ML20236T749
| ML20236T749 | |
| Person / Time | |
|---|---|
| Issue date: | 08/08/1987 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| RTR-REGGD-1.158, TASK-EE-006-5, TASK-EE-6-5, TASK-RE ACRS-2520, NUDOCS 8712020069 | |
| Download: ML20236T749 (50) | |
Text
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Q k $ U ll TABLE OF CONTENTS MINUTES OF THE 328TH ACRS MEETING gg
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AUGUST 6-8, 1987 WASHINGTON, D.C.
ppp ////55 age
[
I.
Chairman's Report (0 pen).........................................
1 II. ACRS Briefing on the Activities of the NRC's Office of l
Special Projects, TVA Projects Division (0 pen)...................
2 III. Fire Protection Research Scoping Study (0 pen)....................
3 IV. August 6,1987 Meeting with NRC Commissioners (0 pen).............
9 l
V.
Proposed Degree Requirements for Senior Reactor Operators (0 pen).
11 VI.
Improved Westinghouse Standard Plant (0 pen)......................
16 VII.
Proposed Final Broad Scope Amendment to General Desi Criterion-4 (GDC-4) Rule (0 pen).....................gn l
17
]
VIII. Nuclear Power Plant Operating Experience (0 pen)..................
19 IX.
Executive Sessions (0 pen / Closed).................................
24 I
A. Subcommittee Reports (0 pen / Closed)............................
24 1.
WasteManagement(0 pen)..................................
24 2.
NewMembers(C1osed).....................................
24 B. Reports, Letters and Memoranda (0 pen).........................
25 1.
ACRS Comments on Proposed Final Broad Scope Rule to Modify General Design Criterion-4 Environmental and MissileDesignBases(GDC-4).............................
25 2.
ACRS Report on Fire Risk Research Study..................
25 3.
ACRS Review of Application for Preliminary Approval of the Westinghouse RESAR/SP-90 Design......................
25 4.
ACRS Comments on the Advance Notice of Proposed Rul'e-making on Degree Requirements for Senior Operators.......
25 5.
ACRS Action on Proposed Regulatory Guide EE 006-5,
" Qualification of Safety Related Lead Storage Batteries for Nucler Power P1 ants".......................
26 6.
ACRS Action on the Proposed Section 3.6.3, " Leak-Before-Break Evaluation Procedures," of the NRR Standard Review P1an.....................................................
26 DESIGNATED oHIGINAL g 20 g 9 871125 certified Br.
2520 PDR
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328 ACRS MEETING MINUTES 11 Page.
7.
'ACRSCommentsonProposedDOEBiiltoEstablish Independent Safety Board.................................
26 i.
C.
Other Comi ttee Concl usions ' (Cl osed).........................
26' i
Ad Hoc Pl a nni ng Subcommittee............................
26 1.
l D.
Future Activities (0 pen).....................................
26 1.
Future Agenda............................................
26
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2.
Futu re Subcomi ttee Acti vi ties..........................
26 i
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.4 iii 1
APPENDICES TO-l MINUTES OF THE 328TH ACRS MEETING AUGUST 6-8, 1987~
WASHINGTON, D.C.
I l
Appendix I List of Attendees Appendix II Future Agenda Appendix-III Future' Subconsittee Activities
. Appendix IV Other Documents Received
)
l l
l 4
--- ---.----- =---.-D u---- -- - --..-.--
--____________1______--__1---
~
Federal Register / vol 52. No.150 / Wedarsday. August 5.1se7 / N:tices((
22%2,,
I r
th:t supports the proposed exemption.
Draft NUREG-1150. Reettor Risk Saturday, August 8,1987 i
/
(
The N,RC staff did not consult other Re ference Document, report dated July.
8 30 a.m.-12.m Noon: Preparation of egincies or persons.
15.1987.
ACRSReports to the NRC(Open)-
Findings of No S6sni5 cant Impace
- Integrated Safety Assessment Discuss proposed reports to NRC Program, report dated Jujy 15.1987.
regarding items considered during this Based upon the foresom, 8 2Mp.m.-3:30p.m.: Meeting with NRC' meeting'
! ~
\\
environmental assessment, we concbde Commissioners (Open)-Discuss items E"~#'
l that the proposed exemption will not noted above 45p.m. bl5p.m.:Requirementsfor have a significant effect on the quality 3:
ge e
onsider provide ov r i t of OE activ ties he is i sd e ned not ropo ed degree e
- # F'*'# # #**# ###"I####"#
prepare an environmentalimpact nuclear power plant operators.
(Open)-Complete discussion of items
' l statement for the proposed exemption.
For further details with respect to the Priday. August 7.19s7 considered during this meeting.
proposed actions, see the licensee's 8:30 a.m.-10:20 a.ma Management Procedures for the conduct of requests for the exemption dated April A//ocation of Resources for Advisory participation in ACRS meetings were 11.1986. August 29.1988, and March 13.
Functions (Closed)-Discuss ACRS role published in the Federal Register on 1987, which are available for public in review of regulatory matters.
October 20.19,86 (51 FR 37241). In inspection at the Commission's Public This session will be closed to discuss accordance with these procedures. oral Document Room.1717 H Street. NW.,
information that involves the internal by rnembers of the pubh,ay be pre,sented r wrinen statemenu m Washington DC, and at the Salem Free
- c. recordmgs personnel rules and practices of NRC County Public Library.112 W.
and information the release of which Will be permitted only during those Breadway. Salem New Jersey 08079.
would represent a clearly unwarranted
,p
'p an uest ns portions of the meeting when a Deted at Betheeds. Marytand this 28th day invasion of personal privacy.
[a eiluly1887 10:30 a.m.-12:00 Noon: Meeting with C
ay a
y e
,, g i
l dStaff i"j'i ke dr For the !*aclear Raguietary Commussion.
NRC Commissioners (Closed)-Discuss Pe s
g to Waher R. Buder.
managemem allocation of moources for atatements should notify the ACRS Director. P:weet Ditwetomie M. Dinsion of advisory functions.
Executive Director as far in advance as ReoctorPm u/L This session wi!! be closed to discuss practicable so that appropriate (FR Doc.
7793 Filed 6-4-a? a-45 araj information that involves the personnel arrangements can be made to allow the sawma ruse ev.e rules and practices of NRC and necessary time during the meeting for information the release of which would such statements. Use of still, motion i
represent a clearly unwarranted picture and television carr. eras during eeting Agenda of the Advisory invasion of personal privacy.
this meeting may be limited to selected l
Committee on Reactor Safeguarda 1:00 p.m.-1:30 p.m.: Standardized portions of the meeting as determined l
In accordance with the purposes of Westinghouse PWR Plant (Open)--
by the Chairman. Information regarding i
s:ctions 29 and 182b. of the Atomic Briefmg and discussion of proposed the time to be set aside for this purpose Energy Act (42 U.S.C. 2039. 2232b) the NRC plan for review of the proposed may be obtained by a prepaid telephone l
Advisory Committee on Reactor Westinghouse improved standardized call to the ACRS Executive Director.
Safeguards will hold a meeting on nuclear power plant design (RESAR SP/
R.F. Fraley, prior to the meeting. In view August 64.1987. in Room 1046.1717 H 90).
of the possibility that the schedule for Street. NW., Washington. DC. Notice of 1:30p.m.-3:00p.m.t Ceneral Design ACRS meetings may be adjusted by the this meeting was published in the Criterion-4 (Open)-Review proposed Chairman as necessary to facilitate the Federal Register on July 20.1987.
revision of GDC-4 to include conduct of the meeting, persons I
considerstic.t ofleak.before break planning to attend should check with the l
Thursday AuFust & 1967 criteria in the design of nuclear power ACRS Executive Director if such B:30 a.m.-a 45 a.m.: Report of ACRS plants.
rescheduling would result in major l
Chairman (Open}--Tbe ACRS Chsirman 3:15 p.m.-5:15p.m.: Nuclear Power inconvenience.
will report briefly regarding items of Plant Operating Experience (Open)-
I have determined in accordance with current interest to the Committee.
Briefing and discussion regarding recent subsection 10(d) Pub. L.92-463 that it is 8:45 a.m.-10:15 a.m.: Tennessee Volley operating events and incidents at necessary to close portions of this Authority (TVA) Nuclear Activities nuclear facilities.
meeting as noted above to discuss (Open}-Review proposed TVA Nuclear 5:15 p.m.-5.45p.m.: Puture A CRS information the release of which would P;rformance Phm-Corporate and Activities (Open)-Discuss anticipated represent a clearly unwarranted proposed restart of'IVA nuclear power ACRS activities and items proposed for invasion of personal privacy (5 U.S.C.
plints.
consideration by the full Committee.
552b(c)(6)) and information related to 10:30 a.m.-12:30p.m.: Fire Protection 5:45 p m.-6:30p.m.: Appointment of the internal personnel rules and (Open)-Discuss proposed scoping plan New A CRS Members (Open/ Closed)--
practices of the agency (5 U.S.C.
f:r a research program regarding fire Discuss qualification of candidates 552b(c)(2)).
protection provisions in nuclear power proposed for appointment to the ACRS.
Further information regarding topics plints.
Portions of this session will be closed to be discussed, whether the meeting 1:30p.m.-1:50p. int Preparation for as required ta discuss information the has been cancelled or rescheduled, the Meetirw with NRC Commissioners release of which would represent a Chairman's ruling on requests for the (Open)-Discuss comments regarding clearly unweranted invasion of opportunity to present oral statements ACRS reports to the NRC oru personal privacy and information that and the time allotted can be obtained by
- Plan for Implementation of NRC involves the internal personnel rules a prepaid telephone call to the ACRS Safety Coals. report dated May 12.1987, and precticos of NRC.
Executive Director Mr. Raymond F.
- #L
, Fedrl R:gWr / Vol. 52. No.150 / W:dnesday, August 5.1987 / N:tices 28183 Fraley (telephone 202/634-32G3).
f requested an exemption for the the availabtfity of a reliable means of.
containment and residual heat removal illumination and whether the path of between 8:15 a.m. and 5:00 p. n a
I (RHR) pit areas where manual, cold travel would be unobstructed and easuy Dated July so 1987.
shutdown operations are required and/
traversed. The licensee has identified lohn C. Hoyle *.
or where possible repairs may be alternate access paths to the required -
Advisory Comminee Monopement Officer.
needed.The staffs evaluation of the equipment or alternate equipment that lFR Doc. 87-C99 Filed 8-4-6. 8 45 am) licensee's request is provided below.
would provide the same functiona. For swwc coot neo-o*
The reason for requiring 8. hour example, the alternate access route to battery powered emergency lighting is to the chargmg pump room is imm the ensure that at least minimallighting is second level of the auxiliary building IDock;t No. 50-261) available for the performance of manual adjacent to the non regenerative and C:rolina Power & Ught Co., H.B.
actions necessary for safe shutdown seal water heat hans, to the Rotnnson Steam Electric Plant, Unit after a fire. Usually manual actions are component cooling water heat No. 2; Exemption required for valve alignment, repairs exchanger from the turbine building via and pump control operations.The the exterior door in the south wall of the I
licensee has stated that a fire on the room, and to the battery room via the The Carolina Power & Light Company grognd level at the south end of the chemical batch addition room.These (CP&l. or the licensee) is the holder of auxiliary building hallway would alternate routes are provided with fixed.
Operating License No. DpR-23 that pervent access to dedicated shutdown
- 8. hour battery powered units.
i authorizes operation of the H.B.
equipment in the charging pump room Furthermore, the alternate access route Robinson Steam Electric Plant. Unit No.
and the component cooling water heat to thi safety injection pump room 2.Thilicense provides. among other exchanger room. Similarly, a fire in the follows the exterior of the auxiliary things, that the H B. Robinson Steam emergency switchgear room would building along the east and north sides Electric Plant. Unit No. 2. is subject to prevent access to the DC distribution of the safety injection pump room all rules. regulations, and Orders of the panels in the battery room. A fire at the exterior door. Portable hand-held Commission now or hereafter in effect.
north end of the auxiliary buildin8 1 ghting will be provided for operator The ststion is a single. unit pressurized hallway on the ground level would access to the safety injection pump w:ter reactor at the licensee's site prevent access to the SI-864 A and B room. Permanent emergency lighting is loceted in Darlington County. South valves in the safety injection pump provided inside the safety injection Cirolina.
room. Also, manual operation of service pump room to operate the required water valve VS-12D. located at the equipment. Portable lights will be g
intake s'.ructure, would require provided in the control room for On November 19.1980, the emerger.cy lighting. The licensee has performing the required functions at the 1
Commission published a revised Section stated that due to the numerous service water intake structure.These 50.48 end a new Appendix R to 10 CFR alternate access pathwsys, a large portable lights will provide adequate Pert 50 regarding fire protection features number of fixed emergency lightmg units illumination for the operators to access cf nuclear power plants.The revised would have to be installed and the the intake structure and operate valve I
i 50.48 and Appendix R became routing of associated cabling to provide V6-12D*
effecth e on February 17.1981. Section the necessary electrical power for Since the only manual actions racticable.
