. Commissioner Bernthal said that this is clearly ALARA and ALARA has been manifestly rejected in most quarters by the U.S. Congress. .

H. W. Lewis spoke of the difficulties the Staff is having regarding questions of uncert'ainty bound up with the issue of the interpretation of a safety goal. He noted that the Subcommittee heard a divergence of views. He suggested that the issue has not been sufficiently addressed as yet. He pointed out that the safety goal must be treated as part of a whole package which includes the backfitting rule and the Severe Accident Policy Statement. He took exception to the Staff's continued use of median versus mean values.

Comissioner Asselstine indicated that. it was his sense that there are fundamental questions and disagreements within the Staff and on the Committee about some of the key issues involved in this safety goal. He wondered whether the disagreements are so significant that perhaps the Commission ought to get more directly involved now before the process works its course. D. Okrent hoped that the ED0 would resolve issues in its recommendation to the Commission.

Comissioner Bernthal suggested that it was his understanding that at least one school of thought in the Staff wants to move away from a definition based on core-melt to a definition based on loss of primary system integrity. C. P. Siess thought that it makes sense to consider the core out of the primary system in the containment where it can challenge the last barrier to the environment (containment). It puts some emphasis on containment performance criteria that one does not get if you do not consider various core-melt scenarios. Comissioner Bernthal suggested that one might get a cleaner analysis from a numerical point of view because of the uncertainty in core-melt phenomena. C. P. Siess noted that one would now be confronted with a calculation of where the primary '

coolant goes. He wondered whether this would simplify or complicate the analysis. W. Kerr suggested that it might be more difficult to calculate the loss of primary system integrity than to 1 calculate the conditions under which core-melt is likely to occur.

Chairman Palladino indicated that this discussion has been very ,

worthwhile. He looked forward to reports frc,m the Committee.

VIII. EPA Standards for High Level Waste Repositories (0 pen)

[ Note: 0. S. Merrill was the Designated Federal Official for this portionofthemeeting.] ,

D. W. Moeller indicated that, for high level radioactive waste repositories, EPA has the responsibility for developing standards L (40 CFR 191) and the NRC for the development of overall performance l

PROPOSED MINUTES OF THE 303rd ACRS MEETING objectives, regulatory guides and regulations (10 CFR 60). 00E sets system guidelines (10 CFR 960). He noted that the NRC has folded in conservatisms, dose, and risk limits into its performance objectives. Both the DOE and the NRC efforts have occurred without finalized EPA Standards. DOE has gone ahead in the absence of the EPA Standards toward making plans for constructing a high level waste repository in accord with 10 CFR 60.

D. W. Moeller indicated that the NRC has been looking at the EPA standards and the discussions have been primarily directed to the resolution of jurisdictional disputes. The NRC Staff has already concurred with the draft EPA standards issued in 1983.

D. W. Moeller indicated that the Subcomittees on Waste Management and Site Evaluation met on June 18 and 19,1985 to review Revision 6 of the EPA Standards. He noted that the Comissioners will shortly be providing coments and recomendations to the EPA on these standards. He suggested that the ACRS offer advice to the Comission as it did on the DOE Mission Plan. ,

i D. Okrent urged that the Comittee encourage the Comission to re.cocmend that EPA officials use a risk-based approach in the development of these standards. The Subcomittee suggested that the proposed EPA Standards are unduly restrictive and may result in l the rejection of some suitable sites and the expenditure of funds  !

that might be better applied to other environmental problems. He i noted that the EPA has refused to relax the Standards and the NRC ,

has concurred. He stated that the Subcomittee review revealed that the Standards appear to be flawed. The Subcomittee believes that the Standards should be expressed in terms of dose equivalent and/or health effects limits as contrasted' to radionuclide release limits, which have little relationship to health risks. The release limits given in'the proposed Standards do not appear to be directly related to the proposed limitation on health effects.

. Since the generic environmental model was used to estimate the ,

population doses resulting from the stated releases, the Subcommittee questions whether the resulting estimates will be ,

applicable to specific sites selected for a repository. It also  ;

questions whether these estimates will be applicable to disposal methods other than a geologic repository.

D. Okrent indicated that the Subcomittee report lists the pitfalls of the EPA Standards and recommends that the Comission make these shortcomings known to the EPA. He suggested that it is most important to tell the Comission that its Staff should stipulate i whether these Standards are practicable. E. Goldberg, OPE, noted that the Commission has no formal endorsement function or concurrence function regarding the EPA standards. D. W. Moeller pointed out that there are also problems with the confidence limits  ;

given in the proposed standards. He suggested that DOE will not be able to prove that they can meet the release limits in the standards with adequate confidence. D. Okrent expressed concern i that the release limits in the EPA standards may not make it l practicable to license a high level waste facility.  ;

Y

PROPOSED MINUTES OF THE 303rd ACRS MEETING i

f

. IX. ANL - West Survey of Control Room Habitability (0 pen)

[ Note: J. O. Schiffgens was the Designated Federal Official for  !

this portion of the meeting.]  ;

~

D. W. Moeller indicated that the Subcomittee on Air Systems met on l June 17, 1985 to discuss two NRC reports: NUREG/CR-4191 " Survey of i Licensee Control Room Habitability Practices" (prepared by  !

ANL-West, consultants to the NRC/NRR Staff), and NUREG/CR-3551, and  ;

" Safety Implications Associated with In-Plant Pressurized Gas  !

Storage and Distribution Systems in Nuclear Power Plants" (prepared  !

by ORNL or consultants to the NRC/AE00 Staff). With regard to ,

NUREG/CR-3551, the consultants recomended that compressed gas j cylinders ought not to be allowed in critical areas unless it can i be assured that should they become missiles they would not damage l critical plant equipment. They also recomended that flow i controlling valves should be installed on hydrogen lines where high  !

pressure cylinders are handled. All high pressure gas lines should i be color-coded.

W. P. Gamel, NRR, explained that in June 1983 the Staff was asked l to. prepare a program plan to address control room habitability i concerns of the ACRS. The Control Room Habitability Working Group l Report, which wus reviewed by the Subcomittee in November 1984, [

will be published as NUREG/CR-1129. The Staff's assessment of i control room habitability practices, which parallels the ANL West [

report (NUREG/CR-4191), will be published as a supplement to i NUREG/CR-1129. He explained that one recomendation of the Working  !

Group is that efforts be increased to obtain industry feedback on control room air systems. This will be accomplished in part by [

increasing participation by the Staff in technical society t

meetings. He noted that the Staff has a new contract with ANL-West  !

to investigate generic issues identified in the original control l room habitability study and there are plans for surveying an l

, additional 12 plants. He mentioned that progress has been slower  !

than anticipated regarding revision of Regulatory Guide 1.52, the standard review plan, and technical specifications. G. A. Reed ,

asked what the average cost for upgrading control room air systems  !

would be for each nuclear plant? J. J. Hayes, NRC, indicated that i the cost would be nominal because most of the upgrade would involve i utility action to meet comitments that it already has in its SAR l and technical specifications. W. P. Gamel added that he expected a ;

cost of $500,000 to $1,000,000 per plant to replace isolation l dampers which have been found to exhibit excessive leakage.

J. J. Hayes explained that three plants were selected for the initial assessment of control room systems, components, operations, procedures and technical specifications. Mechanical and electrical systems were reviewed as well as remote shutdown capability. NRC ,

practices and policies and NRC licensee practices were also explored. It was concluded from the field studies that loss of ventilation and loss of air conditioning events which have occurred at operating nuclear plants should be studied further and their possible contribution to the degradation of plant safety evaluated ,

I l

I

PROPOSED MINUTES OF THE 303rd ACRS MEETING (see Appendix XVI). He mentioned a case of the loss of both air conditioning trains at the Calvert Cliffs Nuclear Plant where spurious signals from instrumentation and the failures of printed circuit boards were found. J. C. Ebersole suggested that a failure of printed circuit boards is a generic issue the ACRS should review by examining applicable procurement specifications. J. J. Hayes noted that LERs currently do not include instrument problems.

J. J. Hayes indicated another conclusion of the field studies was that changes to the action statements and surveillance requirements of technical specifications should be made as needed to insure that the control room HVAC system specifications provide for functioning as designed. He noted that a generic problem identified during this study was the leakage through isolation dampers or valves. As a result, leaky isolation dampers in some plants have been changed to bubble-type dampers. A problem was found during the field studies regarding appropriate laboratory conditions for the testing of charcoal filters in HVAC systems. J.

C. Ebersole suggested that it appears that the control room habitability issue should be expanded to include environmental control since control room temperature limitations affect equipment as well as the operators. J. J. Hayes noted that one problem is the fact that updated FSARs for operating nuclear plants often do not reflect actual systems. This obsolesence occurs because some changes to HVAC systems can take as much as three to five years to be documented. He stated that another general conclusion of the field application studies was that the whole approach to control room HVAC systems lacks a systems approach, but is handled more on a components-approach basis.

X. Long Range Plan for NRC (0 pen)

[ Note: J. C.McKinley was the Designated Federal Official for this portionofthemeeting.]

M. W. Carbon indicated that the Subcomittee on the Long Range Plan for the NRC met on July 10 and 11, 1985 to interview invited individuals from the nuclear industry. These individuals were asked to comment on whether the NRC should have a long range plan and whether the ACRS can be of assistance in the development of such a plan. There was general agreement that the NRC should have some long range plan. R. Mattson, former NRC Staff member, and J.

Ahearne, former NRC Comissioner, both suggested that the Comissioners themselves ought to get more deeply involved in the development of such a plan. Interviewees were in general agreement with the selection of prospective issues to be handled in the long range plan and thought that a plan ought to reflect the expected state of affairs for at least ten years into the future or longer.

M. W. Carbon reminded the Comittee that the Comissinn assigned the Office of Policy Evaluation the task of developing a 5 year plan. The outline for this plan, which is based upcn the current Program Planning Guidance document for the NRC, was briefly

PROPOSED MINUTES OF THE 303rd ACRS MEETING i,

i

described. The Subcomittee plans to review the outline at a future meeting.

M. W. Carbon explained that the Subcormittee concluded that there was need for an additional half-day Subcomittee meeting and an additional half-day of full Comittee discussion regarding the ACRS

strategy for the long range plan. He suggested that there be full

Committee review of the OPE outline at the August meeting. Several i Committee members expressed the opinion that it would be very difficult to provide OPE thoughtful coments on that short a schedule.

i XI. Human Factors and Maintenance Subconmittees on Natural Aptitude l Selection Procedures (0 pen)

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

G. A. Reed indicated that the ACRS Subcomittees on Human Factors and Mainterance Practices and Procedures met on June 18,1985 to explore the use of natural aptitude selection procedures, tests, and evaluations. He indicated that D. Kleinke, Manager of Psychological Services at the Edison Electric Institute (EEI)

discussed EEI's seven Plant Operators selection Selection System testing (POSS p)rojects

, whichwhich include is related to the j Wisconsin Electric Power tests, and the Maintenance Aptitude i, Selection Test (MAST).

) G. A. Reed indicated that he was disappointed in Kleinke's pronouncements in that he did not make an enthusiastic case for j naturai aptitude selection. He was surprised that Kleinke flatly thought that mechanical aptitude could not be learned but he agreed

with him. Physical abilities tests, as part of MAST, showed that i

one could not train someone to have mechanical ability. G. A. Reed j also indicated that the Comittee heard from A. Mascitti,

! Supervisor of Supervisory and Professional Placement, Wisconsin

Electric Power. Both Kleinke and Mascitti thought that selection testing is important. No statement was made, however, as to

whether it would be a good idea for NRC to draft a rule requiring l natural selection testing at nuclear power plants. He indicated i

that he was still in favor of the promulgation of an NRC rule on natural selection. D. A. Ward indicated that he was unable to get i

a clear idea of how many nuclear plants use the EEI tests and how many utilities use some other kinds of tests. W. Kerr thought that i the NRC is already getting too involved in nuclear plant training and he was not sure that a NRC rule requiring natural telection l testing would be such a good idea.

I XII. Briefing Regarding Steam Line Failure in Non-Nuclear Power Plants l

(closed)

[ Note: E. G. Igne was the Designated Federal Official for this l

portionofthemeeting.]

t DLLLIe4/idrlEYmal XIII. Executive Sessions (0 pen)

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

A. Subcommittee Assignments

1. Meeting with the NRC Comissioners The ACRS discussed with the Comissioners the consideration of seismic events in emergency planning. Comissioner Bernthal requested that the ACRS coment on the relative importance of the full range of natural phenomena (earthquakes, tornados, floods, etc.) in terms of their potential impacts on emergency planning. Such an evaluation should be based, so far as practicable, on a probabilistic approach. Comissioner Zech requested that the Comittee develop a strong ACRS consensus view on this rulemaking. Comissioner Asselstine mentioned a July 5,1985 memorandum from the EDO to the Comissioners and certain numbers in that memorandum. He solicited ACRS coment on the numerical values proposed. These matters have been assigned to the Subcomittee on Site Evaluation, Dade Moeller, Chairman, for follow-up.

During the discussion of quantitative safety goals, the Comittee comitted to send to the Comission ACRS consultants' coments on safety goal policy by L. Lave and D.

MacLean.

2. SAlp Evaluation of Licensees A letter from W. J. Dircks, EDO, to C. Dean, Chairman, TVA Board of Directors, dated July 3, 1985 expresses concern regarding TVA management deficiencies based on various indicators which compare TVA plants with other non-TVA nuclear units. The Comittee decided to request that the Subcommittee on Human Factors consider this type of evaluation as a tool to evaluate project management at operating nuclear plants.

3. Incident Investigation Program The Comittee noted two alternate proposals for NTSB type investigations now before the Comission:

NRC Staff Plan for improving the existing program for the investigation of significant operational events i

Use of Atomic Safety and Licensing Boards to investigate significant nuclear events

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PROPOSED MINUTES OF THE 303rd ACRS MEETING r

This issue was assigned to the Subcommittee on Regulatory

. Policy and Practices, H. W. Lewis, Chairman, for follow-up, development of an ACRS position and the preparation of draft coments for consideration for forwarding to the Comissioners.

B. ACRS Reports, Letters, and Memoranda

1. Reports, Letters, and Memoranda ACRS Coments on Proposed NRC Safety Goal Evaluation Report The Committee prepared a report to the Comissioners on its review of the NRC draft Safety Goal Evaluation Report dated April 1985. The ACRS expects to make further coments on the Steering Group report when the E00 has formulated recomendations to the Comission. The Comittee has a range of questions on proposed operating limits and wishes to discuss these matters in detail with the NRC Staff.

Additional coments by F. J. Remick, G. A. Reed, M. W. Carbon, and H. W. Lewis were appended.

2. Provisions for Protection Against Sabotage The Comittee prepared a report to the Comissioners regarding present provisions for protection against sabotage.

Additional coments by G. A. Reed, D. A. Ward, H. W. Lewis, P.

G. Shewmon and F. J. Remick were appended.

3. Long Term Seismic Program Plan For The Diablo Canyon Power Plant The Comittee prepared a report to the Comissioners of its review of the Long Term Seismic Program submitted by the Pacific Gas & Electric Company (Licensee) for the Diablo Canyon Power Plant. Additional coments by D. Okrent, W.

Kerr, and D. A. Ward were appended.

T. EPA Standards for High-level Radioactive Waste Disposal The Comittee prepared a report to the Comissioners which discusses the proposed " Environmental Standards for the Management and Disposal of Spent Nuclear Fuel, High-level and Transuranic Radioactive Wastes" (10 CFR 191), being developed by the U.S. Environmental Protection Agency (EPA).

5. Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees The Comittee prepared a report to the Comissioners of its review of the proposed rule on " Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees" (10 CFR 30,40,and70).

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PROPOSED MINUTES OF THE 303rd ACRS MEETING e

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6. Investigation of Recent Incident at the Davis Besse Nuclear 4 Power Plant The Comittee prepared a report to the Comissioners comending the NRC Staff for its initiative in using an incident investigation team (IIT) to investigate the recent loss of feedwater incident at the Davis Besse Nuclear Power Plant.

7. Control Room Habitability The Comittee prepared a memorandum to the EDO regarding the

, report, " Survey of Licensee Control Room Habitability The ACRS wishes to be kept Practices" (NUREG/CR-4191).

i informed as future developments on this subject take place.

! 8. Materials Research The Comittee prepared a memorandum to the EDO (Attention: Guy A. Arlotto) regarding progress made on two RES projects --

SAFT-UT at PNL and acoustic emission (AE) at Watts Bar with coments on their continued funding and use.

C. Future Schedule

1. Future Agenda The Comittee agreed on tentative agenda items for the 304th j ACRS meeting, August 8-10,1985(seeAppendixII).

2. Future Subcomittee Activities A schedule of future Subcomittee activities was distributed to Members (see Appendix III).

i

D. Sustained Meritorious Service G. A. Reed introduced a draft letter regarding the presentation of several American Nuclear Society (ANS) awards for Meritorious

! Performance in Reactor Operations. The presentations were made on .

June 11, 1985, in Boston as part of the ANS Annual Meeting. The '

Comittee decided, and so informed G. A. Reed, that it would be  !

more appropriate for him to forward the letter as a statement of his own position as a member of the public.

. E. Joint Meeting of Nuclear Safety Comittees D. A. Ward suggested a joint meeting of the ACRS, the Groupe  ;

Permanent, the RSK, and a corresponding Japanese group in the U.S.

in approximately one year. The Comittee suggested that initial j planning of the agenda begin now and that a search be undertaken to

locate a source of supplemental outside funding for this project. L l  ;

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PROPOSED MINUTES OF THE 303rd ACRS MEETING t

F. Proposed Amendments to ACRS By-Laws The Comittee approved three proposed amendments to the ACRS By-Laws. The first pertains to how amendments to the By-laws are proposed and approved. The second pertains to how a Member with a safety-related concern may solicit ACRS action and, if the member does not consider the response to be satisfactory, what additional options and Staff support may be available to him for follow-up. i The third pertains to additional information on conflicts-of-interest.

G. Watts Bar Nuclear Plant

The Comittee heard and discussed the report of the ACRS QA/GC Subcomittee, G. A. Reed, Acting Chairman, regarding QA/QC '

difficulties and proposed corrective actions regarding the IDVP of Watts Bar conducted by Black & Veatch Engineering. Several other questions regarding TVA management of its nuclear plants have recently been identified (see SECY-85-231 Proposed NRC Action with Regard to TVA, dated June 28,198S) and TVA is in the process of reorganizing its nuclear management. The Comittee decided to take no further action regarding this matter until the NRC Staff has completed their evaluation. -

I D. A. Ward indicated that the Comittee intends to take a look at '

the Staff's supplement to the SER when it is in a final draft and decide if any further action is necessary. The decision whether to schedule another subcomittee meeting or just have full Comittee i

discussion could be determined at a later date. J. C. Ebersole asked whether the Staff intends to respond to the many detailed questions submitted by Henry Meyers of the Udall Comittee. The (

Committee expressed interest in seeing a copy of the Staff's j response to the H. Meyers observations when it is available. ,

, H. Indian Point Nuclear Plant l The members discussed a report proposed by Dr. Okrent regarding the .

Indian Point Special Proceeding based on the PRA for the Indian Point Nuclear Station. The Comittee deferred any action on this  ;

matter until backup material can be provided and discussed by the  !

Comittee. This item has been scheduled for discussion during the '

August ACRS Meeting. t 4

I. Licensing Process, Consideration of a National Academy of Nuclear )

Power Safety The members discussed a suggestion by G. A. Reed that the Comittee i consider this subject and provide coments to the Comission  :

regarding this pro by Senator Moynihan since one member  :

(Cormissioner Zech)posal has already expressed support for a training i academy of this nature. The members decided not to pursue this  !

matter.  !

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PROPOSED MINUTES OF THE 303rd ACRS MEETING The 303rd ACRS Meeting was adjourned at 2:45 p.m., Saturday, July 13, 1985 6

N O

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APPENDIXES TO MINUTES OF THE 303rd ACRS MEETING JULY 11 13.1985 um. u,a l

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APPENDIX I Attend:es i t i ATTENDEES 303rd ACRS MEETING

( July 11-13, 1985 ADVISORY COMITTEE ON REACTOR SAFEGUARDS ,

David A. Ward, Chairman Harold W. Lewis, Vice-Chainnan Robert C. Axtmann Max W. Carbon Jesse C. Ebersole William Kerr Carson Mark i Carlyle Michelson ,

Dade W. Moeller

• David Okrent  :

Glenn A. Reed i Forrest J. Remick Paul G. Shewmon i

Chester P. Siess d

Charles J. Wylie '

t i

ACRS Staff i Raymond F. Fraley, Executive Director I M. Norman Schwartz, Technical Secretary .

Herman Aldennan L

] Paul A. Boehnert

! Anthony J. Cappucci j j

Robert Cushman t

Monideep De Sam Duraiswamy  !'

i Medhat M. El-Zeftawy

! John Flack John T. Gilbert l James A. Jeffries i

! Janet Kotra

Morton W. Libarkin (

i Richard K. Major  !

