ML19256F504
| ML19256F504 | |
| Person / Time | |
|---|---|
| Issue date: | 06/16/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19256F505 | List: |
| References | |
| ACRS-1648, NUDOCS 7912190274 | |
| Download: ML19256F504 (299) | |
Text
s /gf T
9S E
2 m.
b b [>N}.Ci' *gg,* r:::
i l i.
j u
y p n[ f, TABLE OF CONTENTS a
230TH ACRS MEETING JUNE 14-16, 1979 I.
Chairman's Report...........................................
1 A. Reviewers...............................................
1 B.
William H. Zimer Nuclear Power Station Unit 1: Alleged False Statement During ACRS Review....................
1 C.
Vi si ts of Foreign Nuclear-Powered Vessels............... 2 0.
Scope and Timing of ACRS Report on NRC Research Programs 2
E.
Bailly Generating Station: Proposed Chan Design................................ge in Piling
................ 2 II. Meeting on the Investigation and Implications of the March 28, 1979 Accident at Three Mile Island Nuclear Plant Unit 2 (TMI-2)................................................... 2 A.
Three Mile Isl and Subcommi ttee Report................... 2 3.
TMI-2 Implications Subcomittee Recort.................. 3 C. Status of TMI-2......................................... 6 D.
Evaluation of Responses to NRC Orders, IE Bulletins, and ACRS Recommendations.................................. 6 E.
Lessons Learned from TMI-2.............................. 6 1.
O v e rv i ew............................................ 6 2.
Operations Lessons Learned.......................... 7 3.
Sys tems and Equipment lessons Learned............... 7 F.
Metropoli tan Edison 's (Met. Ed. ) Presentation........... 7 1.
I n t ro d u c ti o n........................................ 7 2.
Possible Generic Improvements....................... 7 3.
Ope rati on of the Uni t............................... 8 4.
Key Decisions Processes............................. 10 C
90 f'
g 919,1 1623 144 i
Table of Contents 230th ACRS Meeting June la-16, 1979 5.
Additional Questions................................
11 a.
What were the initiating events leading to water hamer and pl an t tri o ?........................
11 b.
What problem in the control room ventilation caused the need for respirators in TMI-l?..... 11 c.
Natural Circulation.............................
12 d.
PORY Slack Val ve Operation......................
12 e.
Hydrogen Subble Calculatior3.................. 12 G.
Sabcock and Wilcox's (B & W) Presentation...............
13 1.
Chronology of Events and B & W Support for TMI-2....
13 2.
Actions Taken to Raduce the Probability of the TMI-2 Ini ti ati ng Sequence............................... 13 -
3.
Small-Break LOCA Phenomenology...................... 13
~
4 Small-Break LOCA Operational Guidelines.............. l a 5.
Transient Experience Review and Failure Modes and Effects Analysis.................................. 14 6.
Other B & W Actions to Date.........................
14 III. Meeting on Millstone Nuclear Power Station Unit 2 (Power Increase)................................................. 15 A.
S ub comi ttee Repo rt..................................... 15 B.
S tatus of the NRC Staff Review.......................... 15 C.
Li cen s ee P res en tati on...................................
15 D.
Commi ttee A cti o n........................................ 16 IV. Meeting with NRC Staff on NUREG-0531, " Investigation and Evaluation of Stress Corrosion Cracking in Piping of Light Water Reactor Plants, February, 1979"...............
16 1623 145
Table of Contents 230th ACRS Meeting June 14-16, 1979 Y.
Meeting witn NRC Staff on Recent Operating Experience, Licensing Actions, Generic Matters, and Future Agenda.....
16 A.
Cuane Arnold: Replacement of Recirculation Nozzles..... 16 1.
S umma ry o f A l l e ga ti on s.............................. 16 a.
Wel der Qual i fi cati ons...........................
16 5.
Radi ographi c Film Quali ty.......................
17 2.
R e s o l u ti o n.......................................... 17 3.
D. C. Cook: Cracking in Feedwater (Carbon-Steel) Pipes.
18 C.
Inadequacies in the Design of Nuclear Power Plant Piping to Withstand Seismically-Induced Loads................
18 0.
Seaver Valley 1: Failure of Steam Dump Valve to Close.. 19 E. Future Agenda........................................... 19 F.
Subcommi ttee Acti vi ti es................................. 20 VI.