!!! of Appendix R contains 15 redundant lighting is not he proposedrequired inside the containment and subsections. lettered A through O. each The licensee justified t RHR pit are for the operation of valves 1
of which specifies requirements for a exemptions for the limited f r cold shutdown, sufficient time is particular espect of the fire protection circumstances where a fire may prevent available for the licensee to take features at a nuclear power plant. One access through the south end of the appropriate action to re-energize the of these subsection.111.J.is the subject of auxiliary building to the above stated normal containment lighting units prior the licensee's exemption request.
areas. The licensee's justification is to containment entry.
Section 1111. of Appendix R to 10 CFR based on the availability of an assured Based on the above evaluation of Pert 50. Emergency Lighting. requires 6 alternate path where 6. hour battery altemate a: cess routes and provison for hour bettery powered lighting units in powered lighting units are provided. In portable, hand-held lighting, tne staff creas nieded for operation of safe the areas where lighting units would not concludes that adequate lighting will shutdown equipment and along access be installed dedicated portable, hand.
and egress routes thereto.
held lighting would be provided for the prevail to access areas and perform operator to perform the necessary necessary safe shutdown functions.
Ug functions.The licensee justifies this Therefore. the licensee's request for By letter dated June 29.1984. as approach on the basis that the exemptions from the requirements of supplemented January 16.1985. CP&L availability of dedicated portable band.
section 111.] of Appendix R for certain requested approval of exemption from held lighting provides a level of paths to the charging pump room.
the technical requirements of section emergency lighting equivalent to that component cooling water heat
!!!.l of Appendix R to to CFR Part 50 required by Section 111.] for the above exchanger room. DC distribution pawis In the battery room and safety injection conceming the need for 6-hour battery areas.
powered lighting units in areas needed The technical requirements of section pump room is acceptable and should be for opIration of safe shutdown 111.] of Appendix R are not expressly met granted. Furthermore, the staff cquipment and along access routes. The at the intake structwe and along the concludes that the installation of s hour exemption request relates to the access access route to the safety injection battery powered emergency hghtmg routes to the charging pump room, pump room because fixed, individual 8 units inside the containment would not l
component cooling water heat hour battery powered !!ghting units are significantly improve the level of fire l
exchanger room. battery roorn. safety not provided for safe shutdown.
protection for this fire ares.nerefore, injiction pump room and the service At the north end of the auxiliary their omission is an acceptable water intake structure. The licensee also building, the staff was concemed about eumption from section 111.] of Appendix I
i p'tKrop i
'o'^
UNIVED STATES 4
NUCLEAR REGULATORY COMMISSION n
l'.; I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o
WASHWGTON, D. C. 20656 i
J Revised July 30, 1987 l
1 l
SCHEDULE AND OUTLINE FOR DISCUSSION 328TH ACRS MEETING AUGUST 6-8, 1987 WASHINGTON, D.C.
Thursday, August 6, 1987, Room 1046, 1717 H Street, N.W., Washington, D.C.
1) 8:30 - 8:40 A.M.
Chairman's Report (0 pen) l 1.1) Opening remarks 1.2)
Items of current interest (WK/RFF) 2)
8:45 - 10:15 A.M.
TVA Nuclear Activities (0 pen) 4 2.1)
Briefing and discussion of proposed i
Corporate Nuclear Perfonnance Plan and-i restart of TVA r.uclear plants (CJW/RPS) i 10:15 - 10:30 A.M.
BREAK' i
- 3) 10:30 - 12:30 P.M.
Fire Protection (0 pen) 3.1) Briefing and discussion of proposed I
scoping studies for a research pro-gram related to fire protection I
features in nuclear power plants (CYM/SD) 12:30 - 1:30 P.M.
LUNCH 4) 1:30 - 1:50 P.M.
Preparation for Meeting with NRC Com-missioners (0 pen) i 4.1) Discuss items for meeting with NRC Commissioners regarding:
4.1-1)
Implementation of NRC Safety Goals, ACRS report dated May 13, 1987 (D0/ DAW /RPS) 4.1-2) ACRS comments on draft I
NUREG-1150, Reactor Risk Reference Document, report dated July 15, 1987 (WK/D0/MDH/RPS) 4.1-3) ACRS corsnents on the Integrated Safety Assessment Program, report dated July 15, 1987 (DAW /RPS) 1 a---
l 4
328th ACRS Meeting Agenda-1 5) 2:00 - 3:30 P.M.
Meetinc with NRC Commissioners' (0 pen) l 5.1)
[1scussion regarding:
i 5.1-1)
Implementation.of NRC Safety j
-Goals, ACRS report dated May 13, 1987 (D0/ DAW /RPS)'
5.1-2) ACRS comments on draft NUREG-1150, Reactor Risk Reference Document report
' dated July 15, 1987-(WK/D0/MDH/RPS) 5.1-3) ACRS coments on the Integrated i
Safety Assessment Program, l
report dated July 15, 1987 (DAW /RPS) 1 3:30 - 3:45 P.M.
BREAK 6) 3:45 - 6:15 P.M.
Requirements for Reactor 0)erators (0 pen) l 6.1)
Discuss proposed (5EGY-87-101)
Degree Requirements for Nuclear PowerPlantOperators(FJR/HA) l Friday, August 7, 1987, Room 1046, 1717 H Street, N.W., Washington, D.C.
7) 8:30 - 10:20 A.M.
Allocation of Resot. ces for Advisory Functions (Closed) 7.1) Discuss impact on ACRS activities of proposed allocation of resources for j
providing NRC advice regarding i
regulation of nuclear waste l
(WK/HWL/RFF) l l
(Note:
This session will be closed to
. i l
l discuss infomation that involves the internal personnel rules and practices of NRC and infomation the release of which-j would represent a clearly. unwarranted
)
l invasion of personal privacy.)
a 8)10:30- 12:00 Noon Meetinc with NRC Commissioners (Closed) 8.1)
Liscuss proposed allocation of resources for providing NRC advice 5
regarding regulation of nuclear waste (WK/HWL/RFF)
(Note: This session will be closed to-discuss information that involves the l
personnel rules and practices of NRC j
and information the release of which would represent a' clearly unwarranted l-invasionofpersonal-privacy.)
l'
- +.e-
328th ACRS Meeting Agenda -
12:00 -
1:00 P.M.
LUNCH 9) 1:00 -
1:30 P.M.
Improved Westinghouse Standard Plant (0 pen) 9.1) Briefing regarding status of review (DAW /MME)
- 10) 1:30 -
3:00 P.M.
Leak Before Break Criteria (0 pen),
10.1) Discuss proposed revision of GDC-4-(PGS/EGI)-.
3:00 -
3:15 P.M.
BRE'K A
- 11) 3:15 -
5:15 P.M.
Nuclear Power Plant Operating Experience:
(0 pen 11.1)) Briefing and discussion of recent' operating events ~and transients (JCE/HA)
- 12) 5:15 -
5:45 P.M.
Future ACRS Activities (0 pen) 12.1) Discuss proposed subcommittee'and' full Committee activities (WK/RFF/MWL)
- 13) 5:45 -
6:30 P.M.
Appointment-of New Members (Closed) 13.1) Discuss candidates for appointment to the ACr" (HWL/NSL)
(Note:
Portions of this meeting will be closed as.necessary to discuss information the release of which would represent a clearly unwarranted invasion of. personal privacy and information that involves the internal ~ personnel rules and practices of NRC.)
Saturday, August 8, 1987, Room 1046, 1717 H Street, N.W., Washington, D.C.
s erlip 14)8:30- 12:00 Noon Preparation of ACRS Reports on:
14.1)
TVA activities (CJW/RPS) 14.2)
Degree requirements for operators (FJR/HA) 14.3)
Fire protection - Scoping studies for research program (CYM/SD) 14.4)
General Design Criterion-4, Leak-Before Break Requirements-(PGS/EGI)
~
12:00 -
1:00 P.M.
LUNCH
.328th ACRS Meeting Agenda
_4_
- 15) 1:00 -
2:00 P.M.
Nuclear Safety Board (0 pen) 15.1) Discuss proposed ACRS coments regard-ing ACRS support of DOE proposed Nuclear Safety Board (WK/HWL/RFF)
- 16) 2:00 - 3:00 P.M.
Miscellaneous (0 pen) 16.1)
CompleteLdiscussion of items considered during this meeting -
i i
I e
1
_________i__=______________._______.____i._____. _ _. _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _.
.___m
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CERTFIEf MINUTES OF THE 328TH ACRS MEETING AUGUST 6-8, 1987 WASHINGTON, D. C.
The 328th meeting of the Advisory Comittee on Reactor Safeguards, held I
at 1717 H Street, N.W., Washington, D.C., was convened by Chairman W.
Kerr at 8:30 a.m., Thursday, August 6,1987.
[ Note:
For a list of attendees, see Appendix I.
Dr. Shewmon was not present on Thursday; Drs. Okrent and Steindler were not present on Saturday.]
The Chairman said that the agenda for the meeting had been published.
He identified the items to be discussed on Thursday. He stated that the-meeting was being held in conformance with the Federal Advisory Comit-tee Act and the Government in the Sunshine Act, Public Laws92-463 and i
94-409, respectively.- He also noted that a transcript of some of the public portions of the meeting was being taken, and would be available in the NRC Public Document Room at 1717 H Street, N.W., Washington, D.C.
[ Note:
Copies of the transcript taken at this meeting are also avail-able for purchase from ACE-Federal Reporters, Inc., 444 North Capitol Street, Washington,DC20001.]
I.
Chairman's Report (0 pen)
[ Note:
R.
F.
Fraley was the Designated Federal Official for this l
portion of the meeting.]
The Chairman announced that Dr. Steindler is now a member of the ACRS.
Dr. Kerr said that the Court has recently issued an opinion concerning NRC's backfitting procedures.
The opinion said that NRC should not consider costs in determining if backfitting is required.
The Chairman also noted that there has been a reorganization of responsibilities within Region IV and several new people have been sent to that Region.
Dr. Kerr informed the Comittee that he had received a memorandum from Commissioner Bernthal concerning the views of members on the safety of Mark I containments in the context of NUREG-1150.
He asked the members to study the memorandum and be prepared to provide their views.
D. Ward, as Chairman of the Thermal Hydraulic Phenomena Subcommittee, informed the ACRS of a recent discovery by a NRR Staff member (W. Lyon) of a potentially unanalyzed LOCA scenario that could result in loss of assured core cooling. Mr. Ward noted that the LB LOCA analyses conclude I
about 3-5 minutes after initiation of pipe rupture with the end of core reflood. Mr. Lyon has postulated that for certain U-tube SG plants with cold leg loop seals (W and possibly CE), core cooling could be lost some hours after core refiliod. The scenario requires refill of the loop seal and a cold leg break whose size (intermediate to large) and location (top of pipe) result in essentially stagnant loop flow during the post reflood recirculation phase.
Steam from core boiling must then be
4 328TH ACRS MEETING MINUTES 2
1 i
forced through the loop seal to exit the RCS.
The resulting " manometer effect" could possibly depress the core liquid level to the puint where core damage may result. Mr. Ward noted NRR is pursuing this issue with the affected licensees and he indicated the Subcommittee is satisfied with the Staff's action for the present.
The Comittee will be kept.
informed as developments warrant.
1 II. ACRS Briefing on the Activities of the NRC's TVA Projects Division j
(0 pen)
[ Note:
R. P. Savio was the Designated Federal Official for this portion i
ofthemeeting.]
l The Committee met with James Keppler and was briefed on the status of j
the NRC's Offices of Special Projects (OSP) activities on TVA's nuclear facilities.