John A. MacEvoy '

j Thomas G. McCreless i John C. McKinley  !

Owen S. Merrill  ;

Austin Newsom i Sidney J.S. Parry i Gary R. Quittschreiber l Richard Savio  :

! Stanley Schofer i j

l O i A/ j

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NRC ATTENDEES  !

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! 303RD ACRS MTG.

i l

1 t j . Thursday, July 11. 1985  !

a

, OFFICE OF NUCLEAR REACTOR t j REGULATION I R. A. Gilbert, DL

{ G. Sege, DST l R. Hernan, PPAS ,

t i W. S. Hazelton, DF

! S. Jsrau, DST  !

l J. F. Stolz, DL l G. Rivenbank, DL  !

4 A. DeAgazio }

0FFICE OF INSPECTION & EN-l FORCEMENT i R. Baer D. A. Powell l

E. Weiss i

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ROGR/0FFICE OF EXECUTIVE i i DIRECTOR FOR OPERATIONS i I 1

! T. Cox  !

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INVITED ATTENDEES 303RD ACRS MEETING ,

i t

i Thursday, July 11, 1985 1

BECHTEL POWER CORPORATION E. Desterle l

PACIFIC GAS & ELECTRIC R. F. Locke L. S. Cluff D. A. Brand l J. B. Hoch B. Norton R. Fray i

R. F. Locke G. G. Sarkisian D. W. Ogden B. S. Lew (

B. Norton r 1

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PUBLIC ATTENDEES 303RD ACRS. MEETING Thursday, Jt.ly 11, 1985 H. G. Hawkins, S. California Edison Comp ny R. Borsum, Babcock & Wilcox R. Fedini, Duquesne Light Company P. Docherty,LWestinghouse M. Beaumont, Westinghouse P. Higg, Atomic Industrial Forum F. Stetson, NUS Corporation W. Shark, Argonne National Laboratory C. Czajkowski, Brookhaven National Laboratory H. Specter, NY power Authority A. J. Pressesky, American Nuclear Society L. M. Muntzing, Dames & Moore J. Nurmi, EPM D. A. Brand L. Connor, DSA J. Berga, Electric Power Research Inst.

G. Sauter, Electric Power Research Inst.

E. Lindemann, McGraw-Hill /Inside NRC D. Holland, General Public Utilities K. Barnes, GPUN 4

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AY

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NRC ATTENDEES 303RD ACRS MTG.

(U)

Friday, July 12, 1985 0FFICE Oc NUCLEAR REACTOR REGULATION E. Adensam, DL T. J. KeKergin, DL R. Hernan, PPAS N. Chokshi, DE C. Tinkler, DSI J. Lane, DSI C. Thomas, DL D. Schletti, SSPB J. Rosenthal, DSI V. S. Pare:ewski, DE M. Rubin, RRAB J. P. Knight, DE D. R. Muller, DSI W. P. Gamill, DSI J. J. Hayes, DSI K. Dempsey, DSI f_s D. Persinko, LQB

'( )

k- R0GR/0FFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS M. Taylor O

e V) k'$

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! APPLICANT ATTENDEES 303RD ACRS MEETING

l i .

,i Friday, July 12, 1985 4

TENNESSEE VALLEY AUTHORITY ' GENERAL ELECTRIC I

H. L. Jones D. Hawkins R. M. Pierce R. Ketchel J. J. Ritts G. Sherwood i R. H. Shell

BLACK & VEATCH '

W. J. Zidziunas l -

l BROOKHAVEN NATIONAL LAB 3 2

1 M. Riech '

l S. Shema ,

R. Taung l J. Pires l GENERAL ELECTRIC

. D. Foreman ,

l R. Vij j R. Villa ,

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PUBLIC ATTENDEES I

, 303RD ACRS MEETING l

1 i

Friday, July 12, 1985 -

i E. Lindeman, Inside NRC i R. Fedin, DuQuesne Light Company l R..Borsum, Babcock & Wilcox  !'

R. Hubbard, MHB Associates A. J. Pressesky, American Nuclear Society

{ J. Berga, Electric Power Research Inst. i

J. Niman, Electric Power Research Inst. {

! L. Peeters, SAIC  ;

M. Wagner, McGraw Hill '

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APPENDIX II a

FUTURE AGENDA APPENDIX A <

!O FUTURE AGENDA

} AUGUST ACRS MEETING 1

San Onofre Nuclear Generating Station 3 hrs i Unit 1 -- SEP review, ACRS comments GESSAR II -- Continue ACRS review 3 hrs '

USI-A-46, Seismic Qualification of Equipment in 3 hrs Operating Plants -- ACRS coments

Maintenance and Surveillance Program Plan -- ACRS 1 hr i coments ACRS Subcomittee report on report of Pipe Crack Deferred study group / proposed changes in the General Design Criteria, regulations, etc. regarding piping design

f - and the DBA ,

i Assignment and makeup of ACRS Subcommittee and Deferred consultants on waste management -- Discussion among i

members

Vogtle' Nuclear Plant Units 1 and 2 -- OL review, ACRS 4 hrs

! coments

l

! ECCS -- ACRS discussion regarding proposed revision 2 hrs -

of Appendix K Tentative Report of Subcomittee on Reactor Radiological Effects Deferred i regarding INPO Radiation Protection Program  ;

Probabilistic Risk Assessment -- Proposed ACRS coments 2 hrs

regarding significance / implementation of PRA results

! for the Indian Point Nuclear Plant and other nuclear

, facilities l

Proposed NRC Accident Source Term -- ACRS comments Deferred

! ' NRC Long Range Plan -- ACRS coments on proposed plan I hr 1

outline Regulatory Guide 1.99, Rev. 2, Effects of Residual Deferred to l Elements on Predicted Radiation Damage to Reactor September Vessel Materials -- ACRS coment regarding publication for public coment Report of ACRS Subcommittee on CESSAR regarding provisions Deferred for decay heat removal in CE system 80 type plants O

i

1

• t t

! Meeting with Director of International Programs -- Deferred  ;

Briefing regarding activities  ;

i i Report of Panel on ACRS Effectiveness 2 hrs  !

Response to Commissioner Asselstine's inquiry regarding Deferred

adequacies / inadequacies in the NRC operator requalification

procedures -- Impact on regionalization, etc. l i . [

IE Inspection Program -- Briefing by IE representa41;ft N Cer,M i

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N-SCHEDULE OF ACRS SUBCOMMITTEE MEETING REVISED M I 2 1985 i

SCHEDULE OF ACRS SUBC0PtilTTEE MEETINGS JULY 17 ATWS (B0EHNERT) - Kerr, Ebersole, Ward, Wylie.

Purpose:

To discuss RPS and scram breaker reliability.

18 & 19 V0GTLE, UNITS 1 & 2 (AUGUSTA, GA) (SCHIFFGENS/MCKINLEY) -

Ebersole, Okrent, Reed, Wyl'le.

Purpose:

To begin review of OL application for Vogtle, Units 1 & 2.

30 JOINT WASTE MANAGEMENT AND PROCEDURES & ADMINISTRATION (MERRILL/FRALEY) - Moeller, Axtmann, Ebersole, Kerr, Mark, Remick, Shewmon, Siess, Ward.

Purpose:

To review the ACRS Role in the Civilian High-Level Radioactive Waste Management Program.

31 REACTOR RADIOLOGICAL EFFECTS (MERRILL) - Moeller, Axtmann, Okrent.

Purpose:

To review INP0's Radiation Protection Program and a recent INP0 report on Excessive Personnel Radiation Exposures.

31 ECCS (BOEHNERT) - Ward, Ebersole, Etherington, Reed.

Purpose:

(1) To continue the review of the proposed revision to O Appendix K of 10 CFR 50.46; 2) To review implementation of GE Appendix K analysis effort; 3) RCP trip issue resolution; and (4) To discuss NRR's ECCS-related issues of ongoing concern.

AUGUST 1 CLASS 9 ACCIDENTS (SAVIO/ HOUSTON) - Kerr, Axtmann, Moeller, Okrent, Shewmon, Siess, Ward.

Purpose:

The Subcommittee will discuss with the NRC Staff and will continue the review of draft NUREG-0956, " Source Tenn Reassessment" and discuss a SECY paper describing regulatory initiatives related to the source term reassessment.

2 CLASS 9 ACCIDENTS (SAVIO/ HOUSTON) - Kerr, Axtmann, Moeller, Okrent, Shewmon, Siess, Ward.

Purpose:

To discuss with the NRC and IDCOR the status of programs related to extending the results of the reference plants and how this relates to the ACRS reconnended search for outliers program.

6 QUALIFICATION PROGRAM FOR SAFETY-RELATED EQUIPMENT (CAPPUCCI) - Wylie, Ebersole, Michelson, Reed, Shewmon, Siess, Ward.

Purpose:

To discuss NRC Staff resolution of USI A-46,

" Seismic Qualification of Equipment in Operating Plants."  ;

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Page 2 l

SCHEDULE OF ACRS SUBCOMITTEE MEETINGS 1 0 '

AUGUST _(CONT'D) 7 LONG RANGE PLAN FOR NRC (MCKINLEY) Carbon, Lewis, Moeller, Remick, Siess, Wylie.

Purpose:

The Subcomittee will con-tinue discussions on developing comments on a long range plan i for the NRC. Topics to be discussed are primarily technical issues related to the regulation of nuclear power plant safety and safety regulation over the next 5 to 10 years. The outline of a comprehensive long range plan being developed by the EDO and OPE will also be reviewed.

7 GESSAR II (MAJOR) - Okrent, Ebersole, Etherington, Mark, Michelson, Wylie.

Purpose:

The Subcomittee will continue its review of GESSAR II for a Final Design Approval applicable to future plants.

8 - 10 304TH ACRS MEETING 27 JOINTECCSANDFLUIDDYNAMICS(BOEHNERT)-Ward,Ebersole.

Etherington, Reed.

Purpose:

(1) To review the status of the hydrodynamic loads issue for Mark I-III containment plants; (2) To review the AE0D report on Interfacing LOCAs; and (3) To review the USI A-43 Resolution Proposal.

SEPTEMBER 4&5 METAL COMPONENTS (IGNE) - Shewmon, Axtmann, Etherington, Michelson, Ward.

Purpose:

To review Reg. Guide 1.99, Rev. 2 and other related concerns, and to discuss the status of the NDT of piping program and the HSST program.

9 REACTOR OPERATIONS (MAJOR) - Ebersole, Kerr, Michelson, (1:00 P.M.) Moeller, Okrent, Reed, Remick, Ward, Wylie.

Purpose:

To discuss recent operating experiences.

10 REGULATORY ACTIVITIES (DURAISWAMY) - Siess, Carbon, Kerr (tent.), Ward (part-time),Wylie.

Purpose:

To review: (1)

Reg. Guide 1.23, Rev.1, " Meteorological Measurement Programs for Nuclear Power Plants," (2) proposed Reg. Guide (Task No.

IC 609-5), " Criteria for Power, Instrumentation, and Control Portions of Safety Sytems," and (3) Reg. Guide 1.105, Rev. 2

" Instrument Setpoints for Safety-Related Systems (tent.)."

10 WESTINGHOUSE WATER REACTORS (CAPPUCCI) - Ebersole, (Closed) Etherington, Michelson, Okrent, Shewmon, Ward (part-time).

Purpose:

' To begin the PDA review of the Westinghouse Advanced PWR (RESAR SP/90).

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Page 3 SCHEDULE OF ACRS SUBCOPti!TTEE MEETINGS V

SEPTEMBER (CONT'D) 11 LONG RANGE PLAN FOR NRC (MAJOR) - Carbon, Lewis, Moeller, Remick, Siess, Wylie.

Purpose:

The Subcomittee will con-tinue discussions on developing coments on a long range plan for the NRC. Topics to be discussed are primarily technical issues related to the regulation of nuclear power plant safety and safety regulation over the next 5 to 10 years.

11 RIVER BEND (SAVIO) - Okrent, Ebersole Shewmon.

Purpose:

To review Gulf States Utilities Company's application for an OL.

12 - 14 305TH ACRS MEETING 16 & 17 HUMAN FACTORS TOUR (RUSSELVILLE, AR) (SCHIFFGENS) - Ward, (Closed) Lewis, Michelson, Moeller, Reed, Remick, Wylie.

Purpose:

This will be a tour and examination of ANO-l's emergency procedures (symptom based) and facilities.

23 & 24 JOINT STRUCTURAL ENGINEERING AND SEISMIC DESIGN OF PIPING (IGNE) - Siess, Ebersole, Etherington, Mark, Okrent, Shewmon,

Purpose:

To review the status of research programs on l containment integrity, seismic margins, piping reliability, and other related matters.

OCTOBER 8 RELIABILITY ASSURANCE (VALVES) (MAJOR) - Michelson, Ebersole, Kerr, Okrent, Reed, Ward.

Purpose:

To continue discussions on valve reliability. A risk perspective on valve perfonnance will be sought. Also to be studied is the importance of valves from a safety standpoint.

9 LONG RANGE PLAN FOR NRC (MAJOR) Carbon, Lewis, Moeller, Remick, Siess, Wylie.

Purpose:

The Subcomittee will con-tinue discussions on developing coments on a long range plan for the NRC. Topics to be discussed are primarily technical issues related to the regulation of nuclear power plant safety and safety regulation over the next 5 to 10 years.

10 - 12 306TH ACRS MEETING O

V/P -

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p SCHEDULE OF ACRS SUBCOMITTEE MEETINGS  !

DATES TO BE DETERMINED (Augast/ September) ADVANCED REACTORS (EL-ZEFTAWY) - Carbon, Siess, Mark.

Purpose:

To discuss the proposed policy for regulation of advanced nuclear power plants.

(August / September) HUMAN FACTORS (SCHIFFGENS) - Ward, Reed, Remick, Wylie.

Purpose:

To explore methods for deciding what actions should  :

be automated in nuclear power plant operation.  ;

(September) DECAY HEAT REMOVAL SYSTEMS (BOEHNERT) - Ward, Ebersole, (tentative) Etherington, Reed.

Purpose:

To continue the review of NRR resolution position for USI A-45.

(September / October) ECCS (PALO ALTO, CA) (BOEHNERT) - Ward, Ebersole, Etherington. .

Purpose:

To continue the review of the joint NRC/B&WOG/EPRI/ '

B&W joint IST Prograrn. A visit is planned to the EPRI Stanford Research Institute facilities supporting this Program.

(October) JOINT REACTOR RADIOLOGICAL EFFECTS AND FIRE PROTECTION O (MERRILL/ ALDERMAN) - Moeller, Axtmann.: Carbon, Ebersole, Michelson, Okrent, Reed, Siess, Wylie.

Purpose:

To review the increased N-I6 radioactivity and fire protection problems in using H 2addition to BWRs to reduce IGSCC.  ;

(October) ATWS (BOEHNERT) - Kerr, Ebersole, Michelson, Ward.

Purpose:

To continue the review of the status of ATWS Rule implemen-tation effort and related issues. ,

October (tent.) QUALITY AND QUALITY ASSURANCE IN DESIGN AND CONSTRUCTION (MAJOR) - Remick, Michelson, Okrent, Reed, Siess, Ward, Wylie.

Purpose:

(I) To review the final Rule on "The Important To Safety Issue, and (2) To be briefed on the "NRC Quality i Assurance Program Implementation Plant."

(Fall) RELIABILITY & PROBABILISTIC ASSESSMENT (location to be (tent.) determined) (SAVIO) - Okrent, Kerr, Ebersole, Lewis, Mark, ,

Michelson, Siess, W&rd, Wylie.

Purpose:

To review the probabilistic risk assessment for Millstone 3.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING O

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS JULY 17, 1985 ATWS (BOEHNERT) Kerr, Ebersole, Ward, Wylie Cons.: Davis, Lee, Lipinski PURPOSE: To discuss RPS and scram breaker reliability.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is reauested; by what date is it needed?

To investigate RPS and scram breaker reliability; as soon as reasonably possible.

O what will be done at this meeting?

To begin investigation of RPS reliability (see attached memo).

What would be the consequence of postponing this meeting?

Delay in researching this topic. It is not an urgent priority issue (per ACRS j discussion at May meeting).

! PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING

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DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBER $ ,

JULY 18 & 19, 1985 V0GTLE, UNITS 1 & 2 (SCHIFFGENS/MCKINLEY)

Ebersole, Okrent, Reed, Wylie i Cons.:

PURPOSE: To begin review of OL application for Vogtle, Units 1 & 2.

s LOCATION: AUGUSTA, GA  ;

BACKGROUND:

What action is requested; by what date is it needed?  ;

Subcomittee OL review in time for Comittee consideration at the 304th, Aug. 9-10, ,

1985 ACRS meeting.

at will be done at this meeting?

Review the OL application and draft a letter.

What would be the consequence of postponing this meeting?

It could be postponed -- I think. The current schedule Has the fuel load date set for September 1986.

PERTINEN7 PUBLICATIONS AND THEIR AVAILABILITY:

To be provided.later. The Staff anticipates publishing the SER on, or about June 20, 1985.

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SCHEDULE OF ACRS SUBC0milTTEE MEETING O

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMIERS JULY 30, 1985 JOINT WASTE MANAGEMENT AND (MERRILL/FRALEY)

PROCEDURES & ADMINISTRATION Moeller, Axtmann, Ebersole, Kerr, Mark, Remick, Shewmon, Siess, Ward

[

PURPOSE: To review the ACRS Role in the Civilian High-Level Radioactive Waste .

Management Program.

LOCATION: WASHINGTON, DC  ;

BACKGROUND:  !

What action is requested; by what date is it needed?

Chairman Palladino requested (of SECY) on June 25, 1985 written'ACRS comments on the l Staff's proposal before he votes on this issue; no date specified.

What will be done at this meeting? i (1) Review Staff's proposal to the Comission regarding High-Level Waste Management Oversight Advisory Comittee (Ref.1);

(2) Review OPE's memorandum to the Comissioners regarding same (Ref. 2); and (3) Prepare written comments on Staff's proposal for full Comittee consideration

during its 304th meeting, August 8-10, 1985. ,

What would be the consequence of postponing this meeting?

It would result in postponing the submission of written ACRS coments to Chairman Palladino.

i PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: [

1. SECY-85-197, Advisory Comittee for Overseeing the High-level Radioactive Waste Repository Program, Policy Issue (Notation Vote), dated May 31, 1985

2. Memo for Comissioners from J. Zerbe, OPE, regarding OPE's review of Reference 1 i (preceding),datedJune 24, 1985 I

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SCHEDULE OF ACRS SUBCOMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS JULY 31, 1985 REACTOR RADIOLOGICAL EFFECTS (MERRILL) Moeller, Axtmann, Okrent Cons.: To be determined PURPOSE: To review: (1) INP0's Radiation Protection Program, (2) NRC's 2-year program evaluating the effectiveness of INP0's Radiation Protection Program, and (3) INP0's recent SOER 85-3, " Excessive Personnel Radiation Exposures."

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

p Dan Muller, AD/RP, suggested ACRS review of item (1) above and requested review of item (2). Item (3) was added because of its related significance and recent issuance.

What will be done at this meeting?

Briefing and review of topics named above; preparation of Subcomittee coments and recommendations.

What would be the consequence of postponing this meeting?-

No adverse effects; meeting is being held during a week when Subcommittee members will already be here for other, more urgent Subcomittee meetings -- hence, better utilization of resources.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. INPO Significant Operating Experience Report (50ER) 85-3, " Excessive Personnel Radiation Exposures," dated April 30, 1985 (INP0 Confidential)

2. A Status Report will be prepared before the meeting. .,

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s SCHEDULE OF ACRS SUBCOMMITTEE MEETING

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DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS JULY 31, 1985 ECCS (BOEHNERT) Ward, Ebersole, Etherington, Reed Cons.: Catton, Schrock, Sullivan.

Theofanous, Tien PURPOSE: (1) To review Appendix K revision effort.

(2) To review implementation of GE Appendix K analysis effort.

(3) RCP Trip issue resolution.

(4) Discuss NRR's ECCS-related issues of ongoing concern.

LOCATION: WASHINGTON, DC BACKGROUND:

1at action is requested; by what date is it needed?

(1) September ACRS veview. '

(2) September ACRS review.

What will be done at this meeting?

Review of I & 2 for September ACRS. Items 3 & 4 were left over from a previous meeting, What would be the consequence of postponing this meeting?

Loss of timely review and subsequent schedule impact on the Commission's review of issues 1 & 2 above.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

lo be provided in the near future.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING

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DATE SUBCOMMITTEE MEETING STAFF ENGR, & MEpBERS AUGUST 1, 1985 CLASS 9 ACCIDENTS (SAVIO/ HOUSTON) Kerr, Axtmann, Moeller, Okrent Shewmon, Siess, Ward PURPOSE: To discuss with the NRC Staff and to continue the review of draft NUREG-0956,

" Source Term Reassessment" and to discuss a SECY paper describing regulatory initiatives related to the source term reassessment.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

The Subcomittee has requested the Staff to provide some indication of how the source p term research will be used in the regulatory regime.

