ML20058C578
| ML20058C578 | |
| Person / Time | |
|---|---|
| Issue date: | 08/09/1990 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2717, NUDOCS 9011020004 | |
| Download: ML20058C578 (52) | |
Text
t.
t
' ~*
9' TABLE OF CONTENTS MINUTES OF THE 364TH ACRS MEETING AUGUST 9-11, 1990
)
- I.
Chairman's Report (Open)......................................
1 II.
Requirements for Essentially Complete Design Plant (Open)....
3
- 1 III.
Incident Investigation Team Report on the" Event at Vogtle Plant (Open)................................................
7 IV.
Generic Issue B-56, Diesel Generator Reliability (Open)......
10 V.
Severe Accident Risk Report (NdREG-1150)
(Open)...............
13 i
4 VI.-
EPRI Requirements for Advanced LWRs (Open)...................
15 VII.
NRC Research Program on Organizational Factors (Open).......
18
' VIII.
Solenoid Valve Case Study (Open)...........................
21 IX.
Reactor Operating Experience and Events at Nuclear Plants (Open)......................................................24 X.. Executive Sessions (Open).....................................
25 A.
Reports to'the Commission-(Open)........................
25 Level of Detail Reauired for Desian Certification Under-Part 52 (Report-to Chairman Carr dated Aug. 14, 1990)... 25 l!
Procosed Resolution of Generic Safety Issue B-56, e
D'
-" Diesel Generator Reliability"- (Report to Chairman I
Carr dated Aug. 14, 1990).......'........................
26
'NRC Research~Procram on Oraanizational' Factors
~
l (Report to Chairman Carr dated Aug. 16, 1990)..........
26 Succort Staff for ACNW Activities (Joint ACRS/ACNW Report to Chairman-Carr dated Aug.' 9,'1990)............
27-
'B.
Subcommittee Reports (Open)............................
27 q
1'
- 1.. -Planning & Procedures Subcommittee Recommendations.. 27 s
u 2.
NRC Safety Research Program.........................
29 j
C.,
Summary / List of Follow-Up Matters.............................
30 D.
Future Activities (Open)......................................
32
.1.
Future Agenda.....................................
...... 32 2.
Future Subcommittee Activities...........................
32 l-l Figure 1 - p.124a i
/D D
2717 poc
- ..' i _,-
11 APPENDICES MINUTES OF THE 364TH ACRS HEETING AUGUST 9-11, 1990 I.
Attendees I
II.
Future Agenda III.
Future Subcommittee Activities IV.
List of Documents Provided to the Committee
+
(
s
=i l
J
8h 7
i
,;3.*m C
' Federal Regasur / Vct. 55, No M]/ Tu:sday, July 31,..mo / Notices 81115
{-// 6 A SeverityLedel/V-Violadonsf NRC staff and the nuclear industry will discuss anticipated ACRS subcommittee Inm/vig,forExample "
participate as appropriate, activities and items proposed for 21 on.-12:Mpn.: Alvin W, Vogtle consideration by the full Commlttee.
the(1) Hot particle exposures for which NuclearPlant(Open}-Members of the 5:sp.m.-6:xp.m.--Prepararlon of.
dose to the skin exceeds the limits of incident investigation Team (IIT) will ACRS Reporte (Open)--De Committee i; 10 CPR 30.101 and exceeds the relevant.
brief the Committee on the results of will discuss proposed reports f, the
'oriterion but does not exceed three their investigation of the recent loss of NRC regarding items consid.'.d during
- times the relevant criterion (either the emergency power at Unit No.1 of the this meeting.
beta emission criterion or the skin doso Vogtle Nuclear Plant during a shutdown erlierlon).
mode of operetion Saturday. August 11,1990, Roosa P-110, (1) Failure to make a notification -
1:xpm.-3:xpjn.iRequirements for 7s20 Nwfolk Avenue, Bethesda, Md.
regulred by 10 CFR 30.4031b) or a report Essentially Complete Design (Open}-
a:x on.-12 Noon and2 p.m.-2:M
- (required try to CPR 20.405 for a hot particle exposure that res]is in a skin ne Committee willreview and report p.m.-Pmpamt/on ofACRS Reports on the proposed definition of -
(Open)-ne Committee will continue dose that exceeds the limits of to CPR.
"essendally oomplete design" for discussion of proposed ACRS reports to
- 30.101 and exceeds the relevant criterion standardized nuclear plants-the NRC.
but does not exceed three times the Representatives of the NRC staff and Procedures for the conduct of e':,d relevant criterion (either the beta the nuclear industry will participate as participation in ACRS mect!..;;: were cmission criterion or the skin dose, e criterion)._
appropriate. :
published in the Federal Register on -
M5p.m.-dg.m.!Seveio Acc/ dent R/sA September 27,1989 (54 FR 39594). In B
a SeverityLevel V-Vlolations
- '0*"**
C; a
th Re e Inmirlitg.forExample or written statements may be prese' nted Groupwillbrief theCommittee j
. (1) Pallure to make a notification regarding the results of the NUREG-1150 by members of the public, recording required by to CFR 30.403(b) or a report Peer review, willbe permitted only during those f
l(required by to CFR 30.406) for a hot 8Pa.-6:m pm.:ACRSSubcommIttee portions of the meeting when a transcript is be' kept, and questions -
maybe asked oNy b riembers of the particle exposure that results in a skin Act/villes(Open %e Committee will dose that exoceds thelimits of to CFR hear a report an discuss ACRS Committee, its consuftents, and staff.
30.101 but that does not exceed the subcommittee and members activities l
relevant criterion (either the beta in designated areas, including the Persons desiring to make oral P anning and conduct of ACRS statements should notify the ACRS l
emission criterion or the skin dose
' criterion).
activides.
Executive Director as far in advance as practicable so that appropriate
' Note: No numerleel t.rlierte (beta emleelon Friday, august 10,19eo, Room P-110, arrangements can be made to allow the
' Talees or skin doses) beve been established 7930 Norfolk Avenue, Bethesda,Md.
necessary time during the meeting for as examples for Severtty 1.evels I and 11.
- 8 2 a.m.-d:#am. EPRI such atatements. Use of still, motion Dated et Rockville, Maryland, this 26th day RequirementsforAdmiredLWAS picture and television cameras during
II"IF 188 0 '
4' (Open}-%e Committee will discuss this meeting may be limited to selected For The Nuclear Reguletary Commleelon. -
proposed ACRS comments and rtions of the meeting as determined Samuell.Chilk,'
recommendations regarding the NRC
{y the Chairman. Information mgarding -
Seawaryofthe commission staff report on chapters 1-6 of proposed the time to be set aside for this purpose o
(FR Doc.90-1Meo Piled 7-40-eat 8:48 em]
EPRI Requirements for ALWRs.
may be obtained by a prepaid telephone asAnne esos vs ms
. RMa.m.-12 noon-Organisatlanal call to the ACRS Executive Director, Mr.
FactoreResearch Progress Report Raymond F. Fraley, prior to the meeting.
(Open}-%e Committee will review and in view of the possibility that the Advisory Committee on Reactor
- report on NRC sponsored research schedule for ACRS meetings may be -
safeguards; Meeting Agenda within the human factors program on adjusted by the Chaltman as necessary organizational factors, to facilitate the conduct of the meeting, in accordance with the purposos of sections 39 and 382b.of the Atomic J p.m.-#p.m.,Solenold Valve Case persons planning to attend should check Study (Open}-Representatives of the with the ACRS Executive Directorif Energy Act (42 U.S.C. 2039,2232b), the NRC Office for Analysis and Evaluation such rescheduling would resultin major Advisory Committee on Reactor of OperationalData willbrief the inconvenience.
Safeguards willhold a meeting on
. AugustS-11.1990,in P.oom P-110,7920 Committee on the AEOD report on the
' I have determined in accordance with '
performance of solenoid valves in subsection 10(d) Public Lew 92-463 that
- Norfrik Avenue, Bethesda, Maryland..
nuclear plants.
It is necessary to close portions of this
..Natice of this meeting was published in 2p.m.-d dspa.--Reactor Operating meeting noted above to discuss
' th) Federal Register on June 20,1990.
Experience (Open/ Closed)-A briefing
. Proprietary Information applicable to 4 %ursday, August 9,1990. Room P-110, and discussion of reactoroperating the matter being discussed (5 U.S.C.
7920 Norfolk Avenue, Bethesda, Md.
experience and events will be held 552b(c)(4)).
Including the recent feedwater line Further information regarding toples 8:30 a.m.-6:45 a.m.:Chaltman's failure at the Loviisa Nuclear Power to be discussed, whether the meeting 1 Remarks (Open)-The ACRS Chairman plant and proposed changes in the has been cancelled or rescheduled, the will briefly report regarding items of frequency of turbine stop valve testing Chairman's ruling on requests for the current interesti in Westinghouse nuclear power plants.
opportunity to present oral statements 3:45 a.m.-10:45 a.m.: Cencric Issue B-Portions of this session will be closed and the time allotted can be obtained by j
ad DieselGenemforReliability to discuss Proprietary Information a prepaid telephone call to the ACRS O
(Open)-The Committee will review and applicable to the matter being discussed. Executive Director. Mr. Raymond F.
1 E
report on proposed resolution of this N5p.m.-5:30p.m.-Future ACRS Fraley (telephone 301/492-8049),
l'
- genericissue. Representatives of the Activities (Open}-The Committee will between 7:45 a.m. and 4:30 p.m.
I
I k
.* ~' '.'31116' Federal Regist:r / Vol. 55. No.147 / Tuesday, July 31, 1990 / Notices
=
Dated: July 24.1990. -
violations, the provlsions of the NRC's time, the matter may be referTed to the John C. lloyle, requirements that the Ucensee had Attorney General for collection.
Advisory Comm/tiesMonogementOfficer, violated, and the amount of the civil in the event the Ucensee requests a lFR Doc. 80-1773e Filed 7-30-00; s.48 am]
penalty proposed for the violations.The hearing as provided above, the issues to aussa toot room, Ucensee responded to the Notico by be considered at such hearing shall be letter dated May 9,1990. In its response, whether, on the basis of the violations the Ucensee admitted the violations but admitted by the Ucensee, this Order
[ Docket No.60-213) requested reconsideration of escalation should be sustained.
~
Connecticut Yankee Atomic Power of the base civil penalty based on its
. Co.: Consideration of issuance of asserted identification of the residual Dated at Rockville Maryland this 20th day or july tom Amendment to Facility Operating a[Qa IR pump lem prior For' the Nuclear Regulatory Commisaloa.
Ucense and Opportunity for Hearing; jemee Uebennan, U'
- U'"
asserted extensive cor c Ive action put in place prior to and following discovery Dinctor, Office ofEnfurcement.
In the Federal Register of July 23,1990, of the RHR pump problem.
Appendix-Evaluations and Concluelos in the center colunm of page 29920,
. make the followina corrections:
Ill On Apdl12,1990, a Notice of 1,in the fifth and sixthlines of the After consideration of the Ucensee's Vi lation and Proposed imposition of second complete paragraph of the response and the statements of fact, Civil Penalty (Notice) was issued for document "1DMCl/sm" should be explanations, and argument for violations identified during a special
-e changed to read "1.cmicroCi/sm."
reconsideration contained themin, the NRC inspection at the uoyah 2.In the first line of the fourth staff has determined, as set forth in the Nuclear Plant. Units 1 2.TVA complete p.agraph of the document the Appendix to this Order.that the responded to the Notice in a letter dated date of August 20,1990 should be violations occurred as stated and that May 9,19Ein its rupuse, the licenew changed to read August 22,1990..
the penalty proposed for the violations admined die violeHons, but mquwted Dated at Betbeeds, Maryload, this asth day designatedin the Notice should be reconalderationof the posed civil ofJuly1se0..
Imposed.
penalty.%e NRC sta evaluation and conclusion sogardiac TVA's moponse is l
Devid I. Meyer, ty
,, go)}o,,
chief
- -Pub &etiene amnch, DivisM""/Muodwnafhformet/esand in view of the foregolag and pursuant 1.Restatementof Violoflons
- pu6#cotioneServices,Officeof to section 234 of the Atomic Energy Act L Adm/a/stroaion.
of 1954, as amended (Act),42 U.S.C.
a.10 CFR part 50, appendix B,
- IFR Doc so-treet Filed Nao-sot ens amt.