03P was established in February 1987 as a temporary office to manage NRC's licensing activities on Comanche Peak and on TVA nuclear 1
facilities.
This Office has about 100 perts,el (about 70 on TVA, about' 30 on Comanche Peak) and contains Headqua. u ; and Regional personnel.
l Mr. Keppler is the Director of OSP and Steart Ebneter and Christopher Grimes, respectively, head the TVA Projects Division and the Comanche Peak Projects Division.
OSP has reviewed TVA management reorganization and has reported their findings in a July 1987 SER.
The current schedule for.the restart of Sequoyah, Unit 2 is November 6, 1987.
Sequoyah, Unit 1 is scheduled to restart about six months after Unit 2.
The first of the Browns Ferry plants is scheduled for June 1988. Relatively little work is being done on the Watts Bar plants and no work is likely to be started until more of the issues in the Sequoyah and Browns Ferry plants are resolved.
There was some discussion as to the nature of the problems at the TVA plants.
NRC believes that the management control systems were flawed and that this resulted both in deficiencies in the records which have to be maintained to verify that the plant license conditions are met and in actual equipment deficiencies.
Mr. Keppler stated that TVA management has been improved by the hiring of Steven White to head the nuclear operations and by Mr. White's management reorganization and hiring of new management personnel.
Mr. Keppler does have some concern as to how long these new managers can be kept at TVA and whether adequate replace-ments can be found.
Mr. Ebersole expressed concern as to the lack of communication between the various elements in the TVA organization which Mr. White had restructured.
Mr. Ebersole stated that attention should be given to assuring that these same problems did not occur in the i
present TVA organization.
Mr. Michelson and Mr. Ebersole stated that attention should continue to be given to assuring that an effective i
" safety conscience" is contained within TVA's organization.
Mr. Keppler discussed the TVA and NRC Staff responses to the comments which the ACRS had made in its August 12, 1986 report on TVA.
The NRC
q N
d 3
328TH ACRS MEETING MINUTES 3
Staff's and TVA's actions. are documented. in the NRC Staff's July'1987 SER. The NRC Staff was" generally satisfied with TVA's actions. The NRC-Staff is continuing to give-attention to the issues associated with the j
" span of control" in Mr.- White's line. management organization: and 'any 4
problems associated with TVA's ability to offer compensation which would
)
attract and retain quality managers.
1 1
Mr. Wylie proposed that the ACRS. Subcommittee.on TVA Organizational 1
Issues meet with the NRC Staff and TVA: to discuss TVA. organization and.
j Sequoyah, restart issues in detail..
The. ACRS agreed to this..The.
Subcommittee meeting was te' tatively scheduled to be. held during the lj
~
n week of September 27,.1987.
i 1
III.FireprotectionResearchScopingStudy.(0 pen) i l
[ Note:
S.. Duraiswamy was the Designated Federal Official for this J
portionofthemeeting.]
Mr. Michelson,. Chairman of the. Auxiliary Systems Subcommittee, said that-in the ACRS report of February 19, 1986 to the Congress on the FY 1987 NRC safety research program, and also in its June 11, 1986 report to the Comission on the FY '1988 NRC research program' and budget, the-ACRS expressed concern about terminating the fire protection research at the.
end of FY 1986 and recommended that funding for this research be re '
stored. The Office of Nuclear Regulatory Research (RES). response was to initiate a Fire Risk Scoping Study at Sandia. National' Laboratories (SNL) to obtain information in determining if further fire-related research is warranted.
In the July 16, 1986 report to the. Comission -the ACRS reiterated its concern about. the premature termination of, the fire protection research.
Although the Comission agreed with the ACRS on the importance of fire protection research, it did not restore the funding.
However, it directed the Staff to work closely with the - ACRS to assess further research needs and to consider the priority that H
should be assigned to the fire protection research.
He said that the Auxiliary Systems Subcommittee has been working with
)
i the Staff on this matter. At the. July 23, 1987 meeting, the -Subcomit-1 tee heard presentations from representatives lof RES and SNL.
At this portion of the full Committee meeting, RES and SNL plan to provide the-current status of the Fire Risk Scoping Study.
He suggested that the Committee members focus on the following questions while listening : to i
the presentations:-
Is the Fire Risk Scoping Study being done in a timely manner?
Are there other issues that need to be addressed in the study :that are not being addressed?
Stating that Append.ix R -is supposed to deal with the issues associated with fire protection, Dr. Kerr asked whether. the Scoping Study is-
4 328TH ACRS MEETING MINUTES 4
predicated on' the assumption that the present-Appendix R is seriously deficient.
Mr. Michelson responded that it is not predicated on any.
assumptions concerning Appendix R.-
One of the1 tasks of this Scoping.
Study is to lock at the completeness of Appendix R and other. associated ~
regulations in addressing important fire-related issues.
In response to a question from Dr. Kerr, Mr. Michelson stated that there is no consensus that Appendix R is inadequate. He does not believe'that risks r from fire are-fully understood and he expects.that the Scoping Study will shed more. light to understand the contribution to risk from fires.
Mr. Ebersole commented that.although fires make up about 25% of the-probabilistic estimate of core-melt frequency,;which-is much larger than that from large LOCAs, he does not. understand why more resources are allocated' to research associated with LOCAs and no resources allocated to fire-related research.
He believes that allocation of resources-should be proportional to the contribution to risk.
Need for Fire Risk Research - J. Flack, RES Mr. Flack said that some of the basic questions to be answered prior to deciding the need for additional fire risk research are:
What is the risk of. fire at nuclear power plants?
What is the uncertainty?
j Is fire a major accident contributor?
2
!%s the risk / uncertainty changed over the last three to five years?
He said that the answers to these above questions should incorporate:
i Current state-of-the-art methodology Most up-to-date data Appendix R backfits Consideration of potential fire risk issues.
He stated that the ongoing Fire Risk Scoping Study.at SNL is expected to l
answer the above-mentioned questions and to provide information for use in deciding the need for additional fire-related research.
Mr. Flack said that the concerns expressed by the ACRS in its letter of July 16, 1986, such as consequential effects of inadvertent actuation of fire protection systems, sensitivity of control room components to fire and fire mitigation, fire-induced control system interactions, impact of seismic events on fire mitigation features, etc., are factored into the Scoping Study.
R i
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328TH ACRS MEETING MINUTES 5
j i
1 i
Fire Risk Scoping Study l
i Program Overview and Status - M. Bohn, SNL Mr. Bohn stated that the Fire Risk Scoping _ Study includes the following five tasks:
Task 1: Assess Uncertainties in Four Previous PRAs Task 2:
Requantify Fire Scenarios from Past PRAs Task 3:
Identify Potential Fire Risk Issues i
Task 4: Assess the Risk Significance of Potential Issues f
Task 5:
Evaluate Completeness of Appendix R Requirements He said that Tasks 1 and 2 have been about 90% completed and the
(
preliminary findings of these tasks have been documented in a draft I
report dated July.15, 1987.
Requantification of rast PRA Methodologies - M. Bohn, SNL Mr. Bohn said that in order to assess uncertainties in four previ-ous PRAs, they have requantified fire scenarios from past PRAs j
associated with four plants--Indian Point Unit-2, Oconee, Seabrook, 1
and Limerick.
The PRAs for Indian Point Unit 2, Oconee, and i
Seabrook were done by Pickard, Lowe, and Garrick (PLG) and the i
Limerick PRA was performed by NUS Corporation.
He said that PLG PRAs considered only transient combustible fires.
On the other hand, NUS PRAs considered transient combustible fires, self-ignited cable fires, and electrical cabinet fires.
Based on new data, the requantification of past PRAs was done in the follow-ing areas:
1 Fire initiating frequencies
,l Equipment susceptibility to fire Fire growth and spread j
Time to component failure He stated that both PLG and NUS PRAs indicated that:
Fires could cause LOCAs and transients.
A significant fire could occur once every 10 years.
The three temperatures (component damage temperature, piloted ignition temperature, and spontaneous ignition temperature) used by PLG PRAs are higher than those used in the NUS PRAs.
COMPBRN I and III computer codes have severe limitations In both PLG and NUS methodologies, fire growth is decoupled from suppression.
Engineering judgment is used to construct suppression models based on limited data.
The potential risk from misapplication of fire suppression agents is ignored.
,u-- -, - - -
a 328TH ACRS MEETING MINUTES 6
PLG PRAs give very little credit for suppression.
Results of Tasks 1 and 2 of the Fire Risk Scoping Study - M. Bohn, SNL Mr. Bohn summarized briefly the results from Tasks 1 and 2 of the Fire Risk Scoping Study.
He said that requantification of. fire scenarios included 'in four previous PRAs indicates major uncertainties in the following areas:
Quantification of fire frequencies Fire propagation modeling Fire suppression modeling Based on the comparison of the.requantified initiating event frequencies to those in the original PRAs for Limerick, it was found that the initiating event frequencies for cable fires and electrical panel fires increased, but decreased for oil fires. The main reason for this difference is that SNL used data from 23 events in the requantification process as compared to the Limerick PRA which used data from 5 events.
A comparison of the requantified propagation time (using COMPBRN I
III Code) indicates that COMPBRN III propagation times are considerably less than the COMPBRN I propagation times.
- However, Sandia's experience with COMPBRN III Code indicates that there are several inadequacies in the Code, including convergence problems, non-physical behavior, and inconsistencies.
Therefore, the results of COMPBRN III calculations should be viewed as tentative at this point.
Insights from Past NRC Fire Protection Research - S. Nowlen, SNL Mr. S. Nowlen said that the fire protection research program falls into three categories.
The first one that is intended to provide information to support NRC decisions was terminated at the end of FY 1986 owing to severe budget constraints. The results of'some of the research performed under this program are being documented.
The second one is Risk Methods Integration and Evaluation Program (RMIEP) which is scheduled for completion by the end of FY 1987.
The last one is the Fire Risk Scoping Study which started in FY 1987.
1 Mr. Nowlen said that between 1976 and 1985 several tests, such as the following, were performed:
Cable ignitability tests Twenty-foot spatial separation test
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328TH ACRS MEETING MINUTES 7
CO, suppression effectiveness tests Electrical cabinet tests Control room fire tests Some of the major insights derived from these tests are:
Even qualified or protected cabling can be ignited, burned, or damaged.
1 Water, C0[s,. and halon suppression systems eventually extin-guish fir Penetration seals fail if they have cracks.
Components fail before burning.
Mr. Nowlen stated that, F 3 sed on the available data, they have concluded that:
Fires occur about once every six years.
Fires make up about 25% of the total estimated core-melt frequency.
l Several fire problems have been overlooked.
Dr. Kerr commented that most of the major insights derived from previous tests are so obvious that one could have predicted almost all of them without performing any tests. Mr. Nowlen and Mr. Datta responded that, although some of the major insights look obvious, they are really not that obvious to the fire prctection.comunity.
The tests actually provided more information than one could have predicted.
Dr. Kerr was not convinced that these insights could not have been predicted without tests.
With reference to Mr. Nowlen's statement that fires occur about once every six years, Dr. Lewis asked, does it mean that 20 signif-icant fires occur per year? Mr. Nowlen responded-that is probably
. correct.
He added that several of the fires included in the data base are not serious fires.
Mr. Nowlen discussed briefly some of the fire tests conducted at the Factory Mutual Test Facility.
Based on the results of several tests conducted, they observed that for heptane, prophylene, and cable fires with ventilation from 1 to 10 room air changes /hr and with peak fires of 500-2000 kw, smoke reached about the 6-foot level within 2.5 to 8 minutes.
Based.on this, they believe that ventilation systems at nuclear plants are not adequately designed to remove smoke efficiently.
1 Mr. Michelson asked about the current status of the Factory Mutual Test Facility.
Mr. Nowlen responded that it is still intact.
However, it is up to Factory Mutual to decide whether to; keep this facility or to' dismantle it.
-- - - _ _ _ _ _ _ _ - _ =
328TH ACRS. MEETING MINUTES
'8 Mr. Ebersole asked, since the Underwriters! Laboratories.~arelinter-ested in: some. of; the fire _ data, why ' can't. they provide : financial support' to support the Factory Mutual Test Facility?