Execu ti ve Ses s i ons.......................................... 20 A.
S ubcommi ttee Repo rts.................................... 20 1.
Regul atory Acti vi ti es............................... 20 2.
Meeting with Representatives of Japanese Regulatory A g e n c i e s.......................................... 21 3.
Reliability and Probabilistic Assessment............ 21 4.
LER Subcommi ttee Report............................. 21 3.
TMI-2 Implications...................................... 21 C.
Action on Plants Reviewed by ACRS but for which no CLs Have Been Issued...................................... 22 D.
Sequoyah Nuclear Power Plant Units 1 and 2.............. 22 E.
Palo Verde Nuclear Generating Station Units 4 and 5..... 22 F.
Three Mile Island Nuclear Station Unit 1:
Power................................... Return to
.............. 22 1623 146 fii
Table of Contents 220th ACRS Meeting June 14-16, 1979 G.
TMI Impli cations Subcommi ttee Assi gnment................ 22 H.
Testimony Before TMI-2 Investigatory Groups............. 23 I.
Subcommittee Press Conferences.......................... 23 J.
ACRS As a Researen " User"............................... 23 K.
NRC Research Program Mid-Year Report.................... 23 L.
Pi p e C ra ck S tudi es...................................... 24 M.
TMI-2 Bull etins and Orders Subcommi ttee................. 24 N.
Us e o f Con s u l tan ts...................................... 2 4 0.
Scope of ACRS Acti vi ti es................................ 24 P.
ACRS Repo rts and Letters................................ 25 1.
Summary Comparison of Stainless Steel and Zircaloy Fuel Rod Cladding................................. 25 2.
ACRS Action on Proposed Revisions of Regulatory Guides 25 3.
Letter to D. L. Basdekas............................ 25 iv
TABLE OF CONTENTS--APPENDIXES 230TH ACRS MEETING JUNE 14-16,1979 Ap p endi x I - A ttendees........................................... A-1 Appendi x II - ACRS Future Agenda................................. A-9 Appendix III - Schedule of ACRS Subcomittee Meetings and 'ours.. A-10 Appendix IV - Zimer Nuclear Power Station: Alleged False S tatements auring OL Revi ew...................... A-12 Appendix V - Bailly Generating Station: Proposed Chan o f Pi l i ngs............................ ge in Des i gn
............ A-25 Appendi x VI - TMI Subcomittee Report............................ A-26 Appendix VII - TMI-2 Implications: Consultant's Report.......... A-29 Appendix VIII - TMI-2: Status as of June 14, 1979............... A-33 Appendix IX - TMI-2: Status of Responses to NRC Orders and IE B ul l e ti n s........................................ A - 4 6 Appendix X - TMI-2: Lessons Learned Working Group Functions..... A-89 Appendix XI - Lessons Learned Task Force:
Studies Bein Considered............................g........... A-90 Appendix XII - Lessons Learned Operations Subgroup: Studies Being Considered................................ A-93 Appendix XIII-Preliminary Tasks of Lessons Learned Design and Analysis Subgroup............................... A-96 Appendix XIV - TMI-2: GPU Personnel Nuclear Experience.......... A-97 Appendix.XV - TMI-2: Possible Generi c Improvements.............. A-99 Appendix XVI - TMI-2: Shift and Relief Log Entry Procedures..... A-104 Appendix XVII - TMI-2: Post-Accident Decision-Making Relationships A-ll6 Appendix XVIII - TMI-2: Conditions for Loss of Natural Circulation A-ll7 Appendix XIX - TMI-2: Evaluation of Hydrogen Bubble Volume in Reactor Cool ant System.......................... A-121 Appendix XX - TMI-2: Chronology of Events and B & 'd Support..... A-127 1623 148 I
V
Table of Contents--Appendixes 230th ACRS Meeting June 14-16, 1979 Apoendix XXI - T'il-2: Analysis of Loss of Feedwater Event, Pre-and Post-Aoril 21, 1979........................ A-132 Apoendix XXII - TMI-2: Small-Break LOCA Phenomenology........... A-141 Apoendix XXIII - TMI-2: Small-Break LOCA Operational Guidelines. A-160 Appendix XXIV - TMI-2: Trans'ient Experience and Failure Modes and Effects Analysis for 8 & W Reactors........ A-168 Appendix XXV - TMI-2: Accident and Prevention, Mitigation, and Recovery and 3 & W Response to ACRS Recommendations A-181 Appenaix XXVI - Millstone 2: P roj ect Status Report.............. A-193 Appendix XXVII - Millstone 2: S tatus of Analyses................ A-257 Appendix XXVIII - Millstone 2: Review for Power Increase........ A-271