\.

What will be done at this meeting?

The NRC will provide a status of the work related to possible regulatory initiatives based on the source term work.

What would be the consequence of postponing this meeting?

The NRC Staff has recomended this date as one which would provide adequate time for ACRS coments to be incorporated in the SECY paper.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. SECY paper on Regulatory Initiatives Related to the Source Term Reassessment (to be provided).

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SCHEDULE OF ACRS SUBCOMMITTEE HEETING t

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DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS AUGUST 2, 1985 CLASS 9 ACCIDENTS (SAVIO/ HOUSTON) Kerr, Axtmann, Moeller, Okrent, Shewmon, Siess, Ward Cons.: Bender, Catton, Corradini, Davis, Lee, PURPOSE: To discuss with NRC and IDCOR the status of programs related to extending the results of the reference plants and how this relates to the ACRS recommended search for outliers program.

LOCATION: WASHINGTON, DC BACKGROUND:

~~ ' hat action is requested; by what date is it needed?

ACRS requested the Staff for a status of NRR's work regarding the tulatory use of the source term work.

What will be done at this meeting?

Status report.

What would be the consequence of postponing this meeting?

The Staff is preparing a SECY paper concurrently with Source Term Reassessment to define possible regulatory initiatives related to source term. Delaying this SECY paper would delay issuance of source term work, s

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. NUREG-0956 )

2. SECY paper on Source Term ) To be supplied

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x SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MDBERS AUGUST 6, 1985 QUALIFICATION PROGRAM FOR (CAPPUCCI)Wylie, SAFETY-RELATED EQUIPMENT Ebersole, Michelson, Reed, Shewmon, Siess, Ward PURPOSE: To discuss the NRC Staff resolution of USI A-46, " Seismic Qualification of Equipment in Operating Plants."

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

(NACRSletterontheresolutionofUSIA-46. 303rdACRS(July 1985).

What will be done at this meeting?

Review the proposed resolution of USI A-46. Prepare the report to the full Comittee and suggested report to Commission.

l l What would be the consequence of postponing this meeting?

4 Delay of ACRS coments.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. CRGR Resolution Package (end of May 1985 - tentative).

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SCHEDULE OF ACRS SUBCOMMITTEE MEETIliG

[ I xy DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS AUGUST 7, 1985 LONG RANGE PLAN FOR NRC (MCKINLEY) Carbon, Lewis, Moeller, Remick, Siess, Wylie PURPOSE: The Subcommittee will continue discussions on developing coments on a llong range plan for the NRC. Topics under discussion are primarily technical issues related to the regulation of nuclear power plant safety and safety regulation over the next 5 to 10 years. The outline of a comprehensive inng range plan being developed by the EDO and OPE will also be reviewed.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Currently a report addressed to the Comission or as input into a parallel OPE effort is expected. The current projection is to conclude this effort in October 1985.

What will be done at this meeting?

To be determined.

What would be the consequence of postponing this meeting?

Timeliness of effort would be effected. Would become out of phase with a parallel DPE effort on LRP.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

Dr. Carbon's latest review plan for this effort is available.

A Status Report will be prepared before the meeting.

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l SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMIERS AUGUST 7, 1985 GESSAR II (MAJOR) Okrent.

Ebersole, Etherington, Mark, Michelson, Wylie Cons.:

PURPOSE: The Subcommittee will continue its review of GESSAR II for a Final Design Approval applicable to future plants.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Final Design Approval of GESSAR II; as soon as possible following completten of the Staff's review.

What will be done at this meeting?

Review the resolution of items in SSER 4.

What would be the consequence of postponing this meeting?

Delay the issuance of a forward looking FDA for GESSAR II.

PERTINENT PUBLICATIONS AND THE:3 AVAILABILITY:

1. SSER 4 (NUREG-0979) Safety Evaluation Report related to the final design approval of the GESSAR BWR/6 Nuclear Island Design (Draft SSER #4 received in late June and distributed).  ;

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, SCHEDULE OF ACRS SUBCOMMITTEE MEETING i DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS

. AUGUST 27, 1985 JOINT ECCS AND (BOEHNERT) Ward.

FLUID DYNAMICS Ebersole, Etherington, Reed Cons.: Catton, Schrock, Sullivan.

Theofanous, Tien PURPOSE: (1) To review the status of the hydrodynamic loads issue for Mark I-III containment plants.

(2) To review the AE0D report on Interfacing LOCAs.

(3) To review the USI A-43 Resolution Proposal.

LOCATION: WASHINGTON, DC BACKGROUND:

What action ~is requested; by what date is it needed?

No specific action date needed.

O What will be done at this meeting?

(1) Review the status of hydrodynamic loads issue for Mark I-III containment plants. .

(2) Review AEOD report on Interfacing LOCAs.

(3) Review implementation plan for USl A-43.

What would be the consequence of postponing this meeting?

Loss of timely review of Item 3 per NRR's schedule.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 4 & 5, 1985 METAL COMPONENTS (IGNE) Shenon, Axtmann, Etherington, Michelson, Ward Cons.: Kassner, Odette PURPOSE: To review Regulatory Guide 1.99, Rev. 2, and other related concerns, and to discuss status of NDT of piping program and HSST program.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

ACRS action is requested by the NRC Staff before Reg. Guide 1.99, Rev. 2, is

[v 7romulgated.

What will be done at this meeting?

The Subcommittee will review with the NRC Staff Reg. Guide 1.99, Rev. 2, and develop recomendations for ACRS comments.

What would be the consecuence of postponing this meeting?

None, except that ACRS coments, if any, will not impact Reg. Guide 1.99, Rev. 2.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Reg. Guide 1.99, Rev. 2, "Ef fects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials, due for ACRS review in late June or early July 1985.

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x SCHEDULE OF Alls SUBCOMMITTEE MEETING l I U'

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 9, 1985 REACTOR OPERATIONS (MAJOR) Ebersole, Kerr, (1:00P.M.) Michelson, Moeller, Okrent, Reed, Remick, Ward, Wylie PURPOSE: The Subcommittee will discuss recent operating occurrences.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Review recent operating experience, select incidents of significance for presentation to full ACRS during September meeting, i.E!L?

Review recent operating experience.

What would be the consequence of postponing this meeting?

Would consider events at a later date.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Status Report to be provided.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MDBERS SEPTEMBER 10, 1985 REGULATORY ACTIVITIES (DURAISWAMY Siess, Kerr (tent. . Carbon, Ward (part time), Wylie PURPOSE: To review the following:

1. Regulatory Guide 1.23, Rev 2., " Meteorological Measurement Programs for Nuclear Power Plants" pre-coment).

2. Proposed Regulatory Guide Task No. IC 609-5), " Criteria for Power, Instrumentation, and Control Portions of Safety Systems" (post coment) .

3. Regulatory Guide 1.105, " Instrument Setpoints for Safety-Related Systems (tent.)."

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

- ACRS concurrence in the Staff's posposal to issue item 1 for public coments.

ACRC concurrence in the Regulatory positions of item 2.

What will be done at this meeting?

See purpose.

What would be the consequence of postponing this meeting?

! Delay the issuance of these Guides.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

The above mentioned Regulatory Guides are expected to be made available to the ACRS during July 1985.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING O .

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 10, 1985 WESTINGHOUSE WATER (CAPPUCCI) Ebersole, REACTORS Etherington, Michelson, (CLOSED) Okrent, Shewnon, Ward (part-time)

Cons.: Davis PURPOSE: To begin PDA review of Westinghouse Advanced PWR (RESAR SP/90).

LOCATION: WASHINGTON, DC BACKGROUND:

What action is reauested; by what date is it needed?

CRS letter on PDA approval by 11/86.

What will be done at this meeting?

Begin reviewing design modules.

What would be the consequence of postponing this meeting?

I Delay in the completion of ACRS PDA review.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. RESAR SP/90 Standard Plant Design (50-601). -

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[^ SCHEDULE OF ACRS SUBCOMMITTEE MEETING

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DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 11, 1985 LONG RANGE PLAN FOR NRC (MAJOR) Carbon. Lewis.

Moeller, Remick, Siess, Wylie PURPOSE: The Subcommittee will continue discussions on developing comments on a long range plan for the NRC. Topics under discussion are primarily technical issues related to the regulation of nuclear power plant safety and safety regulation over the next 5 to 10 years.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Currently a report addressed to the Comission or as input into a parallel OPE effort is expected. The current projection is to conclude this effort in October 1985.

at will be done at this meetin ?

o be detertnined.

What would be the consequence of postponing this meeting?

Timeliness of effort would be effected. Would become out of phase with a parallel DPE offort on LRP.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

Dr. Carbon's latest review plan for this effort is available.

A Status Report will be prepared before the meeting.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING p_

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I DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 11, 1985 RIVER BEND (SAVIO) Okrent, Ebersole, Shewmon PURPOSE: To review Gulf States Utilities Company's application for an OL.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is reauested; by what date is it needed?

Issue an ACRS " full power" 0L letter; at the September ACRS meeting.

What will be done at this meeting?

Complete the Subcommittee action on the River Bend OL review in support of an ACRS review at the September ACRS meeting.

~

Sat would be the consequence of postponing this meeting?

Possible delay of River Bend full power operation.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

NRC Staff SER Supplement to be stepplied at the August ACRS meeting, l

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SCHEDULE OF ACRS SUBCOMITTEE MEETING I r\ '

b DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEIEERS SEPTEMBER 16 & 17, 1985 HUMAN FACTORS TOUR (SCHIFFGENS) Ward, (CLOSED) Lewis, Michelson, Moeller, Reed, Remick, Wylie PURPOSE: This will be a tour and examination of ANO-l's emergency procedures (symptom based) and facilities. The Subcommittee wants the opportunity to examine procedures at an operating plant and see how the TMI required backfits such as SPDS Interface. Up to a day and a half is expected. ANO-1 is an 850 MWe, B&W PWR, LOCATION: ANO-1, Russellville, AR (v50 r.iles outside of Little Rock, AR)

BACKGROUND:

What action is requested; by what date is it needed?

Review implementation of TMI required backfits. No schedolar requirements.

What will be done at this meeting?

Tour and review facilities.

What would be the consequence of postponing this r.eeting?

None PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

! 1. One copy of Ah0-1 Emergency Operating Procedures is available for your inspection at the ACRS Office (ask J. Schiffgens for it).

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DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEM ERS :

SEPTEMBER 23 & 24, 1985 JOINT STP.UCTURAL ENGINEERING (IGNE)Siess,Ebersole, t AND SEISMIC DESIGN OF PIPING Etherington, Mark, Okrent, Shewmon l PURPOSE: To review the status of research programs on containment integrity, seismic i margins, piping reliability, and other related matters. [

LOCATION: WASHINGTON, DC l BACKGROUND:

What action is requested; by what date is it needed?

Meeting is requested by the Subcomittee Chairman in order to keep abreast of research needs and justifications. This infonnation will be needed to plan for ACRS coments j on the Research Program and Budget.  !

i:

O'Whatwillbedoneatthismeeting?

e Discuss the status of the research program with NRC researen and regulatory staffs in j order to plan to provide ACRS coments for .research program and budget reviews. l What would be the consequence of postponing this meeting?

None, except that it is required by law that the ACRS provide Congress and the '

Comission periodic reports on the NRC Research Program and Budget.  !

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

I A status report will be provided with pertinent background infonnation prior to the i meeting.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING 1

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS OCTOBER 8, 1985 RELIABILITY ASSURANCE (MAJOR) Michelson, (VALVES) Ebersole, Kerr, Okrent, Reed, Ward PURPOSE: To continue discussions on valve reliability. A risk perspective on valve performance will be sought. Also to be studied is the importance of valves from a safety standpoint. A discussion with Limitorque Co. is also expected.

LOCATION: WASHINGTON, DC j BACKGROUND:

What action is recuested; by what date is it needed?

No action has been requested. This is a self-initiated task.

What will be done at this meeting?

This meeting will conclude a series of three meetings designed to explore the topic of valve reliability.

What would be the consecuence of postponing this meeting?

No adverse impact.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: ,

A Status Report will be issued prior to the meeting.

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N SCHEDULE OF ACRS SUBCOMMITTEE MEETING r

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMERS OCTOBER 9, 1985 LONG RANGE PLAN FOR NRC (MAJOR) Carbon, Lewis, Moeller, Remick, Siess, Wylie PURPOSE: The Subcomittee will continue discussions on developing coments a long range plan for the NRC. Topics under discussion are primarily technical issues related to the regulation of nuclear power plant safety and safety regulation over the next 5 to 10 years.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Currently a report addressed to the Comission or as input into a parallel OPE effort is expected. The current projection is to conclude this effort in October 1985.

What will be done at this meeting?

To be determined.

What would be the consequence of postponing this meeting?

Timeliness of effort would be effected. Would become out of phase with a.pcrallel OPE effort on LRP.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

Dr. Carbon's latest review plan for this effort is available.

A Status Report will be prepare'd before the meeting.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING p]

N DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED ADVANCED REACTORS (EL-ZEFTAWY) Carbon, (AUGUST / SEPT.) Mark, Siess Cons.:

PURPOSE: To discuss the proposed policy for regulation of advanced nuclear power plants.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it naederf?

ACRS coments on the proposed policy statement for advanced reactors; 306th ACRS meeting (October 1985).

Review the revised version of the policy statement. Prepare the report to the full Comittee and suggested report to the Comission.

What would be the consecuence of postponing this meeting?

Delay of ACRS coments to the Comission.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Letter fm R. Fraley to J. Zerbe, dated 4/15/85.

2. SECY-84-453A - Regulatory policy for advanced reactors, dated 2/26/85.

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SCHEDULE OF ACRS SUBCOPetITTEE MEETING Y

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED ' HUMAN FACTORS (SCHIFFGENS) Ward, (AUGUST /SEPTENER) Reed, Remick, Wylie Cons.: Ginny PURPOSE: To explore methods for deciding what actions should be automated in nuclear power plant operation.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Mr. Ward asked" researchers from the University of Illinois to make a presentation to the Subcomittee. ,

What will be done at this meeting?

O t

bd ) hat would be the consequence of postponing this meeting?

No serious consequences from postponement that I can see.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

None at this tim %

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED DECAY HEAT REMOVAL SYSTEMS (BOEHNERT) Ward, (SEPTEMBER) Ebersole Etherington, (tentative) Reed Cons.: Catton, Davis PURPOSE: To continue the review of NRR resolution position for USI A-45.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is reauested; by what date is it-needed?

N/A What will be done at this meeting?

Begin review of NRR's proposed resolution position for USI A-45.

{ knat would be the consecuence of postponing this meeting?

At this time, given the "spongyness" in the schedule, no definitive answer can be given to this question.

PERTINENTPUBLICATIONSANDTHEIRAVfILABILITY:

To be provided when available.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING

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V DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED ECCS (BOEHNERT) Ward.

(SEPTEMBER /0CTOBER) Ebersole, Etherington, Cons.: Catton, Schrock, Sullivan, Theofanous, Tien PURPOSE: To continue the review of the joint NRC/B&W Owners Group /EPRI/B&W joint IST and related programs.

LOCATION: PALO ALTO, CA area BACKGROUND-What action is requested; by what date is it needed?

No specific action date -- part of ongoing Subcomittee review of Program.

Continue Program review. Key discussion topics will largely be determined based on previous Subcommittee meeting in June. Also visit EPRI SRI-supported test facilities.

What would be the consecuence of postponing this meeting?

No significant impact vis-a-vis facility visits.

Uncertain on program discussion pending results of June Subcomittee meeting in Alliance, OH.

PERTINENT DOCUMENTS AND THEIR AVAILABILITY:

To be provided on a timely basis. ]

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,s SCHEDULE OF ACRS SUBCOMMITTEE MEETING U

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED JOINT REACTOR RADIOLOGICAL EFFECTS (MERRILL/ ALDERMAN)

(0CTOBER) AND FIRE PROTECTION Moeller, Axtmann, Carbon, Ebersole, Michelson, Okrent, Reed, Siess, Wylie PURPOSE: To review the increased N-16 radioactivity and fire protection problems in using H2addition to BWRs to reduce IGSCC.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Comments to Staff. Needed by no specific date.

hat will be done at this meetin ?

view of H2 addition regarding N-16 activity and fire hazards.

What would be the consequence of postponing this meeting?

Meeting suggested by Committee members. Postponement of meeting will not have any scrious consequences.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Paper by V. Benaroya regarding subject available.

2. EPRI workshop report on subject, JAJARC #85-WH21 in ACRS office.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING >

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DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED ATWS (80EHNERT) Kerr, (OCTOBER) Ebersole, Michelson, Ward Cons.: Lee, Lipinski, Davis PURPOSE: To continue the review of the ATWS Rule implementation effort.

LOCATION: To be determined.

BACKGROUND:

What action is reouested; by what date is it needed?

'o specific action date.

What will be done at this meeting?

(See Purpose above)

What would be the consequence of postponing this meeting?

No significant consequences.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

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, SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED QUALITY AND QUALITY ASSURANCE (MAJOR)Remick, (October) IN DESIGN AND CONSTRUCTION Michelson, Okrent,-

Reed, Siess, Ward, Wylie PURPOSE: (1) To review the final rule on the, "Important to Safety Issue."

(2) To be briefed on the NRC Quality Assurance Program Implementation Plan.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

, Approval of rule by~IE QAB; by the August full Committee meeting.

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( hat will be done at this meeting?

Review final rule in preparation to bring before the full Comittee for coment/ approval . Discussion of NRC's Quality Assurance Program Implementation Plan, for information.

What would be the conseouence of postponing this meeting?

It could delay issuance of the final rule on Important to Safety Issue.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. " Quality Assurance Program Implementation Plan," SECY-85-65 is available.

2. Current (for public coment) version of the Important to Safety Rule is available.

Revised version and response to public connents expected prior to meeting.

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p SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED RELIABILITY AND PROBABILISTIC (SAVIO)0krent,Kerr, (FALL) ASSESSMENT Ebersole, Lewis, Mark, (ttntative) Michelson, Siess, Ward, Wylie Cons.:

PURPOSE: To review the PRA for Millstone 3 (not an OL critical path item).

LOCATION: To be determined BACKGROUND:

What action is requested; by what date is it needed?

yeviewoftheMillstone3PRA;themeetingistobescheduledafterthecompletionof he NRC Staff's review of the PRA (estimated to be by the end of~May 1985).

k d here is no ACRS action date.

What will be done at this meeting?

R3 view of the Millstone 3 PRA for infonnation.

What would be the consequence of postponing this meeting?

ACRS has stated that this review need not be completed prior to full power operation.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

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1. Millstone 3 PRA (distributed).

2. NRC Staff report on the results of the NRC/LLNL review of the Millstone 3 PRA (expected by the end of May 1985).

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' APPENDIX IV '

STAFF PRESENTATION OF WATTS BAR IDVP AND ALLEGATIONS '

NRR STAFP PRESENTATION TO THE  !

O ACES i

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SUBJECT:

WATTS BAR IDVP AND ALLEGATIONS I

t DATE:- JULY 12,1985  !

O PRESENTER:  !

E. G. ADENSAM '

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i PRESENTER'S TITLE / BRANCH /DIV:

CHIEF, LICENSING BRANCH NO. 4, DIVISION OF LICENSING l

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PRESENTER'S NRC TEL. NO.: 301-492-7831 j I

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FATTS EAR "DVP AXD A1I3 GAT"0XS G3XBAL REQU3EYEES OF AX IDV?

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• TATTS BAR IDVP l STAFF 33V:EY l

E WATTS BAR AIJRGATIOXS  :

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. _ _ _ _ _ __ ,_.- ,-._.---.. .-- ' ~ - - - ~ ~ ' ^ ^ ^ ^ ^ ~ - ' ^ ^ ^ ^ ~ ^ ^ ^ ^

GEN 33AL R3QE33h3N"S 07 AN IBVP REVIEW ORGANIZATION COMPETENT INDEPENDENT O

SCOPE OF REVIEW

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TAILORED TO CONCERNS O

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4 O YATTS BAR ::DVP BLACK & VEATCH AUDIT THE AFW SYSTEM 0F WATTS BAR UNIT 1 TO ENSURE THAT THE SYSTEM HAS BEEN DESIGNED AND -

(] CONSTRUCTED IN ACCORDANCE WITH THE LICENSE APPLICATION AND LICENSE COMMITMENTS t

• TVA

{ s EVALUATE THE BLACK & VEATCH FINDINGS TO DETERMIN THEIR APPLICABILITY TO OTHER WATTS BAR SYSTEMS D

c& -

O STAF7 REV::EW B:lCK & VEATCH 3FFORT (AFFSTSTEX)

B&V BOTH COMPETENT AND INDEPENDENT 97 FINDINGS REVIEWED OUT Or 428 O ONE riNDinc sTitt uNDEa aEviEw

- RESP 011SE TO IE BULLETIN 79-02 TVA GENER:C 3FFOR:.'