Mr. Nowlen '
responded that although the UnderwMters Laboratories are interest-ed. in obtaining. fire data, they are not.willing to? provide.any c i
financial support.
)
Approach to Identify New Fire Risk Issues - S. Nowlen, SNLu Mr. Nowlen said that the ~ following fire risk issues. have not been.
y addressed in past PRAs:
Control Systems Interactions
' Smoke Control L
Seismic / Fire Interation-Manual Fire Fighting Effectiveness!
Spurious Suppression' System'Actuationf
.j Cable / Component Total Environment' Survivability, He said that all of these; issues will be studied under Task 3'-
" Identify Potential Fire Risk Issues,", of the. Fire' Risk Scoping.
j Study.
-1 He stated that they pla' n to solicit opinions of experts 'in-various fields associated with fire safety, fire research, fire protection, risk analysis, plant ' design,- and plant regulation.
He mentioned that they have not been successful in obtaining information: from the Institute of Nuclear Power Operations (INPO). '
Indicating that SNL seems' to look at the effects of: seismically induced fires on non-safety systems :Dr.10krent asked whether they plan to look at the effects of seismically induced fires on safety systems. Mr. Nowlen responded that although it.is not-specifically.
l~
included in the slide, they do plan to.look.at this: issue.
Mr. Ebersole. comented that several plants extend sensitive cir-cuits from the control' room terminal boards to various regions of the plant to provide capability-for remote shutdown.
These extensions are vulnerable; to; fire, and providing,. adequate protection to these extensions is very difficult. He believes that this is a bad. practice and that:.the ~ Staff should. require elimination of such extensions. Mr. Nowlen responded that the.same issue was brought up, by: Mr. Ebersole during the July. 23, 1987-Auxiliary Systems Subcommittee meeting.
He'said that they plan to analyze this issue under the' ongoing Scoping. Study.
Mr. Reed commented that he does not believe-that fire _ protection research is going to solve all the issues.
Each decade there 'may be a major breakthrough in - the fire mitigation equipment and systems.. He believes that fire protection. activities will never
y 1
328TH ACRS MEETING' MINUTES 9
l i
- have an. end; they will continue. to grow forever.
_He 11sf also.
against automatic, fire suppression systems because spurious;actua-tion of: such systems will. create.~many unsafe conditions.i.He
. believes that emphasis should be given to installing. a ~ dedicated.-
I fire-barrier separated decay heat removal system in nuclear power 1
plants.
In his opinion, no additional fire protection research is i
necessary.
1 Mr. Michelson said that the. Subcommittee has : prepared an interim letter on the' adequacy of the scope.and direction of the. Fire Risk-Scoping Study for_ consideration by 1 the ~ full. Comittee.
After-4 completion' of' the Scoping Study, the. Subcommittee! will: prepare another letter and submit it to the full Committee for considera-tion.
IV. Augus't 6,1987 Meeting with NRC Commissioners (0 pen)
[ Note:
R.
F.
Fraley was the Designated Federal Official for this
-)
s portion of the meeting.]
The Committee discussed the format =for its. meeting with the Commission-ers on August 6,1987 in-open session regarding Implementation of NRC.
Safety Goals,. ACRS coments on Draft-NUREG-1150, and ACRS coments on-the Integrated Safety Assessment.
Meeting with the Commissioners:
[Commissionerspresent:
Chairman Zech, Commissioners Roberts,'Bernthal, andCarr]
ACRS Comments on Draft NUREG-1150, " Reactor Risk Reference Document" Dr. Kerr introduced the discussion of the Committee's letter dated July 15, 1987.
He noted that the letter was written in the: context of this being a draft report.
He emphasized the following coments: - (1) the study'did not' include external' events, (2) the manner of using expert opinion in the prediction of uncertainties was obscure and could lead to serious biases, (3) a number of applied analytical codes are -fairly new -
and not well validated, and (4) the Staff appeared to be_ unable to draw general conclusions :about the relationship.between perceived risks and adherence to the regulations.
Chairman Zech ' asked if the last comment had -been discussed with the Staff. Dr. Kerr and Dr. Lewis indicated that considerable' discussion on
-that issue'had been held with the Staff at previous ~ meetings.
Dr.. Lewis indicated that the Staff did not entirely understand. sampling plans and he recomended that some. statisticians be utilized _in future studies.
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I 328TH ACRS MEETING MINUTES 10 i
l Commissioner.Carr made a comparison of. the Committee's statement on -the adequacy of regulations to a comment by Dr. Okrent.that he would not say
' that all current plants would meet the safety. goals.
1 Dr. Okrent elaborated upon the comment indicating that, for, example, he.
was aware of. a very few regulations related.to the containment capabil-ity ' to cope 'with.a severe accident.
He also expressed an opinion that -
regulations are inadequate in. the areas related to the prevention. of j
core melt.
l '
Chainnan Zech indicated that the.Comission Lwould be. interested in
. l receiving the Staff's response-to the Committee's letter.
Commissioner Bernthal. asked ~ whether the Comittee hed - reviewed. the containment capability for. withstanding a core. melt accident in which-the reactor pressure vessel was breached.
Dr. Kerr indicated that the containment response at Peach Bottom (BWR-Mark I) had been fairly well analyzed but that' the Comittee had-i not reached a consensus. on this study or the general capability of J
containments to withstand severe accident conditions.
i Dr. Okrent indicated that the PRAs in NUREG-1150 were incomplete due to the deficiencies or limitations discussed in the Committee's letter and that he would caution one in drawing conclusions from this study about the safety of plants.
Chairman Zech indicated that the Staff 'should carefully - address the Committee's concerns on NUREG-1150 and he encouraged the Committee to continue their interaction with the Staff on this matter.
Commission-ACRS Discussions on Safety Goal Policy Implementation Mr. Ward sumarized the recommendations that the ACRS made in their May 13, 1987 report on implementing the Commission's. Safety Goal Commissioner Bernthal. asked that the concept of.the Safety Goal. Policy.
hicrar-chy and the intended levels of conservatism be ' explained. - The Members stated that it was intended that the ' hierarchy.. proceed from the-Safety Goal Policy goals to the lower levels of _ more directly implementable goals.
Conservatism would 'be added to each level to accommodate.the I
uncertainty associated with implementation. The conservatism should be such as to make the implementation robust but not so large as to change
' the intent of the original policy.
The implemer.tation of the Safety Goal Policy's "reasonab assurance of no core melt" statement and the proposed "less than 10~y/ year frequency for a:large release" guide were discussed.
These concepts appear to need to be further developed before implementation. Dr. Okrent noted that he believed that there was a need to develop a better capability for managing serious accidents and that the ACRS had highlighted this as an implication of the Chernobyl accident.
Dr. Okrent also stated that he'did not think that the' current-l
i
'328TH ACRS MEETING MINUTES 11 knowledge supported the claim that operating plants in general meet the Safety Goal.
Comission-ACRS Discussion of the Integrated Safety Assessment Program-Mr. Ward sumarized the comments made by the ACRS in their July 15, 1987 report on the Integrated Safety Assessment Program (ISAP).
The ACRS 1
believes that ISAP is.a good concept.and that the Program should be made widely. available in the future.. Chairman Zech stated that.NRC's; budget ~
currently provides resources for completing ISAP reviews for.all
]
-Northeast. Utilities plants (Haddam Neck and Millstone 1-3).
ACRS will continue to follow and to review the NRC Staff work in this area.
.'[ Note: 'At a closed August' 7,1987 meeting 'with ' the Commissioners the :
t discussion was about how to provide advice to the Comission on nuclear:
4 wasteissues.]
V.
Proposed Degree Requirements for Senior Reactor Operators (0 pen)..
[ Note:
H. Alderman was the Designated Federal Official?for this portion of the meeting.]
Dr. Remick, Chairman, Human Factors Subcommittee, noted that he had 1
previously given a subcommittee report to the full Comittee.
He cited the advance notice of proposed rulemaking of May 1986, which indicated the Comission was considering:. requiring Senior Reactor; Operators licensed after January 1,1991 to possess a baccalaureate degree 'in engineering or physical science.
He noted that the Subcommittee on Human Factors discussed this. in July 1986 and decided to defer coments - until the public coments were received.
He remarked that the Staff had issued. SECY-87-101,. which discussed the public coments.:
The Human Factors Subcommittee reviewed SECY-87-101 in June.1987 and decided to recomend a letter to the Comission.- Time constraints
- 4 prevented discussion during - the full. Cemittee's Jul further discussion was scheduled for the August meeting. y meeting,. so Dr. Remick introduced Paul Collins of the KMC Corporation for the first presentation.
Paul Collins, KMC Corporation Mr. Collins stated that he was here representing a group of'19 utili-ties, known as the Qualification of Reactor Operators Utility Group. He noted that this group has interfaced with the. NRC since 1981u on a variety of issues involving qualification of reactor. operators. He said the. working members of the group. consist of training managers and' operational personnel.
l
.aq
s 328TH ACRS MEETING MINUTES 12 Mr. Collins listed some objections' to the' proposed rule:
Lack of technical justification for.the proposed rule.
Those people responsible for the development of perfonnance-based.
~
training have not prepared any analysis ~of the skills and knowledge required. to perform as a senior operator;: in turn, they, have' not used these' data to indicate that a degree is;necessary.
NRC-sponsored studies indicated that a. degree. is notJ a necessity for ensuring the health and safetyfof the generalLpublic..
Reviews of-abnormal occurrences from 1981 to the present have' not revealed any. indication that a-lack - of1a degreed senior operator contributed to the occurrences or prevented - proper mitigating actions.
A concern that the requirement for a degreed shift supervisor will
-increase the turnover on shift..
The high turnover will result in the senior ~ operator positions-being staffed with recent college graduates.
People in charge'of shifts will. have less experience' than those now in charge.
Due to the high. turnover rate, senior operators will have' only the; minimum qualifications.
The senior operator position, instead - of being a capstone to an individual's. career will be a starting position with a high turn-over rate.
Mr. Collins noted that this could affect the quality of-the operators and decrease the overall operating experience, including" the operating experience of reactor operators.
~
Mr. Collins discussed the survey of operators, seniori operators' and shift technical' advisors. regarding their perception-of the advance notice of proposed rulemaking.
He noted that heihad interviewed 18 operators, 18 senior operators and 13 shift technical advisors..
To the question "Do-you believe all senior ' operators should be required to have a B.S. degree?" the majority said no.'
In response to the question about the ~ impact of degreed senior topera--
tors, the response was that there would be a high turnover in~ the senior.
operator position.
Some operators added-that there would be an adverse' effect on operator morale and a loss of-experience in'the senior opera-tor position.
-)
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-328TH ACRS MEETING MINUTES 13 i
l To the' question "Do-you believe senior operators should be required to.
receive technical training in engineering reactor theory that addresses -
l design basis events.'and severe accidents?" the : response was that.
additional training was necessary but.it did ~not have to be at the college level.
In response to the question "Do' you believe that a B.S. degreed senior operator on shift would enhance safety?" the majority did not believe-that he would.
In response to the question "Do you believe having-a degreed' senior -
operator on shift with only. two years' of nuclear experience would be detrimental to safety?" the majority. replied yes.
.l l
Dr. Kerr asked,."If. you had - your. choice in selecting. future senior j
operators responsible :for a ' plant,. would you choose someone with : or without a degree and put him-through aL training program to ultimately become an operator?" Mr. Collins responded:
"I don't think I would use-1 that as a criterion for selection."
Dr. Kerr asked if a high salary paid to to senior operators would reduce the turnover. Mr. Collins-replied that he couldn'.t say.
I Mr. Brinsfield, Senior Analyst, Delian Corporation 4
Mr. Brinsfield noted that he was one of the ' authors.of-a report-on operator response to incidents.
This report used a probabilistic. risk i
perspective and was prepared under contract to KMC and was published in:
i l
September 1986.
He noted that the purpose of' the study was to determine, using risk analysis techniques, any possible insights on whether attributes poten--
tially associated with a bachelor of-science. degree or equivalent in engineering or related s'cience ' would uniquely -enhance ' an experienced staff's ability to respond -to events that might occur. at a plant.- He stressed that these attributes are potential attributes.
.He said the scope of the study was to use existing data and infomation 1
from PRAs to define key accident " states" that hypothetical'lly could; occur.