_ Appendix XXIX - Intergranular Stress Corrosion Cracking Study.... A-304 Appendix XXX - Duane Arnold: Weld and Safe-end Design........... A-325 Appendix XXXI - D. C. Cook 2: Cracks in Feedwater Pipes......... A-333 Appendix XXXII - Seismic Analysis of Piping: Improper Use of Algebrai c Summation of Loads.................. A-341 Appendix XXXIII - Beaver Valley 1:
Failure of Steam Dum to Close...........................p Valve
.......... A-345 Appendix XXXIV - Requests, Agreements and Comitments During 230th ACRS Meeting............................ A-349 Appendi x XXXV - Conduct of Members............................... A-352 Appendix XXXVI - Follow-Up Items for ACRS Consideration.......... A-353 Appendix XXXVII - Sumary Comparison of Stainless Steel and Zircaloy Fuel Rod Cladding................... A-360 Appendix XXXVIII - ACRS Action on Proposed Revisions of Regulatory Guides...................................... A-363 Appendi x XXXIX - Post-TMI NRC Actions............................ A-364 Appendix XL - Adcitional Documents for ACRS ' Use................. A-365 1623 1 0 vi
Federal Ri.,ter / Vol. 44. No.109 / Tuesday. June
.979 / Notices 323:5
~
of established records of distmguished Station and the implications regarding Friday.fune 75.1979 service: and (3) shall be so selected as to nudear power plant design cf the Three 00mn2 AX MnW md provide re, presentation of the views of Mile Island Nuc! ear Station Umt.
Metropoliton Edison Company (C;en)--
scientific seaders m all areas of de Accident wh2ch occurred on March 23.
The Committee win hear presectauens Nanon.
1979. A portion of this session will be by and hold discussiens witn The terms of eight Members of the dosed as required to discuss P epnetary representatives of de Metropolitan Nattenal Science Board will expire on Infor=ation related to this matter.
Edison Company regarding the accident May 10. Isao. All Members of the 1980 H:JG AR-LWPM. Meet 2rg with which occurred at TMI-2 on March 28.
class are eligibile for reappointment NRCStoff(Open)-The Committee will 1979. activ ties following the accident.
except Dr. Norman Hackerman and Dr.
meet with members of the NRC Staff 2 and the current status of the plant.
Crever E. Murray who have been hear reports on the status of the Three Portions of dus session willbe closed as Members of de Board for two six-year Mile Island Nuclear Staton Umt 2 and required to discuss Propnetary teres. Section 4(d) of the Act states that the NRC Staff evahiation of industry Information related to' dis =atter.
Any person. other than the Directer, rephes to 1&E BuUenns 7945.79-05A.
f,
(( g3 I
who has been a member of the Board for 79-05B. 79-06.79-06A. 79468. and 79-g,3 cy f
c, twelve consecut:ye years shall 08 NRC orders to the operators of
/CpenWe Committee wtB hear, thereafter be inelig:ble for appoint =ent Babcock and Wilcox nuclear plants. and dunng the two. year pened foUowing the NRC Start action in response to ACRS pjes nta ons by and ol
, the a co and recom=endatiens regarding de March expiracon of such twelfth year.
W icox Company regarding NRC-I&E The Board and the Director soucat and
.8.19 9 acadent at the Three Mde Bulletms No. 7945 and 79-05A. '9-05B; evaluate nocunations for submission to Island Nuc! ear Station Umt :(TMI-21 the President. Nommacons 2W PR-230PR Deeur:ve Session t ons m m e se B cock and accompamed by biographical (Open)--Re Committee will consider Wilcox nuclear steam supply systems; information may be forwarded to the items to be discussed dunng its meeting and ACRS recommendauons resultmg Chairman. National Science Board.
w th the NRC Commissioners including frr m the March 28.1979 accident which Washington. D.C. 20550, no later than ACKS Intenm Reports No. 2 and No. 3 occurred at 3C-2.
August 15.1979.
regarding implications of the March 28.
Portions of this session will be closed Any questions should be directed to.