INSPECTION REVIEW DEDICATED REVIEW GROUP O

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OBJECTIVES OF REVIEF GROUP

1. TO DETERMINE IF THE TVA PROGRAM TO

ADDRESS THE FINDINGS OF THE B&V REPORT WAS ADEQUATE WITH RESPECT TO EVALUATION O OF GENERIC APPLICABILITY OF THE FINDINGS AND CORRECTIONS MADE TO THE PLANT DESIGN '

AND CONSTRUCTION

2. TO ADDRESS RECENT ALLEGATIONS REGARDING IDVP

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DEDICATED REVIEW GROUP REVIEW 0F WATTS BAR IDVP B&V REPORTS TVA POLICY COMhDTTEE REPORT AND RELATED

() DOCUMENTS NSRS REPORTS AND RELATED DOCUMENTS SITE VISIT

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PREPARE SER I l hh

"YPES OF CONCERNS RAISED O

GENERAL MOST PREVIOUSLY IDENTIFIED IN NSRS REPORTS, NCR's, CDR's, OTHER TVA CORRESPONDENCE, AND INSPECTION REPORTS CONCERNS REGARDING B&V CLOSEOUT OF 50D ITEMS ONLY ONE CONSTRUCTION SPECIFICATION LOOKED AT BY B&V B&V DID NOT KNOW HOW THE PLANT WAS BUILT B&V COMPARISDN MADE WITH REGARD TO DESIGN CRITERIA NOT REGULATORY CRITERIA OTHER TYPES OF ISSUES:

MATERIAL TRACEABILITY RECORDS OF TOTAL LOADS STRUCTURAL STEEL WELD REQUIREMENTS CABLE PROBLEMS PROCUREMENT PROBLEMS VOLTAGE REGULATION FOR BUSES UNISTRUT USE INADEQUATE D/G LOAD MARGINS

< sm

O REVIEW PIE ASSIGN RESPONSIBLE ORGANIZATION SCREEN FOR SAFETY SIGNIFICANCE AND IMPACT ON f LICENSING i i

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REVIEW RESPONSES TO MAY 16, AND MAY 30, 1985 LETTERS l

REVIEW AND EVALUATE ISSUE i

l PREPARE SER OR INSPECTION REPORT  !

D l l

l APPENDIX V

.. I CONCERNS RE TVA CONSTRUCTIO  !

<,[ p ung e,

! UNITED STATES

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; NUCLEAR REGULATORY COMMISSION wassincton o.c.nosss

...../ May 16, 1985 Docket Nos: 50-390, 50-391 and 50-438, 50-439 Mr. H. G. Parris Manager of Power Tennessee Valley Authority 500A Chestnut Street, Tower II Chattanooga, Tennessee 37401 -

Subject:

Concerns Regarding TVA Construction Sites Enclosure I lists eleven concerns about your Watts Bar facility that have been comunicated to the NRC. We ask that you review these concerns and take appropriate steps to assure that your programs and implementation of those programs-in these areas are adequate to meet applicable requirements and to support safe operation of the facility.

any generic implications of these issues.Furthermore, we ask that you address are not very specific. We recognize that some of these concerns ,

to assume there is no basis for concern.However, that lack of specificity should not le

(" broad (Q) enough for you to certify the safety significance of these concerns.Yo Pursuant to Section 182 of the Atomic Energy Act of 1954, as amended, we ask that you provide the results of your review as soon as possible to assist us in our evaluation of these concerns Enclosure 2 lists a number of question that we have regarding these concerns..

soon as possible. -

Please provide us with your response ass We also ask that you identify any outstanding cases currently under review j

by intimidation.

or TVA's Office of the General Counsel regarding employee harassment, reprisals Watts Bar and Bellefonte facilities.We recently received some additional c

! no new issu:*s related to safe plant operation have been identified.You We regret should some omissions occur, but this is how they were received by us.

l I suggest wetomeet responding as soon as you are prepared to discuss your schedule for this letter.

refer them to E. Adensam of my staff on FTSShould 492-7831.

you have any questions on this ma l

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- 2-The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten resondents; therefore, OMB clearance is not required under

! P.L.96-511.

Sincerely.

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g L. Thompso J r. frect6r D sion of Licensi Office of Nuclear Reactor Regulation

Enclosures:

I As stated ,

cc: See next page ,

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WATTS BAR Mr. H. G. Parris Manager of Power i Tennessee Valley Authority 500A Chestnut Street Tower II Chattanooga, Tennessee 37401 cc: Herbert S. Sanger, Jr. , Esq.

General Counsel Tennessee Valley Authority 400 West Sumit Hill Drive, E 11B 33  :

Knoxville, Tennessee 37902  !

i Mr. D. Checcet

- Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230

• Mr. Ralph Shell Tennessee Valley Authority 400 Chestnut Street, Tower II Chattanooga, Tennessee 37401 fq Mr. Donald L. Williams, Jr.

(V) Tennessee Valley Authority 400 West Sumit Hill Drive, W10B85 Knoxville, Tennessee 37902 Resident Inspector / Watts Bar NPS

• c/o U.S. Nuclear Reg 01 story

'c- Commission Rt. 2 - Box 300 Spring City, Tennessee 37381 Regional Administrator U.S. Nuclear Regulatory Comission, i

Region II 101 Marietta Street, N.W., Suite 2900 l Atlanta, Georgia 30323 l

i Mr. David Ellis

' Tennessee Valley Authority 400 Chestnut Street Tower II Chattanooga, Tennessee 37401 Mr. Mark J. Burzynski Tennessee Valley Authority ~

t Watts Bar NP [~

P.O. Box 800 .

Spring City, Tennessee 37381 v

1 BELLEFONTE Mr. H. G. Parris Manager of Powar Tennessee Valley Authority 500A Chestnut Street, Tower II Chattanooga, Tennessee 37401 cc: Herbert S. Sanger, Jr. , Esq. J. Nelson ,erace General Counsel Tennessee Valley Authority U.S. Nuclear Regulator.s Comission, Region II 400 West Sumit Hill Dr., E11B33 101 Marietta Street, Suite 3100 Knoxville, Tennessee 37902 Atlanta, Georgia 303?3 Mr. H. N. Culver Tennessee Valley Authority 400 West Sunnit Hill Dr., 749A HBB Knoxville, Tennessee 37902 Mr. William T. Watters Lic_ensing Engineer Tennessee Valley Authority 400 Chestnut Street, Tower II Chattanooga, Tennessee 37401 Mr. R. A. Wallin O Babcock & Wilcox Company P.O. Box 1260 Lynchburg, Virginia 24505 Mr. Robert B. Borsum -

Babcock & Wilcox Company g Suite 220 7910 Woodmont Avenue Bethesda, Maryland 20814 Mr. Donald L. Williams, Jr.

Tennessee Valley Authority 400 West Sumit Hill Drive W10R85 Knoxville, Tennessee 37902 Resident Inspector, Bellefonte NPS c/o U.S. Nuclear Regulatory Comission P.O. 8tox 477 Hollywood, Alabama 35752 G

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ENCLOSURE 1 CONCERNS RELATED TO WATTS BAR

1. A concern has been expressed that there is no material control of ASME bolts smaller than 1"; and, therefore, the bolts < 1" are mixed up and no one EB knows where the good ones are.

2. A concern has been expressed that electrical hangers have been modified 7 '

after their initial inspection and not reinspected.

3. A concern has been expressed that field modifications have been implemented '

on components, piping, supports, structures, and embedments that resulted in no accurate records of total loads on these elements.

4. A concern has been expressed that the cumulative effect of tolerances has not been factored into the design and drawings, especially with respect to ,

hanger location. -

I Several concerns have been raised regarding the Independent Design O, 5. Verification Program conducted by Black & Veatch. These are: .

a) a concern regarding the close out of about 500 items,  ;

b) a concern that only one construction tvecification was looked at by -

C Black & Veatch in their review, c) a concern that Black & Veatch did not know how the plant was actually ,

i built, and d) a concern that Black & Veatch only compared the system's design and l construction to its design criteria, not to the underlying regulatory criteria.

6. A concern was expressed with respect to the use.of three Q lists that all  ;

differ, e i .:

7. A concern was expressed that the method of identifying NCR's and the use of the Inspection Rejection Notice (IRN) system effectively negated the j

NCR process. It was submitted this was so because an NCR was only generated when 1) the equipment / component / system /etc. had been previously inspected and accepted; 2) the records for that inspection were in the  ;

vault, and 3) there was a subsequent discovery that something was wrong;  ;

however, if there were a problem identified in an initial ins @~ction an j l IRN is generated.

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8.

A concern has been expressed with respect to structural steel welding requirements in that TVA is using a different code than the code normally used by the industry for structural steel welding.

9. A concern has been expressed that in FSAR amendment #53 TVA lessened the experience requirements for the plant manager.

10.

A concern has'been expressed regarding weld filler material control, especially in the area of storage and issuance of materials.

• 11.

A concern has been made that the Quality Assurance (QA) organization at construction sites lacks the independence required by NRC regulations.

- Also the statement was made that inadequate QA organization independence problems were identified in a Management Analysis Company (MAC) report,

" Assessment of Organizational Change in the Tennessee Valley Authority Power Program and the Nuclear Quality Assurance Program."

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ENCLOSURE 2 QUESTIONS ON WATTS BAR CONCERNS l

1.

With respect to concern 1 our initial inspection during the week of April 29, 1985, identified instances where unmarked bolts were installed in the facility on ASME components and supports. In addition, the staff also learned that two NCRs (1979 and 1981) have been issued regarding the purchasing and installation of bolts without required markings. Describe the reasons for the apparent QC breakdown in bolt control, the reason for the repeated NCR in 1981 and your evaluation of whether this occurred in other QC areas. In view of the above, what is your basis for detersining compliance with criterion VIII of Appendix B to 10 CFR 50? ,

t If documentation does not exist which demonstrates compliance with this

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regulation, describe the process and provide sample documentation which leads you to conclude that you comply with this regulation. Demonstrate that bolts less than 1", which have been installed, comply with all applicable ASME Code' requirements related to identification and control.

2.

With respect to concern 2, please review your records to determine if such modifications have occurred and, if they did, what assurance you have that the modifications have been properly reinspected. This review ,

should include applicable work requests. Verify that documentation exists to demonstrate that modified electrical hangers were designed and con-structed pursuant to applicable FSAR commitments. To the extent such documentation does not exist what is the basis for concluding that electrical Appendix 8, hangers, Criterion asX?curren,tly installed, comply with 10 CFR 50,

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3.

f With respect to concerns 3 and 4, 50.55 Interim Report No. I en NRC WBNCEB8419 states that TVA's drawing series 47A050 includes several tolerances (e.g., location of concrete anchorages, movement of attachments, (

3 and modification of baseplates) such that the cumulative effect of these

tolerances may result in sitlificant increases in baseplate stresses and anchor bolt loads. Interim Report No. 1 also states that there is no I evidence that these potential increases due to cumulative effects were considered in the design of various supports and that the cumulative effect

• j of these bolt load tolerances by 50%. could increase baseplate stress by 150% and anchor I Provide the TVA engineering specifications establishing ec.ceptable i.rJ anchorages. dimensional installation tolerances for supports, baseplates, Provide the analytical bases for establishing the above proced res which demonstrate that the effect the tolerances have on the i stresses on loads in interfacing components and structures will cause these values to exceed their allowable limits. Provide the process used when i

TVA engineering specification installation tolerances has! i l to demonstrate continued compliance with design allowable parameters, i l

Provide used. sample documentatitn which demonstrates how this process has been

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t i r ra ed wh h u u de i e Pn t d d o ducted n T c se ws o sistent with your licensing comitments .

and safe operation of the plant  !

In addition, please address the iollowing questions:

The NRC staff can identify only one General Construction Specification (GCS-G-32) in the documents reviewed by 84V. How many other General Construction Specifications are applicable to the auxiliary feedwater system?

If you identify other applicable General Construction ~

Specifications that were not reviewed by B&V, how could TVA use 88V to support a conclusion that construction complied with the FSAR comitment? ,

What corrective actions (i.e., design, hardware, procedural modifications)

- have been taken as a result of the B&V review? Does B&V agree that these actions resolve the concerns expressed in the 88V findings? What '

specific actions have been taken in systems other than the Auxiliary Feedwater System to determine the extent to which deviations found by l B&V in the AFW system existed in other systems? How have such actions 1 been documented? t r

5.

With respect to concern 6, identify the documents that demonstrate your a Q list from the date of the construction permit (CP). complia  !

6.

With respect to concern 7, is this a proper description of the NCR process?

Please verify and certify that reporting of deficiencies meets your licensing comitments and the regulations and that IRN's l and NCR's are properly controlled.

and their resolutions?. Is there a master file of IRN's -

C Technology Company Employee Response Team solicit employee v regarding improper use of the IRN process in lieu of the NCR process. '

7.

With respect to concern 8 please verify your code use for this welding to assure regulations and licensing comitments have been met.

your implementation of other types of welding conformed to the Verify accepted standards.

i Please provide memoranda or other documents indicating provided to problems NRC. with TVA's AWS welding program not previously i 8.

I With respect to concern 10, now does your program assure the ASME Code requirements of 10 CFR 50.55a are met and 10 CFR Part 50,  !

Appendix 8 Criterion VI!! traceability requirements are met. To what version of the ASME Code was TVA comitted in the CP7 this version of the ASME code require traceability of fillerDid  !

( material to welds by heat and 1ot numbers? What internal or  ;

external approvals for your program were required and received? - .

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3-What version of the specification GCS-G29M, Process Specification 1.M.3.2(RO) & l.M.3.1(R7) was used prior to 1/12/83 and 1/13/837 There appears to be an inconsistenc specifications in that 1.M.3.2(RO) y for between powerthese boilers two is process more stringent than 1.M.3.1 (R7) for nuclear plants. Describe how TVA is implementing these process specifications in the current version of GCS-G29M in the field?

9.

With respect to concern ll, describe the adequacy of the indepen.dence of the QA organization as it applies to Watts Bar. In addition, describe actions taken by TVA to resolve problems identified in the MAC report and actions TVA is taking with respect to the recort's recommen-i dations. Provide any analysis which has been conducted by TVA to determine

the extent to which Watts Bar design and construction quality may have been report.

compromised as a consequence of deficiencies enumerated by the MAC If no analysis has been conducted, do you intend to conduct such an analysis? If not, why not?

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Enclosure 3 Electrical, IAC and Diesel Generators Electrical and I&C Regulations (Reg. Guides, NUREGs, Bulle: ins and Notices) have been ignored and violated to a very large degree at all plients.

Caused by a lack of knowledge by personnel Caused by a poor attitude toward safety and regulations by personnel Caused by a lack of knowledge of industry positions on regulations

, 5% voltage drop at each plant causes problems Cycles diesel generators unnecessarily, degrading reliability Too many plant shutdowns TVA compensates by operating buses at higher than nonnal voltage ratings, anticipating voltage reductions, stressing equipment and components unnecessarily and reducing their lives and re11 abilities Inadequate voltage regulation for buses Diesel Generator margins inadequate TVA has added DGs to.8F. Sequoyah and Watts Bar -

C -

Each time a question is raised, TVA must conduct another study TVA adds [ illegible) without upgrading licensing documentation l

• Diesel generator reliability problems Requires reliability upgrading program Requires reduction in number of starts "

Requires much attention given to testing program Requires preventative maintenance upgrading program Requires more interaction with INP0 and other utilities, as well as

\ vendors, to establish resolutions to problems &

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} Electrical separation and physical separation of redundant wiring and cabling and for equipment and components are all inadequate at all plants Detailed reviews need to be made (They are so extensive that a consultant probably should be used, providing independence from TVA)

Environmental is inadequate Qualification at all plants of electrical and I&C equipment and components I

Qualification was often not done If done, or records replacement of do not exist in many cases, resulting in requalification items l Current upgrade programs needs scrutiny WBN - (maybe other plants) Class IE and Non-Class IE Batteries are unacceptably supported (no battery tie-downs)

- Unistrut supports unacceptably used j

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• Human Factors engineering and/or reviews have not been implemented for control panels and stations at WBN (possibly other plants also) - Violation.

of intent of NUREG-0700.

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Too many poor engineering practices in this area i

j Out of service tags for valves, electrical equipment, etc., at Bellefonte have been violated everywhere Extremely serious personnel safety problem I

• Thermal overload bypass and indication problems at WBN - probably have similar problems meeting Reg. Guide 1.97 at other plants c

! There are cable ampacity problems at WBN where derating was not properly considered Probably problems at other plants ~ ~ '

fr loads, including diesel generator loadsInadequate management, contro f'b

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-m Inadequate control of or preparation of calculations for loads Inadequate management and centrol of load margins, including electrical loads and mechanical loads (heat. bhp, etc) that translate into electrical loads Cable tray fill criteria of 601 for I&C cables is inadequate -

National Electrical Code allows 40% and 50% on exception basis. TVA violates code Industry practice is 40%

The situation is even worse with the addition of spray-on fire retardent materials which take up space in trays Cable pull tension monitoring is lax Cable bending radii problems

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Computer cable routing program inadequate and its status system is inadequate Cable trays are too heavily filled; cables [ illegible)

C Cable megger readings are not stored as QA records, losing traceability Construction with G- Test and Installation Specs (Called General Construction Specs numbers) are often incomplete and inadequate Electrical testing and planning inadequate

1. Engineering either does not address testing or does so inadeouately Acceptance criteria is inadequate to nonexistent Electrical incorporated Standards and Guides in design criteria are treated as guides and are no.t adequately as requirements e (a)

• Electrical in generaldesign criteria, where it exists, is not complate, is vague, and is inadequate

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Cabling is routed outside trays, coiled on tray supports or floors tied on sides of trays and supports, tied on bottoms of trays, etc. All this and more exists at WBN, where extremely bad cable practices exist such as the above and 90' wire bends [illegiblel Between 400 and 500 breakers were unacceptably set at WBN. EN DES practices and attitudes concerning these were poor.

good engineering practices were violated. The National Electrical Code and

- Many cable trays at WBN are full, some exceeding 100% tray capacities, and they are not identified at site or in computer status as full Wall penetrations of WBN -

able trays are not identified by name and/or number at Lighting fixtures at WBN are not properly restrained and caDed to prevent them from becoming missiles or swinging missiles during seismic events O*WBN-(andPossiblyotherplants)-Unistrutmateriali instruments, skids, pipes, instrument conduit, lines CO control stations and panels, fluid piping on piping, lighting, etc., 2 fire protection piping, fire protection water ,

All unacceptable use for Seismic Category I support

[Items supported as such may either fail or.become missiles to cause other Illegible]

TVA comitments in FSAR, SER, and NRC Question Responses are treated lightly and are not being met in a wide number of areas Personnel do not follow regulations and commitments, and do not think they even need to report deviations or change comitments and obtain NRC acceptance TVA safety and licensing evaluations by EN DES (Including NEB) are inadequate and appear too much in cover up mode e

• TVA personnel have attitude problems in meeting regulatory comitments tw

Too many crafts and others on site at WBN Gross at lack of knowledge of regulations and their seriousness by TVA personnel all levels Lack of frequent visits to sites by Designers Communications problems among designers, constructers and operation personnel Procurement specs, drawings and. vendor supplied documents not per as-built and/or as delivered configurations TVA inadequately reviews vendor work TVA receipt and inspection of equipment are inadequate (Example:TVA in many cases does not inspect until ready to install - not when received) ss Construction vendor process requirments/ does not always follow EN DES requirements documents or instructions These do not always get included on as-built documents .

,_ Too much after-the-fact approval QC inspection is often inadequate - (It only takes a walk thru a plant such as WBN to see examples everywhere)

Engineering (EN DES) inadequately addresses and considers operation, main-in the design processtenance, testing and construction requirements and general! '

There are no' forced interactions with other utilities There is no identified to formal INPO system to track and assign connitments for problens ,

There is poor tracking of NRC experience information l

4 Improper reporting of events at operating plants or in design / construction O. TVA personnel are inadequately trained and not knowledgeable in what is reportable 1

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Lack sites of adequate (or any) configuration control (management) or at i

Poor interface control between systems Lack of traceability of design requirements Standard answer is "Its TVA Practice" Design / installation requirements drawings do not always represent esign or include d Design them guides or standards are utilized only when designer wa o use Design guides / standards inadequate in many areas These are misused - applicable parts are [Illegiblel p

• Material centrol is poor O -

Traceability of requirements, paperwork, and materials a equate are in d Paperwork for quali,ty records is poor .

l Storage requirements implementation is poor Handling poor. of equipment in storage and during and after constr WBN uction is and filth equipment in many cases is in poor condition

! inside and outside y dirty Equipment receipt and inspection is inadequate (identified pre i v ously)

These problems exist at Bellefonte and WBN (probably re) elsewhe Lack of adequate tracking for EN DES comitments nges and design cha

• Lack of good status system (punch lists) for completion of c tion, pre-op, etc. status is poorcompletion ofPla.nt-construc- NRC actions, and co e

/m' Project industry operating Engineering inadequate (or nonexistent)

[ Illegible) to incorp b) orate TVA and k-6 &

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• Calculation Problems Some are not ever prepared Some are inadequate in scope and quality Some are not stored as quality records, but are destroyed Traceability of design requirements is impacted due to above problems There is inadequate _ interface control and control of calculations

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TVA a large haspercentage set up design criteria (WBN) and, after the fact, have inactivated of criteria As-Built cases drawings and documents are nonexistent or in poor condition in many

/ TVA does not adcquately (or at all) independently verify vendor calculations or designs.