He. remarked that a " state" is definod as a certain condition of i
the reactor defined by measurable-parameters at the plant.
The idea was..to identify circumstances in which operator response could' be enhanced by the attributes assigned to a bachelor's degree, and.then to compare that effectiveness with alternate methods ef achieving the same effect.
He stressed that this was a qualitative evaluation rather than a numerical study.
Based,upon an infomal survey among NRC, INPO,. Delian staff PRA ana-lysts and human. factors experts, fou'r. attributes potentially associated p
with a degree were developed. These four were:-
l l
~
l 328TH ACRS MEETING MINUTES 14' Engineering fundamentals the. ability to' predict' and project '
trends-General problem. solving - the ability.to accept a Llot of informa -
tion and use alternative. approaches in solving-problems.-
Integrated plant view - determing how the:effectsLof.one system may offset other things within the plant' Communication skills - the ability to communicate the1need, or-instructions,' to carry out actions to a' variety of personnel within..
the plant.
- Mr. Brinsfield observed that they had taken a number of. sequences. from PRA or: PRA-like studies for both PWRs :and BWRs.
These-sequences:were looked at-in detail with. operator-action-event _ trees. : Screening was.
performed to determine whether the attributes that were _ associated with the degree could be brought - to bear and. could possibly.- enhance. the.
operator's ability to respond.
The first screen was a timing screen 'in which it 'was assumed that any action that needed to be-taken less than 10 minutes into the accident would be:one of instant response based upon training already received.
Therefore, the above-listed four attributes would not really.be brought to bear within that time.
Anything that was greater than two-hour response time was assumed to be' responded to by the technical support center which'would then-be staffed by degreed experts.
Situations screened were those where the procedures 'were adequate 1to inform the operator what to do, or where the infonnation to the operator was very clear.
The study concentrated on identifying. accident states' where: ambiguous information and complexity of action might require some enhancement of operator skills.
~
~
Mr. Brinsfield. reported that for the states considered, although the degree might provide enhancement' of skills, the enhancement could be just as effectively achieved through some other alternative, such as a focused training effort.
Many of the states that were studied-had times too 'short in which to respond, or times long enough that the technical' support center could handle them.
The ability to achieve the necessary attributes a'ssociated with a degree is uncertain and focused training and improved procedures might be a j
_c
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328TH ACRS MEETING MINUTES 15 better way to achieve the desired ability-to. deal with unusual situations.
l J. Persensky,- Division of Licensen Performance and Quality 'Evalhation, Human Factors Assessment Branch, NRR Mr. Persensky discussed the advance' notice of proposed 1rulemaking.
He-noted that it would require that all applicants for a Senior Operator-License after January 1, 1991 would'have to hold a baccalaureate degree in engineerin degrees on a.g or physical. science.There would be an option for other..
case-by-case' basis.
The candidates for SRO license would:
have to. have two years ':of experience,' one year at 1 greater than 20-percent power.. The' senior operators that ~ are licensed as of that date would be grandfathered.
Those people applying' for a license just prior to that date would be allowed one reexamination, if they failed.
Concurrent with the proposed. rule.would be~a policy statement that would-encourage industry to provide career paths for the degreed licensed operator.
Mr. Persensky noted that 200 responses were received regarding > the advance notice.
Five _ favored the proposed rule and 195 were against it.;
The negative responses noted that.the rule was not necessary, experience was more important than education,..it would have a negative: impact on L j
safety, there would be increased turnover, and it would block the career-path of reactor operators who did not have _ degrees.
Mr. Persensky noted that the NRC established a. peer panel of experts to determine whether a degree was necessary for reactor operators.. The peer panel concluded that a degree was - not - necessary for. ' reactor I
operators.
The-peer panel suggested that the STA should be replaced with a shift engineer who would be licensed and would possess a degree.
The advantages of a degree were listed as opening career paths to upper
~
management, enhancing professionalism on
- shift, and as enhancing accident' management on shift.
s i
The disadvantages of the degree rule would be high turnover, the undesirability of shift work, blockage of the career path of non-degreed operators ~, and damage to the morale of the operating staff.
Mr. Persensky noted three options in SECY-87-101.
1.
Proceed with the proposed rulemaking 2.
Have a degreed senior manager on shift j
3.
Use a policy statement that goes beyond the current SR0/STA.
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'l i
e 328TH ACRS MEETING MINUTES 16 The Comission has voted in; favor'of proceeding with the rulemaking and-
{
changing the effective date to four years.after the rule is issued.
Committee Discussion Dr. Remick noted that, following the Human Factors Subcommittee meeting, one. ACRS member indicated. that he wouldi go - along with: the degree i
requirement and the z others were. opposed.
Dr. Remick' remarked that-perhaps the Comission should articulate the goals.it' wishes to-accomplish, and ' stimulate industry to undertake consideration of how:
q these goals should be accomplished.
1 1
Dr. Remick also noted that the Subcommittee came to.the ' conclusion that i
there.wasn't any technical rationale for degree requirements-for. Senior i
i Reactor Operators as this was a policy decision.
t Mr. Reed remarked that a college degree requirement will. have as many ills as it has advantages and should not be a factor in-operator selec--
tion.
There was discussion on the choice of engineer or physical science degrees and why.wasn't a nuclear engineering.. degree specified.
The consensus was that the choice of. engineering or science degrees was a
" filter" to specify a basic level.
Dr. Moeller expressed disappointment that the Professional Reactor Operators Organization had not set up standards to advance the qualifi-cations of reactor operators.
The Committee decided to write a letter to Chairman Zech'regarding this subject.
.VI.
Improved Westinghouse Standard Plant (0 pen).
~
[ Note:
M. M. El-Zeftaw portion of the meeting.]y was' the Designated Federal Official for this Mr. Ward, Chairman of the Westinghouse Reactor. Plants Subcommittee; briefly sumarized the purpose of-the discussion as to inform the.
Comittee' in regard to the current NRC Staff review schedule for the Westinghouse Advanced Pressurized Water Reactor (WAPWR).
H. Berkow, Project Director, NRR, gave a brief-status ' report on the review of RESAR-SP/90.- Initially Westinghouse -(W,
started submitting its design in modules to the' RR)C-Staff for review.in Oct According to W, the purpose of this. modular approach is to speed up the-licensing proces.s.
W completed its final-submittal (Modules 1 through
- 16) in March 1987. Todule 16.is PRA/ severe accidents studies.
' l i
4 328TH ACRS MEETING MINUTES 17-In September 1986, the NRC Staff indicatedL that, ~due to. budget and resources' limitations, they had no integrated schedule to continue' and '
complete the WAPWR, review..
Currently' due to the NRC-Staff reorganization,The Staff has resumed and reactivated its review with a draft SER ' expected to be issued in April 1988.
The ' NRC Staff -- is -
requesting several ACRS: meetings, starting in.- June 1988.
The preliminary design approval '(PDA) ' decision December 1988.
date~ is expected to be Mr. Ward stated that the review of this program is highly;important and' recomended that it will be appropriate.for the-ACRS. Subcommittee on W' Reactor Plants to meet with the NRC. Staff and W representatives in a series-of meetings beginning well. in advance of the:1ssuance of a draft.
'SER.
. 1 i
Dr. Okrent comented that the ACRS would like to review the scope' and j
results'of the PRA used by W in support of its; design..He also comment-1 ed that a comparison of the design with modern plants of.a'similar type' y
I in 'the U.S. and abroad (e.g.. Vogtle, Paluel, Sizewell' B, and KONVOI l
plants)'would be useful. The rest of the members. agreed.
]
Mr. Reed commented'that a comparison with the EPRI requirements document i
for future PWRs will also be beneficial.
Mr. Wylie agreed with that concept.
j The Committee decided to write a letter. to Chairman Zech regarding this subject.
VII.
Proposed Final Broad Scope Amendment to General Design Criterion-4 (GDC-4) Rule (0 pen)
[E. G. I meeting..gne was the Designated Federal Official for this portion of the
]
Dr. Shewmon, Chairman of the Metal Components. Subcommittee, reported.on-the change to the GDC-4 rule that will eliminate the dynamic effects.of ~
l postulated pipe breaks in all' high energy lines. This concept hinges-on the idea that you can demonstrate that pipes would > leak and be detected i
before the pipe would break.
If this is so, then you can eliminate some l
awkward, heavy. pipe whip restraints which make inspection of. the pipes.
difficult.
The Subcommittee has reviewed the matter, and felt that i
adoption of this concept would increase plant safety by making the piping system more accessible for inspection.
The Subcommittee has-approvedtheleak-before-break (LBB) concept.
Mr. Reed felt that LBB should'not apply to BWR piping systems because of '
corrosion problems even if the pipes were subjected to stress improve--
ment techniques and the plant' employed hydrogen water chemistry..He stated that a long history of unfailed samples is needed before LBB is applied.
In response to a question by Mr. Ebersole, it was stated that -
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328TH ACRS MEETING MINUTES 18 LBB at the present time applies. to pipe.- sizes about six '. inches or L greater.because of. leak detection. requirements and pipe; crack stability t
considerations.
I L
R; Bosnak, RES, discussed GDC-4 ' rule changes with respect 1 to public comments received, and issues raised by the CRGR: and the ACRS. :He stated that the objectives at this-point are'to request ACRS_ approval toL publish the amended ~GDC-4' in - the Federal Register and ' to seek ACRS approval to issue ~ Standard Review Plan (SRP) Section 3.6.3 for public comment. SRP 3.6.3 implements the procedures to apply GDC-4.
Overall, 23 letters were received-from ' the public comenting: on ' the proposed rule.
Major comments and their resolutions ~ are listed as follows:
Factors of 10 on leakage detection capability and 'of 21on crack stability are considered too. conservative.. Because of too much' uncertainty," however, the Staff will stay with the given factors.
With respect to equipment qualifications, the Staff will, on: a case-by-case basis, entertain. a relaxation Lin its environmental requirements.
The design basis for the containment and-ECCS remains unchanged, although it was. stated that the Staff was studying this matter for possible relaxation.
Lower-bound material properties - are acceptable instead of from archival specimens.
1'i' Creep temperatures have been chan 800*F for austenitic. materials, ged to 700*F for ferritic and to instead of using 750'F for both materials.
LBB applied to piping subject..to intergranular. strt!ss corrosion
~
cracking (IGSCC) will be reviewed by the Staff on a. case-by-case basis.
If the piping material:' susceptible to IGSCC has flaws of
-less than 10% of the wall thickness:and has provided two mitigating
. methods (stress improvement, water chemistry) LBB may apply; a
similar approach will be taken if-the. pipe is fabricated with nonsensitized materials.
Pipes' with weld overlay repairs are not acceptable for LBB consideration.
The Connittee questioned the application. of LBB to pipes.with a history of water hammer.
R. Bosnak stated that if the applicant or licensee can satisfy the Staff'of an extremely : low likelihood of water hatiner, the Staff will allow the LBB concept to apply; however, this may be diffi-cult to show.
- With respect to heavy component support. redesigns of plants!- already built, R. Bosnak specifically pointed to redesign only of snubbers and-high-strength threaded fasteners.
Since the LOCA load has been
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328TH ACRS MEETING MINUTES 19 I
i eliminated in the LBB concept,. the snubber and threaded fastener
{
capacities. can be reduced with corresponding benefit -of more_ reliably j
designed snubbers and fasteners not subject, to.: stress : corrosion cracking.
With these lower, capacity. snubbers, a. seismic event that-is.
beyond the design basis could fail the heavy component. lateral supports.
With these supports assumed to-fail, and accounting for the mechanistic failure geometry of.the components,; analyses have. been -performed indicating that the.. piping system.. attached - to the component will not j
fail.
In new plants. the SSE : loads will be increased, based on recent-D studies done on seismic hazards that resulted in properly designed.
' supports.
Pressurization of structures: protecting. essential' equipment is retained :
even when LBB is demonstrated.
Energy release time is three seconds'and break size is equal to a single' pipe cross-sectional area.
In reply to Mr. Michelson's question, regarding high energy line break inside of containment -that could damage.the containment, R.. Bosnak stated that support shields or pipe whip. restraints would stay.
They cannot be removed as the containment design basis has not changed.
The Comittee did not object to issuing for public comment SRP 3.6.3 concerning the implementation of the amended GDC-4 rule.