1979 accident at DC-2. use of as required to discuss Propnetary Miss Vermee Anderson. Executive quantitative risk assessment as a In.ormation related to this matter.
Secretary. National Science Board (202/
regulatory basia 2nd the need for ACRS 632-5840).
review of proposed regulations 2:00PX4 0PR Meeting with Norman Hackennan, regarding.ransportation of spent NRCStoff(Open)-ne Committee will nuclear fuel.
heat presentations by and hold Ocirman. NotionalScience Boarti discussions with members of the NRC 130PR4JOPR Meeting with
&U
~
NRC Coaumssioners(Open)(Room Staff regarding recent operatmg HJ01-Tha Committee wdl meet with experience and licensing actions the NRC Commissioners to discuss including a failure of the condenser items noted above.
steam dump contrel valves to close NUCLEAR REGut.ATORY J:J0 PR-dm PR: Lecutive Session foUowing a load rejection at the Beaver COMMISSION (OpenJ--The Committee mil hear and Vadey Nudear Plant Umt No.1.
discuss the report of its Subcommittee replacement of reactor pressure vessel noules in the Duane Arnold Nudear Advteory Committee on Reactor on Operating Reactors and consultants 8"D"***;
IIM who may be present regarding the P! ant, and cracking in the feedwater In accordance mth the purposes of request for an increase in power level piping at the D.C. Cook Nudear Station.
Sections 29 and 182b. of the Atomic for the Millstone Nuclear Power Station The Committee ml! also hear a brief Energy Act (42 U.S.C. 2039. 2:32 b.). the Unit 2.
report from the NRC Staff regarding Advisory Committee on Reactor Portions of this session mil be closed inadequacies in the desig 1 of the piping Safeguards will hold a meeting on June if necessary to discuss Propnetary of several nuclear plants related to their 14-16.1979. In Room 1048.1717 H Street. Information related to this matter.
abdity to withstand seismic
[
NW, Washington. D.C. Notice of this 42 PR--J:Jo PR Millston, disturbances..
i meetmg was published on May :4.1979 NuclectPower Station Unit 2(Open)-
he NRC Staff will also report to the 1
(44 FR 30177).
De Committee will hear and discuss Comnuttee on NUREC-0531 I
The agenda for the subject meeting
" presentations by members of the NRC
" Investigation and Evaluation of Stress will be as follows:
Staff and the applicant regarding the Corrosion Cracking in Piping of Light request f r an increase in power for the Water Resctor Plants" dated February Dursday. June H* 1979 Midstone Nudear Power Station. Umt :.
1979, and NUREG-0396. "A Modified 8:JO AR-U:JO AR Lecutive Portions of this session will be closed Basis for the Development of State and l
Session (Open/ Closed)--ne Committee if required to discuss Proprietary Local Covernment Radiological will hear and discuss the report of the Information related to this matter.
Response Plans In Support of Light l
ACRS Chairman regarding 5:J0PR4JOPR LecutiveSession Water Nudear Power Plants."
=iscellaneous matters relating to ACRS (OpenJ -The Committee wdl hear and The futun schedule for ACRS activities. A portion of this session wdl discuss reports ofits Subcommittees on activities mil also be discussed.
l be dosed to discuss classtfled matter related to nuclear power plant 5:30 PR430PR Lecutive Session t
mformation related to the operation of safety including proposed revisions to (Open/--Discuss proposed ACRS nudear powered naval ships.
NRC Regulatory Guides, use of comments and recommendations The Cocumttee mil hear and discuss probabilistic assessment as a regulatory regarding B0-2:impucations of the g
the report of its Subcommittees on the requirement and return to operation of accident at 30-2 on the design of status of the Three Mile Island Nudear the Fort St. Vrain Nudear Station.
nuclear faclities, and the preposed l
1623 150 i
, ys Federal Registe ' Vol. 44. No.109 / Tuesday. June 5.19-
/ Notices I
power levelincrease for the Millstone I have determined in accordance with Point Nuclear Power Station Unit 2.