2282. and to '.:Fil 2.206,1 Is hereby CriterionXVI. Corrective Action, on/emithat:
requires la part, that measures shall be amountof Seeenty Dvenousayit l'n the The Ucensee pay a civilpena established to assure that conditions adverse to quality, such a failures, 10oekeiso-a27and esm Weense eofthe deviations and nonconformanceare Dollars (g75,000) within 30 dadraft, or Nea.DN77and DN79 EAM11 date of this Order,
- check, promptly identified and corrected, egey 8equ yah-
"g tgUn 0
a
&dw a
the Reaed p Losa, May imposing Cive Monetary Poneny Director Of6ce of Enfonement.U.S.
alertedlicensees to a significant Nuclear Regulatory CW% Atta:
condition adverse to quality that I-Documcat Control Desk. Washington, involved the potentialfor the
- Tennessee Valley Authority -
DC 20555, deadheading of one ormore pumps in N
(Ucensee)is the holderof O Ucense No. DPR-77 and No.perating -y DPR-79 mini w ine c auna to two or more
' i i
issued by the Nuclear Regulatory
%e Ucensee may request a hearing pumps or other piping configurations Commiselon (Commission or NRC) on within 30 days of the date of this Order, that do not preclude pump-to pump September 17.1990 and September 15 A request for a bearing shall be clearly interaction during mialGow operation.
1981, respectively.The licenses marked as a *% quest for an Ucensee enginwring calculation DNE j
authorize the Ucensee to operate the Enforcement Hearing" and shall be SQN-74-D053, deled July 22,1988,.
Sequoyah Nuclear Plant, Units 1 and 2, addressed to the Director.OfGce of determined that RHR pump damage.
i et Soddy Daisy, Tennessee. In Enforcement. U.S. Nuclear Regulatory would occur for a pump that was run -
accordance with the conditions Commission. Attn: Document Control deadheaded for greater than 11 minutes, specifled therein.
Desk, Washington, DC 20655. Copies to CFR 50.9 requires,in part, that also shall be sent to the Assistant information provided to the Commission II_
Ceneral Counsel for Hearings and by a licensee, be comp &ete and accurate An inspection of the Ucensee's Enforcement at the same address, and to in all material reeps.;ta.
. activities was con 6ted on January s-the Regional Administrator. Region 11.
Ucensee letter to the NRC in response 12,1990.%e renus of this inspection
- Atlanta, Georgia, to NRC Dulletin 88-04, dated August 2.
Indicated that 6e Ucensee had not if a hearing is requested, the 1988, stated that the potential existed for t
- conductedits 2ctivities in full Commission willissue an Order deadheading a safety related RHR pump -
compliance w nh NRC requirements. A designating the time and place of the due to pump-to pump Interaction under written Notice of Violation and ;
hearing. lf the Ucensee falla to request a miniflow conditions when the head Proposed imp osition of Civil Penalty hearing within 30 days of the date of this differential between the pumps (Notice) was served upon the Ucensee Order, the provisions of th1 Order shall exceeded 11 pounds per square Inch by letter dated April 12,1990.The be effective without further proceedings. (psi). The letter also stated that recent Notice stated the nature of the If payment has not been made by that r.urveillance test data demonstrated that
c asm
/o
'o,,
UNIT E D STAT ES I
NUCLEAR REGULATORY COMMISSION e
i, j
ADVISORY COMMITTEC ON FIEACTOR SAF EGUAllDS o
w AwiNo1ON, D. C. 20LLt>
g e..
+
August 2, 1990 (REVISED)
SCl!EDULE AND OUTLINE FOR DISCUSSION 364TH ACRS MEETING AUGUST 9-11, 1990 Thursday. Auaust 9.
1990. Room. P-110. 7920 Norfolk Avenue. Be nesda, Md.
8:45 A.M.
Chairman's Remark's by ACRS ChairmaD (Open) 1)
8:30 1.1)
Opening Remarks (CM/GRQ) 1.2)
Items of current interest (CM/RFF) j 2) 8:45
- 10:45 A.M.
Reauirements for Essentially Camelete Desian (open) 2.1)
Comments by ACRS Subcommittee Chairman regarding proposed definition of l
" essentially complete design" for' standardit9d nuclear plants (CJW/MME) 2.2)
Meeting with representatives of the NRC staff and the nuclear industry, as appropriate 10:45
- 11:00 BREAK l
- 3) 11 90
- 12:30 P.M.
IIT for Alvin W. Voatle Nuclear Plant (Open) 3.1)
Comments by ACFS Subcommittee Chairman (JCC/PAB) 3.2)
Briefing by representatives of the NRC Incident Investigation Team and the NRC staff, as appropriate, regarding the IIT investigation of the loss of emergency power at Unit No. 1 of the Vogtle Nuclear Plant during a shutdown mode of operation.
l 12:30 1:30 P.M.
LUNCil 4) it30 3:15 P.M.
Generic Issue B-56. Dioel Generator Reliability (Open) 4.1)
Comments by ACRS Subcommittee Chairman regarding proposed resolution of this generic issue (CJW/MME) 4.2)
Mooting with representatives of the NRC staff and the nuclear industry, as appropriate 3:15 3:30 P.M.
BREAK
m 2
4:45 P.M.
Egvere Accident Risk Report (NUREG-1150) 5)
3:30 (open) 5.1)
Comments by ACRS Subcommittee Chairman (WK/MDH) s.2)
Briefing by Chairman of the NUREG-1150 Peer Review Group 5:15 P.M.
ACRS Subcommittee Activities (Open) 6)
4:45 6.1)
Report of ACRS Planning and Procedures Subcommittee Meeting on August 8, 1990 regarding conduct of ACRS activities (CM/RFF)
Friday. Auaust 10. 1990. Room. P-110. 7920 Norfolk Avenue. Bethesda. Md.
9:30 A.M.
EPRI Reauiremen a for Advanced LWRs 7) 8:30 (Open)-
7.1)
Discuss proposed ACRS comments and recommendations on the NRC staff evaluation of Chapters 1-5 of pro'-
posed EPRI Requirements for Advanced Light Water Reactors (CJW/MME) 8)
9:30 12:15 P.M.
NRC Research Procram on Oraanizational-(10:30-10:45 - BREAK)
Factors (Open) 8.1)
Comments by ACRS Subcommittee Chairman (JCC/HA) 8.2)
Meeting with representatives of the NRC staff regarding NRC research within the human factors program on organizational factors 12:15 1:15 P.M.
LUNCH 2:15 P.M.
Solenoid Valve Case Study (Open) 9)
1:15 9.1)
Comments by ACRS Subcommittee Chair-man regarding an AEOD report on the performance of solenoid valves in nuclear plants (HWL/HA) 9.2)
Briefing by representatives of the NRC Office for Analysis and Evalua-tion of Operational Data 3:00 P.M.
Reactor Oneratina Experience (Open/ Closed)
- 10) 2:15 (3:00-3:15 - BREAK) 10.1) Comments by ACRS Subcommittee Chairman (JCC/PAB) 10.2)
Briefing by and discussion with members of the NRC staff regarding operating experience and events at nuclear power plants including the l
3 Loviisa N Jclear Plant feedwater line failure.
(Portions of this session will be closed as necessary to discuss Proprietary Information provided in confidence by a foreign source.)
3t45 P.M.
Future ACRS Activities (Open)
.11) 3t15 11.1)
Discuss anticipated subcommittee j
activities (RPS/RFF) i 11.2)
Discuss topics proposed for con-sideration by the ACRS (CM/RPS) 11.3)
Discuss scope and nature of ACRS 1
annual report to the U.S. Congress and/or the NRC on the NRC Safety Research Program (IC/RFF/SD) 5:45 P.M.
Preparation of ACRS Reports to NRC (Open)
- 12) 3:45 12.1)
Discuss proposed reports to the NRC ont 12.1-1)
Requirements for Essent'ially Complete Design (CJW/MME) 12.1-2)
GI B-56, Diesel Generator Reliability (CJW/MME)
, Saturday. Auaust 11. 1990. Room P-110. 7920 Norfolk Avenue. Bethesda, Md1 13) 8:30
- 12t30 A.M.
Preoaration of ACRS Reoorts (Open) 13.1)- Continue preparation of ACRS reports to the NRC on:
13.1-1)
EPRI Requirements for Advanced LWRs (CJW/MME) 13.1-2)
Research on Organizational Factors (DAW /HA) 13-1-3)
Requirements for Essentially Complete Design (CJW/MME) 13-1-4)
GI B-56, Diesel Generator Reliability (CJW/MME)
MINUTES OF THE 364TH ACRS MEETING AUGUST 9-11, 1990 The 364th meeting of the Advisory Committee on Reactor Safaguards (ACRS) was held at Room P-110, 7920 Norfolk Avenue, Bethesda, Md.,
on August-9-11, 1990.
The purpose of this meeting was to discuss and take appropriate actions on the items listed in the attached agenda.
The entire meeting was open to public attendance.
A transcript of selected portions of the meeting was kept and is available in the NRC Public Document Room.
(Copies of the transcript are available for purchase from Ann Riley & Associates, Ltd., 1612 K Street, N.W.,
Washington, D.C.
20006.)
I.
Chairman's Reoort (Open)
[ NOTE:
Mr.
R.
F.
Fraley was the Designated Federal Official for this portion of the meeting.)
Mr. Michelson, the full Committee Chairman, convened the meeting at 8:30 a.m. with a brief summary of the planned meeting schedule and the provisions under which the discussions were to be held.
He stated that the Committee had received neither written comments nor requests for time to make oral statements from members of the public.
Items of current Interest Mr. Michelson stated that the following items are of current interest:
The NRC has published a proposed rule related to renewal of e
nuclear plant licenses for public comment.
Mr. Michelson suggested that the cognizant ACRS subcommittee follow-up on this matter and keep the full Committee abreast of further developments.
The Justice Department has decided not to prosecute Mr. Stello on charges that he purjured himself in testimony before the Congress.
The NRC also plans no further action, e
The NRC Inspector General (IG) has concluded that the NRC staff misstated the facts and also provided incomplete information to the Commission on the emergency preparedness program for the Pilgrim nuclear power plant.
The IG states that the staff relied too heavily on the utilities' paper trail and assurances that progress was being made in resolving outstanding emergency preparedness issues.
Also, the staff did not evaluate the conflicting information provided by various credible sources.
Chairman Carr has directed the Executive Director for Operations (EDO) to recommend a course of action on this matter.
e m
m
- m
I 364th ACRS Meeting Minutes 2
Mr. Michelson noted that the ACRS, in its September 14, 1988 report to the Commission, pointed out the inadequacies in the emergency preparedness program for the Pilgrim plant and recommended that a
clearly defined program for early correction of these inadequacies be available and approved by the staff prior to approving start-up of this plant.
Comanche Peak and Seabrook nuclear plants will be ready to begin commercial power operations in the very near future.
The main steam isolation valves at Salem Units 1 and 2 were e
found to have deficiencies in that they will not close fast enough to meet the requirements specified in the Salem Technical Specifications.
Based on a survey of six BWR units and on the evaluation of e
the motor-operated valve (MOV) data supplied by the BWR Owners Group on the remaicing BWR units, the staff has determined that a concern exists regarding the capability of certain MOVs to perform tueir design function to isolate a downstream pipe break.
The staff has developed an action plan to address the MOV operability cc.cerns.
Also,. the staff is performing a sensitivity analy s.,s on the NUREG-13 50 models based on the MOV deficiencies to gain insights on this-problem and will determine the need for more detailed modeling.
The staff is currently preparing a draft 50.54(f) letter that will be discussed with the BWR Regulatory Response Group during August 1990.
Mr. Michelson asked Mr. Igne, ACRS staff, to obtain more information on this matter and provide it to the Committee, The staff has initiated an investigation of the cable conduit e
deficiencies at Sequoyah 'Jnits 1 and 2.
They have selected 15 " worst case" conduits fo't this investigation. As requested by the staff, the Tennessee Valley Authority has submitted a revised justification for contir.ued operation of Sequoyah Units 1 and 2.
o On July 13, 1990, a 4 kV "c" lius breaker explosion at D.
C.
Cook Unit 2 killed one worker and injured three others.
This incident was due to personnel error.
Mr. Carroll informed the Committee that the NRC has issued an interim standard on hot particles.
He noted that this standard is based on the National Council on Radiation Protection and Measurements (NCRP) recommendations that are included in the NCRP Report-106, as recommended by the ACRS.
y.,
I 364th ACRS Meeting Minutes 3
II.
Recuirements for Essentially Comolete DesicD (Open)
(NOTE:
Dr. M. El-Zeftawy was the Designated Federal Official for this portion of the meeting.)
Mr. Wylie, Chairman of the Improved LWRs Subcommittee, stated that the purpose of this session is to continue review of the NRC and industry proposals for the completeness of designs issue for evolutionary and passive plants.
ynious Levels of Detail Recruired for Desian Certification Proposed kv_.the Staff - Mr. M. Virgilio, NRR l
Mr. Virgilio, NRR, briefed the Committee regarding SECY-90-241,
" Level of Detail Required for Design Certification Under Part 52."