There are no design reviews of vendor design cr TVA does not ccnduct ihdependent design reviews of its work.

QA has not effectively audited the design and construction process Lack of coordination of effects of upcoming (near or long term) design changes with all disciplines and site construction inadequate evaluation of impacts (not under configuration control)

Lack of accountability of TVA personnel and management for not following '

procedures, regulations, etc. and ~for not doing adequate and acceptable job

• Too an ethicmuch to blame do it on QAthe right forfirst quality problems versus emphasizing and demanding time.

struction Put quality into design a~nd'c6n-O

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8-e Commitment (action) system in TVA nonexistent No action party and schedule .

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• I Lack of effective comunications and interface control among organizations with EN DES - Branches, Projects Procurement, etc.

Protective and defensive attitudes of NEB and various Branch / Project groups concerning problems rather than an attitude to admit [ Illegible]

Lack of proper environments and fire protection in equipment storage areas Lack of knowledge (on site and in EN DES) as to status of OCIRs and IRNs ,

Untimely closecut of ECNs Lack of knowledge of status of ECNs or designs affected

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,, UNITED STATES j  ;

n NUCLEAR REGULATORY COMMISSION l WASmNCTON, D. C. 20555 May 16, 1985

...../

Docket Nos: 50-390, 50-391 and 50-438, 50-439 Mr. H. G. Parris Manager of Power Tennessee Valley Authority 500A Chestnut Street, Tower II Chattanooga, Tennessee 37401

Subject:

Letter on TVA Construction Sites Dated May 16, 1985 In Mr. H. Thompson's letter to you dated May 16, 1985, " Concerns Regarding TVA Construction Sites", there was an error of omission in Question 3 of Enclosure- 2. The second request wtlich reads " Provide the analytical bases for establiahing the above procedures which demonstrate that the effect the tolerances have on the stresses on loads in interfacing components and structures will cause these values to exceed their allowable limits." should read as follows:

{'x

,V) " Provide the analytical bases for establishing the above procedures which demonstrate that the effect the tolerances have on the stresses or loads in interfacing components and structures will not cause these values to ,

exceed their allowable limits." '

t We apologize question above.for any incorwenience and ask that you respond to the restated Sincerely, l4:n.h% Y. W ^* .

Elinor G. Adensam, Chief -

Licensing Branch No. 4 Division of Licensing -

cc: See next page I

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APRENDIX VII

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N BOO 7' g# DESCRIPTION OF DAVIS-BESSE LOSS OF est" 8ve 3844 MFW AND AFW UCENSEE EVENT REPORT (LEA) TEXT CONTINU AT)ON *=ovia ow ie sm-mos y aum mies

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c 15 lo lc lc l 314 !6 8l5 0lI]3 - - O I 0. 0 l3 0F 2 l2 mw .~ e ama n Description of Occurrence: Davis-Besse Unit 1 was operating at 90 percent of full power with the No. 1 Main Feedvater Pump, MFP, operating in automatic and the No. 2 MTP in =anual control. This cenfiguration was established to limit the susceptibility of the No. 2 MTP to control proble=s which-had previously occurred. The control proble=s occurred only after a reactor trip and appeared to be-connected to the cutoestic mode of operation. This configuration, therefore, per=itted automatic feedvater control during operation and offered i= proved availability of at least one MFP in the event of a reactor trtp.

5t 0125 hours0.00145 days <br />0.0347 hours <br />2.066799e-4 weeks <br />4.75625e-5 months <br />, the No. 1. MTP tripped en overspeed due to an unrelated control probles. The control Roos operators increased the No. 2 MFP speed, but it did not have adequate capacity, for the existing reactor power. The reactor tripped en high teactor Coolant Syste=, RCS, pressure at 0135:30 hours, tripping the turbine.

Reactor power was at approximately 80 percent of full power at the time of the trip.

Ie=nediately folleving the trip, a spurious Stean and Feedvater Rupture Control Syste=, STFiS, low steam generator level full trip occurred on Channel 2, an STRCS full trip alarn was received, and both main steam isolation valves, MSIVs, closed.

An actual low stea= generator level did not exist at this ti=e. This sperious trip .-

resulted in a partial actuation of the STECS components since only the MSIVs actuated.

{ hen the MSI7s closed, the main steam supply vaa isolated to the MF?s. The No. 2 5P continued to supply feedvater until approximately 0140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> at which time its diacharge pressure was not high enough to supply feedvater to the steam generators.

The level in the steam generators which was being e.aintained at the lov level li=it ssepoi=t (35 inches) began to decrease. SFRCS Actuation Channel No. I then auto =ati-cally initiated on low steam generator level, starting the No.1 Auriliary Feedvater Ptsp, ATP, to feed the No.1 Steam Generator (see Attachment 1 for a diagram of SFECS actuated couponents).

At 0141:08 hours, a Control Roon eperator atte=pted to manually initiate the SFRCS, hevever, he incorrectly actuated the SFRCS on lov stea= pressure instead of the desired lov stea generator level. Therefore, each SFRCS actuatien channel sensed thte its respective'steas generater was depressurized. STRCS Actuation Channel No. I then atte=pted to align ATP Ho I to feed Steam Generator No. 2. STRCS Actuation Channel No. 2 atte=pted to elign ATP No.'2sto.faed Steas Generator No. 1.

Both ach:ation channels closed their respective Auxiliary Feedvater Containment Isolation Valves (AF599,' AF606), which prevented any auxiliary feedvater flow from ~

renching the stea= generators. At 0141:31 hours. ATP Fo. I tripped on overspeed.

At 0141:44 hours, AFF No. 2 tripped on overspeed.

At 0142:00 hdtirs, an operator recognized the santal initiation error and reset the low pressure SFRCS buttons, and pushed the lov staan generator level SFRCS =anual c'etuation buttons. Since both SFRCS actuation channels were already tripped on lov l

stes= generator level, the SFRCS automatically began to realign the ATFs when the I

low preesdre buttons were reset. However, the Auxiliary Feedvater Contaissent Golation7alves(AF599,AT608)didnotautomaticallyopen. The operators attempted open these valves from the Control Roon by operating their control switches and by reinitializing the SFRCS. These atte:: pts failed to open the valves. Equipment 1

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UCENSEE EVENT REPORT (LERJ TEXT CONTINUATION *=ovie om io vw-a*

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l Isvis-Besse' Unit 1 .

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010 Operators were sent to open these valves locally, and when the valves were moved off their closed seats utilicing the manual handvheels, the motor operator responded and fully opened the valves. During this period, attempts were also being made to rastart the AT?s and preparaticus were underway to start the motor operated Startup Teedvater Pump.

The RCS average temperature was increasing due to the lack of pri=ary to secondary brat transfer. RCS pressure was increasing dce to th,e decreasing density of the RCS vater and increasing pressuri:er level. RC! pressure increased to the Power Operated Relief Valve, PORY, setpoint (2425 psig)- ne PORV cycled a total of three times, relieving pressurizer pressure to the Quench' Tank. Following the third opesing, the PCR7 failed to reclose at the proper RCS pressure. The Control Room operator observed the primary plant conditions and closed the block valve en the PORV. RCS pressure was at approxicately 2075 psig when the block valve closed. The Quench Tank contained the discharges from the POR7.

At approx 1=ately 0131 hours0.00152 days <br />0.0364 hours <br />2.166005e-4 weeks <br />4.98455e-5 months <br />, the operators placed the Startup Feedvater Pump in op2 ration to supply the steam generators. Steam Generator No. 1 pressure had d2 creased to appreximately 750 psig. Steam Generator No. I repressuri:ed to approxi-mately 900 psig from the Startup Teedvater Pump. Steam Generator No. 2 had decreased

) 920 psig. At 0152 hours0.00176 days <br />0.0422 hours <br />2.513228e-4 weeks <br />5.7836e-5 months <br />, the No. 2 AFP was returned to service by the operators Leally. Maxi =u: RCF. temperature had reached approri=ately 592 degrees Fahrenheit.

L,c 0155 hours0.00179 days <br />0.0431 hours <br />2.562831e-4 weeks <br />5.89775e-5 months <br />, the No.1 ATP was returned to service locally by *he operators.

Control of the ATP turbines was maintained locally by an operator at the turbine trip throttle valve. At 0158 hours0.00183 days <br />0.0439 hours <br />2.612434e-4 weeks <br />6.0119e-5 months <br />, RCS average temperature was restored to the normal post trip te=perature. The cooldown of the RCS lowered RCS pressure to a mini =u cf approxis.ately 1720 psig. Operators manually started the No. 1 Eigh Prcssure Injectien. EPI, Pu=p in the piggyback mode (Decay Heat Pump No I supplying tha suction to the EPI Pu=p No.1) in precautionary anticipation of the rapid cooldown.

Only a slight amount of water (less than 50 gallons) needed to be injected.

' Sevaral other equipment calfunecions occurred which did not affect the physical plant response.

One source range nuclear instrumentation, NI, chtsnel was inoperable prior to the trip. The re aining source range NI channel failed to indicate properly whin it was autenatically energized after the trip. The display units far.the Safety Parameter Display System, SPDS, were inoperable in the Control Roem at the i

time of the trip. At 0158:40 hours, the suction of the No. 1 AIP automatically transferred.fre= the Condensate Storage Tank, CST, to the Service Water Syste=. The I cperator ma=ually realigned the pump suction back to the CST. No significant amount of service water was added to the stean generator during the recovery from the transient. It was noticed that M pneumatic operator on one main turbine bypass valve was di= aged, preventing the valve from being opened. This did not affect the post transient respense of the plant.

Additions? details of the plant transient and corrective actions vill be provided in

• 3 restart report response to the Region III Confirre.atory Action 1.etter (85-06),

ach=ent 2 provides a chronological listing of the event. This report is being mitted in compliance with paragraph 50.73(a)(2)(i), 50.73(a)(2)(iv), 50.73(a)(2)(v),

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1 10 0l5 0F 2 l2 rm u- . :--% an.wm 50.73(a)(2)(vi), and 50.73(a)(2)(vii). This report also satisfies the reporting  ;

~requireseats for a Tnergency Core Cooling System Actuation Special Report, Section 6.9.2(a) of Technical Specifications. This was the fourth high pressute injection l cetuation cycle to date.

Designation of Anarent Cause of Occurrence: This transient was initiated when the No. 1 MFP developed control problems and tripped on overspeed. The plant tripped on high RCS pressure due to inadequate feedvater being supplied from the No. 2 MT?

during the plant runback. D e cause of the MFP overspeed tripping was defernined to be due to a bad speed su= nation and valve lif t reference circuit )oard card in the MFP control. A frequency to voltage converter chip had failed. -ne board is being ret'urned te General Electric for further analysis on the root cause of the failure.

He root cause of the MSIV closure has not yet been deteruined. It is pr'esently believed that the MSITs properly responded to a momentary lov level SFRCS trtp.

Further investigations vill follow once an action plan is completed.

The cause of the SFRCS spuricus trip on lov stea= generator level has not yet been positively determined. Troubleshooting vill begin in accordance with the action plan. Hoverer,*'it is presently believed that the stess generator level sensing I

-hannels are sensing an extre=ely rapid secondary side pressure transient that hecurs in the secan generator following the turbine stop valve closure on a turbine trip. These level transmitters share a cc=non set of sensing lines with transmitters which were replaced during the 1984 Refueling Outage. Prior to the 1984 Refueling Outage, Ea11ey BY level transmitters were installed which have nov been replaced by Rosesent Model 1153. Since these Rose =ont transmitters have no significant displace-nest required for operation, while the Bailey BYs required a volume displacement to operate the bellows, it is postulated that the responsiveness of the sensing line and trans=1tter arrangement has been greatly increased by this change. This increased responsiveness allowed the STRCS to sense the rapid secondary side pressure transients which previously were undetected. yurther analysis of this condition is underway.

The cause of the incorrect manual SFRCS initiation was personnel error attributt.d to a poor switch layout. n ese STICS manual initiation pushbuttons had been identified in the Detailed Control Roen Design Review as one of the principal items needing hinr.an engineering improvements. nere are two adjacent vertical columns of buttons with five buttons in each column (see Attachment 3 for arrangement details). Each -

column represents one STECS actuation channel. To manually initiate both channels of the_SFRCS for steam generator low level, the operator shculd have depressed the fourth button from the top in each column; instead, the two top buttons were depressed.

A design change had been developed prior to this event *to improve the switch layout and vil.1 be implemented during this outage.

The cause of the AFPs tripping on overspeed after initiation has not yet been positively determined. Water flashing through the nozzles of the AFP turbines is hought to be i contributor. ne governor was inspected on both AFP turbines, and l h. contributing factors to the overspeed were seen. Further investigations and

-csting are planned.

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UCENSEE EVENT REPORT (LER) TEXT CONTINUATION an. cam ens ic mm imau emen enese see ' DUGLIT EIU.ed h G' gg t gyggg e og .aad 13 f

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3 6 3[5 -0 l1l3 0;0 0;6 or 2l2 m o - . - o mu e Th2 cause of the Auxiliary Teedvater Valves (AF599 and AT608) not opening by the actor operstor was deteruined to be a cothination of a high differential pressure cnd an improperly set torque switch bypass li=it switch. V,ith this torque svitch bypasa limit switch i= properly set, the motor operator was allowed to torque out during the opening stroke. These valves are open during normal operacion and were closed by the incorrect manual initiation of the STRCSc If the ATPs had been op3 rating at the time the valves attempted to open the diff erential pressure across the valves vould have been significantly lover, and the valves should have opened to allow the auxiliary feedvater flow to occur. These valves were stroked follovips the transient and ability of the valves to open (without significant differential- -

prnssure) was verified. Recent testing also verified that the valve operators torque out unde'r high dif ferential pressure with the improperly set torque switch bypass li=1t switch. Further investigations are in progress. .

De cause of the control proble=s with the AT?s af ter the overspeed was reset is presently attributed to the difficulty in opening the trip throttle valves. No mechanical deficiencies were found while investigating the resetting of the overspeed trip device 711nkage. Turther investigations are in progress.

f9e cause cf the POE7 not properly reseating has not yet been positively identified.

erator observations at the time of the transient indicate that the electronic C urrels sig=al was calling for the valve to reclose. A visual inspection and disasse=bly of the PORY failed to identify the cause. Further investigations are in progress.

The two. independent SPDS display units were inoperable due to separate but similar fcilures in the data transmission syste= between the Control Roon terminals and their respective processors. The failures are of an inter =ittent nature and the exact cause is still under investigation.

The ause of the scurce range NIs inoperability has not been positively identified.

Tha failure of the source range NIs has been a repetitive proble= at Davis-Basse with repeated investigations failing to dete mine the root cause. Since 1977, the boron trifluoride detectors. prea=p. ud cable in Containment have been replaced, s Qqng with the modules in the Reactor itotection System and a reworking of the grounding on the preaty and count rate a 91fier 1 module connections.

• No positive effect on the total el%stion of the sgking nor the erreneous/ elevated count rste has occurred frem these corrective ae.iens. Further review is being perfor=ed on the-possibility of ground loops, induced current or voltage from adjacent cables, or inter =1ttent probless with the count rate amplifier module.

~

The cause of the Turbine Eypass Valve 2-2 da= age has not been identified. The valve v:s disasse= bled and the actuator stem extension piece was found bent. four parts were missing, and the valve internals were found loose. Several valve parts were rhipped to'the vendor for further analysis. Further review of the turbine bypass

[ ]ve failure is underway.

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• 2 12 ex . - . - - wa awem The cause of the inadvertent Ar? No. I suction supply transfer from the condensate Storage T:nk to Service Water has not yet been determined. Testing and other investigations are currently being perfomed.

The cause of MS-106 apparently cycling in about one third of the expected stroked time is still under investigation.

, Analysis of Occurrence: This event involved a temporary loss of feedvater to the stea: generators. This event was bounded by the analyses previously performed (see Toledo Idisen subnittals to NRC Serial No. 506 dated May 22, 1979, and Serial No.

517. dated June 15,1979), which analyzed a los4 of all feedvater for 30 ninutes following a reactor trip. These analyses showed that as long as either:

~

1) Auxiliary Feedvater is restored within 30 =inutes of the loss of main feedvater.

OR

2) Vichin ?O minutes, at least one makeup pu=p and the PORY are available for primary cooling .(feed and bleed) and the Startup Feedvater Pu=p is avail-

/ ) able to supply a steam generator, U

fuel cladding te=peratures would re=ain within a few degrees of saterated fluid temperature and no cladding rupture or cetal vater reaction would oc:ur.

Operator interviews indicated that the shif t was fully aware of the core status and v2re prepared to implement the " feed and bleed" core cooling nethod if the auxiliary fctdvater vas not restored. The Startup Feedvater Pump was available throughout the svent and in fact was placed in service within ten cinutes of the tripping of the AFPs. Auriling feedvater was restored within 12 minutes of the loss of feedvater.

Th se response times and equip =ent availability are well within the loss of feedvater analyses.

At no time during the event was the required subcooled margin (20 degrees Fahren-heit7 lost. The nacar coolant pu=ps continued to operate throughout the event.

The pri=ary code erfety valves were not challenged and at no time during the event did the ECS pressure or te=perature exceed the allovable values. The maximum -

temperature reached was below the normal operating temperature for the het leg te=perature. There is no indication of any fuel cladding degradation based on the recetor coolant radioche=1 sty analysis.

An anclysis has been performed by Babcock & Wilcor to determine if the trassient cdversely affected the stem:n generators. Conditions and conponents specifically analyzed imelude: (1) Msin Teedvater Nottles, (2) Auxiliary Teedvater Nozzles. (3)

Steas GeneYator Tubes (4) Tube to Shell Delta T's, and (5) Lover Tubesbeets results show that the transient had no adverse structural effect on the steam merators.

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UCENSEE EVENT REPORT Mm TEXT CONTINUATION ==eno ow ==a =*as y trnags gatgg mama m oocasy saJanus a gas asynsess e,i past L3 l

SeDavis,B, esse Unit ! n .. - M *;;'. FC:;

, , _ y Corrective Action: he failed circuit board vill be replaced in the No. 1 MFP. As a precautionary measure, the No. 2 MFP speed control circuit vill also be inspected for a similar failure. .

n2 corrective actics on the MSIV closure has not yet bean determined since trouble- j shooting has not yet begun. j The corrective actims for the SFRCS spurious trip on low steam generator level have not yet been detemined since troubleshooting has not yet begun, n e proper method of ca'nual ' actuation of the SFRCS buttons vill be reviewed with all licensed operators.

He svitch layoot is being asdified to add additional denarkation of the actuation buttons, and to add actuation guards over the switches (see Attachment 3).

ne corrective actions to be taken to prevent the ATP trip on overspeed have not yet b u determined.

ne torque svitch bypass limit svitch will be reset on the Auriliary Teedvater Velves AFS H and AF608. Maintenance personnel vill receive additional instruction, and the procedure for setting the motor operator valve li=it svitches vill receive Sditional clarification.

is-Besse Other nuclear safety related motor operated valves at vill b'eevaluated.

The corrective actiou to correct the control problems with the AT?s after the overspeed was reset have not been identified.

Corrective actions to be taken on the POR7 have not yet been identified.

Corrective actions for the repair of the data transmission systems affecting the S?DS Control Roon displays have not yet been identified, ne corrective actions for repair of the source range N1s have not yet been deter =ined, ne pneu=a:ic actuator for Turbine Bypass Valve 2-2 will be replaced. Additional corrective actions may be necessary af ter furth6r investigation to determine the rcot cause of the f ailed valve actuatore J-A tabulation of the causes and corrective actions determined to date is su:marized in Attachment 4.

,, Corrective action details for the No.1 AFP suction supply transfer from the conden-sate Storage Tank to Service Water has not yet been identified.

Corrective action details for MS-106 have not yet been identified. Further investi-gation is in progress.

lure Data: h is is the first occurrence at Davis-Eesse of a loss of both main d auxiliary feedvater.

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_O!1I3 010 0l9 0F 2l2 rocv u .  :.= w am.wm B is is the first failure that has occurred at Davis-Besse on the NI? turbine electronic controllers which has caused an overspeed trippin5 of the pumps. A new electronic control systen for main feedvater pumps was installed during the 1981.

Refueling Outage.

Spurious closures of the ESIVs have occurred previously at Davis-Besse before ti=e d212ys vere added to the steam to feedvater pressure differential trip circuitry.

An SFRCS spurious half trip on lov stean generator level has occurred on two previ-oss trips since the 198'4 -Refueling Outage. Spurious trips on lov stea= generator level have not occurred prior'to the 1984 Refueling Outage.