Subsection 10(d) Pl.92-463 that it is The Coaunission has determined that 802d E## 18' #
necessary to close portions of this the issuance of this amendment wdl not Y
&Jo A3f.-4:00 PJf.: F. rec::tive meeting as noted above to protect result in any sigmScant endrenmental Session (Open)-The Committee will Proprietary Information (5 U.S.C. 552 impact and pursuant to 10 CFR 51.3(d;(4) discuss proposed ACRS comments and b(c)(4}). to preserve the conEdentiality an environmental impact statement. or recommendatiens regarding TMI-2; the of dasstDed information and de negative declaration and environmental implications of the accident which arrsngements for physical protection of impact appraisal need not be prepared occurred at TMI-2 on March 28.19"?.
the Palo Verde. Sequoyah~and Mdistone in connection with issuance of this amendment.
and the requested power level increase nudear plants (5 U.S.C. 552b(c}[1)] and for the Md! stone Nuclear Power Stanon. to permit discussion of matters involved For further details with respect to this Unit 2.
in an adjudicatory proceeding (5 U.S.C.
action. see (1) the licensee's submittal 552 b(c)(10)).
dated May 18.197'3. [2) Amendment No.
ne Committee will diset:ss propcsed 2a to Ucense No. CPR-22. and (3) the ACRS reports to NRC regarding the Further information regarding topics Commission's related Safety Evaluation.
request for a Construction Permit for the to be discussed. whether the meeting All of these items are avadable for Palo Verde Nuclear Generating Station has been cancelled or rescheduled. the Units 4 and 5. and an Operating Ucense Chairman's riding on requests for the public inspection at the Commission's for the Sequoyah Nuclear Plant Units 1 oppornmity to present oral statements Public Document Room.1717 H Street.
and 2.
and the time allotted therefor can be N.W., Washmgton. D.C. and at the ne Committee will discuss proposed obtained by a prepaid telephone call to Hartsydle Memonal Library. Home and ACRS comments /posttions regstding the ACRS Executive Director.Mr.
Fifth Avenues.Hartsvdle. South Carolina. A copy ofitems use of stairdess steel fuel element Raymond F. Fraley (telepboce 202/634-may be obtained upon requ(2) and (3) 1371). between S;15 A.M. and 5.00 P.M.
est cladd!ng, stress corrosion cracking in EDT.
addressed to the U.S. Nuclear nuclear power plant piping and a Regulatory Commission. Washington, modified basis for development of Dated: May 31.19m D.C.20555. Attention: Director. Division radiolog: cal emergency response plans John C. Hoyta, of Operating Reactors, m support oflight.wster nudeer power AdF/80fyCOmm/frJeMGnogemMiC# or.
Dated $t Sethesda, Maryland, this 24th day C
pa s.
og g,y, gg73, Committee mH condnue same cooe rsom discussion of other items considered For de Nudur Regulatory Commluion, durtng this meeting including regulatory A' 8d*""'*
W et O and other deficiencies identified as a
@*** ** N I
- "'"penting Reactan Bmach No. l.
result of the March 28,1979 accident at D*"## A**###"'
TMI-2.
Caronna Power & Ugrit Co.t laauance
[
Portions of this session will be closed of Amendment to Facility Operating coon Ucense
~
!$ format on. p vis oY:
secunty of the facilities involved.
The U.S. Nuclear Regulatory cal W88"_'**t -Js11_ _ _
50 classified Information and matters Comminion (the Commission) has issued Amendment No. 38 to Facdity Caronna Power & Ugftt Co.t fasuance involved in an adjudicatory proceeding.
Operating Ucensa No. DPR-23 to of Amendment to Facill'y Operating Procedures for the conduct of and Carolina Power and ught Company (the Ucc1se participation in ACRS meetin?s were licensee) which revised Technical
~
published in the Federal Register on Specif! cations for operation of the H. B.
na U.S. Nucleat Regulatory October 4.1978 (44 FR 45926). In Robinson Steam Electric Plant Umt No.
Commission (the Commission) has accordance with these procedures, oral 2 (the facdity) located in Darlington Issued Amendment No. 37 to Facility or wntten statements may be presented County, near Hartsville. South Carolina.