He stated that the ongoing activities are to maximize standardization and preserve feasibility with appropriate degree of flexibility.
The main objectives of SECY-90-241 are tot e
Define a level of design detail for 10 CFR Part 52 design certification that will maximize standardization, and be achievable without vendor specific data for components in traditional A/E scope (e.g., pumps, valves, heat exchangers, etc.).
e Estimate the engineering effort required to develop the defined level of detail.
The staff's approach to this issue is to describe the design process for plants licensed under Part 50 and Part 52, and then define the engineering products that must be completed at the time of design certification.
The degree of design detail completed will vary depending on structures, systems, and components.
Mr.
Virgilio stated that 10 CFR 52 implies three bodies of design information, namely:
Submitted in the application and certified by rulemaking e
Submitted in the application and not certified and e
e Completed and available for audit.
The flexibility is provided through changes to certified material via 10 CFR 50.12, and. changes to uncertified material after combined operating license (COL) via 10 CFR 50.59.
The staff has examined four levels of detail, the corresponding degree of standardization achieved, compliance with 10 CFR 52, and the safety and economic benefits devised from each.
These four lovels are:
j 364th ACRS Meeting Minutes 4
e Level i The degree of standardization resulting from this level of detail and the certification process will provide identical physical, functional, and performance charac.teristics of all structures, systems, and components affecting safety, except i
for site-specific characteristics.
e Level 2 The degree of standardization.resulting from this level of detail and the certification process will provide physically
- similar, and identical functional and performance characteristics of all structures, systems, and components affecting safety, except for site-specific characteristics.
e Level 3 The degree of standardization resulting from this level of detail and the certification process will provide identical functional and performance characteristics of all systems, structures and components, except for site-specific characteristics.
e Level 4 i
The degree of standardization r9sulting from this level of detail and the certification process will provide at least a product line type of standardization.
The staff does not believe that the design detail necessary to realize a Level 1 degree of standardization is consistent with 10 CFR 52 regarding the content of the application.
Mr. Virgilio commented that Level 1 is probably not commercially 1
. feasible, bemuse the level of deta:.1 required in Level I
certification would make it difficult to ensure continued availability of components with all.the certified attributes over the life of certification.
Level 2 provides the maximum degree of standardization while avoiding to some extent the Level 1 concern.
Level 3 characterizes the industry proposal (incorporating the two-tiered approach) as the staff understands it.
The fourth level of detail (product-line standardization) would not constitute an acceptable application for design certification under the current provisions of 10 CFR 52, because it is not sufficient to allow the staff to reach its final conclusion on all safety issues in a one-step process.
Inspigtion, Tests. Analyses, and Acceptance criteria (ITAnc) - Mr.
E.
Imbro, NRR Mr. Imbro, NRR, stated that the staff licensing review of an application for design certification for all levels will deviate n.
- i 364th ACRS Meeting Minutes 5
from the traditional practice, with the addition of (ITAAC).
M.e staf f believes that the ITAAC will provide reasonable assurtsnce that a plant which references the design will be built and operated in accordance with the design certification.
Information normally contained in procurement specifications and in construction and installation specifications and that would be audited by the staff would be included or referenced in the application for a design certification if it is necessary for the staff to make its safety findings.
In Levels 1 and 2, the entire application will be essentially certified.
In Level 3, the design certification will contain much less detail than in Levels 1 and 2, plus the rulemaking approval of Tier 2 along with the industry-proposed Section 10 CFR 50.59-type change mechanism.
Mr. Michelson commented that it is not clear which document the NRC staff will be reviewing and certifying for the future designs.
For c4 ample, is it the Standard Safety Analysis Report (SSAR) only or does it include the references also?
Mr. Michelson expressed concern that the NRC staff did not perform a comparison with the contents of the EPRI-ALWR Requirements Document.
Dr.
Siess commented that there is no clear indication that maximizing standardization would improve plant safety.
o Two-Tier honroach Pronosed by the Industry - Ms. R. Nease, NRR Ms. Nease, NRR, described the industry "two-tiered" approach for design certification.
In this approach, the top tier certified design would include essential safety performance criteria that, once certified, could only be changed by receiving an exemption through 10 CFR 50.12.
A second tier of material would include more detailed design information.
This second tier would be associated with the rule certifying the design (but not be part of the certification itself) and would include a change process like the current 10 CFR 50.59, that would allow changes Vithout prior NRC review as long as no unreviewed safety question is presented.
The advantages of the two-tiered approach are:
Ability to incorporate technological improvements e
Easier and less costly design process e
Easier to accommodate unavailability of equipment e
Allows owner input to the procurement process.
e u
I 364th ACRS Meeting Minutes 6
l-The disadvantages of this approach are:
e Potential loss of standardization Greater chance for adjudication at COL e
o Less certainty on design details and cost Minor decrease in regulatory stability.
e Pns_entation by Muclear Manaaement Resources Council (NUMARC)
Mr.
W.
Rasin and Mr. Rowden, NUMARC Mr. Rasin, NUMARC, described the "two-tiered" approach proposed by industry for design certification.
He stated that the two-tier approach to the structure of the Design Certification Rule is based on the specific requirements of 10 CFR 52 which distinguish between what will be submitted for NRC review in the design certification application and what will be contained in the Design certification (DC) Rule itself.
Mr.
Rowden stated that the two-tier structure which industry recommends is simply a means for formatting and documenting in the DC Rule the certified and the non-certified parts of the design, and for specifying the change mechanisms governing each in accordance with the 10 CFR 52 requirements.
The first tier would contains A description of the certified design based on SSAR section e
1.2, with detail comparable to that in current SERs.
The full array of inspections, tests, analyses, and acceptance e
criteria required by 10 CFR 52.
The second tier would Reference the entire SSAR design description.
The SSAR is the e
primary technical document of the design certification application and will be the basis for the NRC's Final Design Approval and Design certification reviews. By referencing the SSAR in the DC Rule's second tier, the NRC would document the features and commitments that were the basis for NRC approval (beyond those certified in the first tier) and document the
" matters... resolved in connection with the issuance... of a design certification" (per $52.63 (a) (4)).
The second tier would contain also the
" validation attributes," which the NUMARC report proposes as a bridge to demonstrating compliance with those first-tier acceptance criteria that are not readily measurable or otherwise verifiable by direct field inspection or test.
364th ACRS Meeting Minutes 7
Mr. Minnick expressed concern regarding the effect of common-mode failures if standardization would be encouraged.
He noted also that standardization of nuclear units is inherently limited in any event by differing site characteristics and inevj, table variations in operating experience.
Mr. Michelson commented that differences in local pipe and cable routing to support vendor-specific component configurations should be looked at very carefully.
Af ter further deliberations, the Committee provided a report to the Commission as discussed in Section X of this document.
III.
NRC Incident Investiaation Team (IIT) Briefina on the March
- 20. 1990 Voatle Loss of Vital AC Power / Loss of Decay Heat Removal (DHR) Event (Open)
[ NOTE:
Mr.
P.
Boehnert was the Designated Federal Official for this portion of the meeting.)
Mr. Carroll, Chairman of the Plant Operations Subcommittee, stated that representatives of the IIT that investigated the Vogtle loss of vital AC power / loss of decay heat removal (DHR) event would brief the Committee regarding the findings and recommendations resulting from their investigation.
Mr. Jordan, Office for Analysis and Evaluation of Operational Data (AEOD), introduced the IIT members present and noted that this IIT differed from past IITs in that representatives from industry (INPO) were included as members of the Team.
A set of recommendations for NRC follow-up has been transmitted to the EDO and the Regional Offices.
The staff will discuss the status of their follow-up activities with the ACRS.at a future meeting, if l
the Committee so desires.
Mr.
Carroll suggested that NRC consider participation by 1
l repressntatives of industry for Augmented Investigation Teams (AITs) as ws:11, in order to reduce the duplicative effort l-experienced oy licensees when both NRC and the industry (INPO) l converge on a plant to investigate an incident.
Mr. Jordan indicated he would take this comment under advisement.
l Ettp3.Dtation by the IIT - Mr. A. Chaffee, Region V Mr. Chaffee, IIT Team Leader, provided the details of the IIT investigation of the Vogtle event.
He stated that on March 20, 1990, Vogtle Unit i lost all vital AC power and DHR for 37 minutes while at mid-loop conditions.
The initial power loss was caused by a service truck backing into a power pole resulting in loss of the single vital AC transformer in operation (the other i
l 364th ACRS Meeting Minutes 8
transformer was being serviced).
The only available emergency diesel generab,or (EDG) started on demand, then tripped; successful j
restart of this EDG took ~ 37 minutes, after override of various trip functions.
The reactor coolant system (BCS) temperaturo increased 40*F during the pr r loss.
factors Leadina to the Incident o
Risk manacement of Shutdown activities was lacking:
The licensee did not control access to switchyard; movement of the truck involved in the accident was not controlled.
There was a history of incidents at this site related to switchyard work causing power outages.
The conflagration potential of the truck in question was not considered.
In response to a question from Mr. Michelson, Mr. Jordan stated that the issue of generic concern with potential fire sources in plant protected areas is an item for staff follow-up.
Kinimum electrical cower sucolv redundancv Maintenance wa e
unnecessarily scheduled on one of the two transformer.
supplying vital power while at mid-loop conditions. Technica specifications allow minimum (2
of 4) electrical powc supplies during any shutdown condition.
e Diesel aenerator reliability was coor:
The sensor ("Calcon" water
- jacket temperature sensor) responsible for the trip of the available EDG had evidenced an extensive failure history (64 failure events since 1985).
The licensee had not pursued an aggressive root cause investigation of this problem.
In response to a question from Mr. Carroll, Mr. Chaffee said the calcon sensor is reliable, providing one accounts for both its sensitivity to foreign materials and the peculiarity of the. calibration procedure required.
Dr. Kerr raised a concern that this one incident is not a reliable indicator that the diesel generator reliability was unacceptable. Mr. Jordan agreed, from a numerical standpoint.
However, he made the point that this problem was a common-cause failure mechanism that wa:3n't sought out and fixed; further, the calcon sensor problem was known to the industry.
l
- i 364th ACRS Meeting Minutes 9
Licensee Handlina of the Event generally sound resoonse instincts:
The licensee response to e
the loss of DHR was prompt and inclusive (closed containment and RCS).
Actions were taken in accordance with the lessons learned from NRC Generic Letter 88-17 concerning loss of DHR at mid-loop conditions; for
- example, redundant level indication and core exit thermocouples were available.
e command. control, and communications:
There were problems encountered in notification of offsite authorities due to the loss of AC power and the licensee's unfamiliarity with backup and alternate notification methods.
The transition between the licensee's outage and emergency planning organizations was also difficult.
e Containment eauipment hatch closure issues It took the licensee 79 minutes to close the equipment hatch, which was ample time for this event;
- however, the licensee had identified 54 minutes as the necessary closure time for the limiting event.
Further, no consideration was given to the impact of loss of (all) power on the hatch closure time (closure takes 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> without power).
Mr. Carroll asked when the Committee can expect a staff response to its request for information on the capabilities of licensees to perform ef ficacious closure of containment equipment hatches, as necessary.
Mr. Jordan said that this item is am IIT follow-up issue and the results of the follow-up will be reported to the ACRS.
e EDG suDoort systems:
The design of the EDG system impedes diagnosis of problems; for example, the "first out" alarm indications were misleading, alarm indications are not recorded.
Interaction of the EDG control and load sequencer circuitry is complex.
The licensee's understanding of the operations of the control and sequencer systems was incomplete.
Potential Generic Imolicationg e
Shutdown risk manaaement needs to be imoroved:
Over the last 35 years, the occurrence rate for loss of RHR at mid-loop conditions and/or loss of power during shutdown has been - 6 ovents/ year.
Existina analyses and auidance on loss of RHR at mid-loon and loss of power at shutdown needs to be incorporated into licensee's trainina and crocedures:
For example, the lessons
. - = _
364th ACRS Meeting Minutes 10 learned from Generic Letter 88-17 need to be universally adopted by all licensees.
o Additional analysis of shutdown coolina confiaurations is needed:
The effect of RCS configurations on core cooling options is not fully understood.
e Classification eroblems related to emeraency olannina are evident:
Regarding classification of an event by a licensee, guidance is lacking both to ensure uniformity in the selection of the proper severity level and the applicability of the emergency plans to operations during cold shutdown.
e Technical specifications need to be reviewed:
The current technical specifications are not based on risk analyses.
Licensees' understandina of EDG system and trio indication slesian should be reviewed Mr. Jordan indicated that the principle issue of concern to NRC is the efficacy of shutdown risk management.