Incorrect manual initiation of the STRCS has not previously occurred at Davis-Pesse, ne AT?s tripping on overspeed af ter initiation has not previously occurred at Davis-Besse.

The Auxiliary Feedvater Talves AI599 and AF608 are nem. ally open. One previous occurrence of one cf these valves not opening with high differential pressure occurred after the March 2, 1984 reactor trip.

I (aoperatersdonotnorallyattempt to control the ATP turbines locally. Proble=a with centrolling these pumps do not appear to have been repetitive, however, some proble=s have been experienced previously with proper resetting of the trip throttle valve.

ne POTW has not been challenged since the pressure setpoint was reised in 1979.

Prior to 1979, several deficiencies were noted in the valve operation. In September 1977, the valve stuck in the open position, causing an overpressurization of the quench tank.

Th2 diversity of the SPDS display sources (Ramtek and Chromatics display devices) has nor elly allowed at least one SPDS display to remain operable. The failure rate of these units is higher than is acceptable. Efforts are underway to increase the systet reliability. s, y Th2 failures of the source range NIs have been a repetitive occurrence at Davis-Eesse ,

cven though exhaustive evaluations and corrective actions have been taken.

Th2 damaged pneu=ntic operator on the turbine bypass valve has not previously occurred at DavisJEesse.

n are have been several cases where the AFP suction inadvertently transferred from

~

the Condensate Storage Tank to the Service Water supply.

1

.coort No: NP-33-SS DVZ No(s): 85-088 m ,2.. m.

-.+.m..ai.~- . . . . . r .~k.,.

. a_.., , . . , _ _ , . , _ _ , , _ _ ._, _,

, APPENDIX VIII I, , RECENT SIGNIFICANT EVENTS i

1

~

AGENDA FOR ACRS MEETING

{ JULY 11, 1985 4:30 P.M.

l ROOM 1046, H STREET RECENT SIGNIFICANT EVENTS l

i l

Facility / Title Event Date Presenter / Phone M

1. Hatch 1 Stuck May 15, 1985 G. Rivenbark, NRR Q Open SRV. (492-7136) j

2. Oyster Creek Scram Discharge Volume .

June 12, 1985 D. Powell, IE (492-8373) 7 i

Isolation Valves Failure .

i-

3. Rancho Seco - RCS High June 23, 1985 H. Wong, IE Point Vent Leak- (492-4536) f t

l 4: Sequoyah Unit 2 Reactor Trip due to May 22, 1985 E. Weiss, IE (492-9005)

/[

Improper Use of Test i Instrument j 5. Davis-Besse loss of June 9, 1985 E. Jordan, IE /1

j. ' MFW and AFW.* (492-4848) 4 1 A. DeAgazio, NRR t l (492-8945)  !

l l

i

.*The I.I.T. will discuss this event in detail at a later meeting.

ee 1

4 1 ,

l 1

L-

L- -

._.a.. A;_E _ _ ,. .. . ..

HATCH UNIT 1 - STUCK OPEN SAFETY RELIEF VALVE OF MAY 15, 1985

-(G, RIVENBARK)

A SYSTEMS INTERACTION EVENT ,_,

UNIT 1 OPERATING AT FULL POWER CONTROL' ROOM EMERGENCY VENTILATION SYSTEM CHARC0AL FILTER DELUGE VALVE ACTUATED WATER LEAKED THROUGH VENTILATION DUCTS INTO A HATCH UNIT I ANALOG TRANSMITTER TRIP SYSTEM (ATTS) INSTRUMENT PANEL CAUSING SRV TO OPEN REACTOR MANUALLY SCRAMMED FEEDWATER PUMP REC 0 VERS REACTOR WATER LEVEL SRV CLOSED . WITHOUT OPERATOR ACTION CAUSE LOSS OF INSTRUMENT WATER SUPPLY CAUSING DELUGE VALVE TO OPEN TOGETHER WITH PLUGGED DRAINS-NOT SURE HOW WATER CAUSED THE SRV TO OPEN ACTION REPLACED ATTS POWER SUPPLY, CLEANED PLUGGED DRAINS AND INSPECTED DP.AINS IN REDUNDANT FILTER UNIT

..~* LICENSEE PROPOSES TO ADD CLEAN 0UT CHECK PROCEDURES FOR PLENUMS AND THEIR DRAINS

~

ORAB WILL DEVELOP TIA TO COORDINATE:

IE NOTICE

.,FURTHER INVESTIGATIVE EFFORTS GENERIC REVIEW

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0YSTER CREEK - UNCONTROLLED LEAKAGE OF REACTOR COOLANT OUTSIDE CONTAINMENT JUNE 12, 1985 (D, POWELL, IE) i f  :

I WITH REACTOR AT 997. POWER, FAILURE OF THE ELECTRIC PRESSURE i

d REGULATOR CAUSED A TURBINE BYPASS VALVE TO OPEN RESULTING IN '

i A REACTOR PRESSURE DECREASE, FOLLOWED By MSIV CLOSURE AND .  !

REACTOR SCRAM.

l SCRAM DISCHARGE VOLUME DRAIN VALVES FAILED TO FULLY SHUT /

SEAT CAUSING REACTOR COOLANT TO BE DISCHARGED.TO THE  !

REACTOR BUILDING DRAIN TANK.  :.

[

j -

RELEASE OF' STEAM FROM FLOOR DRAINS AN6 PAINT BLISTERING ON f i t l HOT PIPE CAUSES PORTION OF REACTOR BUILDING DELUGE SYSTEM

t l TO ACTIVATE j l -

SCPAM S'IGNAL NOT RESET F'OR 36 MIN ALLOWING CONTINUOUS .

REACTOR COOLANT FLOW TO THE DRAIN TANK. CAUSE WAS-600 PSI l

INTERLOCK ON MSIV CLOSURE / LOSS OF CONDENSER VACUUM.

- l SAFETY SIGNIFICANCE - (1) LOCK OUTSIDE CONTAINMENT, l

! t (2) POTENTIAL EQUIPMENT MALFUNCTION DUE TO FIRE DELUGE -

[

l i SYSTEM, (3) EXCESSIVE CRD SEAL,TEMPERATUP.ES. f CAUSE-VA,LVE SPRING ON VALVE V15-134 (VELAN) VALVE UNDERSIZED

[

-VALVE V15-121 (VALTAK) STROKE DISTANCE INSUFFICIENT [

, 5 I

.TO TIGHTLY SEAT THE VALVE. (1/8" OPENING) l

[j;

-!MPROPER POST-INSTALLATION TESTING OF VALVES

' CORRECTIVE ACTIONS - REPLACED 400 LB SPRING WITH 1100LB [ l SPRING, ADJUSTED VALVE STROKE ~ DISTANCE CHECKED CRD SEALS FOR DAMAGE, CHECKED '

0 EQUIPMENT, NO DAMAGE FOUND.

l NRC FOLLOWUP ACTION - IE NOTICE IN PREPARATION. l

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RANCHO SEC0 - RCS HIGH POINT VENT LEAK JUNE 23,1985 (H. WONG. IE)

.:wanown PLANT IN HOT STAND 0Y-RESTARTING FROM A REFUELING OUTAGE 20 GPM NON-ISOLATABLE PRIMARY COOLANT LEAK ON HIGH POINT i

VENT ON B STEAM GENERATOR HOT LEG i -

TMI MODIFICATION INSTALLED 1983 REFUEll5G OUTAGE i -

120* THRU WALL LEAK AT WELD CAUSE APPEARS TO BE MISSING SUPPORTS AND FATIGUE FAILURE i l!CENSEE ACTIONS: .

! STRESS ANALYSIS TO IDENTIFY OVERSTRESSED AREAS (BOTH HOT -

j LEG VENTS)

REPAIR SYSTEMS

, INSTALL SUPPORTS I

WALKDOWN TO INSPECT AND EVALUATE OTHER SYSTEMS l

REGION V, IE TEAM PARTICIPATING IN WALKDOWN.

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t AND AUXILIARY FEEDWATER [,

JUNE 9, 1985 (A. DEAGAZIO, NRR) j,!

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i -

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, operators close block valve. -

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I I

Operators restart auxiliary feedwater pumps and restore  ! l normal post-trip conditions.

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No indication that subcooling margin was lost or that

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SSINS No.: 6835 IN 85-50 UNITED STATES NUCLEAR REGULATORY COMMISSION L/ OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 o July 8, 1985 IE If60RMATION NOTICE NO. 85-50: COMPLETE LOSS OF MAIN AND AUXILIARY FEEDWATER AT A PWR DESIGNED BY BABC0CK & WILCOX r I

P ADDRESSEES:

. All nuclear power facilities holding an operating license (OL):or construction permit (CP).

Purpose:

This information notice is being provided to inform licensees of a significant reactor operating e. vent involving the loss of main and auxiliary feedwhter at a pressurized water reactor. Information in this n6tice is preliminary and was obtained from the special NRC fact finding team wnich is investigating the event. A comph te report of findings will form the basis for further communi-cations or actions related to this event. The NRC expects that recipients

~

will review this notice for applicability to their facilities. Suggestions p contained in this notice do not constitute NRC requirements; therefore, no

-(-)specificactionorwrittenresponseisrequired.

Description of Circumstances:

On June 9, 1985, the Davis-Besse plant was operating at 90% power with Main !7 Feedwater Pump 2 in manual control because. problems in automatic had been i exper.ienced. A control problem with Main Feedwater Pump 1 occurred, and it E tripped on overspeed. Reactor runback at 50% per minute toward 55% power was automatically initiated. Nevertheless, 30 seconds,later, the reactor tripped ~[f at 80% power on high pressure in the reactor coolant system. r One second after reactor / turbine trip, one channel of the Steam and Feedwater [

R,upture Control System (SFRCS) was automatically initiated due to a spurious signal indicating low water level in Steam Generator 2. Both Main Steam ,

Isolation Valves (MSIVs) closed. Three seconds after the actuation, the SFRCS '

automatically reset. Closing of the MSIVs isolated the turbine of the operating- .

main feedwater pump from its source of steam. The pump continued to supply. '

'feedwater to the steam generators for a few' minutes as it coasted down.

Four and.a half minutes a.fter reactor trip, water level in the steam generators began'to fall from the normal post-trip level which is 35 inches. After MSIV [

closure, steam release to atmosphere continued to remove decay heat. 'One minute b later, Channel 1 of SFRCS actuated when the water level in Steam Generator 1 P actually reactied the SFRCS setpoint at 27 inches (See Figure 1). SFRCS started Auxiliary Feedwater Pump 1 and initiated alignment of it to Steam Generator li

) -

- - c i f

• IN 85-50 '

July 8, 1985 Page 2.of 4

\

(d  :

Within seconds after automatic initiation of Channel 1 of SFRCS, the operater actuated both channels of SFRCS; however, he inadvertently actuated both SFRCS channels on low steam pressure instead cf low water level. When an SFRCS ,

channel is actuated on low steam pressure, a rupture of the steam line asso'ciated with that channel is presumed to have occurred. The SFRCS closes the steam  ;

generator isolation valves, including a valve in the auxiliary feedwater line, l and aligns the auxiliary feedwater pump to the other steam generator. Because  ;

4 both channeis had been manually actuated on low steam pressure, both steam i 1 generators were isolated from both auxiliary feedwater pumps. Five seconds  : '

after the operator's inadvertent actuation of both channels on low steam pressure, SFRCS Channel 2 received an actual low water level actuation signal.

Because low pressure initiation takes precedence, alignment of the auxiliary .

feedwater pumps remained unchanged. At six minutes into the event as both i auxiliary feedwater pumps were accelerating, they tripped on overspeed.

In summary, all main feedwater had been lost, both steam generators were isolated ,

from feedwater and were boiling dry, all auxiliary feedwater pumps were tripped, pressure of the reactor coolant system was rising, and reactor coolant system .

temperature was increasing.  ;

Within one minute after the operator's inadvertent actuation of the SFRCS on i low steam presTure, the mistake had been recognized and the SFRCS had been reset. If equipment had performed in accordance with system design recuirements, the operator's error might not have had a significant impact on the event.

The auxiliary feedwater isolation valves should have reopened automatically,

,_ , but the valves did not reopen. The operator then tried to reopen the valves e from the main control panel, but the valves would not reopen. Operators were dispatched to locally start the auxiliary feedwater pumps, open the auxiliary [

feedwater isolation valves, start the nonsafety-related motor-driven startup feedwater pump, and valve it to the system. L '

}

Pres.sure and temperature in the reactor coolant system continued to rise [

l because there was not sufficient water in the steam generators to provide an '[  ;

adequate heat sink. At 13 minutes after reactor trip, reactor coolant system j pressure reached 2425 psig, and the Pilot Operated Relief Valve (PORV) opened three times to limit the pressure rise. Ort the third lift, the valve remained i  ;

i open. The operator closed the PORV block valve and reopened it two minutes [ ^

later after the.PORV had clused. . I i Approximately 16 to 18 minutes after reactor trip, the operators had the star' tup and auxiliary feedwater pumps running and the valves aligned. Water levels were  ;

beginning to rise in the steam generators. Reactor coolant temperature reached i l a maximum of 594* F and then started to decrease to normal. Refilling of the,

~

j.  !

steam generators caused the reactor coolant system to fall to 1716 psig and  ;

i about 540 F before returning to normal (See Figure 2). [L ,

4

, i

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At 30 minutes after reactcr trip, plant conditions were essentially stable. '

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IN 85-50 -

July 8, 1985 Page 3 of 4

~

Discussion:

For several minutes after reactor trip, the cteam generators were unable to cool.the reactor coolant system adequately.

The first problea contributing t'o this event was the loss of all main feedwater due to closure of the MSIVs. The licensee's hypothesis, based on information from Babcock & Wilcox, is that turbine trip causeo a pressure transient upstream  ;

from the turbine stop valves which caused the outputs of the redundant steam i generator level instrumentation channels to oscillate'widely for several ,

seconds. The licensee beTieves that this caused a spurious low level actuation i.

of SFRCS which closed the MSIVs.

\

Three additional-problems contributed to this' event by affecting the availability l of both trains of the auxiliary feedwater system. The first occurred when the '

reactor operator pressed the wrong SFRCS buttons. The second occurred when Li both auxiliary feedwater pum;is tripped nn overspeed. The third occurred when '

both auxiliary feedwater isolation valves did not reopen when SFRCS was reset.

r Control buttons for the SFRCS are arranged in two vertical columns. E'ch a

column of buttons controls one SFRCS channel. The operator should have pressed the fourth button from the top in each column. Instead, the operator pressed I the top buttoni causing isolation of both steam generators.

Both auxiliary feedwater pumps are driven by Terry turbines which tripped on overspeed early in the event. When this occurred, steam was being supplied to +

~j the turbines via crossover lines, which are longer than the normal supply lines i and include long horizontal runs. The licensee believes that significant condensation may have occurred in the crossover lines. Further, the licensee believes that the quality of steam arriving at the turbines may have been ,

affected significantly by the configuration.of the crossover lines and may have  ;

caused the overspeed trips. -

t The auxiliary feedwater system isolation valves have Limitorque motor operators.

The motor operators have torque switches which prevent overtorquing of the valves by disconnecting power to the motors.. When the valves are being opened, +

additional torque is required to overcome friction while the gates are being unseated and while a significant pressure differential may exist across the gates. During the initial part of the opening stroke, the torque switch in the t i motor operator is bypassed by a bypass switch so that full motor torque is' developed if necessary. The licensee believes that these bypass switches went ,

[:

off bypass too early. The valves did not reopen until an operator unseated ,  !

th.em by hand. '

?

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. . . _ . . . .. _ u u_ . . . _ _ . . ~ __ _..;.. ._ _ _ _ _ . __ m . _- _ ___ .

4 IN 85-50 July 8, 1985 l Page 4 of 4 No specific action or written response is required by this information notice.

If you have any questions about this matter, please contact the Regional  ?

Administrator of the appropriate NRC regional ~ office or this office. '

s A -

oward Jordan, Director i Divisio of Emergency Preparedness i and Engineering Response '

i Office of Inspection and Enforcement t i Technical

Contact:

R. W. Woodruff, IE 2 I

t (301) 492-4507 .

i Attachments: -

1

1. Figure 1 - Steam Generator 1 Level and Pressure ,

2. Figure 2 - RCS Temperature and Pressure

3. List of Recently Issued IE Information Notices'. . '

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• Attachment *3 IN 85-50 July 8, 1985 LIST OF RECENTLY ISSUED Q IE INFORMATION NOTICES Information Date of '

Notice No. Subject Issue Issued to 85-49 Relay Calibration Problem 7/1/85 All power reactor facilities holding an OL or CP 85-48 Respirator Users Notice: 6/19/85- All power reactor' Defective Self-Contained facilities holding Breathing Apparatus Air. an,0L or CP, research, Cylinders and test reactor, fuel cycle and Priority 1 material Ticensees 85-47 Potential Effect Of Line- 6/18/85 All power. reactor Induced Vibration On Certain t. facilitie's holding Target Rock Solenoid-Operated .an OL or CP

_ Valves 85-46 Clarification Of Several 6/10/85 All power reactor

' Aspects Of Removable Radio- facilities holding

. active Surface Contamination an'OL

~ Limits F,or Transport Packages 85-45 Potential Seismic Interaction 6/6/85 All power reactor Involving The Movable In-Core facilities holding Flux Mapping System Used In an OL or CP Westinghouse Designed Plants 85-44 Emergency Communication 5/30/85 All power reactor [

System Monthly Test .

facilities holding an OL 85-43 Radiography Events At Power 5/30/85 All power reactor Reactors facilities-holding ,

an OL or CP 85-42 Loose Phosphor In Panasonic 5/29/85 All power reactor 800 Series Badge Thermo- facilities holding luminescent Dosimeter (TLD) an OL or CP .

Elements '

OL = Operating License t

CP = Constructi.on Permit

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(! APPENDIX IX ~

l5 3I Fd LD RANCHO SECO ISOMETRIC DIAGRAM RCS I; HIGH POINT VENT C 4 .

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APPENDIX X GE PRESENTATION ON GESSAR-II

.{f- ~

CONTAINMENT CAPABILITY e

'{ \

i GESSAR II CONTAINMENT CAPABILITY t

f A PRESENTATION TO THE ADVISORY .

COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D.C.

[

i i

i e

i I

GENERAL ELECTRIC COMPANY  :

i i

JULY 12, 1985  !  !

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. . . . s .. .m. 2 . . .u m m - ..._: . . . .

U i MK III CONTAINMENT DESIGN ,

COMPARISON {

ULT.

DES. PRESS CONT. CONC.  !

. CONT. PRESS CAPAB. DOME DIA FILL PLANT TYPE (PSIG) (PSIG) CONFIG. (FT) a BASE

t i

GESSAR II STEEL + 15 85 TORI- 120 YES (238) CONCRETE SPHERICAL SHIELD BLDG. q o

GRAND GULF LINED 15 67 HEMI- 123 N/A b l (251) REINF. SPHERICAL k; '

CONCRETE PERRY STEEL + 15 94 TORI- 120 YES (238) CONCRETE SPHERICAL SHIELD BLDG, RIVER BEND STEEL + 15 90 TORI- 120 YES (218) CONCRETE SPHERICAL

SHIELD BLDG.

.. CLINTON LINED 15 95 HEMI- 124 N/A

! (218) REINF. SPHERICAL i

CONCRETE

- RSV15 JULY 1985

POTENTIAL FOR UNSTABLE PROPAGATION OF AN UNDETECTED FLAW o MAXIMUM STRESS IN THE CONTAINMENT FOR ULTIMATE PRESSURE CAPABILITY (85 PSIG) LOADINGS CONSIDERED.

o LOWER BOUND FRACTURE TOUGHNESS PROPERTIES CONSIDERED.

o PLATE MATERIAL o WELDMENT o HEAT AFFECTED ZONE o BASED ON A CONSERVATIVE FRACTURE ANALYSIS, A POTENTIAL CRACK f '

') 0F UP TO 1/2 INCH DEEP.(>25% OF WALL THICKNESS) AND 3 INCHES LONG CAN BE TOLERATED WITHOUT PROPAGATION TO FAILURE.

o WELDING PROCEDURES LIMIT FLAWS T0 s10% OF WALL THICKNESS.

o UNLIKELY THAT ANY WELD DEFECT SIGNIFICANTLY CVER 2% OF WALL THICKNESS AT WELD JOINTS WILL ESCAPE FULL RT OF WELDS.

UNDETECTED FLAWS WILL NOT AFFECT CALCULATED ULTIMATE PRESSURE CAPABILITY OF THE CONTAINMENT f

(

l RSV15 - 2

/db 7'd fd JULY 1985

.l.