Operating Ucense No. DPR-23, issued to by members of the public. recordinge The amendment la effective as of the Carolina Power and Ught Company, vni! be perrtutted ordy during those date ofits issuance.
which revised Technical Specificatzons portions of the meeting when a transcnpt is being kept, and questions ne amendment deletes pressurtzer for operation of the H. B. Robmson may be asked only by members of the level as an input to safety inlection Steam Electric Plant. Urut No. 2 (the Committee. its consultants, and Staff.
actuation, and requires acnistion of facility] located in Darlington County, Persons destrms to make oral safety injection based on two out of South Carolina. De amendment is statements should notify the ACRS three channels oflow pressurtzer effective as ofits date of issuance.
pressure.
ne amendment incorporates Executive Director as far in advanca as practicable so that appropriate ne application for the amendment Commission requested changes arrangements can be made to allow the complies with the standards and regarding the qualifications of the Envircemental and Radiation Control necessary time during de meeting for reqturements of the Atomic Energy Act Supervisor.
such statements. Use of stul. monca of 1954. sa amended lee Act), and tha Comunission's rulee and regsdations. ne ne applicatfon for the amendment picture and television cameras during this meeting may be limited to selected Commission has made appropriate complies with the standards and portions of the meeting as deternune<i findings as required by the Act an f the requirements of the Atomic Energy Act Commission'a rules and regulations in 10 of1954, as amended (the Act), and the by the Chairman. Information regarding CFR Chapter L which are set forth ist the Commission's rules and regulations. The the time to be set aside for this purpose may be obtained by a telephone call to license amendment. Prior public notice Commission has made appropriate the ACRS Executive Director (R.F. of this amendment was not required findings as required by the Act and the since the amendment does not involve a Comnussion's rules and regidations in 10 Fraley) prior to the meeting..
signiSygnt hazards consideration.
CFR Chapter 1. which are set forth in the
!!canse amendment. Pnor public notice ma
~.
g
\\S\\
-m+
pew -
==w+
= geme me e-.um e w
+ **- = = = = -
ee-e em
.. N.
1 T
^
,l("j27 t
c 3
. [
G( m ' --:
F r2 it o
N..
r_
/
^ k
[' f!
3 "C
MINiRES CF THE 230TH ACRS 'EETL'G JJNE 14-16,1979 NASHI'JGTON, D. C.
The 220th meeting of the Advisory Committee on Reactor Safeg tards, held at 1717 H S treet N.
'.-l., Washington, CC, was convened at 8:30 a.m., Thursday, June 14, 1979.
[ Note : For a list of attendees, see Appendix I.]
The Chairman noted the existence of the published agenda for this meeting, and the items to be discussed.
He noted that the meeting was being held in conformance with the Federal Advisory Committee Act (FACA) and the Government in the Sunshine Act (GISA), Public Laws92-463 and 94-409, respectively. He noted that no requests had been received frcxn members of the public to present oral statements.
He also noted that copies of the
~ transcript of some of the public portions of the meeting wuld be available' in the NRC's Public Document Rocxn at 1717 H Street N. W., Washington, CC,
___ _within accroximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
[ Note:
Copies of the transcript taken at this meeting are also available for purchase from ACE Federal Reporters, Inc., 444 North Capitol St N.W.,
Washington, CC 20001.]
I.
Chair. nan's Recort (Cpen to Public)
[ Note: Raymond F. Fraley was the Designated Federal Employee for this portion of the meeting.]
A.
Reviewers te Chairman named Messrs. Lawroski and Mathis as reviewers for the 230th ACRS meeting.
B.
William H. Zimmer Nuclear Power Station Unit 1:
Alleced False Statement Durina ACRS Review te Chairman noted the matter identified in the memorandum from J.
G.
Keppler, Director, Region III, IE, to D.
Thompson, IE, dated April 10, 1979, regarding alleged erroneous information provided to the Committee during its CL review of the Zimmer Nuclear Station (see Appendix IV). The Committee requested that the NRC Staff report to the ACRS regarding resolution of this matter (see Apperidix :CCCV). o
MIWlES OF ""4E 230T'd ACES VEETING June 14-14,1979 C.
Visits of Foreign Nuclear-Powere Yessels 2e Chairman noted the planned visits to US ports of 2 foreign nuclear-pwered naval vessels. The Committeedecided not to review these visits.
D.
Scoce and Timinc of ACRS Recort on NRC Research Procrats te Chairman reported on a discussion wi2 Recresentative M.
K.
Udall, Chairman of the House Subcommittee on Energy and the Envi -
rorment, Committee on Interior and Insular Af fairs, and H.
R.
Myers, Special Consultant to de Committee on nuclear matters, regarding the scope and timing of the annual ACRS Report on the SRC's Research Program. De ACRS' plans for an early repet to the NRC Comisnior.ars in July, to be followd by the full annual report in January, appeared to be satisfactory.