An Action Plan to address the Team recommendations is being prepared and the staff will brief the Committee on the particulars during a future ACRS Meeting.
Mr. Carroll praised the IIT members for an ex:ellent investigative effort, and a fine report as well.
IV.
Generic Issue B-56, " Diesel Generator Reliability" (Open)
(NOTE:
Dr. M. El-Zeftawy was the Designated Federal Official for this portion of the meeting.)
Mr.
- Wylie, Chairman of the AC/DC Power Systems Reliability
. Subcommittee, briefed the Committee regarding the resolution of Generic Issue (GI)
B-56,
" Diesel Generator Reliability."
He indicated that GI B-56 is related to the Station Blackout Rule (10 CFR Part 50, Section 50.63).
The staff issued Regulatory Guide (RG) 1.155, " Station Blackout," to provide guidance for compliance with this Rule.
RG 1.155 identifies the need for ensuring reliable sources by means of a reliability program designed to maintain and monitor the reliability level of each power source over time for assuranco that selected reliability levels of 0.95 or better are being achieved.
Presenta1.joA bv the RES Staff - Mr. W. Minners, RES Mr. Minners, RES, pointed out that resolution of GI B-56 will be accomplished through the issuance of Revision 3 to RG 1.9, "Solection, Design, Qualification, Testing, and Reliability of i
364th ACRS Meeting Minutes 11 Emergency Diesel Generator Units Used as Class 1E Onsite Electrical Power Systems at Nuclear Power Plants."
This RG accomplishes the following:
Integrates the pertinent guidance previous [y addressed in RG e
1.108, Revision 2, and Generic Letter 84-15.
Defines the principal elements of an EDG reliability program that is for the most part consistent with current industry practices.
Defines testing requirements, eliminates cold fast starts, and e
reduces accelerated testing.
As a result of the staff's discussion of this matter with NUMARC's B-56 Working Group, NUMARC has revised its NUMARC-8700 document,
" Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," Appendix D, to provide guidance similar or identical to the reliability section of Revision 3 to RG 1.9, but in more detail.
The staff indicates that Revision 3 to RG 1.9 utilizes definitions from INPO's U.S.
Plant Performance Indicator Program (PPIP) to enhance reporting consistency.
Revision 3 to Regulatory Guide 1.9 will be applied to all operating plants for purposes of monitoring EDG reliability levels and reviewing EDG reliability programs with respect to meeting the Station Blackout Rule.
Mr. Serkiz, RES, stated that resolution of GI B-56 through the issuance of Revision 3 to RG 1.9 will not introduce any regulatory requirements beyond those currently required for compliance with
.the Station Blackout Rule.
Mr. Serkiz pointed out that EDG maintenance problems continue to exist despite reported high levels of EDG reliability and he cited as examples the problems experienced at Vogtle, Cooper, D.C. Cook, l
Calvert Cliffs, and Zion plants.
Mr. Minners provided histograms of EDG failures on demand for 1976 through 198p.
Although the average failure on demand observed is about 2x10', there is a significant spread from the highest to the t
lowest demand failure rate.
There were 145 instances in which multiple EDGs were simultaneously failed, unavailable, or showed some degradation.
Dr. Lewis expressed concern regarding the staff's failure for optimal use of the collected data on diesel failures, and commented l
l
364th ACRS Meeting Minutes 12 that it is not clear what problem the reliability program proposed by the staff is supposed to solve.
l In Revision 3 to RG 1.9, the staff references the.. revised Appendix D to the NUMARC-8700 document, as appropriate, and includes guidance for an EDG reliability program (Section C.6). The revised Appendix D was submitted on May 2, 1990, and has been reduced in scope.
A 50.54 (f) Generic Letter (GL) has been prepared by the staff to determine the course of action that licensees and applicants plan to pursue.
The staf f met with the CRGR on July 25, 1990 to discuss this issue.
The CRGR recommended that Section C.6.1 (monitoring) remain in the main body of Revision 3 to RG 1.9.
However, the illustrative examples and considerations contained in Sections C.6.2 through C.6.7 should be included in an Appendix to this RG.
The CRGR recommended also that the 50.54 (f) GL be revised to more clearly note that NRC is requesting a response as to whether licensees plan to implement guidance provided, or to identify other actions that they plan to implement for monitoring and maintaining EDG reliability.
Dr. Kerr questioned the staff's proposal to use Revision 3 to RG 1.9 as a tool for backfitting purposes.
He questioned also the enforceability of such an approach and indicated that the staff is attempting to advocate and impose a maintenance rule on the industry.
Comments by..NUMARC - Mr. W. Rasin, NUMARC Mr. Rasin, NUMARC, stated that NUMARC has revised the NUMARC-8700 document, with the following changes:
e Initiative 5 of Appendix D of the NUMARC-8700 document has been revised to include monitoring of EDG reliabilities against the target reliability selected for Station Blackout, and also to address-actions for a problem EDG experiencing 4 I
or more failures in the last 25 demands.
NUMARC has revised Appendix D to the NUMARC-8700 document-and the current version has been reduced in scope.
The previous guidance dealing with surveillance
- needs, performance monitoring of important EDG parameters, data
- systems, maintenance, failure analysis and root cause investi5 : ion, problem closecut and methodology for determining progr. amatic deficiencies is now included in a topical report entitled,
" Effective Elements of an EDG Reliability Program."
This Topical Report will not be submitted to the NRC for review.
NUMARC intends to provide this Topical Report to utilities, as needed.
l
i 364th ACRS Meeting Minutes 13 Mr. Rasin stated that HUMARC opposes strongly the NRC staff's proposal to issue a
50.54(f)
GL, and believes that it is unnecessary and unwarranted in light of the. established high industry performance. He pointed out that invoking 10 CFR 50.54 (f) in a GL makes the guidance in Revision 3 to RG 1.9 de facto regulatory requirements.
He stated also that the staff has not satisfied the backfit rule requirements, and recommended that the staff focus on programs instead of performance.
Mr. A. Marion, HUMARC, stated that the indJstry actions addressing resolution of GL B-56 provide the NRC staff withs e
A docketed commitment to maintain the chosen target reliability of 0.95 or 0.975 A commitment to a standard set of trigger values, acceptable e
to NRC, from which to monitor EDG target reliability.
NUMARC believes that GI B-56 has been resolved satisfactorily by the industry without the need for regulatory action.
Dr. Lewis expressed concern regarding the lack of expertise among the NRC staff to handle mathematical and statistical interpretation of the. data.
After further deliberation, the Committee provided a report to the Commissi; m on this matter, as discussed in Section X of this document.
V.
Severe Accident Risk Reoort (hUREG-1150) (Open)
(NOTE:
Mr. D. Houston was the Designated Federal Of ficial for this l
portion of the meeting.)
l
,Dr.'Kerr, Chairman of the Severe Accidents Subcommittee, said that the Committee had held a number of meetings (including subcommittee j
l meetings) over the past few years to discuss NUREG-1150, " Severe i
Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,"
l dated June 1989.
In a report to the Commission, the Committee l
recommended that NUREG-1150 be subject to a peer review.
On July 7,1989, the Commission established the Special Committee to Review the Severe Accident Risk Report.
Dr. Kerr introduced Dr. Kouts, Chairman of the Special Committee, and noted that the review by the l
Special Committee was nearly complete.
He also indicated that a l
separate review had recently been completed by an ANS committee.
Review of NUREG-1150 by the Special Committee - Dr. Kouts Dr. Kouts, Chairman of the Special Committee, reviewed the charter and membership of the Special Committee.
He then discussed the key
364th ACRS Meeting Minutes 14 features of the report and the methodology applied during the study.
He noted that the special Committee had adopted the terminology of the Probabilistic Safety Analysis (PSA) rather than the PRA.
Dr. Kouts indicated that the Special Committee found NUREG-1150 to be a good, extensive, state-of-the-art report and declared it to be a large improvement over the first draft and WASH-1400.
Even so, he indicated that several weaknesses had been identified by the Special Committee, as noted below:
Risks from events that occur during lower power or shutdown o
conditions were not analyzed.
Risks from events resulting from loss of instrument air and o
steam line breaks were not considered.
Pressure vessel failures were not considered.
e Errors of commission were not well done.
Recovery actions were not fully considered.
e Dr. Kouts discussed the Limited Latin Hypercube (LLH) sampling
- process, XSOR codes, and seismic analysis.
He presented a description of the expert opinion process and indicated that, while this process was done as well as one could expect, ';he Special Committee still had some concerns about the process and the balance of personnel on the panels.
He noted that the human reliability analysis (HRA) was the weakest component of the study.
In summing up this section, he listed the following recommendations:
NUREG-1150 is a good report and should be published as soon e
as possible, improvements to the report could be made after it has been published.
External events should always be included in the PRAs since e
risk contribution from such events is substantial.
seismic modeling issues should be resolved.
e Human reliability analysis should be developed.
Care should be exercised in using the XSOR codes.
e Dr. Kouts concluded his presentation with a discussion of cutoff levels, means or medians, and a comparison of fission product releases as calculated in WASH-1400 versus NUREG-1150.
For cutoff levels, the Special Committee recommended a cutoff on probabilities at 1E-07.
Regarding means or medians, Dr. Kouts indicated that
e.+-
364th ACRS Meeting Minutes 15 both are displayed in the current report and both are useful l
depending on how you intend to apply them.
For fission product
- releases, all the figures discussed by Dr.
Kouts showed a
significant reduction in releases as calculated for NUREG-1150 as compared to WASH-1400.
Dr. Wilkins questioned the voting of members on the Lawrence Livermore National Laboratory (LLNL) seismic panel where 80 percent of the panel gave no value of goodness for a model developed by another member of the panel.
Dr. Kouts indicated that the Special Committee certainly recognized the concern but never came to a e
conclusion on this matter.
Dr. Siess expressed an impression that one of the objectives of the i
expert panel process was to see how large one could make the uncertainties.
Dr.
Kouts disagreed and indicated that the objective was only to see what the uncertainty range was.
Dr.
Siess further asked if specific areas of further research had been identified.
Dr. Kouts discussed a few issues where this was the case, for example, testing the EPRI seismic methodology against observed earthquakes.
e Dr. Lewis asked if conclusions drawn from this sample of five plants, especially in regard to meeting the safety goals, could be extended to the population of operating plants.
Dr. Kouts replied J
that he would not go that far in extending the results.
This also seemed to be the position of the ANS committee.
Mr. Minnick asked if the Special Committee had developed a position on containment venting to relieve high internal pressures that might result from a severe accident.
Dr. Kouts stated that the Special Committee had not taken a position on this issue, either way, but.his personal opinion was that one would be reluctant to use such a system.
In his closing remarks, Dr. Kerr proposed, and the Committee agreed, to prepare a draft report to the Commission on NUREG-1150 and submit it to the Committee for consideration during the September 6-8, 1990 ACRS meeting.
The Committee also suggested
'that the ACRS Subcommittees on Extreme External Phenomena and on Severe-Accidents hold a joint meeting to discuss the adequacy of consideration of seismic and fire issues in NUREG-1150.
VI.
EPRI Reauirements for Advanced LFRs (Open)
(NOTE:
Dr. M.
El-Zeftawy was the Designated Federal Official for this portion of the meeting.)
Mr. Wylie, Chairman of.the Improved LWRs Subcommittee, noted that the Electric Power Research Institute (EPRI) in conjunction with
~
+ -
m W
i 364th ACRS Meeting Minutes 16 the utility-sponsored ALWR Steering Committee, has prepared a compendium, referred to as the EPRI-ALWR Requirements Document, of technical requirements applicable to the design of an ALWR plant.
The document represents a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond.
Mr. Wylie said that the NRC staff has prepared draft SERs on Chapters 1 through 5 of the EPRI-ALWR Requirements Document.
Currently, there are approximately 160 open items that have to be resolved.
The Improved LWRs Subcommittee met with the NRC staff and EPRI representatives on July 11, 1990 to discuss Chapters 1 through 5.
This matter was also discussed during the July 12, 1990 full Committee-meeting.
The Committee decided then to continue its discussion at the August 9-11, 1990 ACRS meeting.
l Status of EPRI-ALWR Recruirements Document - Mr. G. Vine, EPRI Mr. Vine, EPRI, briefed the Committee regarding the status of the EPRI-ALWR Requirements Document.
He stated that the goals of the ALWR program for future nuclear generation are to achievet Significant improvements in safety for both public acceptance i
and utility / investor confidence Stabilized regulatory basis e'
Standardization that meets utility and regulatory requirements e
Schedules that meet urgent utility baseload capacity requirements.
t Mr.
Vine stated that the utilities must lead in the ALWR development, and the key to standardization is customer consensus F
via' common requirements.