~

j CESSAR II Containment Structural Analysis 1

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, Mark III contairunent au11 dings i of Standard Plant {

RSV15 - 3

{

p/d[ _

JULY 1985 i

t s

Q --

. DOMINANT CONTAINMENT FAILURE MODES o LOADING AND FAILURE MODE

1. H2 DETONATION IN CONTAINMENT 1. LOCAL: CONTAINMENT FAILURE o SHOCK WAVE AB0VE WATER LINE o INTERNAL PRESSURE ON GLOBAL: A) CONTAINMENT FAILURE CONTAINMENT ABOVE WATER LINE o EXTERNAL PRESSURE ON B) DRYWELL CEILING DRYWELL FAILURE

2. H2 COMBUSTION IN CONTAINMENT 2. A) CONTAINMENT FAILURE AB0VE o INTERNAL PRESSURE ON WATER LINE CONTAINMENT B) NO DRYWELL FAILURE SINCE o SMALL EXTERNAL PRESSURE NO SIGNIFICANT LOADS ON DRYWELL (s5 PSIG)

3. H2 SLOW BURNING 3. A) NO CONTAINMENT FAILURE o INTERNAL PRESSURE ON SINCE-PRESSURE IS LOW CONTAINMENT B) NO DRYWELL PRESSURE SINCE o SMALL EXTERNAL PRESSURE N0 SIGNIFICANT LOADS ON DRYWELL (s5 PSIG)

4. STEAM AND/0R NON-COMBUSTIBLE 4. A) CONTAINMENT FAILURE AB0VE GAS OVERPRESSURIZATION WATER LINE AT CONTAINMENT o SMALL INTERNAL PRESSURE ULTIMATE PRESSURE ON DRYWELL $5 PSIG CAPABILITY o INTERNAL PRESSURE ON B) NO DRYWELL FAILURE SINCE CONTAINMENT NO SIGNIFICANT LOADS O

RSV15 - 4 7 . . . - . . - . . . .

7 d'8A JULY 1985

= _ = . - a a.. -. :. a- - ..-. - - . . _ _ . x. .. .

i i

i o VALIDATION OF FAILURE MODES .

i i

1 o BY ANALYSIS I l b r

i.

o LOAD TYPES AND APPLICATION I o CONTAINMENT AND DRYWELL STRUCTURAL CONFIGURATION o CALCULATED STRESSES '

j o HIGHEST STRESSED POINTS ASSUMED TO FAIL FIRST 1 l l q o NO FAILURES ASSUMED WHERE LOADS ARE SIGNIFICANTLY l i LESS THAN THE DESIGN LOADS.

1

i i  !

i '

! i I

i o DRYWELL STRUCTURE, HEAD, PERSONNEL  !

I LOCK NOT CHALLENGED.

4 i

l 0 SUPPRESSION POOL BYPASS DUE TO i  :

j DRYWELL BOUNDARY FAILURE SHOULD i l NOT BE A CONCERN. l 1

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_! . ,z ,. g . _ _.- . .-- _ . , 7 JULY.1985

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FAILURES WITHIN DRYWELL o FAILURES IN DRYWELL IGNORED ON THE BASIS OF LOW PROBABILITY o REACTOR PRESSURE VESSEL ONLY.

o RPV INSPECTION o AS PART OF FABRICATION -

0 100% RADIOGRAPHY o 100% UT OF ACCESSIBLE PORTIONS (ALL WELDS ACCESSIBLE IN BWR/6) o ISI o 25% UT WITHIN 10 YEARS o 100% UT IN PLANT LIFE o CRD HOUSINGS o AS PART OF FABRICATION o 100% PENETRANT TESTING OF CRD HOUSING TO VESSEL WELDS -

0 100% UT OF CRD HOUSING TO VESSEL WELDS o ISI  :

L l

o EXCLUDED o WELD IN COMPRESSION '

o FAILURE OF ONE HOUSING PENETRATION POSTULATED O  :

RSV15 - 6

['/d f JULY 1985 by,c 7 y m n:;,wn:m  : -

=- -- - -

{~ u .;. . .n.mu ' _; _ ._. ___.=_n....-

DRYWELL HEAD AND CONNECTION DRYWELL PERSONNEL LOCK o DESIGNED TO ASME, SECTION III, NA-3352, t

! o MATERIAL AS ALLOWED BY ASME, SECTION III, NE 2000, i o FABRICATION AND INSTALLATION TO ASME, SECTION III, NE 4000, i

I o TEST PRESSURE o 30 PSIG FOR.HIGH PRESSURE TEST.

o 3 PSIG FOR LOW PRESSURE TEST.

o INSPECTION AND TESTING o INSPECTION TO ASME SECTION III, NE 5000.

o REPEAT SHOP TESTS AT 30 PSIG PRESSURE FOR LEAKS.

1 o FIELD TESTS

! o PREOPERATIONAL STRLCTURAL PROOF TEST AT 30 PSIG.

4

! o INTEGRATED HIGH PRESSURE LEAK RATE TEST.

i

o INTEGRATED LOW PRESSURE LEAK RATE-TEST. .

o PERIODIC LEAK RATE TEST AT LOW PRESSURE. i i .

j o AFTER EACH CLOSING 0F DRYWELL HEAD AND PERSONNEL .-

l LOCK, CONNECTION TESTED FOR LEAKS AT 30 PSIG.

lO

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RSV15 - 8 j O JULY 1985 9

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0 F A (6/80)

RSV15 - 9 M /// JULY 1985

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MANilAL LEVEM DRYWELL 9LRE l novas s suo virus or parvsLL sracervan rensmmit Arm RSV15 - 10 .

~: , . - - -

[-//r . ......

JULY 1985 l

e tw;- .: : a = --- ~ . ,2....~-:-=-.~.. .. , . -.~ . a. . - : . - . . . :- . - -

EFFECT OF MOLTEN CORE ON DRYWELL STRUCTURES 0

MOLTEN CORE ASSUMED TO BURN 6 FEET DEEP llOLE IN BASEMAT.

0 TEMPERATURE OF MOLTEN CORE ASSUMED TO BE 4000'F.

0 HEAT CONDUCTION ANALYSIS AND LINEAR ELASTIC STRESS ANALYSIS

PERFORMED.

t 1

! O RESULTS '

0 TEMPERATURE IN DRYWELL WALL = sl50'F i

0 DEFLECTION OF DRYWELL WALL =10.5 INCHES i

0 STRESSES IN DRYWELL WALL 0 4 3500 PSI l AT BASEMAT IN CONCRETE (F'C=4000 PSI) i I

! O t40 KSI 114 STEEL (FY = 46 KSI) 4 l

! NO DANGER EXPECTED TO DRYWELL WALL OVERALL STABILITY i

j i I

i RSVIS - 11 i

JULY 1985 i

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['p_* , . s't r ,'

• e

APPENDIX XI [

BNl u9 CONTAINMENT

, STRUCTURAL VER l*

l l

SIRKTURAL YERIFICATION STUDIES.: , j i

~

9 TORISPHERICAL STEEL CONTAIREST i

e DAYELL (STEEL) HEAD . , ,

O RE!WORCEB COERETE DRYELL Est Sam

? o j 0 RELIABILITY EVAlttilm W SEEL CERTAINENT I

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_._ __ . _ . . . . . . . ,, ,,m_,. ,, ,

1

. I l

. i I

(

t i

l

[. ;pSHIELD l

BUILDING

,s. .

)

" -CONTAINMENT

.,F.",

, ~ <.

0-

y. ~

c.

?

DRYWELL ,

'.a HEAD ," .

I,'<

_, j

,.3,,

,,, );. -UPPER POOL g

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f w* mig &wm w.g g % y .< i IivN.JN

'.i Nhbhhhx l L.

r&w .k ..

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~@me - . .. .

REACTOR- -.

.=. 4 w-,e f. .!

J> ,,

• ee g,g .. ,.

n.

F *

\

,"- .....! . g , 3,,, -REACTOR Z.

. [ 2

~

g "'j f, SHIELD WALL i r

1 *- -.S% -

/ ,.

.! I p

-re h*

../ '.

a.

a.

. ,. .e

= . .

/. ,.

, .f .[- 1 I .r

',, -WEIR WALL f 7.k '.*

6., .2 $/).(

~

d' .'.i .$ 'f[? hl T. $

4 -

-hor 2cNrAt f *'

. '. "[] l;;,;.a.e[h,* *.3 . ' .' ,e.

I,i .,*,*; e p, f gg.i f'.,,,,,

• • -$ VENTS

^ * * '

l;, ,',,

• L_.,,,'..,....?f",,,' . . t * , b ,' 'j , ,,.,, .,

• • • _ J

. . SUPPRESSION

,, p , ., ;, g. . ,

Poog.

. . :(.,'.. l.,,4. , . . .', *;) * : ..-. ,:;. , ,

qs 13., '. . . . . .. ,.. , . . . .];

,p.*f %

b,

-fb l f f r

~

I dL NW .,

i ;

g ,

i

.e Figure 15.1 Principal features of MARK III containment .

l I

(

GESSAR II SSER 2 15 '44

. i

~ . - . . r ' (, . : [ *~ T " O ,,r

"}"L".T*'*****'~ ,

~~ ~ 7 ~. Tw--

3 ~ y:

. .e.

. +.=c-=.----~... .

5 .,

I

x. .

, .... . .. . .. u n. . g . : -e. .. . r 1 GESSAR2 TQUXLP CORCON-M' ARCH 2-151 5-21-85 800 .

a wo- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - _ - - - - - - - - - - - - - -

t m . .

. j D ,

g . INTER-  :

, 00 0 *s -

E D. -

y g DI M

w w.

~CORCON

' 2 g

h

's < 400- f.<

n. ll

M ec

-) o u

q 30 0- - . _ _______

,. +

J _ _ _ _ _ er, g . ,

[

A I.

@ 20a. {

i i 3 i  % '

I 10 0 ',

00 2000 4000 600 0 000 0 1000 0 3200 0 It000 1600 0 1000 0 2000 0 TikIE - (EIINUTE)

VOLUME No.1 a :t=,.

, .: .. c . . -- .

, ....-:...-,-. - -' ^ ~ '

1

~

O

'l;

\*g 8

r D*

i t

e - 00  :

0 Dog I-o VESSEL D

T:s 1700 K d'

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o 9, 3 06 .

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a,*a h o ,*

u* OO a o 4 1.8M

• O b D

3 3 C 0 o o n 6 0O o

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• 0 '

1. # 0 0 '

o >' 0 r # #

o ** ~

6.lM Oo o 3,6M D C

< 7.7H O 2 0 0

\

oo

\ L 0, TO WETWELL CORIUM q O

' 1 gM o D O o ,

o I 9 Ooo o

0 O

o f1 b o

'a o o J 3 0 3

3 e 3M ._ 0B .Q D o

O o 0 e - o o O o e g g 4.6M TO PENETRATE 0

0 ,

D j og 0 0 BASEMAT d  !

0 g 0 0 0 0 D 0

{

o g0 0 0 o o o O o D

00b g o g 0 '

0 0 g t O

d S, , , ,> -

a. ;_;

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

('

ABLATION RATES l {

t !.

S ,'Cr&ONCRETE)(I ABLATION - I INITI AL) + k ABLATION h

$g:10-20W/cM

=

2 (LOWER CAVITY)  ;

i y=12-3 W/cM2 (SURROUNDING)

{

p = 2.5 gM/cM3, C = 1 J/ 3 M/K, AS 240 J/GM  !

[ = ABLATION INITIAL RATE = 10-20 cM/HR (LOWER CAVITY)

=

12-3 cM/HR (SURROUNDING) ,

N 10 HRS % 120 CM AXIAL

& 140 CM RADIAL, PEDESTAL INTEGRITY DOUBTFUL ,

O THERMAL GRADIENT '

~

f T- *( 8 T/g  ;

1 HR T SURFACE 1700K I INITIAL = 300K ___._ __

+ 25 cM g L l 20 HR  ;

T INITIAL e  ! ,

i

• -- 60 cM e F P

l l i s  ;-

Yh  !

,7---- .~,,.

-~ q.7, 7,4, . y 7 : .,..q.,, ...

va,' -

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,,,, -SHIELD '-

[. BUILDING

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• r j

. i

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. pCONTAINMENT  :

?.

. DRYWELL .'

HEAD t'

.; . 3

.~ t

^_ j j. pUPPER POOL F

.,, . . g- . .; ,..,

,,, , g.

-. v7 ,

i  ;', -f; *eW-sa? -wggg* .' .

7' .

I' Q ;WeMW31 *f4 7p #t---g[.'. i

.. $$?**UM

~~ '

^

BrdCMSQ&

• gy;. . . "W ' ' ' - -B -... ..,.: .'  ?

• pSRV D!SCHARGE f'

., em,; REACTOR . .

L g.e. A$.

Ae; ,,

".t

.......! s =  :

-REACTOR f,

.r

• n .

"s

'f.

. . [ 3' ,-,g f, SHIELD WALL L 1

i >- .

. i.

.s .. r, ,.. .

./.. . '.;s, .

oRywEtt i 1

?

' . '/*

.. , l'v .

. c ..

. p l .*.

I-

? I ."{ J -WEIR WALL j

,:-_....1

'\ .T- +,f: i- J. :., t 7.7 .

-g . .

G FJ P. 1. [i

.. . ,,,,, HOR 20NTAL l~,

M2  %'.....1W W F

.; :l.:;;!

. . .... vsNTs

.r .

• = = x..:...:. .:.. . . . . ... . . .- :... . ....~.5... w d ,'

' '. - ,suPPREsseN cot

. = .. ,

. L t.

r F

Figure 15.1 Principal features of MARK III containment .

J i-tr l 5' I w GESSAR II SSER 2 15-44 0,

_e---y-+,.-- - , , . - , . . . , . . . . . , , , . . . , . . ,

, , , , , , , {

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

A . .- _ . . . w. .a.w.j S= ~ , u. ~:. . .

i I

DOMINANT CONTAINMENT FAILURE MODES I t-OVERPRESSURE DUE TO NON-CONDENSIBLE GAS GENERATION HYDROGEN DEFLAGRATION AND DETONATIONS -.

P FAILING STRUCTURE,-SEALS AND PIPING PENETRATIONS [-;

e BY PRESSURE AND TEMPERATURE L Y

PHENOMEN0 LOGICAL ISSUES NOT CONSIDERED: IJ >

.t :.

STEAM EXPLOSIONS (l

DIRECT HEATING MECHANICAL FAILURES NOT CONSIDERED:

n RPV FAILURE F DOUBLE MSIV FAILURE LBLOCA WITH STUCK OPEN VACUUM BREAKERS .

VACUUM BREAKERS FAIL CLOSED Y, MECHANICAL FAILURES CONSIDERED -

/. 1-I-E3, 1-T-L3 RELEASES SPAN RANGE OF EARLY

[b-w h AND LATE CONTAINMENT WETWELL FAILURE DUE ,

i; TO POTENTIAL FABRICATION FLAW P r.,

EXCESSIVE DRYWELL-WETWELL' LEAKAGE CONSIDERED 5 e

{~

El, E2, SEQUENCES MODEL TOTAL BYPASS OF POOL BY

\

f'1. ,

g VAPORIZATION RELEASE '.

i- ,

DRYWELL HEAD FAILURE '

~

12, 120 SPAN RANGE V.

f I.'

ts

! l;..

g '/ ) k

, C' * - - *

, 5 . E5$ h N' I ~ -

'. _ _ _ . _ _ _ . _ _ _ - - . - - - - __ _ . . . .' . x J ., . c-l r

i l t

Table 15.1 ..

Conditional consequences predicted by the staff for

• internally initiated events and probability of  !~ '

occurrence with and without UPPS, per reactor year Release Early Early Latent Probability _

category" Person-fatality injury fatality rems p

w/o UPPS w/UPPS 1-T-L3 0 0 [

40 7 x E5"" 3 x E-6

' 9 x E-7  !

1-T-E3 0 0.0005 200 3 x E6 8 x E-6 1 x E-6 1-T-I2Q 0 j.,

3 200

, 3 x E6 1 x E-5 1 x E-6 2-T-B3 0 .

0 300 5 x E6 4 x E-6 4 x E-7 ATWS 0 '

1

[1 400 6 x E6 ' F 3 x E-6 3 x E-6  !-

l'T-I2 0 6 I 500 8 x E6 3 x E-6 3 x E-7 1-5B-El 0.006  !'

( 10 600 9 x E6 1 x E-9 1 x E-9

" Sea definitions in Table 15.15.

""7 x E5 = 7 x 105 ' '

Notes: ' l- ,

• c (1)

All conditional'mean consequences were calculated using the upper p t 1

i range BNL source term values described in SSER 2.  ;._

> .-(2) e The calculations assumed the Shippingport site, with public evacuation within 10 miles and relocation 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after plume passage. L t, . ~. ;

(3)

Mean consequences were computed over 91 different weather conditions.

[~.

p I.

p i g

e f-

r=

i i

• h "ssaa e g g.,,; y ssea 4.__:

_ J:L - - .m.m. -d i

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

- . - - . - - - . . - . . . - - - . - . ~ - . . . . u = 4

. . t..

i' L

r F

pi l'

.i i l'_ [

t

> i,c - :r,

l. u,>

p CONSIDERATION OF HYDR 0 GEN ISSUES i

.!-t:;

f i' f I t..l t

CURRENT REQUIREMENT o

[

[,< ,6

- .o. i QUANTITY :100% ZR - H O 2 CLAD. EQUIVALENT f o

78000 LB, ZR l'

,  ! -i j L i 3400 LB. H 2 '

t I i

i:# {

} .

RATE  : ACCEPTABLE TO STAFF h.l i

i r e

. t v t s!

i i 9 b . .,  ; ..

L+

l4 *

t. e i  !

v .!,

L ;

(" !

i l

! l '. <

6 L 4 2 i

t -

l

~ -[

i  ;

l . -

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e-l l

h L*

i 1

I,

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t,..

l I

t.

Is

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,eP

;. ;!

,n ,, -

---.-..c,., , . -

,. .[, f .[iN. i# '

.. '- -. ~

L - 4 +1 - D- ~-"- * ' * ~ " ' - - ' ~ ~ ' ~ ~ ~ ~ ~ ' '

.r. . .. . . n. ..= ,nm.

~

1

/"'N l (v)

GESSAR CONTAINMENT VOLUME (WETWELL + DRYWELL) = 1.4E6 CU. FT.

MASS AIR (e 80F INITIAL) = 1.05E5 LB.

MASS 02 = 2.2E4 LB.

MAX MASS H2 THAT COULD BURN = 2.7E3 LB.

(BASED ON MASS 02 AND ASSUMING COMPLETE COMBUSTION)

CORRESPONDING MASS ZR OXIDIZED = 6.2E4 LB.

FRACTION CLAD OXIDIZED TO PRODUCE = 79%

SUFFICIENT H2 TO BURN ALL 02 IN CONTAINMENT O)

(

~

FRACTION CLAD OXIDIZED TO PRODUCE = 65%

SUFFICIENT H2 TO BURN 02 TO 4% LOWER LIMIT (2.ZE3 LB H2 )

I n'

l-o

1. 7J'J' l .

L- .

_______.x... . _ _ _ _ . .o - ---.c- u- -

e O GESSAR - H2 RATES UNMITIGATED SCENARIOS, I.E. TOTAL LOSS OF INJECTION IN-VESSEL PRIOR TO SLUMP FOLLOWING EPG WITH SUCCESSFUL ADS 25 LB / MIN. PEAK 500 LB. TOTAL FAILURE OF ADS WITH 2SRV'S OPEN 85 LB./ MIN. PEAK 1400-1700 LB. TOTAL IN-VESSEL AT CORE SLUMP

+ 250 LB. H 2 AT ~ 400 LB./ MIN, EX-VESSEL CCI 2.5 TO 6 LB. H 2/ MIN. .

40 TO 150 LB. C0/ MIN.

^

CORCON CALCULATIONS t HEAD FAILURE = 150 MIN.

I CONT. FAILURE = 1700 MIN. .

. i AT ASSUMED 72 PSIA I 4200 LB. H2 [-

58000 LB. CD  :

t 99000 LB. C02 i' ea

--- -- -. . - ~

.+ .

, w - -

_, , . , , , ,) :, , ..,..-.-a,..,+ .-- - - - ~ ~ ~ - - - - - - - ~ ~ -

~- -

-'^^. a.

. s . -

2

> 0 w

-- N g g C E b E = 5 8= o s

FUNCTION

• S =S g = 8 gh 3 2 w $  %

Eu g5

= .n .,80s 3C e .d! be W W S.

=

3 Se= EE oo o 1 - =c e 08 =E "E ES EE *'* e7; E- s 0C $55 & $5 -E =>t =

uw ac ic C .g Z o = =uc >ow c to Oo Om OR w m Zu a 6 c cw zu te zo gg FAILURE SYMBOL 3 , y. y. 5* 5 5" v2 o

L3 E.SE.7

. 0.05 1E-4 _ L2 5.5E.11 O W A 0

. O D 6.9E4 1.0 0

/N 2E.7 1.1E.5 0.s E20 3.3E.7 /

0.8 024 0.5 E2 3.3E.7 /

)

0.2 o $

L3 1.854 0.33

' l0,g1 1.EE.8 Lo d 0.7 /

. ,[**- / u O

! 0.95 .