E.
Bailly Generatina Station:
Procosed Chance in Pilina Desian The Chairman noted the receipe of a request frcm the Commissioners for the Committee to consider the safety implications of the use of shorter than originally proposed pilings under major struc-tures at the Bailly Generating Station, dated June 8, 1979 (see Appendix V). Re Committee agreed to considea this matter during the 231st ACRS Meeting (July) (see Appendix XXXIV).
II.
Meetino en the Investiaation and Imolications of the March 23, 1979 Accident at Three Mile Island Nuclear Plant Unit 2 (TMI-2) (Cpen to Public)
(Note :
Richard K.
Major and Ragnwald Muller wre the Cesignated Federal Dnployees for this portion of the meeting.]
A.
Three Mile Island Subcomittee Recort Mr. Etherington, Subcxmittee Chairman, reported on the Subcxmit-tee Meeting held in Harrisburg, PA, June 5-7, 1979.
He discussed the current status of mI-2, the interface betWen the NRC Staff and the I.icensee, Ntropolitan Edison Company, the details of the accident and the initiating events, and general operating problens that were experienced at the plant (see Appendix VI).
He noted that the Subcomittee members visited the mI site. He noted also, that although it had been suggested, no press conference was held because the Subcommittee Chairman believed that the full Committee should first determine policy regarding NRC's press conferences.
1623 153 -__
'C4'IES CF THE 230TH ACES ME.CI';G June 14-16, 1979 3.
7:d!-2 Imolications Subco :nittee Recort Messrs. Etherington and Sisss reported for the NI-2 Implications Subcommittee.
Mr. Etherington noted that the NI-2 St bcecmittee met on :d.ay 31 and June 1, and tnat 3 Committee Members and 4 Consultants attend-ed. Proposed studies of hydrogen problems, including hydrogen generation and removal, detonation of hyd rogen-exygen mixtures inside the reactor vessel, combustion data, and a possible program of selected expariments were discussed.
A question was raised concerning the passibility of hydrogen embrittlement of the NI-2 primary system components; the NRC Staff plans to investigate this matter.
We status of boiling water reactors with respect to TMI-2 acci-dents, and the responses from IE Bulletin 79-08 were discussed.
21s Bulletin requires lic asees of EWRs to review and repo.r_t the following items:
e the applicability of NI-2-type accidents to their plants, e
containment isolation, consequences of loss of main feedwater, e
level indication systems, e
operating and training procedures, T~ ~ Z -~ [ ] -- ~ ]
e valve positioning indication, e
the possibility of inadvertent release of radioactive fluids, e
maintenance and test procedures, e
procedures for notification of NRC of unexpected conditions, e
procedures for dealing with abnormal hydrogen generation, e
and appropriate chances in technical specifications.
e The applicability of NUREG-0560, Staff Recort on the Generic Assessment of Feedwater Transients in Pressuri::ed Water Reactors Desicned by the Babcock and Wilcox Cc=cany, ICRS recernmendations 1623 154
CL*"ES OF S'E 230T*4 ACRS '4EITING June 14-15, 1979 were requested of GE and of al' 9WR licensees. ?.e NRC Staff plans to issue a generic P4R report similar to."JREG-0560 in July, and to take appropriate actions.
Lessons learned from the NI-2 accident, including de combination of human erro r, equipment malfunction, and design deficiencies were discussed; detailed discussions will be presented to the full Committee at this meeting.
Babcock and Wilcox (3 & W) discussed their efforts i=ediately following de NI-2 accident and their current efforts to supoort the operations of 3 & N plants currently operating, including 2e develognent of modifications to 2ese plants and the retraining of operato rs.
A device, newly installed on the 3 & N simulator, to read out margins of saturation from pressure and temperature impur, was demonstrated. Actions taken to reduce the prbability of acci-dents, such as the "MI-2 accident, were discussed.
Small-break IfCA phenomenology and operating guidelines were also discussed.
In resconse to a request by the NRC Staff, B & W provided an _
analysis, dated Msy 12, small 3reak in the Pressurizer PORV with no Auxiliary Feedwater and Sincle Failure of the ECCS with Realistic Decay Heat. This analysis concluded that the a & W system can survive this extreme condition indefinitely with no feedwater.