The ALWR Requirements Document presents fundamental engineering improvements in safety and availability, through increased margin and simplicity.
These improvements would not. have. been. achieved by regulation.
Mr. Vine indicated that there is international participation on the EPRI-ALWR Requirements Document, and currently there are agreements on this matter with Taiwan Power Company, Korea Electric Corporation, Kanshi Electric Power, the. Netherlands, ENEL Italy, and EdF France.
t Dr.'Siess commented that it is not clear that standardization, per se, would improve safety and reduce the risk of an accident.
j
'Mr. Wylie asked if the EPRI-ALWR Requirements Document would result in a reduction of limiting condition for operation (LCO) issues.
A representative of EPRI stated that he does not have an answer to this question.
1:
l 364th ACRS Meeting Minutes 17
]
Mr. Michelson expressed concern regarding the legal standing (if any) of the EPRI-ALWR Requirements Document and asked what the staff's commitments are regarding this issue.
The staff responded that currently there is no requirement or specific guidance from the Commission regarding the legal standing of the EPRI-ALWR Requirements Document.
i Mr. Michelson commented that it is not clear if the fire or internal flood events are considered as severe accident issues, and g
asked how they will be handled in the PRA analysis.
Mr. Vine indicated that the EPRI Requirements Document consists of three volumes.
Volume I, Executive Summary, is a management-level synopsis of the Requirements Document, including design objectives
.and philosophy, the overall physical configuration and features of a future plant design, and the steps necessary to take the proposed design criteria beyond the conceptual state to a completed, functioning power plant.
Volume II addresses the evolutionary designs that include overall performance and design requiremente"(Chapter 1) and requirements for systems and structures (Chapters 2 through 13).
Volume III addresses the passive plant designs.
The EPRI-ALWR Requirements Document applies to the entire nuclear plant and incorporates resolutions of generic safety issues and optimization issues.
The document reflects the industry consensus on principal safety, performance, and design issues.
Mr. Vine stated that the EPRI-ALWR Requirements Document enhances accident prevention and improves mitigation capability.
Examples of accident prevention features ares Improved station blackout capability e
Improved decay heat removal system reliability Increased RCS and secondary inventory e
Improved reactor pressure vessel integrity (ring forgings)
~ o Improved embrittlement resistance.
e Examples of the mitigation capability are:
More stringent containment integrity requirements e
e Periodic leak check Cavity / lower drywell floor area sized for spreading of core e
debris to promote cooling Reactor depressurization to avoid high-pressure melt ejection
364th ACRS Meeting Minutes 18 PWR cavity arranged to de-entrain core debris to avoid debris e
transport to the upper compartment.
EPRI has committed to an end-of-August 1990 transmittal of evolutionary plant requirements (Rev.
1) and passive plant requirements (Rev. 0).
VII.
NRC Research Procram on OraaniEational Factors (Open) r (NOTE:
Mr.
H.
Alderman was the Designated Federal Official for this portion of the meeting.)
Mr. Carroll noted that the Human Factors Subcommittee met with representatives of the NRC staff on July 31, 1990 to discuss two matters:
NRC research program on Organizational Factors Procedure violations resulting from the Chernobyl accident.
e The presentation to the full Committee on procedure violations will be scheduled at"a future date.
Mr. Carroll remarked that, during the presentation yesterday by Dr.
Routs on NUREG-1150, the Committee heard the views of the Kout's committee on the importance of management and organizational factors in the safety arena and the need to find better ways to I
quantify the impact of organizational factors in the evaluation of the safety.of nuclear power plants.
Raghground of oraanisational Factors Research - Dr. F. Coffman, RES Dr. Coffman, Human Factors Branch, RES, discussed the background of the organizational factors research.
Last summer, the staff briefed the Commission on the Human Factors Research Program Plan.
The Commission asked that the staff provide the Commission'with a progress report on organizational factors research with emphasis on the focus and direction of the program.
Prior to meeting with the Commission, the staff will brief the ACRS and would appreciate any comments.
Dr. Coffman noted that, in 1988, the EDO concluded that:
" problem plants are typically associated with weak and poor management, while good plant performance normally results from good corporate and _ plant management. "
The NRC uses the information from the SALP program and the semiannual senior management meetings to evaluate the effectiveness of utility management. The SALP program receives input from the Region-based inspections, team inspections, safety evaluation reviews, and follow-ups and daily contacts with utility management. The semiannual management meetings receive input from:
J 364th ACRS Meeting Minutes 19 SALP, inspection findings, performance indicators, PRA insights, and operating events and their follow-ups.
Dr. Coffman stated that the RES role is to provide.for improvements in these programs, and to confirm that the programs are based on a firm technical basis.
Dr. Shewmon asked who attends the semiannual management meetings, where are they held, and what the output is.
Dr. Coffman replied that the semiannual meetings are held either at headquarters or in 1
the regions and they involve Regional Administrators, the Director of NRR and some of his deputies.
The output of the meetings is usually a grading of the plants.
Some of the plants may require increased regulatory attention. A diagnostic evaluation team visit might be a result.
Dr. Kerr remarked that the licensed operators are subject to a rather significant screening process by the NRC.
He noted that they have a licensing process and a relicensing process,and that they are under almost constant surveillance as they operate.
He noted that when - they make mistakes they are subject to anything from a fine to a possible prison sentence.
He noted that this might have some influence on their performance and asked if this was considered. Dr. Coffman responded that the part of the program that looks at the operators is a project that deals with hew to measure team performance.
This looks at licensed operators in the control ~ room performing as a team.
The purpose of that project is to develop measures of their performhnce.
Mr. Michelson noted that the human error involved in maintenance operations could have a finite probability of leading to core :
damage.
He asked if this was part of the organizational factors study.. Dr. Coffman said that it was part of this study.
Q11311s of the Oraanisation Factors Research Program - Dr. T. Ryan, RES Dr. Ryan, RES, said that the organizational research is taking a top-down look at what is going on at the plants, from the plant level down to the natural subdivisions but not necessarily-including individual workers.
He stated that they are interested in:how the rest of the organization influences what a particular individual may-or may not do.
.Also, they are interested in the influences of. technical support programs, such as maintenance and-training, on overall performance.
Dr. Ryan said there were three reasons for doing the research:
l t
364th ACRS Heeting Minutes 20 RES has received requests from other offices and the regions e
to look at organizational factors with regard to:
PRAs, the accident management area, some reviews, inspections, and evaluation activities ongoing in NRR and AEQD and in support of AEOD's performance indicator program.
Industry experience revealed that there were institutional and e
management related issues in many incidents.
Organizations outside RES have strongly recommended work in e
+
this area, Dr. Ryan stated that the program is composed of 11 individual projects that fall into three activities j
e Process and outcome modelina.
This concerns methods for characteriting not only the structure, but also the operating characteristics of the human element, the human organization of a nuclear power plant,and its subdivisions, e
Data collection.
There are two types of data collectient Active, where data is collected at the plant sites Passive, where data is derived from the data collected by the NRC.
Dr.
Ryan discussed Nuclear Organization and Analysis Concept (NOAC).
This is based on a request from Dr. Murley, when he was Region I Regional Administrator, for a methodology to integrate organization and management factors into a PRA.
This concept was tested at the Pittsburgh fossil fuel facility in California.
The NOAC'was also tested at Diablo Canyon.
The plant management at Diablo Canyon is using some of this information to make some management-related decisions and changes.
Dr. Ryan discussed some work at Pacific Northwest Laboratory (PNL)
-to look at measures or indicators of team skills, focusing on the control room teams during off-normal operations.. Some of this work was done at the Technical Training Center to promote team performance.
A lot.of simulator data was collected and evaluated and this information will be published in-PNL Report 7250.
In response to a question by Dr.
Siess as to what was an
. organizational factor, Dr. Ryan said what they are trying to do is to take the characterization of the management and the organization and, given that, determine the likelihood that mistakes are going to be made.
L There was considerable Committee discussion on how the results of I
this research would be used.
Dr. Ryan was asked if the PRA numbers L
l L
a
364th ACRS Meeting Minutes 21 would be changed.
Dr. Ryan said that the staff's effort was to develop a methodology.
Mr. Minnick commented that the most valuable par.t of the program will be the development of insights as to what is important in management and then communication about them to plant management.
Dr. Ryan talked about the AEOD performance indicators.
These are direct indicators of safety.
NRC's Office of RES has been asked to look at the feasibility of developing " intermediate indicators" that would forecast how the direct indicators would change in about
- a year.
A linear analysis model has been developed that relates structural and operating characteristics of an organization with outputs and makes predictions about safety.
Dr. Ryan summarized some of the ongoinc works Trainina arga.
The staff ir looking at the remediation and e
instructor student ratios.
o Nonnuclear indicators.
The staff is looking at information from outside the nuclear industry.
One example is mechanical integrity indicators that are basically related to maintenance.
A project on the relationship between organizational factors.
VIII.
Solenoid Valve case Study (Open)
'(NOTE:
Mr.
H.
Alderman was the Designated Federal Official for this. portion of the meeting.)
Preamble - Mr. T. Novak, AEOD Mr. Novak, AEOD, noted that the staff was presenting a preliminary case study. to the ACRS on the operating experience of solenoid-operated valves (SOVs).
The study has been submitted to valve manufactures, NUMARC, INPO, EPRI, NRC offices, and the regional offices for comments.
The commenters were asked to comment only on the-body of the report, not the conclusions.
The impetus.for the study was an event at the Brunswick station where two sets of SOVs failed to close upon receiving the signal and there was no containment isolation.
Mr. Carroll asked if the recommendations were so " ironclad" that the conclusions would not be changed even if they received worthwhile comments.
Mr. Novak assured him that this was not the caSO.
i
t c'
364th ACRS Meeting Minutes 22 ABOD Case Study on 80V F_g.ilures - Dr. H. Ornstein, AEOD Dr. Ornstein, AEOD, said the case study was started during January j
-1988.
The licensee event reports (LERs) and the Nuclear Plant Reliability Data System (HPRDS) data bank wefe searched for evidence of SOV failures. There were 900-1,000 SOV failures during the period from 1985 to 1988.
He noted that SOVs are used in almost all safety systems in all nuclear power plants in this country.
Mr. Michelson pointed out that SOVs can be used for direct control or indirect control of a process.
Fe asked if both methods of control were considerod.
Dr. Ornstein stated that both methodt of control'had been considered.
1 Dr. Ornstein categorized the nature of SOV failures:
e Damian nonlication:
One example of this is not properly 1
considering ambient temperature.
He cited one case where localized steam caused heating of the colenoid and failure of the valve. A second example concerned electrical heat-up of the coils and malfunctioning of the valve.
e Maximum ooeratina pressure differential:
The valve will not be able to function reliably if the pressure differential from the inlet to the exit port is higher than a prescribed value.
If the pressure differential from one side to the other is greater than.that specified in the design, the valve will pop at the wrong tinc.
e Directional Sovst In this case, the valves are installed t
contrary to the proper direction of installation.
Maintenance: Dr. Ornstein cited the paucity of manufacturers' e
information on maintenance for SOVs.
Some SOV manufacturers give almost no information whatsoever to the users as to what-
.to expect of their valves and what to do with regard to life expectancy under certain conditions.
He noted that many failures have been reported in which the licensees have rebuilt SOVs unsatisfactorily.
Mr. Michelson asked if the valves were rebuilt based on a schedule or after they have failed to function properly.
Dr.
Ornstein said it varies from manufacturer to manufacturer.
The time varies from a recommended time to a suggestion that the valves be rebuilt on a periodic basis.
Some manufacturers sell a kit to enable the utility to rebuild the valve in situ.
s l-I i
w
,e--
w
--7---
___~
h g. s, i
364th ACRS Meeting Minutes 23 Dr. Lewis asked which parts of the valve normally degrade.
Dr. Ornstein said that seals, coils, and springs degrade with time.
e contamination:
This is where the valve is exposed to a contaminant such as oil or debris in an air line.
He cited a case in the North Anna plant where the air system was unintentionally washed with service water and about 48 SOVs failed.
Lubrication:
Many of the valves are lubricated at the f actory prior to shipping.
Valves are lubricated during rebuilding and during preventive maintenance.
Improper lubrication has been responsible for valve failures.
Qualificatica: There br<e been SOVs selected for applicationa where they are not capable of operation under the operating conditions.
e Manufacturina defects: There have been errors in the assembly of the valves,. choice of lubricants, choice of elastomers, and internal clearances.
i Dr. Ornstein listed areas of concern e-Potential for common-mode failure:
The SOV failures cut across systems as well as individual trains of a particular
- system, gamoonent failurA_ rates:
The component failure rates appear e
to;be higher than previously estimated.
e Incioient failure detection:
There are no techniques available today to detect incipient failure of SOVs.