2E.7 l

- /

0.01 L1 0 0.76 E3 7.SE.7 i J E2 7.9E.9  ;

i

SUMMARY

L1 negligible E1 0 ,

i R 1.8 E.8 2E.7 L3 2.4 E.6 0.01 E1 0 E1 negligible '

- E2 .E.7 1.0E 5 1

'E2 1.0E.10 E2O 3.3E.7 . .

D 7.SE4 *

[

Nete: 1,S E E

• 1.8 10 8 f $,/jk y Figure 15.2 CTI-P, best estimate containment event tree GESSAR II SSER 2 15-45 7,- ,

- . . _ _ . . . < x .m . t __ _ _ __ . . _;. . _ .. _ . _

GESSAR - H2 THREATS TO CONTAINMENT f

1 ADIABATIC BURN OF ALL 02 CAN OVERPRESSURIZE  !

CONTAINMENT ("10 ATM) i 1-T-E3 TYPE RELEASE i CONTINUOUS BURN 3000 ts, suRN OvER 1 HR. AT ORAND GULF i

- 110 PSI A,

• 600 F  !

i f

h I

s'. . i L

t I

i l

i f

i i

i f l r 4  :

4 1 I L t

I

}

l

l s

m_ . ,._ ,- .-. ,.

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[ '., [ ; ? .; . #

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s _, _..2L.. . . - .2.' L..,_ . ___ . ... i .*. i. .. .. [_U J . v *

1 ',.. ' _ , _ J,

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

- - - - - - . . . -- u- ._

..+_ .. _

r p- , l C')

\v ...

C 8

=

o l 6 %

Z .

- z .

~ m 1 2 s D

y N -h J %

f C o 4 e >

a -E o

D o N

Z ~$~d e3 J -@ 5 a

_) d 5

,i U .$ ta i N 2

, , e6 8

!s g w a

I D -

o I O -g -

s '

o s Z . p*

Y 4 I g .

C -b I_ .

i e

o O b b b b b b O '

}-

8 S  ? 8 8 'E E R '

2 5

- 3805S38d 1N3W18....; YdWOD 1V101 .

O C

\, -

e

. p /r1 l

'. r

r

.(~

s *

~~

.. \

4

, r

'[ ]

i  !-

........,.~......,e..,,...~. .

• j 8

.i 700 0 GRAND GULF 2 CELL 112 50 LB/ MIN -'

l -

j g s000-

+

'l D v .. ,

1-. < 5000-

> CC

'! taJ

'! o. .

i

,d 4000-

.j H

~j 't p i E 300 0-k e , n

E .7 . I

$ *C 2000- l C.

! ':E .

I 100 0-1 4

00 , , , , , , , , , ,

g 00 100 0 2000 300 0 4000 500 0 6000 700.0 8000 900 0 1000 0 1I00 0

~

TIME - (MINUTE)

VOLUME NO. 2

. WeYc'"A$

- 8

_..

• _sal '_m _J ....--..;--m.. - -

a - ~

a ~

/+

C DEFLAGRATION

$LOBAL - NO IGNITORS

- FAILURE OF WETWELL SEAL ASSUMED UNIT PROBABILITY

- SMALL PROBABILITY OF RPV PIPE BREACH OR DRYWELL SEAL FAILURE LEADING TO El OR E2 RELEASE, OTHERWISE D RELEASE

- ABOUT FACTOR OF 3 IN PERSON-REM CONSEQUENCES O LOCAL - WETWELL SEAL MAY FAIL

- SMALL PROBABILITY OF RPV PIPE BREACH OR DRYWELL SEAL FAILURE LEADING TO E2 OR E3 RELEASE OTHERWISE L3 RELEASE

- ABOUT FACTOR OF 4 IN PERSON-REM CONSEQUENCES O

V g -M/

. . . , .-u . L . . w . ~ . , . = - . . :. - ~. u.

i LOCAL DETONATIONS 1

l

1-SB-El PORTRAYS DRYWELL AND WETWELL EARLY FAILURE i

CREDIT FOR PRIMARY SYSTEM RETENTION AND POOL i

. SCRUBBING OF VOLATILES i

i t

PERSON-REM CONSEQUENCES ABOUT AN ORDER OF MAGNITUDE k GREATER THAN 1-I-L3 1

1 1 - I - 12, 1 - T - 120 PORTRAYS DRYWELL HEAD FAILURE DUE TO DETONATION SHOCK LOAD ABOUT FACTOR OF 3 IN PERSON-REM CONSEQUENCES DEPENDING

'I ON FAILURE LOCATION

, e l' e I

l t

I t

l e

i g

a. . u K . ~.W. . - x;4L'_:. . . .. ..~.. ~ :.L L. .: .iiL i.a . - -

a i

HYDROGEN CONSIDERATIONS:  !

4 I

OPTIMUM IGNITION SOURCES HC0G TEST PROGRAM TO CONFIRM ADEQUACY OF GLOW PLUG l 4

, POWER SOURCE 4

DIVERSEPOWERSOURCERECbMMENDEDFORIGNITORS i .

I; LIMITATIONS OF IGNITION SOURCE 1

f

. STATUS OF HC0G CONSIDERATIONS 1

1 i

h 8'.

i i

i i

l l

1 i

i i

I l

l I

e l

l l

l 0-/(r3  :

w- u a w.w .-+ ..; . :- 5E: -

APPENDIX XIII-NRR COMMENTS ON SAFETY G0AL EVALUATION REPORT O '

~

NRR COMMENTS 0N SAFETY GOAL EVALUATION REPORT THOUGHTFUL, WELL WRITTEN.

ONE SERIOUS DIFFICULTYi CORE MELT FREQUENCY GUIDEIINE OF 10-N/RY IS TOO LENIENT:

50% CHANCE OF SERI]US REACTOR ACCIDENT NEXT 20 YRS.;

10% CHANCE OF 2 OR MORE SUCH ACCIDENTS. ..

CONTAINMENT SYSTEMS MUST FUNCTION BEYOND DESIGN .

, CONDITIONS; TOO MUCH RELIANCE ON KNOWLEDGE OF FR BEHAVIOR, CONTAINMENT PERFORMANCE.

THE 10-4/RY GUIDELINE NOT WELL DEFINED (!.E., WHETHER IT REPRESENTS SEVERE CORE DAMAGE OR EVEN EXTENSIVE -

t CORE MELT WITH NO RCS PENETRATION, OR EXTENSIVE CORE MELT WITH RCS PENETRATION AND CONTAINMENT CHALLENGE).

PROPOSED ALTERNATIVE:

LARGE-SCALE FUEL 8 FP RELEASE FROM RCS< 10-5/py i

I l

i .

.. pd

c. - . -

c ___ _ . . . _ _ _ _ _ . . - - .. - -- --- --- -

W 7 ~s

! .)

x _./

OTHER COMMENTS INCLUSION OF AVERTED ON-SITE LOSSES: NRR AGRESS STRENGTHEN SUPPORTING DOCUMENTATION PROVIS10NAL IMPLEMENTATION GUIDANCE:

EXCLUDES SAFETY IMPROVEMENT AT CORE MELT FREQUENCIES OF 3 x 10-5 TO 10-3 UNLESS MORTALITY RISK ODOS ARE NOT MET.

(x _./') -

FAILS TO ADDRESS PRA OMISSIONS, BIASES, EXTENT OF DIFFERENCES BETWEEN ESTIMATES AND ODOS, ETC.

ALARA DISMISSAL CONFLICTS WITH SEVERE ACCIDENT POLICY, ST,UDARDIZAT10N POLICY.

COMMISSIONER ASSELSTINE'S PROPOSED SAFETY GOAL: REJECTION FLAWED.

GREATER EMPHASIS ON PREVENTlHG TMIS IS WARRANTED:

REDUCE CM FREQUENCY GUIDELINE.

SHOULD NOT SEEM TO SHIFT PRIMARY SAFETY RESPONSIBILITY FROM LICENSEE TO NRC.

t

~N

)

a f

1 i

L!KELIHOOD OF CORE MELT ' '

ACCIDENT PROBABILITY IN N )/RY MfA CORE NELT PROBABILITY (f '

NEXT 20 YRS IN POPULATION 0F 100 PLANTS, PA l  :

8

! 3 x 10-4 45%

1 x 10-4 20%

3 x 10-5 6%

, 1 x 10-5 2%

1 i

l

[

1 I

i.

t l l

'i e

r h'/b $

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

. . - . . . . _ . . . . . . _ .._ :__',.L _ _ _

l SAFETY G0AL IMPLEMENTATION DIAGRAM i TOTAL CORE MELT FREQUENCY (MEDIAN)

A ,

CONSIDER SAFETY

^

> IMPROVEMENTS 3

~

T

-~

d t .-

EXCLUDE CONSIDERATION OF IF MET; 3 REGULATORY ACTIONS TO T~

LOWER RISK FOR PLANTS WIT A CORE MELT FREQUENCY k9 '

~4 IN THIS RANGE, UNLESS --

10 pCHECKMORTALITYd G0ALS IF NOT MET WITH REASURABLE M CONFIDENCE

(

3 x 10 -5 10

-5 r CHE K DOMINANT

(  ; SEQUENCES; IF ONE r F 3 SEQUENCE HAS FREQ.

>10~ 00 MORE

( ANALYSIS

./

I I

e f

9x 10' 100 SEO @ 9 x 10 -6 l

, #/4 ?

4

-m_.

~ . . _..

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

._m

, - . - 7_. _ , _7,., . _ ~7. _ y - ~_u 4

l E dode  :

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I, -

t Mea;a, i

i 1

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.i e  ;

i

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l. s !,

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if l l j g I i

4 5 3 =.-= % b:s w:b h m

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w., ~ \ Vetevik W a) ((

F

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0 4

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b $. L y .n . g a:sn ba%) {. t

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APPENDIX XIV

-- - F. R0WSOME COMMENTS ON SAFETY GOAL EVAULATION REPORT 1

4 NRR COMMENTS ON SAFETY G0AL EVALUATION REPORT THOUGHTFUL, WELL WRITTEN.

l ONE SERIOUS DIFFICULTY:

4 CORE MELT FREQUENCY GUIDELINE OF 10-4/RY IS T00 LENIENT:

- 50% CHANCE OF SERIOUS REACTOR ACCIDENT NEXT 20 YRS.;

10% CHANCE OF 2 OR MORE SUCH ACCIDENTS.

1

- CONTAINMENT SYSTEMS MUST FUNCTION BEYOND DESIGN CONDIT, IONS; T00 MUCH RELIANCE ON KNOWLEDGE OF FP BEHAVIOR, CONTAINMENT PERFORMANCE.

- PROPOSED ALTERNATIVE:

LARGE-SCALE FUEL & FP RELEASE FROM RCS <10-5/RY.

. CANCER RISK GUIDELINE IS T00 LENIENT AS SURROGATE FOR SOCIETAL RISK.

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{ - CONSIDER AN AGGREGATE SOCIETAL RISK GOAL.

INCLUSION OF AVERTED.ON-SITE LOSSES: NRR AGREES, i

- STRENGTHEN SUPPORTING DOCUMENTATIOf!.

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! - TREAT AS FAVORABLE COST IMPACTS, TO ARRIVE AT NET COST.

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PROVISIONAL IMPLEMENTATION GUIDANCE: TWO PROBLEMS:  !

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- EXCLUDES SAFETY IMPROVEMENT AT CORE MELT FPEQUENCIES OF i 3x10-5 TO 10-3 UNLESS MORTALITY RISK QDOS ARE NOT MET,  !

- TOO SIMPLISTIC: FAILS TO ADDRESS PRA OMISSIONS, BIASES, EXTENT OF DIFFERENCES BETWEEN ESTIMATES AND CDOS, ETC. ,

i ALARA DISMISSAL CONFLICTS WITH SEVERE ACCIDENT POLICY, l STANDARDIZATION POLICY, >

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! COMMISSIONER ASSELSTINE'S PROPOSED SAFETY GOAL: l l REJECTION FLAWED,  !

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- GREATER EMPHASIS ON PREVENTING TMIS IS WARRANTED: i REDUCE CM FREQUENCY GUIDELINE,

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UNITE.

CONSIDERATION OF POTENTIAL COMPL I' 0 'r, EFFECTS OF EARTHQUAKES ON EMERG

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I WASHINGTON, July 5, 1985

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O MEMORANDUM FOR: Chairman Palladino Comissioner Asselstine Comissioner Bernthal Comissioner Zech FROM: William J. Dircks Executive Director for Operations

SUBJECT:

CONSIDERATION OF POTENTIAL COMPLICATING EFFECTS OF EARTHQUAKES ON EMERGENCY PLANNING The purpose of this memo is to infor1n you 'of our progress and direction in preparation of the subject final rulemsking package which I plan to submit to you by early August 1985.

On December 21, 1984, the Comission published a proposed rule change to 10 CFR Part 50 that relates to Emergency Planning and Preparedness at Production andUtilizationFacilities(49FR49640). The proposed rule stated that neither emergency response plans nor evacuation time analyses need consider the -impact of earthquakes which cause or occur proximate in time with an accidental release of radioactive material from a nuclear power reactor.

To date, 61 coment letters have been received. Twenty five (25) letters favored the promulgation of the proposed rule. The letters favoring the pivpu=ed s uie were from utilities, consulting firms representing utilities.

2 private citizens and the Department of Energy.

Thirty-four (34) letters opposed promulgation of the proposed rule. Many voiced strong displeasure, shock or disbelief at the position the Commission was taking in the proposed rule change. The majority of these letters were from private citizens, and environmental groups. <

Additional input was also received from Japan, France, Sweden, Germany and Taiwan, all of which stated that the potential complicating effects of earthquakes were not specifically considered in their nuclear power reactor emergency planning, j Several issues raised in the public coments (and in particular in coments t

.from The Union of Concerned Scientists) will require substantial technical analysis prior to going forward with proraulgation of a final regulation. For l

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! The Commissioners 2 1 O example the staff needs to: (1) assess whether there are sufficient facts j to support the staff's belief that the complicating effects of earthquakes on

! emergency plans are adequately taken into account by the flexibility that

! exists in all emergency plans; (2) deal with the issue that defects in j seismic design and quality assurance in construction can substantially ,

j undermine the seismic strength of plant systems and structures; (3) evaluate i the limited existing information on the contribution of seismic events to

! overall core melt risks, recognizing that only a few PRAs assess seismic

! risks and the treatment entails many uncertainties; (4) deal with the question why emergency plans should not consider the complicating effects of l[

very severe earthquakes (i.e., 2 to 4 times the SSE) whose return frequency is 10E(-4) to 10E(-5) while current emergency plans concern themselves with

! plant accidents whose estimated return frequency are also in this range.

l These complex analyses, which are underway, are not expected to be completed before late July, 1985.

{ After careful review of both the San Onofre and Diablo Canyon decisions

! regarding the complicating effects of earthquakes on emergency planning, as

! well as the issues identified above, the staff is considering 3 alternative approaches:

j

) Alternative 1: Adoption of the proposed rule into a final rule with minor j but important word changes, for example, "no additional emergency prepared-

ness measures need be established to account for severe, low frequency

! natural phenomena than is already required in 10 CFR 50.47 and Appendix E."

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Alternative 2: Leavina the issue open for adjudication on a case-by-case j basis; accomplished by withdrawing the proposed rule or by requiring consid-i eration of earthquakes, i Alternative 3: Promulgation of a final rule which clarifies the original j

intent of the Connission to require that emergency response plans shall l

assure that the following capabilities exist relative to the complicating impacts of severe, low frequency natural phenomena.

I 1. Ability to transport necessary personnel to the plant after the event in order to augment the original staff to cope with degraded modes of plant operation.

2. Ability to obtain damage estimates to the plant and to be able to commu-nicate these estimates to offsite authorities. The information should i be available to factor into the decisionmaking process, including reco-4 unendations for protective actions after severe, low frequency natural phenomena.

3. Emergency plans for offsite authorities should take into account various l degrees and locations of damage to the plant environs. This shall be O

a M x aA The Comissioners 3 limited to knowing alternate routes of travel as well as establishing cHteria for detennining whether to shelter, relocate or to evacuate.

Having considered all of the above, as well as all comments received, past operating reactor and emergency preparedness experience, I am leaning toward a recomendation that a final rule be promulgated which would embrace the concepts of Alternative 3. This alternative would be a clarification and articulation of the Comission's original intent as to what is specifically required to assure the necessary flexibility to cope with the complicating effects of severe, low frequency natural phenomena on emergency planning.

/

J Willia .. Dircks Executive Director for Operations cc: SECY OGC OPE M. Cutchin n

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ASSESSENT OF FIELD APPLICATIONS OF iQ i

lM j CONTROL ROOM HABITABILITY PRACTICES

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! CONCLUSIONS  !

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A. CONTROL ROOM SYSTEMS, COMPONENTS, OPERATIONS, PROCEDURES, TECHNICAL SPECIFICATIONS n i

% B. NRC PRACTICES AND POLICIES AND NRC LICENSEES' PRACTICES i ..

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\ CONTRIBUTION TO THE DEGRADATION OF PLANT SAFETY SHOULD BE EVALUATED N

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A. 2 CHANGES TO THE ACTION STATEMENTS AND SURVEILLANCE REQUIREMENTS OF TECHNICAL SPECIFICATIONS SHOULD BE MADE AS NEEDED TO ENSURE THAT Tile CONTROL ROOM HVAC SYSTEMS SPECIFICATIONS PROVIDE FOR FUNCTIONING AS DESIGNED (1) FLOW RATE SPECIFIED FOR THE PRESSURIZATION TEST (2) ALLOWABLE LEAKAGE RATES FOR ISOLATION VALVES OR DAMPERS (3) ASSURANCE THAT FLOW IS NOT OCCURRING THROUGH THE ADSORBER UNIT WHEN THE SYSTEM IS SUPPOSED 10 BE ISOLATED N

01) APPROPRIATE LABORATORY CONDITIONS FOR THE TESTING OF ACTIVATED CARBON (5) SPECIFY THE LOCATIONS OF TEMPERATURE MONITORS AND WHETHER CONTROL ROOM TEMPERATURE LIMITATIONS ARE FOR HUMAN PERFORMANCE OR EQUIPMENT PERFORMANCE AS APPROPRIATE u

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A, 14 ALL NRC RELATED CONTROL ROOM HABITABILITY CRITERIA AND THEIR BASES SHOULD BE INCORPORATED INTO ONE DOCUMENT TO ASSIST THE N

% NRC STAFF AND THE APPLICANTS AND LICENSEES IN UNDERSTANDING THE BASES FOR REQUIREMENTS l

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i A. 5 THE CAPABILITY OF THE REMOTE SHUTDOWN FACILITIES TO BRING

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N THE PLANTS TO COLD SHUTDOWN NEEDS TO BE DEMONSTRATED i

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. l B1 THE ADVANTAGES OF CURRENT CONTROL ROOM DESIGNS ARE NOT IMMEDIATELY EVIDENT. ALTHOUGH PRESENT CONTROL ROOM DESIGNS HAVE BECOME MORE COMPLEX, THEY DO NOT NECESSARILY AFFORD Q

k THE CONTROL ROOM OPERATORS BETTER PROTECTION THAN THE SIMPLER, OLDER DESIGNS s

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i B. 3 SOME LICENSEES ARE MAKING IMPROPER 10 CFR 50.59 EVALUATIONS I

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! B4 THE UPDATED FSAR DOES NOT REFLECT PRESENT CONTROL ROOM

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O U.P_I)1TfJ) FSAR_'sJ1D_S_Y_S_T.EM_ Cal (F_LG_llRATLR THE UPDATED FSAR's FOR THE OPERATillG PLAllTS DO NOT REFLECT THE ACTUAL AS-BUILT. SYSTEMS, NOR DO THEY ACCURATEL,Y DESCRIBE HOW THE SYSTEMS ARE OPERATED. THE OBSOLESCENCE OF THE FSAR's APPEARS TO BE DUE TO INADE00 ATE OR UNTIMELY C0llFIGURAT10N CONTROL.

MODIFICAT10llS TO BOTH THE HARDWARE AND THE METHODS OF OPERATING THE SYSTEMS HAVE BEEN MADE WITHOUT COMPLETE.

C0llFIGURAT10N CONTROL. AREAS LACKING IN CONFIGURATION CONTROL !!1CLUDE PLAllT PROCEDURES, TRAINING MATERI AL, AND PLAlli DRAWINGS.

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APPENDIX XVIII '

ADDITIONAL DOCUMENTS PROVIDED ACRS  !

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ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE

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1. Memorandum, C. P. Siess to ACRS Members, Core-Melt Frequency Guidel'ine and i

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Cost-Benefit Analysis, July 11, 1985

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2. Memorandum, A. L. Newsom to ACRS Members, NRC Appropriations Bill for FY 1986, July 11,1985 f;

3. Memorandum, H. W. Lewis to ACRS Members, There is No Way to Measure a Median Failure Rate, July 12, 1985

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