VIII. Nuclear Power Plant Operating Experience (0 pen)
)
[ Note:
H. Alderman was the Designated Federal Official for this portion of the meeting.
Introductory Remarks - Mr. Ebersole Mr. Ebersole noted that five events, selected from a list covering the I
past two months, would be discussed.
He noted that if there were time left over, a few other events would be discussed..He then turned' the discussion over to Pat Baranowsky.
P. Baranowsky, NRR, Events Assessment Branch I
P. Baranowsky said that the first event concerned a reactor trip breaker failure at the McGuire plant and some-s'ubsequent malfunctions that occurred with similar reactor trip breakers at other plants.
T. Peebles, Region II T. Peebles noted the onset of the problem was on July 2,1986 when reactor control rod drive tests were being conducted at McGuire, Unit 2 as part of the startup testing following the refueling outage.
After several rod drives had been successfully tested,. the 2B reactor' trip breaker failed to open during a manually-initiated trip from the 1
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328TH ACRS MEETING MINUTES 20
)
main control panel.
The trip attachment had overheated, shorted and opened.
Three operators in the control room stated that the reactor trip lights showed that the reactor trip breaker had tripped.
Subse-quent observation of the events recorded printout indicated that it had not opened.
Initial results indicate that a combination of high cycle wear and a broken weld led to the jamming of the mechanism.
The cause of the erroneous status light indication is not known.
A second event was at Catawba on July 24, 1987.
Visual inspection of the welds on a Catawba breaker resulted. in noticing problems with welds on the pole shaft.
One weld indicated a small crater crack.
One weld showed a quarter-inch long lack.of fusion where the center pole was attached.
The NRC is. sending the pole shaft to the - Franklin Research Institute'for evaluation of the welds.
Dr. Kerr asked if there was y indication that the Catawba breaker was defective when installed.
T. Peebles responded that there are several i
p'ossible contributors, one of which was that the welds were of question-1 able quality.
Mr. Thomas (Duke Power) added that, based on preliminary visual inspection, they believe the welds were of poor quality, i
Procedural Inadequacies Regarding Switch-Over Circulation of the I
Westinghouse Plants - R. Lobell l
R. Lobell's discussion concerned the procedures for switch-over during a large break LOCA.
Suction is initially taken from the.RWST.
Operator actions are necessary to take suction from a sump when the RWST runs low.
This issue arose during Florida Power and Light Turkey Point design basis reconstitution efforts.
As part of this. effort, Turkey Point looked at the basis for the time allowed the operators for switching l
from the RWST to the sump during a large break LOCA.
Their procedures allowed them to turn off all ECCS pumps for a period of 10 minutes while they were making the switch-over.
Westinghouse was asked the basis for the 10 minutes. The basis was somewhat obscured, so Westinghouse reanalyzed the basis using current information and found they couldn't support the 10-minute number.
The current basis allowed two minutes for all ECCS pumps to be shut off.
The previous-assumption was that the water in the core boiled at a-certain rate, depending on the decay heat.
The latest data showed that not only would the water boil off but it would leave the break from entrainment, due to the difference in heads from the colder water in the downcomer and the warmer water in the core.
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I 328TH ACRS MEETING MINUTES 21 l
Turkey Point then changed their procedures so they left the high pres-I sure pumps running.
They turned off the low head pump, switched the i
suction from the sump to the low head pump, started that pump and then switched the high head pump so there was a continuous source of water to the vessel.
l d
Westinghouse was contacted to see if any other plants might be affected.
Westinghouse supplied a list to the NRC and the NRC has contacted the
{
other plants.
All the utilities involved have made changes to their i
procedures to accommodate this potential effect, j
l In response to a question regarding what type of code was used Mr.
Lobell stated that it was not an Appendix K calculation but rather a j
{
best estimate code.
1 Mr. Ebersole noted that the subcommittee would be interested in hearing
)
about the follow-up on a generic basis, i
i Containment Spray Degradation, Pilgrim, J. Carter J. Carter discussed containment spray nozzle and header system degrada-l tion at the Pilgrim plant.
Pilgrim was in the process of modifying their containment spray nozzles and header system in June of this year.
They had decided to remove the existing nozzles from the upper and lower spray within the drywell and replace them with plugged nozzles.
Six out of the seven holes in the nozzle would be capped, and this would -reduce the flow through the nozzle in order to have a better flow distribution within the drywell and also to provide adequate cooling for their drywell wall.
When they removed the nozzles, they found that rust had accumulate' d within the header and in the nozzles themselves to a depth of perhaps approximately one-half inch.
The Staff was concerned about the generic implications. The concern was that there might be other plants with plugged spray nozzles' in the containment area.
In addition, concern was expressed as to whether the testing that was done to verify the operability of the spray system was adequate.
Mr. Carter noted the piping within the drywell is carbon steel.
The environment is somewhat humid, and the spray system is wetted from time to time.
One of the concerns was that the air in the system with the wet carbon steel piping might be the source of the rust.
Several utilities were asked if they would make a quick check of their nozzles and their piping.
None of the utilities that checked found any appreciable rust.
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.j 328TH-ACRSLMEETING MINUTES 22.
j s
j Pilgrim - and ' the NRC - Staff: are' independently studying the ' problem - to -
determine the source of the rust.
Mr. - Reed suggested the' possibility of nitric acid' being produced as a
)
result of neutron irradiation.
He asked if anyone was' checking this ~
possibility.
Loss - of Rea:: tor Coolant Inventory During Maintenance at' North Anna, L.
- {
Engle The event at North Anna occurred over a three-day time span. ' North ' Anna l
was coming out.of: a 59-day 1 refueling outage on June. 17, 1987.. 0pera-l tions were commenced to leave mode 5. and enter mode -4..
Two reactor cooling pumps were initiated.
After several hours, the A pump tripped
.i and it was determined that the motor had grounded and would need to be replaced.
The other ' pump was' secured.
In order to facilitate motor.
removal, reduction in system temperature = and pressure was-started.
The pressurizer was vented.
The-primary level was reduced to about ' 20-percent and the PORV closed.-
~
I On June 18 the pump was l uncoupled.
A small leak of about 1.5 gpm developed on the pump shaft.from the seal injection.-
The operators continued. to monitor' at the. 20% pressurizer level, and went to a procedure of volume controlL tank float. The operators assumed-that they could maintain makeup in this system for balance in inventory-due to the seal leakage by maintaining the pressurizer level at 20% and watching the makeup control tank.-
On June 22, Westinghouse requested that, if the pump se'al leakage could be reduced, it would help repairs.. The operators decided to raise the pressurizer level to cycle the PORV to vent and then reduce the pressur-izer level to generate a slight vacuum.
- 1 In the process of cycling, they determined something unusual was going i
on because the pressurizer relief tank.went from positive to negative i
and they realized that they had had a vacuum for some time, i
l They immediately began a quick inventory of reactor coolant.
They 1
immediately rutored makeup and, by the time they restored inventory, I
they had to add about 17,000 gallons of. borated water.
.)
1 During the event there was adequate inventory at all times for residual j
4 heat removal.
l.
Mr. Engle pointed out that:
Using a pressurizer level for' monitoring reactor coolant system
.inventorycanbemisleading(aswasseenatTMI).
4
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328TH ACRS MEETING MINUTES 23 During modes 4 and 5, there isn't much instrumentation available.
During these modes, Reactor Coolant System Inventory should be carefully measured, mass flow balances should be checked and all available level indications should be constantly measured.
If RVLIS had been used during this period, the operator would have been alerted to this problem.
The operator should have realized that the coolant was " cold" and did not have the residual heat to maintain a slight bubble in the pressurizer.
As follow-up to this event, the licensee issued an event description to INPO, and carefully examined their procedures.
The procedures will require that RCS makeup in this regime be carefully monitored and every piece of available instrumentation be watched.
RVLIS is going to be j
l included in the procedures.
1 l
Questionable Procedures During Shutdown, Fermi Power Plant, P.
Baranowsky P. Baranowsky told the Committee of a situation at the Fermi plant in which an operator trainee was assigned to record reactor coolant system temperature as a function of time.
The plant was going through some corrective maintenance on the RHR system and they wanted to monitor primary coolant temperature. The trainee monitored the temperature hour after hour.
No one had ever told him what to do when it goes to above 200 degrees.
Some one finally realized they should check the tempera-i ture.
This situation showed some flaws in the way operations were carried out during shutdown.
Millstone - Operations During Shutdown, P. Baranowsky This discussion concerned monitoring temperature in a RKR system that was' isolated from the reactor coolant system.
The plant was in cold shutdown.
The operator erroneously used the valved out RHR temperature to monitor reactor coolant temperature.
Reactnr coolant temperature increased above 200*F for about 5 minutes, causing mode change from cold S/D to hot S/D.
Scram With Complication, Brunswick, P. Baranowsky Mr. Baranowsky discussed a scram event at Brunswick which involved a common mode failure of some safety relief valves.
In this event a voltage regulator problem resulted in a turbine ger. orator trip.
As part of the recovery from this event the operators were using RCIC to control the level.
They were using safety relief valves in manual to
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i 328TH ACRS' MEETING MINUTES 24 control pressure.
It turned out that one of the three safety relief.
valves that were being~ used to control pressure would not open on demand I
when they tried to use it.
They were successful in getting down to I
shutdown with the other two.
-l The concern is that the safety relief valves are part of the automatic I
depressurization system which is sometimes necessary for operations at BWRs, and they play a part in certain loss of coolant accidents.
The NRC requested the licensee to test these valves at startup and, when l
they did, a different valve failed.
1 The licensee had just had maintenance on the solenoids for these safety i
relief valves.
The licensee replaced the newly maintained solenoids j
with some older ones that they knew were operational and sent the newly l
maintained solenoids to Wylie Labs. for testing.
]
When the solenoids were disassembled, they found a substance on them j
that was identified as "locktight" that was stuck around. the plungers.
)
- 1 What had occurred at Brunswick was that a new technician at Target Rock
)
had worked on the maintenance of these valves.
He was cautioned not to j
let the excess "locktight" drip.
The "locktight" is used. to hold nuts in place on the valve.
Unfortunately it was allowed.to drip on the l
plunger.. The "locktight" sets up where there is a lack of air.
The valves were tested while they were going up in power.
They worked at that point. When the plant was inerted, the locktight set up and caused i
the plunger to seize.
1 IX. Executive Sessions (0 pen / Closed)
A.
SubcommitteeReports(0 pen / Closed) 1.
WasteManagement(0 pen) 4 Dr. D. W. Moeller, the Subcommittee's Chairman, summarized the trip to the University of Arizona, to the University's field facilities near Superior, Ariz. where research is being conducted on the evaluation of groundwater flow and transport methodology in an unsaturated medium, and to the DOE's waste management activities at. the Nevada Test Site where research is being conducted on problems. pertaining to the design, i
construction and operation of a HLW geologic repository in a i
tuff medium.
He indicated that each ACRS Consultant that
{
participated in the visit will be preparing a - report to the ACRS on his views of the activities observed.
A further report to the ACRS is likely.
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0-jj 328TH ACRS MEETING MINUTES 25 i
2.
New Members (Closed)~
The nomination of candidates to fill the existing and future l
vacancies was discussed.
Nominations are' to be submitted prior to the September full Comittee meeting.
Mr. Fraley agreed. to explore the possibility. that exceptions- 'may be allowed to - the current limit. on the number of terms ' that a member may serve.
B.
Reports, Letters and' Memoranda '(Open) l 1.
ACRS -Coments 'on Proposed Final Broad Scope Rule to Modify-General Design Criterion-4. Environmental and Missile Design -
Bases (GDC-4). (Report to Chairman Zech dated August. 12, 1987)
The Comittee endorsed' the proposal to modify GDC-4 to allow exclusion of the dynamic effects-of postulated pipe ' rupture 4
from a' nuclear power ' plant's design - basis.
Additional com-ments.by D. Okrent and G. Reed stated that it was not prudent
- to permit application of " leak-before-break"' to BWR high energy piping at this time.
j 2.
. ACRS Report on Fire Risk Research Study (Report to; Chairman Zech dated August 11, 1987).
l l
The Commfttee reported that the Fire Risk Scoping Study being conducted by - Sandia National.. Laboratories for the NRC -is progressing satisfactorily and is targeting the various concerns expressed by the ACRS.