Lack of manufacturer cuidance.
There is a limited amount of I
e information available regarding maintenance, replacement, and longevity._
Dr. Ornstein listed some of the problem areas:
l The rebuilding of_the valves has led to many problems e
i e
Contamination has contributed to common-mode failure j
Improper or wrong lubrication has created problems e
i e
Surveillance has not detected failed valves The manufacturers are not receiving operational information.
e
1 364th ACRS Meeting Minutes 24 Mr. Novak said that they expect to issue the final report of thic
)
case study in September and would probably return to brief the ACRS in October 1990.
IX.
En. actor Ooeratina Exoerience and Events at Nuclear Power Plants (Open)
(NOTEt Mr.
P.
Boehnert was the Designated Federal Official for i
this portion of the meeting.)
Mr. Carroll, Chairman of the Plant Operations Subcommittee, noted that Mr. Benaroya, AEOD, would present the details of the feedwater line rupture event that occurred at Unit 1 of the Finnish Loviisa (PWR) nuclear power plant.
.Mr. Benaroya noted the following regarding the subject events Following the May 28, 1990 event, Drs. Remick and Kerr and e
himself visited the plant.
The Loviisa site has two Russian-PWR VVER units, each i
generating 465 MWe.
The containment is of ice-condenser design.
German instrumentation and controls are used.
Each plant has six (horizontal) steam generators and two turbines.
Unique ECCS features includet two (each) accumulators / hot and cold leg; containment spray and hydrogen ignitors are also I
- present, i
-Particulars of the feedwater system were noted.
Figure 1 e
L shows the layout of this system, including the rupture site.
The break occurred immediately downstream of a flowmeter
.(sharp-edge design) that was part of a (German) spool piece.
The break was guillotine in nature, occurring with the reactor at ~60% power.
Erosion / corrosion was given as the cause-of one millimeter the rupture; the piping was eroded to (original thickness was 18 mm) at the rupture location.- The erosion terminated at the weld junction of the spool _ piece to the (Russian) piping.
Discussion ensued regarding the material composition of the g
spool piece relative to the piping.
NRR indicated that the l-piping contained chrome, which was lacking in the spool piece.
Dr. Shewmon indicated the presence of chrome stabilizes the oxide
- coating, making the metal more resistant to l-erosion / corrosion.
l l
The break flow was automatically isolated on the suction side, e
and was manually isolated on the discharge side.
Approximately 15,000 gallons of water were discharged, with L
l
~
Yr' J$
~
j
~
~
kn
~
r a
m t
1 re a
D s
t i
a n
C i
w v
de o
L eF e
s h
~
2 t
0 s
f r
s o
o
{
t p
a yr m
m r
re u
e e
at p
t
)0 n
ia s
e lws r
y g
idp e
s xem t
m ueu a
r a
Afp w
e d
t e
j e
a t
e w
S D
F d
f C
ee F
kna pi t
1 lI
) i r
e Ja e
r t
u r
a g
gI w
i lI d
F eeF
~
I I
g8 c5-
"F f w-l I
i
~bma I
l!l
e 364th ACRS Meeting Minutes 25 some local-damage to cabling.
No safety-related system functions were lost, and the plant was shut down without
~
further' incident.
No one was injured.
In response to a questic:. fram Mr. Michelson, Mr. Benaroya said he didn't know either the extent of physical ceparation the pipe saw at the bceak site, or if the fluid exited in an alluvial fan, y
e The sharp-edge flowmeter orifice was identified as the cause of the erosion.
Upon inspection, 9 of 10 of the similar flowmeters also needed repair.
All will be replaced with a different design (and different material) in the near future.
C Inservice inspection. (ISI) was being conducted on the secondary system - piping prior to the event, but for some unknown reason the flowmeters were not inspected.
The ISI L
program will be expanded on both units, beginning with the next refueling outage.
e-Mr.
S.
Koscielny, NRR, noted that he had performed some E
analyses of the Loviisa feedwater piping system, using the EPRI " Checkmate" code, to determine the sites most susceptible to erosion / corrosion.
He said the code results tuaverged on the flow orifices as being the sites-expected to 2xperience the highest rates.of' erosion / corrosion.
7 Prior to recess,-there was some discussion of the details of the design of'the Russian steam. generators.
X.
Executive Sessiong (Open)
-1.
Reoorts to the CommissioD e-Level of' Detail Recuired for-Desian Certification Under Part 52 (Report to Chairman Carr, dated August 14, 1990)
The committee recommended that the commission adopt the
. Level 2 option cc lineated in SECY-90-241,
" Level of Detail Required for Design certification Under Part 52."
iMwever, it does not believe that all of the information required under Level 2 should be included in the design aertification rule.
The Committee suggested that some form of the two-tier approach proposed by NUMARC is essential and that the staff and industry. develop criteria to define the division between the two tiers.
Thc Committee expressed its willingness to review such criteria, as progress is made in this effort, and report to the Commission if the Commission wishes that this be done.
p
o.,c
?4 LP. hCRS-Meeting Minutes 26 Additional retarks provie by ACRS member Lawrence Minnick were ajpended to this report.
e Erocosed Resolution of Generic Safety Issue B-56. " Diesel Generator Reliability" (Report to Chairman Carr, dated August 14, 1990)
The Committee stated that the proposed resolution of Generic Issue B-56 includes unjustified imposition of i
maintenance requirements on licensees.
The Committee WL noted its belief that the licensees' commitments to monitor and maintain diesel generator reliability above the chosen target levels, coupled with the industry initiatives, are sufficient to ensure acceptable diesel generator reliability under the Station Blackou". Rule.
The Committee recommended several changes in the proposed Revision 3 to Regulatory Guide 1.9 and its promulgation and use.
-Additional comments provided by ACRS member Harold W.
Lewis were appended to this report.
e-NRC Research Procram on Oraanizational Factors (Report
,to. Chairman Carr, dated August-' 16,.1990)
The Couinee stated that the NRC research on i
Organizational Factors appears to'be focused on agency needs and can make a contribution to future improvements in the : effectiveness of nuclear. power plant organi-
-tions.
However, it expressed concern-that this research i
program seems to be-directed toward.the need to consider
~
. operator performance in PRAs in a m o r e.- q u a n t i t a t i v e manner.
The Committee noted its belief that more emphasis should be plar.ed on communicating to nuclear
. power plant licensees the insights developed on effective managerial approaches.
The: Committee-stated that i
continued support ara encouragement for this research-program from the Commissioners and the NRC' staff
. management will be necessary, and that it would follow the progress of this research program with interest.
Additional comments provided by ACRS member Harold W.
Lewis were appended to this report.
l
s
,.364th-ACRS Meeting Minutes 27 Euonort Staff for ACNW Activities (Joint ACRS/ACNW Report e
to Chairman Carr, dated August 9, 1990)
In response to a request, dated June,14, 1905
~ rom former NRC Chairman Zech regarding ACNW/ACRS comments on the effectiveness-and efficiency of continuing a
centralir.ed support staff for the ACNW and ACRS versus a separate staff for the ACNW, both Committees concurred in a
joint report to the Commission stating that continuation of the combined staff is preferred.
B.-
Subcommittee Reoorts (Open) 1.
Plannina and' Procedures Subcommittee Recommendations Mr. Michelson, Chairman of the Planning and Procedures Subcommittee, provided a report-of the August 8, 1990 meeting of the Planning and Procedures Subcommittee. The Committee's actions on the recor,mendations of the Planning and Procedures Subcommittee are as follows:
o Revisions to ACRS Bylaws Related to Added Remarks
.The Committee
- aproved, with minor
- changes, revisions.'to ACRS Bylaws proposed by Dr. Lewis to provide a mechanism for additional members to join the author (s), after' the meeting has ended, who prepare additional remarks to ACRS reports.
e Letter /Recort'by Individual ACRS Members The Committee approved the revision.to ACRS Bylaws.
proposed by Dr. Lewis that a. member may decide to write'a letter / report on his own initiative if, after discussion with the' Committee, he feels that the importance1.of the subject warrants prompt
.gf action.
o-Uodate of ACRS Bvlaws As recommended by the' Committee, Dr. Lewis agreed to prepare an updated version'of:the ACRS Bylaws to make.them-consistent with' present activities, nomenclatures, etc.,. and submit it to the full Committee'for consideration.
i t
?
=
'364th ACRS Meeting Minutes 28 t
o Editina of ACRS ReDorts The Committee agreed to the. revised procedures proposed by the Planning and Procedures Subcommittee wherein the cognizant Subcommittee chairman will have the final responsibility for editing ACRS reports.
e Adooted Plant Activities The committee endorsed the following procedures proposed by the Planning and Procedures subcommittee regarding activities by members who have adopted plants:
Members should visit at least one adopted plant per year.
Arrangements.for plant visits should be made in accordance with the established procedures through tne ACRS office.
1 Plant visits should be primarily for information gathering purposes, not for review or resolution of problems.
(Mr. Wylie was asked to prepare a protocol for such visits.)
Plant visits by a single member should include a cognizant ACRS staff engineer as practicable.
Plant visits by more than one member should always' include an ACRS staff engineer to ensure.
[
.that Federal Advisory Committee Act. (FACA) requirements are met.-and to prepare a report of-the plant visit.
The ACRS list for selective distribution of documents should be revised to reflect the types of documents applicable to adopted plants so as to enable adopted plant members to select -
the types of documents that they wish to receive in support of their activities.
Members should inform the full Committee-of any areas of concern or particular interest
-observed during plant visits, u
e r
~
364th ACRS Meeting Minutes 29 n
2.-
NRC Safety Research Proara]B (OPen)
(NOTE:
.Mr.
S.
Duraiswamy was the Designated FederaA
' Official for this portion of the meeting.)
Dr.
- Catton, Chairman of the Safety Research Program Subcommittee, stated that he met with NRC Chairman Carr on July 31, 1990 and discussed the ACRS intent to expand the scope and content of the annual ACRS report to the Congress on the NRC Safety Research Program and budget.
Chairman Carr stated the following:
t The ACRS has a statutory obligation to provide an e
annual report to the Congress on the NRC Safety Research Program and budget.
The scope, content, and nature of such a report should be determined by the committee.
The Nuclear. Safety Research Review Committee
[
(NSRRC), established as recommended by the National
.Research Council, has been reviewing the NRC Safety-
- Research Program and providing advice to the RES Director.
He does not believe that the ACRS should i
review the activities of the NSRRC.
e
-ACRS comments on new areas of research are always
- welcome, i
e.
In his opinion, it is not advisable for Dr. Catton to meet with.the staff members of the Congressional Committees to obtain their. views on the nature of the ACRS report to the Congress.
They may ask certain things that the ACRS may have difficulty in complying with.
.Dr.
Siess asked ' whether. Chairman Carr expressed any interest in' receiving an annual report from the ACRS'on l.
L the NRC Safety Research Program.
Dr. Catton stated that he.does not believe that chairman Carr wants such a report. 'However, he seems to be. interested-in receiving u'
ACRS comments on new areas of research.
After further discussion,.Dr. Catton agreed to prepare l
an: outline fer the ensuing ACRS report to the Congress l'
and submit it to the full Committee for consideration during the September 6-8, 1990 ACRS meeting.
i
I
,s 364th ACRS Meeting Minutes 30 l
C.
Summarv / List of Follow-UD Matters In its report to the Commission related~to the level of e
detail required for design certificati.on under 10 CFR 4
Part 52, the Committee suggested that some form of the two-tier approach proposed by NUMARC is essential and that the NRC staff. and industry develop criteria to define the division between the two tiers. The Committee expressed its willingness, to review such criteria, as progress is made in this effort, and report to the a
Commission if the Commission wishes that this be done.
(Dr. El-Zef tawy has the follow-up action on this matter.)
In its report to the Commission on the NRC Organizational e
Factors-research program, the Committee stated that it wculd follow the progress of this research program with interest.
(Mr. Alderman has the follow-up action on this matter.)
e Dr. Lewis agreed to prepare an updated version of ACRS Bylaws and submit it to the ' full Committee for consideration.
(Mr. Fraley has the follow-up action on this matter.)
e Mr. Wylie, Chairman of the Adopted Plant Activities Subcommittee, agreed to develop detailed guidance for use by those members who adopted plants in carrying out their adopted plant activities.
(Mr.-Quittschreiber has the follow-up action on this matter.)
e The Committee instructed the ACRS staff to revise the list for selective distribution of documents to reflect the' types of documents applicable to adopted plants so as to enable - the -members to identify the types of documents that they wish to receive in support of their adopted' plant' activities.
(Mr. Schofer has the follow-up action on this matter.)
Dr. - Catton, Chairman of the Safety Research Program-Subcommittee, agreed to propose an outline for the ensuing ACRS report to the Congress on the NRC Safety Research Program and' submit it to the full Committee for consideration during-the September 6-8, 1990 ACRS meeting.-
(Mr. Duraiswamy has the follow-up action on this matter.)
v 364th ACRS Meeting Minutes 31 The Committee suggested that the ACRS Subcommittees on e
Extreme External Phenomena and on Severe Accidents hold a joint meeting to discuss the adequacy of consideration of seismic and fire issu(s in NUREG-115D.
(Mr. Houston has the follow-up action on this matter.)
The Committee instructed Dr. Savio to find out why Mr.
e
- Morris, NRR, did not file a " differing professional opinion" to bring his concerns related to the testing and replacement of reactor trip breakers to the attention of the NRC management and the ACRS.
(Dr. Savio has the follow-up action on this matter.)
Dr. Siess, Chairman of the Generic Items Subcommittee, suggested that cognizant Subcommittee Chairmen review the appropriateness of the priority rankings proposed by the NRC staff for various Generic Issues and provide comments in writing to Mr. Duraiswamy on or before August 20, 1990.
He suggested further that the members also review the, proposed priority rankings of those Generic Issues t
that were not assigned to them and bring their comments, if any,-for discussion during the September 6-8,c1990 ACRS meeting.
(Mr. Duraiswamy has the responsibility in coordinating this matter.)
e Dr. Siess and Mr. Michelson suggested that cognizant Subcommittee Chairmen review the items assigned to them during previous full Committee meetings and also the rules and policy matters that were assigned to them previously and propose a plan of action to the full 2
Committee as.soon as possible.
They-suggested also that the subcommittee Chairmen contact cognizant ACRS staff (identified in the tables passed out during the meeting)
,1f they need additional information. -(Mr. Duraiswamy has the responsibility for coordinating this task.)
In view of the fact that extensive changes have been made to-the EPRI-ALWR Requirements Document, the Committee decided not to comment on this matter until after completing its review of the revised version - of this docoment that is expected to be made available by the end of August 1990.
(Dr. El-Zef tawy has-the follow-up action -
on-this matter.)
During the discussion of the IIT report related to the loss of vital AC. power / loss of decay heat removal event at Vogtle Unit 1, Mr. Jordan, AEOD, made the following commitments.
(Mr. Boehnert has the follow-up action on this matter):
. V a
's.
364th ACRS Meeting Minutes 32 l
Provide the ACRS with a copy of the proposed staff
]
actions, that have been submitted to the commission by the EDO, for dealing with the IIT recommenda-tions.
Provide a
briefing to the Committee at the appropriate time in the future on the status of the staff actions on this matter.
During the discussion of the Solenoid Valve Case Study, e
Mr. Novak, AEOD, committed to provide a briefing to the Committee after the industry comments have been received and resolved.
(Mr. Alderman has the follow-up action on this matter.)
Dr. Kerr requested, and the Committee endorsed, that the e
staff provide information on the following.
(Mr.
Boehnert has the follow-up action on this matter):
Temporary instructions on inspection activities associated with plant-specific emergency operating procedures.
Details of the
" Engineering Managers' Forum" initiated during 1988 by licensees in NRC Region-V that is mentioned in the August 6, 1990 letter from Mr. Thomas T. Martin, Regional Administrator, Region I, _ to Mr. Feigenbaum, Seabrook nuclear power plant.
D.
' Future Activitics'(Open) 1.
Future Acenda The Committeo Agreed on a tentative schedule for the next Committee meeting.(Appendix II).
2 '.
Future Subcommittee Activiting A list-of future subcommittee meetings was distributed to the members (Appendix III).
The meeting'was adjourned at 11:45 a.m. August 11, 1990.
1
'g f..
. e =.
i
- r APPENDICES MINUTES OF THE 364TH ACRS MEETING AUGUST 9-11, 1990 I.
Attendees II.
Future Agenda III.
Future Subcommittee Activities IV.
List of Documents Provided to the Committee j
r 4
. k e
- r APPENDIX I 364TH ACRS MEETING MINUTES AUGUST 9-11, 1990 ATTENDEES PUBLIC ATTENDEES NRC ATTENDEES Thursday. Auaust 9, 1990 R. A.
Szalay, NUMARC J. G, Glitter, NRR
.J. MacEvoy, Bishop Cock Charles Miller, NRR E. Fotopoulos, SERCH Licensing, Bechtel R. Van Houten, SECY Margo Barron, NUS/LIS Gene Imbro, NRR J. F.
Quirk, GE Marylee Slosson, OEDO Stan Rittenbusch,-ABB/CE Tom Cox, NRR Kay Ng, NUMARC.
L. Norrholm, OCM/KC D.. L. Rehm,. Duke Power Co.
I.
Yoshida, NRR Bill Rasin, NUMARC D. Trimble, OCM/JC Adrian Meymen. NUMARC Mat Taylor, OEDO Marc Rowden,. Fried, Frank Cherie Siegel, AEOD Stephen Floyd, CP&L Warren Lyon, NRR V..Pareto, DEVONRUE Fred Allenpach, NRR L. N. Rib, AECL Tech.
Al Serkiz, RES H. Eckert, DOE Paul Norian, RES Rich' Stark, NUS.
Ed Tomlinson, NRR Alan Nelson, NUMARC Brian Sheron, RES
. Tony pietrangelo, NUMARC O.
Chopra, NRR Alex Marion,.NUMARC Al Chaffee, NRC Alan Rubin, TENERA M. Eisenberg, NMSS Herbert Kouts, Defense Nucl.ac. Safety Bd.
J. A. Murphy, RES
~
Fridav, Auaust 10. 1990
-D..M. Chapin, MPR T. Ryan, RES
. Dave. Norman, SERCH Licensing,Bechtel V. VanHouten, SECY i
David Strawson, MPR T. Kenyon, NRR Margo Barron, NUS/LIS J. G.
Spraul, NRR i
Gary Vine, EPRI-C. Miller, NRR-L.'V.-Toth, Gasser Assoc.
J. Glitter,fNRR
'Gabrielle Williams, Bethesda Licensing Off.
I. Yoshida, NRR F.-Coffman, RES Carl Johnson, RES Joel Kramer, RES Jesse Arildsen, NRR i
f.
J. Murphy, RES l
Elise Heumann, OC M. Taylor, OEDO L
J. Rosenthal, AEOD Hal Ornstein, AEOD P.
Lam, AEOD P. Cross-Prather,AEOD i,
L P. O'Reilly, AEOD l
T.
Novak, AEOD V.
Benaroya, AEOD S.
Koscielny, NRR l
-i
- t
-+
8 b
APPENDIX II 364TH ACRS MEETING MINUTES AUGUST 9-11, 1990 FUTURE AGENDA Tentative schedule for the 365th. Seotember 6-8, 1990. ACRS Meetina Briefing and discussion on e
Reactor ODeratina Experience (Onen) information and experience gained from nuclear power plant operations, including the Westinghouse Owners Group study justifying reduction in turbine stop valve testing frequency.
Members of the NRC staff Will participate, as appropriate.
Discussion regarding the e
NRC Ouantitative Safety Goals (ODen) status'of NRC plans for implementation of Quantitative Safety Goals for nuclear power plants.
e Activities of ACRS Subcommittees (Ocen) - Reports and discussion regarding the status of designated subcommittee activities, including reports of the Thermal-Hydraulic Phenomena and Human Factors Subcommittees, e'
Reactor Ccolant Pumn Seal Failures (Generic Issue 23) (ODen)
Review and comment on proposed resolution of this generic issue.
Members of the NRC staff and the nuclear industry will
-participate, as. appropriate.
e Nine Mile Point Nuclear Plant. Unit 1 (ODen) - Briefing and discussion of the status of Nine Mile Point, Unit i restart by representatives of the NRC staff and'the licensee,'as appropriate.
e:
- Performance-Based Ouality Assurance (ODen) - Information briefing and discussion regarding use of performance-based quality assurance requirements and the associated revision of the NRC Standard Review Plan.
o.
Forelan Activities (Ocen/ Closed) - Report and discussion of NRC-
-USSR. meeting on aspects of nuclear: power plant safety related to radiation embrittlement and annealing, severe accidents, erosion / corrosion attack of piping and components,-and fire protection for nuclear facilities.
4 e.
. Generic Safety Issues (Ocen) - Discussion and comment regarding
-proposed. priority rankings for a group of new generic safety issues.
Members of.the NRC staff will participate, as appropriate.
e-Yankee Rowe Nuclear Power Plant (ODen) - Discussion and report
.regarding the status of.the Yankee'Rowe reactor pressure vessel radiation embrittlement.
Representatives of the NRC staff and the licensee will participate, as appropriate.
L e
Systematic Assessmest of Licensee Performance (Onen) - Discuss proposed.ACRS comments on the NRC SALP program.
r-1 i '
4-364th ACRS Meeting Minutes II-2 L
Discuss e
Reactor Risk Assessment Document (NUREG-1150) (Ocen) proposed ACRS comments on the NRC report, NUREG-1150, Severe l
Accident. Risks:
An Assessment for Five U.S.
Nuclear Power Plants.
-e Future Activities (Open). - The members will discuss anticipated ACRS subcommittee activities and items' proposed for i
consideration by the full Committee.
Discussion with e
Conduct of ACRS Activities (Open) representatives of the Office of the General Counsel regarding-i
~ conduct of ACRS-subcommittee and subgroup activities.
Discussion among members regarding limitations on ACRS use of "outside" legal consultants.
e
. Nomination of ACRS Members (Ocen/ Closed) - Discuss qualifications of candidates proposed for appointment to the Committee.
50 4
l' f
1 L-l
{'
{
l
~..
- 5. ' -
'=
364TH ACRS. MEETING APPENDIX III-FUTURE SUBCOMMITTEE ACTIVITIES ACRS/ACNW COMMITTEE & SUBCOMMITTEE MEETING August 10,'1990 Joint Decay Heat Removal Systems and Thermal Hydraulic Phenomena, August 28, 29 and 30, 1990, at the Westbank Inny 475 River Parkway,
_each day.
The Idaho Falls.
ID _(Boehnert),
8:30 a.m.
subcommittees vill:- (1) discuss the details of the modifications made to the RELAP-5 MOD-2 code as specified in the MOD-3 version (August 28), (2) explore the use of feed and bleed for decay. heat removal in PWRs (August 29), and (3) review the proposed resolution
'of Generic Issue 23, " Reactor Coolant Pump Seal Leakage" (August 30 r
- a.m.).
Attendance by the following is anticipated, and reser-vations have been made at.the Westbank Inn (208/523-8000) for the nights of August 27, 28 and 29:
Dr. Catton Mr. Wylie Mr. Ward Mr. Davis NONE Mr. Carroll' Dr. Plesset Dr. Kerr Mr. Schrock
-Mr. Michelson Dr. Sullivan 23rd' ACNW Meetina, August 29, 30 and 31,1990, 7920 Norfolk Avenue, Bethesda. MD (Major), 8:30 a.m.,
Room P-110.
Attendance by the L
following is' anticipated, and reservations have been made at the hotels-indicated for the nights _of August 28, 29 and 30:
-Dr. Moeller HOLIDAY INN Dr. Steindler HOLIDAY INN
~
Dr. Hinze HOLIDAY INN Dr. Okrent
. HOLIDAY INN Dr. Pomeroy
. COSMOS CLUB p-
>Plannina and Procedures, September 5, 1990,-7920 Norfolk: Avenue, Bethesda, MD (Fraley), 2:30 p.m. until 5:30 p.m. (tentative), Room P-422.
The Subcommittee will review panel of candidates nominated Is for ACRS - positions s'heduled during. 1991.
Attendance by the
'following is anticipated, and reservations have been made at the h
hotels indicated for the night of September'4:
Mr '. Wylie HOLIDAY _ INN Mr. Carroll HOLIDAY INN i
i 1
l.
. c-
'e 2
MgLterials and Metallurav, September 5, 1990, 7920 Norfolk Avenue, Bethesda. MD (Igne), 8:30 a.m., Room P-110.
The Subcommittee will Review Yankee Rowe Reactor Pressure Vessel Embrittlement Issues.
Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of September 4:
Dr. Shewmon NONE Mr. Ward HYATT Dr.-Lewis EMBASSY Dr. Wilkins HOLIDAi INN Dr. Carroll HOLIDAY INN Mr. Wylie HOLIDAY INN Dr. Catton HOLIDAY INN Mr. Bender HOLIDAY INN Mr. Michelson DAYS INN (CONGR)
Dr. Bush HOLIDAY INN Mr. Minnick HOLIDAY INN Mr. Hanstad NONE Dr. Slass HOLIDAY, INN Dr. Odette NONE
' 365th ACRS Meetina, September 6-8, 1990, Bethesda, MD, Room P-110.