3.
ACRS Review of Application for' Preliminary Approval of the Westinghouse RESAF/5P-90 Design (Report to Chairman Zech dated J
August 11,1987)
The Comittee reported that it believes that the review of i
this application 'is highly important and expects to partici-pate extensively in the review.
The ' letter identified some-i subjects the Comittee would like to have addressed 'in the review.
4.
ACRS Comments on the Advance Notice of Proposed Rulemaking on Degree Requirements for Senior Operators sReport to Chairman j
Zech. dated August 12,1987)
I The Comittee recommended that. the proposed rule. be. recon-sidered.
Additional comments _ by G. Reed said that he ap-plauded the ACRS letter and wished to add further support.
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4 328TH ACRS MEETING MINUTES 26-i 5.
ACRS Action on Proposed Regulatory Guide EE 006-5, "0ualifica-I tion of Safety Related Lead Storage Batteries for Nuclead
.i Power Plants" (Memorandum to Mr. Stello dated August 11,,
j 1987) i The Committee concurred in the proposal to issue the Guide for
'public comment and said that it expects to' review the. final.
i version together. with the public coments and' the NRC Staff's ~
response.
{
6.
. ACRS Action on the Proposed Section 3.6.3, " Leak-Before-Break j
Evaluation Procedures." of-the NRR 5tandard Review Plan (Memorandum to Mr. 5tello dated August 12, 1987)
The Committee concurred.in the proposal to issue the proposed section of the' NRR Standard Review Plan for public coment and said that it expects 'to review the final version of this sec-
- {
tion together with public comments and the NRC Staff's response.
7.
ACRS Comments on Proposed DOE Bill to Establish Independent.
j Safety Board (Memorandum to Mr. Malsch dated August 12, 1987)
The Comittee provided 'the Deputy General Counsel with its
-)
comments on those portions of the proposed legislation that pertain to ACRS activities.
C.
OtherCommitteeConclusions(Closed) 1 1.
Ad Hoc Planning Subcommittee.
.)
An Ad Hoc Planring: Subcommittee on advice to the Commission on waste management was formed.
Dr. Moeller. is Chairman with Drs. Siess, Steindler, and Remick as members.
The subcomit-tee is to attempt'to develop a plan whereby a group of members can directly advise the Commissioners on matters related 'to I
waste management.
D.
FutureActivities(0 pen) 1.
Future Agenda-The Comittee agreed on the tentative' agenda for. the 329th meeting, September 10-12, 1987, as shown in Appendix II.
2.
Future Subcommittee Activities
. J.
A. schedule of future subcommittee activities was distributed to members (Appendix III).
The 328th = ACRS Meeting was adjourned at 11:20 a.m., Saturday, August 8, 1987.
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APPENDICES TO MINUTES OF THE-328TH ACRS MEETING 4
AUGUST'6-8, 1987
'1 l
WASHINGTON, D.C.
i Appendix I
. List of Attendees j
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Appendix 11
' Future' Agenda Appendix III Future Subcommittee Activities
)
' Appendix IV Other Documents Received l
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321 322 323 324 325 326 327 328 329 330-331 332' ACRS MEETING DATE:
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ATTENDEES Dr. William Kerr, Chairman Dr. Forrest J. Remick, Vice Chairman v
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Mr. Jesse C. Ebersole V
Dr. Harold W. Lewis-V i
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Dr. Carson Mark V
/
7 Mr. Carlyle Michelson y
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Dr. Dade W. Moeller-y Dr. David Okrent V
V Mo Mr. Glenn A. Reed v
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i Dr. Paul G. Shewmon DC Dr. Chester P. Siess
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Mr. David A. Ward
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Mr. Charles J. Wylie
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q ATTENDEES 328TH ACRS MEETING Thursday, August 6, 1987 j
NRC ATTENDEES I
l R. Hernan, NRR.
1 John A. Zwolinski, OSP j
B. D. Liaw, OSP i
B. Rotella, OSP l
B. Zalcman, OSP l
T.d.
$ito onsultant}
PUBLIC ATTENDEES I
B. Whitlock, Ace-Federal l
S. Savage, NUS R. Borsum, B&W j
Eve. Fotopoulos, SERCH, Bechtel l
Jo Owen, Aeres I
Nancy Caldwall, Wash. Ind. News i
David Pace, AP Dave Airozo, McGraw-Hill Phil Shott, Wash. Ind. News Steve Nowlton, Sandia Labs.
Michael P. Bohn, Sandia Labs.
James Stoda, Bishop Cook, Purcell & Reynolds L. Connor, DSA P. Collins, KMC Wes Brinsfield, Delian i
Dave Natepka, Wisconsin Public Service Bruce A. Snow, Rochester Gas & Electric V. L. Kilpack YAEC David E. LaBarge, Yemont Yankee / PROS Bill Gott, Public Service E&G Mark Phillis, Bishop, Cook Purcell & Reynolds Art Bivens, NUMARC David L. Lambert, TVA R. L. Gridley, TVA W. H. Hannon, TVA C. H. Fox, TVA I-2.
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i 328TH ACRS MEETING ATTENDEES Friday, August 7, 1987 Saturday, August 8, 1987' NRC ATTENDEES D.-L. Lambert, TVA 1
R.'Hernan, NRR R. Bosnak, RES J. O'Brien, RES H. Berkow, NRR T. Kenyon, NRR C. Sellers, NRR B. Elliott, NRR K. Wichman, NRR S. Lee, NRR P. Norian, RES R. Johnson, RES T. King, RES G. Toalston, NRR T. Cintula, AE0D D. Hood, NRR W. Orders, R. II T. Peebles, R. II R. Lobel, NRR T. Carter, Jr., NRR L. Engle, NRR PllBLIC ATTENDEES B. Whitlock, Ace-Federal A. Wyche, SERCH, Bechtel M. Shannon, Westinghouse T. VandeVenne, Westinghouse W. M. Sch'.vley, Westinghouse P. Riddle, Techmark S. Reynaud, Techmark W. F, Guerin, Westinghouse J. C. Scarborough, NUS K. H. Nam, Bishop, Cook, et al.
S. A. Bernsen, Bechtel D. Airozo, McGraw-Hill J. E. Thomas, Duke Power Co.
M. A..Schoppman, Florida Power & Light Co.
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1 8
I APPENDIX II FUTURE AGENDA 1
September 10-12, 1987 ECCS Evaluation Models (DAW /PAB).- Discuss proposed revision of l
ID CFR 54 Appendix K regarding uncertainty methodology.
Generic Issue 93, Steam Binding of Auxiliary Feedwater System Reliability (DAW /PAB) - Discuss requirements for auxiliary feedwater system reliability.
Generic Issue 124, Auxiliary Feedwater System Reliability (DAW /PAB)
- Discuss requirements for auxiliary feedwater system reliability.
.]
Improved Safety Features for LWRs'(CJW/RKM) - Discuss response to Chairman Zech_'s inquiry.
l USI A-47, Safety Implications of Control Systems'(JCE/MME) -
Discuss concerns identified by D. Basdekas.
Zion Full Field Exercise (DWM/CYM/EGI) - Hear reports and discuss the exercise at the Zion Nuclear Plant.
i l
Generic Issue 23, Reactor Coolant Pump Seal Failure (DAW /PAB) -
Status report to be given.
Qualification of Nuclear Power Plant Components (CJW/RKM) -
Triefing by NRC Staff on seismic walk-through of the Zion Nuclear Power Plant.
EmergencyPlanning(DWM/EGI)-DiscussproposedACRScomments regarding emergency planning.
Trojan Nuclear Station (JCE/HA) - Briefing regarding excessive thinning of main feedwater piping.
USI A-44, Station Blackout (CJW/MME)
Briefing on proposed rule on station blackout.
I Waste Management (DWM/OSM) - Discuss August 17-19 subconinittee meeting and report resulting from discussions during that meeting.
October 8-10, 1987 Renewal of Nuclear Power Plant Licenses (CJW/RKM) - Discuss NRC policy on extension of licenses beyond 40 years Technical Specifications (CYM/EGI) - ACRS convrents requested on l-interim policy statement.
APPENDIX II
_m.___________
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'II-2 License ' Requirements for' St'orage of Spent Fuel (CPS /0SM). - Discuss proposed revision of 10CFR72.
AE0DAssessmentofOperatingEvents(HWL/RKM)'-Discuss-AE00 assessment of selected operating events.
i PWR Seismic Design Margins (D0/RPS) - Briefing by NRC Staff (RES)..
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AUE 88 887 tryisto ACRS SUBCOMMITTEE MEETINGS Waste Management, August 17-19, 1987, 1717 H Street, NW, Washington, DC, r
(Merrill), 8:30 A.M., Room 1046. The Subcommittee will review several pertinent HLW, LLW, and related research with the NMSS and RES Staffs.
Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the nights of August 16, 17 and 18:
Dr. Moeller LOMBARDY Dr. Carter NONE Dr. Mark LOMBARDY Dr. Krauskopf NONE Dr. Remick NONE Dr. Orth NONE Dr. Shewmon NONE Dr. Parker NONE Dr. Steindler NONE Dr. Trifunac NONE Auxiliary Systems, August 18, 1987, 1717 H Street, NW, Washington, DC (Duraiswamy), 8:30 A.M., Room 1167. The Subcommittee will discuss the heating, ventilating, and air conditioning (HVAC) system malfunctions and their impact on safety systems.
In addition, it will discuss problems associated with instrument air systems AE0D findings concerning the instrument air system malfunctions and its recommendations to alleviate this problem. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of August 17:
Mr. Michelson DAYS INN Mr. Reed DAYS INN Mr. Ebersole DAYS INN Mr. Wylie DAYS INN Advanced Reactor Designs (Closed), August 25-26, 1987, Idaho Falls, ID (El-Zeftawy), 8:30 A.M.
The Subcommittee will visit and tour the EBR-II, l
FMF, ZPPR, and TREAT facilities to gain information on the problems per-taining to the design, construction, and operation of liquid metal reac-tors. In addition, the ANL representatives will brief the Subcommittee regarding the overview of EBR-I and EBR-II design and operating history, IFR program overview, metal fuel performance and fabrication, and fuel cycle demonstration. These briefings may be held in open sessions.
Attendance by the following is anticipated, and reservations have been made at the Westbank Motel (208/523-8000), 475 River Parkway, Idaho Falls, ID for the nights of August 24, 25 and 26:
Mr. Ward Dr. Mark Mr. Ebersole Dr. Shewmon Regional and I&E Programs, August 28, 1987, Region V, 1450 Maria Lane, Suite 210, Walnut Creek, CA (Boehnert), 8:30 A.M.
The Subcommittee will review the activities under the control of the Region V Office. Lodging will be announced later. Attendance by the following is anticipated:
Dr. Remick Mr. Ward (tent.)
Mr. Michelson Mr. Wylie Dr. Moeller Mr. Reed APPENDIX III
\\
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1 A
i 1 i
'.Y Future LWR Designs, September 8, 1987, 1717 H Street, NW, Washington, DC 1
(Major),1:00 P.M., Room 1046. The Subcommittee will discuss its reply
'w
'l to the 4/22/87 Staff Requirements Memorandum regarding the feasibility.
benefit, and cost effectiveness of selected and combined systems as recom-mended in the ACRS letter of 1/15/87 on Improve LWRs. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of September 7:
1 Mr. Wylie DAYS INN Mr. Reed DAYS INN Mr. Michelson DAYS INN Dr. Siess ANTHONY Dr. Okrent ANTHONY Generic Items, September 9, 1987, 1717 H Street, NW, Washington, DC l
(Duraiswamy), 8:30 A.M., Room 1046. The Subcommittee will continue the discussion on the effectiveness of the programs that address generic issues and USIs. Also, it will discuss with selected licensees the contri-bution to plant safety resulting from the implementation of the resolved 4
l generic issues and USIs. Attendance by the following is anticipated, and I
reservations have been made at the hotels indicated for the night of September 8:
Dr. Siess ANTHONY Dr. Moeller LOMBARDY Mr. Ebersole DAYS INN Dr. Remick NONE Mr. Michelson DAYS INN Mr. Wylie DAYS INN 329th ACRS Meeting, September 10-12, 1987, Washington, DC, Room 1046.
Extreme External Phenomena, September 17, 1987, 1717 H Street, NW, Washing-ton, DC (Savio), 9:00 A.M., Room 1046. The Subcommittee will discuss the NRC Staff's Seismic Design Margins Program and the application of the methodology to Maine Yankee. Lodging will be announced later. Attendance by the following is anticipated:
Dr. Siess Dr. Moeller Dr. Lewis Mr. Wylie 330th ACRS Meeting, October 8-10, 1987, Washington, DC, Room 1046.
Auxiliary Systems, September 29, 1987, 1717 H Street, NW, Washinoton, DC (Duraiswamy), 8:30 A.M., Room 1046. The Subcommittee will discuss the criteria used by the utilities to design Chilled Water Systems, associated regulatory requirements, and the criteria being used by the NRC Staff to review the Chilled Water System design. Lodging will be announced later.
Attendance by the following is anticipated:
Mr. Michelson Mr. Reed Mr. Ebersole Mr. Wylie Dr. Moeller III-2
l I
. I Joint Waste Management and Quality and Quality Assurance October 14-16,
.T 7'
1987, 1717 H 5treet, NW, Washington, DC (Merrill/Igne), 8:30 A.M., Room 1046. The Subcommittees will review QA Experience in Readiness Reviews as j
applied te nuclear power plants, HLW geologic repositories, and monitored i
retrievable storage (MRS) facilities. Also to review pertinent HLW, LLW, I
and associated research topics to be determined at an agenda planning I
session with NMSS and RES staffs (date not yet set).
Lodging will be announced later. Attendance by the following is anticipated:
Dr. Moeller Dr. Shewmon j
Mr. Reed Dr. Siess Dr. Kerr Mr. Wylie Dr. Remick Planning Subcommittee, October 22-24, 1987, Location to be determined. The i
Committee will discuss the future role of ACRS and membership needs.
Attendees will be announced later.
1 Safety Research Program (Closed) (Tentative), December 2, 1987, 1717 H 5treet, NW, Washington, DC (Duraiswamy), 8:30 A.M., Room 1046. The Subcom-mittee will discuss the proposed NRC Safety Research Program and budget for FY 1989, and possibly the OMB mark on the NRC budget, so as to have a clear perspective of the overall NRC Safety Research Program. Also, to gather information for use by the ACRS in the event it needs to write a report to the Congress. Lodging will be announced later. Attendance by the follow-ing is anticipated:
Dr. Siess Dr. Remick Dr. Kerr Dr. Shewmon Dr. Mark Mr. Ward Mr. Michelson Mr. Wylie Dr. Moeller Decay Heat Removal Systems, Date to be determined (September), Washington, DC (Boehnert). The Subcommittee will continue its review of the NRR Resolution Position for USI A-45.
Attendance by the following is antic-ipated:
Mr. Ward Mr. Wylie Mr. Ebersole Dr. Catton Mr. Michelson Mr. Davis Mr. Reed l
t III-3
1 4
]
p.i-Thermal Hydraulic Phenomena, Date to be detemined (September / October),
'4 version of revised ECCS Rule, and (2) the status of the RES themal hydrau-
~
)
washington, DC (Boehnert). The Subcommittee will review: (1) the final u
lic research program. Attendance by the following is anticipated:
Mr. Michelson Dr. Catton Mr. Ebersole Dr. Schrock i
Dr. Kerr Mr. Sullivan i
Mr. Ward Dr. Tien i
Mr. Wylie 1
l GE Reactors (ABWR), Date to be determined (September / October), Washington l
DC (Major). The Subcommittee will review the Review Guidelines between GE 1
aiid the NRC Staff. Attendance by the following is anticipated:
J l
Dr.ReriI3 Dr. Okrent Mr. Ebersole.
Dr Showmon Dr. Kerr
- r. Wa P l
Mr.Michh Mr. Wy Standardization of Nuclear Facilities, Date to be detemined (October),
Washington, DC (Alderman).
The Subcommittee will review the Staff SER and Chapter I of the EPRI Requirements document. Chapter 11 may also be dis-cussed. Attendance by the following is anticipated:
Mr. Wylie Mr. Reed Mr. Michelson Dr. Siess Babcock & Wilcox Reactor Plants, Date to be determined (October / November),
Washington, DC (Major). The Subcommittee will continue its review of the long-tem safety review of B&W reactors. This effort was begun during the summer of 1986; initial Comittee coments offered on July 16, 1986 in a l
1etter to V. Stello, EDO.
Attendance by the following is anticipated:
Mr. Wylie Mr. Michelson Mr. Ebersole Dr. Okrent Dr. Kerr Mr. Reed Dr. Lewis Mr. Ward Metal Components. Date to be determined (October / November), Charlotte, NC l
(Igne). The Subcommittee will review the status of the NDE of cast stainless steel piping. Attendance by the following is anticipated:
Dr. Shewmon Mr. Ward Dr. Lewis Mr. Rodabaugh Mr. Michelson Dr. B. Thompson III-4
1 i
7" 1
Diablo Canyon, Date to be determined (late November /early December),
Location to be detennined (Igne). The Subcommittee will review the status 1
of the Diablo Canyon Long-Term Seismic Program. Attendance by the follow-ing is anticipated:
1 Dr. Siess Dr. Page Mr. Ebersole Dr. Maxwell Dr. Kerr Dr. G. Thompson Dr. Lewis Dr. Trifunac 4
Dr. Moeller Dr. Scavuzzo Structural Engineering, Date to be determined (late November or January q
1988), Albuquerque,NM(Igne). The Subcommittee will review the results of l
the model concrete containment test. Attendance by.the following is anticipated:
i Dr. Siess Mr. Bender Mr. Ebersole Dr. Pickel Dr. Okrent Mr. Rodabaugh j
Dr. Shewmon Joint Seabrook/0 occupational & Environmental Protection Systems / Severe Acci-dents, Date to be determined, Washington, DC (Igne/ Houston / Major). The s
subcommittees will review Seabrook Emergency Plans and other related
)
matters.
Attendance by the following is anticipated:
l l
Dr. Kerr Mr. Reed Dr. Lewis Dr. Remick Dr. Mark Dr. Shewmon Mr. Michelson Dr. Siess Dr. Moeller Mr. Wylie Dr. Okrent Dr.Catton(tent.)
Seabrook Unit 1, Date to be determined, Washington, DC (Major). The Subcommittee will review the application for a full power operating license for Seabrook Unit 1.
Attendance by the following is anticipated:
Dr. Kerr Dr. Moeller Dr. Lewis Mr. Michelson 1
III-5 L
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t.,
APPENDIX IV OTHER DOCUMENTS-RECEIVED.
328TH ACRS MEETING.
AUGUST 6-8,~1987, WASHINGTON, D.C.
1.
Memo to ACRS Members from R. Savio,
Subject:
"Aug. 6, 1987.ACRS Briefing on the Status of the-NRC Staff's TVA Projects Division,"
dated 7/31/87
-(Aug.7-9,1986)yfromV..Stello, Subject': '
Memo to R. Frale "326th ACRS Meeting 2.
Follow-Up. Items," dated 11/5/86-I '
3.
Letter to B. Your.gblood, NRC from S. A. White, TVA, dated 9/26/86 l
4 SECY-87-173. "Stetus of Staff Actions Required Before Restart 'of.
Sequoyah 2," dated.7/16/87 5.
SECY-86-334, "TVA Preliminary Lessons Learned," dated 11/12/86 6.
Memo for ACRS Members from S. Duraiswamy,
Subject:
" Status Report:
- Auxiliary Systems Subcommittee: Report and Discussion of. Fire Risk.
Scoping Study - 328th ACRS Meeting,-Aug. 6-8, 1987," dated 7/27/87L 7.
Memo to ACRS Members from R. Savio,
Subject:
" August 6 1987 ACRS Discussions with the Comission on Safety Goal Policy-Implementation" 8.
NRC Policy Statement on. Safety Goals [7590-01]
l 9.
. Integrated Safety Goal Matrix l
- 10. Draft Memo for the '.D0 from the NRC Secretary,
Subject:
"Comission Guidanca on Implementation of the NRC's Safety Goal (undated)
- 11. Comments on Safety Goal Implementation by. Commissioners Roberts 6/15/87 Carr 7/1/87 Bernthal 1/27/87,5/6/87 i
- 12. Status Report - Discussion with Commissioners on NUREG-1150
- 13. Memo to ACRS Members from R..Savio,
Subject:
" August 6, 1987, ACRS Di::cussions with the Commission on ISAP," dated 7/30/87 -
- 14. Status Report - Discussion of Degree Requirements for Senior Reactor Operators-15.
(Official Use Only) Status. Report on Meeting with Commissioners l
'16.. Status Report on Westinghouse Advanced PWR Design APPENDIX IV i
l
ag IV-2 i
- 17. Status Report on the Amendment of GDC-4 rule which applies to the leak-before-break concept
- 18. Status Report on Nuclear Plant Operating Experience Slides North Anna Loss of RCS Inventory (4)
Monticello - Loss of Electrical Power to Certain ECCS Components McGuire Failure of Reactor Trip Breaker to Open (3)
. Brunswick Safety Relief Valve Failure Westinghouse - Post-LOCA ECCS Switchover Procedures Pilgrim - Potential Degradation of Containment Spray System
- 19. Memo to W. Kerr, et al. from S. J. Chilk,
Subject:
"SECY-87-91, Advice to the Commission on Waste Management," dated 7/24/87
- 20. Memo for T. Rothschild from R. Fraley,
Subject:
" Proposed DOE Bill-to Establish Independent Safety Board," dated 7/22/87
- 21. Memo for W. Kerr, et al. from R. Fraley,
Subject:
"ACRS Review of DOE Facilities," dated 7/21/87
- 22. Legislative Referral Memorandum,
Subject:
" Energy's Draft Proposal to Establish a Nuclear Safety Board, dated 7/15/87 HANDOUTS l
1.
NRC Information Notice No. 87-36:
Significant Unexpected Erosion of Feedwater Lines 2.
NRC Bulletin No. 87-01: Thinning of Pipe Walls in Nuclear Power Plants 3.
Memo for W. Kerr from Commissioner Bernthal,
Subject:
" Agenda for Periodic Comission Meeting with ACRS Scheduled for August 6,
~!
1987," dated Aug. 3, 1987 4.
Memo for E. Igne, et al. from R. J. Bosnak,
Subject:
" Resolution of ACRS/CRGR Comments on Final Broad Scope Amendment to GDC-4,"
l dated July 29, 1987 5.
Memo to Commissioners from the Chairman,
Subject:
" Agenda for Management Meeting with the ACRS August 7, 1987," dated Aug. 4, 1987 l
6.
Memo to W. Kerr from V. Stello,
Subject:
" July 15, 1987 ACRS Coments on Draft NUREG-1150, dated Aug. 6,1987 l
IV-2
IV-3 7.
Memo for ACRS Members from R. F..Fraley,
Subject:
" Future ACRS Activities - 329th ACRS Meeting, Sept. 10-12, 1987," dated Aug. 5, 1987 8.,
Summary Report of ACRS Waste Managenent Field Trip to University of Arizona and the Nevada Test Site - July 28-30, 1987 PRESENTATIONS Slides "ACRS Briefing on TVA Performance Plan Review," Janes'G Keppler
" Insights from Past Fire Protection Research for NRC at Sandia" "The Fire Risk Scoping Study by M. P. Bohn"
" Fire Risk Research Briefing to the 328th ACRS Full Committee" by John Flack
" Qualification of Reactor Operators Utility Group Comments on Advanced Notice of Proposed Rulemaking Degree Requirements for Senior Operators" by P. F. Collins
" Operator Response to Incidents - A Probabilistic Risk Perspective" by Wes Brinsfield "SECY-87-101 Issues and Proposed Options Concerning Degree I
Requirement for Senior Operators" by J. J. Persensky-
"The Testing of a Reinforced-Concrete Containment Model" by James Costello "RESAR SP/90 Review Status Report" by Thomas Kenyon
" Agenda for ACRS Meeting on August 7, 1987" "McGuire 2-Failure of Reactor Trip Breaker to Open on Demand, July 2, 1987" 1
" Final Broad Scope Rule to Modify GDC-4: by R. J. Bosnak s
IV-3 1
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