24th ACNW Meetina, September 19-20, 1990, Bethesda. MD, Room P-110.
. Advanced Pressurized Water Reactors, September 20,
- 1990, 7920 Norfolk Avenue, Bethesda, MD (El-Zeftawy),-8:30 a.m.,
Room P-422.-
The' Subcommittee will review, the draft SER for the Westinghouse SP190 design - Lodging will be announced later.. Attendance by the following is anticipated.
-Mr. Carroll Mr. Minnick Dr. Catton:
Dr. Shewmon Dr. Kerr Mr. Wylie Mr. Michelson Advanced Pressurized Water Reactors, September 21,
- 1990, 7920 Norfolk Avenue,'Bethesda, MD (El-Zeftawy), 8:30 a.m.,
Room P-110.
The subcommittee will continue its review the CE System 80+-design and related safety issues.
Lodging will be announced later.
Attendance by the.following is anticipated.
-Mr. Carroll-Mr. Minnick Dr. Catton Dr. Shewmon
~Dr. Kerr
.Mr.'Wylie
-Mr. Michelson
.;L66th ACRS Meetina, October 4-6, 1990, Bethesda, MD, Room P-110.
..y..
c,..
3 TVA' Plant Licensina and Restart, October 24-25, 1990, Huntsville.
AL (Houston).
The Subcommittee will conduct a site tour and review
.the planned restart of Browns Ferry Unit 2 (site visit 10/24 p.m. ).
Attendance by the following is anticipated:
Mr. Wylie Mr. Minnick Mr. Carroll Mr. Ward Mr. Michelson 25th ACNW Meetina, October _24-26,.1990, Bethesda. MD, Room P-110.
~i-
-Joint Severe Accidents'and Probabilistic Risk Assessment, Date to be determined (August / September), Bethesda, MD (Houston).. The Subcommittees will continue their review of NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants.
Attendance by the-following is anticipated:
Dr. Kerr.
Mr. Ward
'Dr. Lewis Mr. Wylie Dr. Catton Mr. Bender iMr. Michelson Mr. Davis Mr.'Minnick Dr.. Lee Dr. Shewmon Dr. Okrent Dr. Siess Dr. Saunders Joint' containment = Systems and Structural-Encineerina,.Date to be determined (August / September), Bethesda. MD (Houston /Igne).
The Subcommittees will develop containment design criteria for futu're plants._ Attendance by the following is anticipated:
Mr. Ward Mr. Minnick Dr. Siess Dr. Shewmon
'Dr. Catton Mr. Wylie Mr. Carroll Dr. Corradini
.Dr.-Kerr.
Mr. Bender.
Plant Operations, Date to be determined (September / October),
Bethesda. MD (Boehnert). ' The Subcommittee will begin review of-the NRC: staff's Action Plan to evaluate the risk from nuclear power plant shutdown operations.
Attendance. by - the following is j
' anticipated:
Mr.. Carroll Mr. Michelson Dr. Kerr-Mr. Minnick i
Dr. Lowis Mr. Wylie i
i
4 Joint Advanced !?ressurized Water Reactors and Advanced Boilina Water-Reactors, Date to be determined (October), Bethesda. MD (El-Zeftawy/ Alderman).
The Subcommittees will discuss the licensing review basis documents for CE System 80+ and GE ABWR designs.
Attendance by the following is anticipated:
Mr. Carroll Mr. Ward
- Mr. Michelson Mr. Wylie Dr. Catton Dr. Showmon Dr.-Kerr i
Materials and Metallurov, Date to be determined, Bethesda. MD (Igne).- The Subcommittee will. review the proposed resolution of Generic Issue 29, " Bolting Degradation or Failure in Nuclear Power Plants."
Attendance by-the following is anticipated:
. Dr. Shewmon Mr. Ward.
- Dr. Lewis Mr. Bender Mr. Michelson Dr. Kassner Quality and Ouality Assurance, Date to be determined, Bethesda. MD (Igne).
The Subcommittee will discuss the performance-based concept of quality, what it means, its implementation, -and
- preliminary.results.
Attendance by the following is anticipated:-
Dr. Siess Dr. Stevenson Mr.' Ward Mr. Cerzosimo (tent.)
Mr. Wylie.
Joint ComDuters in
' Nuclear Power Plant Ooerations and Instrumentation-and control
- Systems, Date to be determined, (Boehnert/El-Zeftawy).. The Subcommittees will discuss the use of
-computers. and solid state control logic-in nuclear power plant operations.
Attendance by the following is anticipated:
Dr. Lewis Mr. Wylie Dr. Kerr Mr. Davis Mr. Carroll Dr. Lipinski 4 Mr. Michelson
3.y
, o s l
5 Auxiliary and Secondary Systems, Date to be determined, Bethesda, MQ (Duraiswamy).
The Subcommittee will discuss:
(1) criteria being used by utilities to design Chilled Water Systems, (2) regu-
-latory requirements for Chilled Water Systems design, and (3) criteria being used by the NRC staff to review the Chilled Water Systems design.
In addition, the Subcommittee may discuss at this meeting matters concerning fire protection and mitigation in nuclear power plants.
Attendance by the following is anticipated:
Dr. Catton Mr. Wylie Mr. Carroll Dr. Quintere Mr. Michelsen
?
Joint Reculatory Activities and Containment Systems, Date to be determined,-Bethesda, MD (Duraiswamy/ Houston).
The Subcommittees
.will. review the proposed final. revision to Appendix J to 10 CFR' Part 50, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,"
and an associated Regulatory Guide.
Attendance by the followi.rj is anticipeted:
Dr. Siess Drs Kerr-Mr. Ward Mr. Michelson Mr. Carroll Mr. Minnick Dr.tCatton Mr. Wylie Occucational and Environmental Protection Systems, 'Date to be determined, Bethesda, MD (Igne).
The Subcommittee will review the Advance Notice of Proposed Rulemaking on hot particles. Attendance
- by the following is anticipated
J
'Mr. Carroll Dr. Lewis Mr. Wylie Dr. Moeller a
i I
a m
1
,w
, we
a-
=
. s.
,-,q APPENDIX IV MINUTES OF THE 364TH ACRS MEETING AUGUST ~9-11, 1990 i
LIST OF DOCUMENTS PROVIDED TO THE COMMITTEE MEETING NOTEBOOK 13d2-2 REOUIREMENTS FOR ESSENTIALLY COMPLETE DESIGN e
Presentation Schedule Status Report _with Attachments e
Attachment I - 10 CFR Part 52, Early Site Permits; Standard Design Certifications; and Combined
+
Licenses-for Nuclear Power Reactors, dated April 28, 1989 Attachment II - ACRS_ Report dated August 12, 1986, Re:
ACRS Comments on Proposed NRC Standardization Policy Statement
- Attachment III - ACRS Report dated October 15, 1986, Re:, ACRS Comments on Draft NUREG-1225,.
L
" Implementation of NRC policy on Nuclear Power Plant L
Standardization" Attachment.IV - acrs report dated June 7, 1988, Re:
NRC Proposed Rule on Early Site Permits, Standard Design Certification, and Combined Licenses for Nuclear Power Plants Attachment V'- ACRS Report dated January 19, 1989, Re:
Draft Final Rule on Standardization and Licensing Reform, 10 CFR Part 52 At?.achment VI - ACRS Report dated February 15, 1989, Re.
Draft Final. Rule cnt Standardization and Licensing Reform, 10 CFR-Part 52 Attachment'VII - ACRS Report dated November 24, 1989, Re:
Module 1 of the Draft Safety Evaluation ~ Report for the. Advanced Boiling Water = Reactor Design Attachment VIII - Remarks.by Commissioner J. Curtiss ont the 1990 Nuclear Power Assembly,. Washington, D.C.
E May.22, 1990 Attachment IX - SRM M900427'from S. Chilk, Secretary L
-for J'1 Taylor, EDO, Briefing on Evolutionary Light H
-Water Reactor Certification Issues and Related Ei Regulatory Requirements, dated June 20,_1990
'~
< Attachment X -'SECY-90-241,1 Level of Detail Required for Design Certification under Part.52, dated 7/11/90 l
e is
-r e
0
- gg 3
364TH ACRS. Meeting Minutes Appendix IV-2 3
NRC INCIDENT INVESTIGATION TEAM BRIEFING FOR ALVIN W. VOGTLE NUCLEAR PLANT - Loss of Power / Decay Heat Removal Capability e
Presentation Schedule Status Report with Attachments:
e Memorandum dated July 18, 1990 from P.
Boehnert, ACRS to ACRS Members,
Subject:
Background Material:
NRC Incident Investigation Team (IIT). Presentation of Investigation of the Loss of Power / Decay Heat Removal Event at the Vogtle Plant - August ACRS Meeting SRM dated June 28, 1990 to J. Taylor, EDO from S.
Chilk, Secretary,
Subject:
Briefing on IIT Report on Vogtle Event Memorandum dated June 20, 1990 from P. Boehnert, 4
ACRS to J.
Carroll, ACRS Plant Operations Subcommittee..
Subject:
Commission Meeting on June 8, 1990 - NRC Incident Investigation Team Report - Loss of Vital AC Power and Residual Heat Removal at Vogtle Unit 1 Memorandum dated June 8, 1990 from P.
Boehnert ACRS, to J.
Carroll, ACRS Suubcommittee on Plant Operations,
Subject:
NRC Incident Investigation (IIT) Report - Loss of Vital AC Power and RHR During Mid-Loop Operations at Vogtle Unit 1 Excerpts from Advance Copy of subject IIT Report Presentation materials provided during the meeting.
e 4.
GENERIC ISSUE B-56. DIESEL GENERATOR RELIABILITX o
Presentation Schedule-
+
Status Report with
Enclosures:
e Letter. dated May 3, 1990 from W. Rasin, NUMARC, to E.
Beckjord, RES, NRC, re NUMARC efforts relating to GI B-56, Diesel Generator Reliability w/ encl Encl. A: -Response to CRGR Comments Encl. B:. Reg. Guide 1.9,.Rev. '3 Encl. C:
Proposed Generic Letter Encl. D:
Backfit Analysis Encl. E:
FRN draft notice Encl. F:
NUMARC-8700, Rev.
1, App. D dated 5/2/90 Encl. G:. Memo to Project Managers Memo from A. Thadani'To W. Minners, dated 7/6/90 Revised Section C.5, Reg. Cuide 1.9, Rev. 3 dated 7/9/90 Attachment I - Letter from W. Rasin, NUMARC, for C.
Wylle, ACRS, Re:
Generic Issue B-56, EDG Reliability, g
L dated July 26, 1990 Attachment II - Memorandum for R.
Fraley, ACRS, from l
W.
Minners, RES, Re:
CRGR Review of B-56 Resolution, L
dated July 27, 1990 l
Presentation materials provided during the mooting.
e i
.o.
' 7 '.a
_ -c e*
364th ACRS !!eeting Minutes Appendix IV-3 5
SEVERE ACCIDENT RISK REPORT (NUREG-1150)
Tentative Agenda e
Status Report e
Charter of the Special Committee to Review NUREG-1150 e
List of Membership of Special Committee e
DRAFT Report of Special Committee, Chapter 7, " Conclusions e
and Recommendations (INTERNAL COMMITTEE.USE ONLY)
DRAFT Report of Special Committee, Chapter 8,
" Answers to i
e Commission Questions" (INTERNAL COMMITTEE USE ONLY).
Presentation materials provided during the meeting.
8 NRC'RESEARCH PROGRAM ON ORGANIZATIONAL FACTORS o
Tentative Schedule Status Report with
Attachment:
o DRAFT SECY Paper, Organizational Factors Research Progress Report (INTERNAL COMMITTEE USE ONLY)
Presentation materials provided during the meeting.
e 9'
SOLENOID VALVE CASE STUDY e
Tentative Schedule o
Status Report l
Presentation materials provided during the meeting.
e-l 10,. REACTOR OPERATING EXPERIENCE AND EVENTS AT NUCLEAR PLANTS e
Presentation Schedule' e'
- Status Report with
Attachment:
Finnish Report dated June 19, 1990, "Feedwater Pipe Rupture at Loviisa. Nuclear Power Plant Unit 1" i
Presentation materials provided during the meeting.
{
e 11.1' LIST OF FUTURE ACRS SUBCOMMITTEE MEETINGS AND'ACRS/ACNW MEETINGS
-l UANDOUTS i
11.2 Memorandum to ACRS. Members from R. P.
Savio,
Subject:
Future ACRS Activities, 365th ACRS Meeting - September 6-8, 1990.
1 Ls s_ _