However, during the early stages of heat renoval by boiling, the safety valves would have to pass water of equal volute to de steam generation. B & N has undertaken to verify that these valves have sufficient capacity to pass this quantity of water without signifi-cantly over-pressurizing 2e system.
Representatives of a
& W responded to the ACFdi reca mendations:
Additional analyses have been made and reported on both small break I4CAs and natural circulation.
Recommendations have been made to B & W customers to use wide--range hot-leg temperature indicators, installation of the new B & W indicator to show the margin of saturation, and improved displays of key operating data, A supplementary. training program is being offered.
e e
Matters of reactor vessel level indication, reactor coolant venting, and expanded safety research are being reviewed.
(For a consulta0t' repo rt regarding natural circulation, the steam generwat model, pressurizer sizing, and operator awareness, see Appeu."x VII.)
-- 1623 155
MINtTlIS CF ~"4I 22C':'M ACRS 4EITING June 14-16,1979 Mr. Siess noted that the NRC Staf f identified research needs relating to 2e 74I-2 accident, and proposed that this research be initiated in Fiscal Year 1980.
"'his research would require a supplemental budget of approximately $29 million. 'Ihis sucplemen-tal research would include Transient small-LOCA events, $131/2 million. of which o
$91/2 million would be for tests and approximately $4 million for analyses, Inhanced operator capability, 53.5 million, o
Plant resconse under accident conditions, $5 million, Post-cortem examination and plant recovery, $2.7 million, o
Improved risk assessment, S2.4 million, and e
Improved reactor safety, $2.2 million.
e If btriget supplements are not feasible, this research could be
~~-~
financed by reallocation of existing research funds.
Commission-- -
- ~
~ and/or CorxJressonal approval wuld be required.
If the realloca--
tion option is adopted, lower priority programs, such as tha
' reeder reactor, advanced converter reactor, safeguards, and mor.t c
of the fuel cycle envircrrnental research programns would be elimi-nated.
A c:xnbination of a supplenentary budget and reallocatien could be adopted.
A fourth alternative, reorientation of currant programs to the maximum-extent practicable and requesting the Electric Power Research Institute (EPRI), Department of Energy (DCE), and industry to find the remainder of the research funds, has not been given much consideration by the NRC Staff.
It is possible, when consideration is given of the lessons learned from the NI-2 accident, that the NRC's entire research progran may be redirected, and the supplenentary research funds obtained in this manner.
Mr. Siess notad that 4en the Committee included in its Interim Report Number 3 that there was a need for additional research brotz!ht out by the NI-2 accident, he believed that that research would be of an exploratory, rather than confirmatory, nature.
Mr. Siess noted that the NRC Staff is considering the Semiscale Facility for some of the confirmatory research; several members expressed the view that the Semiscale Facility was inadequate for this type of research. 1623 156
'CPRES CF tie 230TH ACRS MEETING June 14-15,1979 C.
Status of TvI-2 R. H. Vollmer, NRC Staff, discussed the current status of the 24I-2 plant (see Appendix VIII).
D.
Evaluation of Resconses to NRC Orders, IE Bulletins, and ACRS Recommendations D. Ross, NRC Staff, provided a status report on the responses from NRC Orders and IE Bulletins emanating from the 74I-2 accident (see Acpendix IX).
E.
Lessons Learned from TVI-2 1.
Overview R. Mattson, NRC Staff, discussed the functions of the Lessons Learned Task Force and provided a flow chart describing the information processed by the tassons Learned Working Group and the Bulletins and Orders Task Force with their interfaces (see Acpendix X).
He also identified the personnel that make up the Lessons Learned Working Grouo. He informed the Commit-tee that its reccomendations with regard to the T4I-2 accident would be answered by the Task Force in writing. Resconsibility has been assigned to specific individuals for each recormnenda-tion.
R. "attson said that the working group is still attempting to define the long-term objectives.
These objectives will be discussed with the Committee 'ahen they are defined.
':he Chairman cuoted from the Committee's 1976 re;ntt on 24I-2, and noted that Metropolitan Edison has indicated that no aork has been done on tnese reccerendations or on Regulatory Guide 1.97, which deals with instrumentation to follow the course of an accident. He said that the representatives from Metropoli-tan Edison have indicated that no pressure was exerted free -he NRC Staff to see that the recommendations were c:rnplied ath.
Mr. Bender suggested that the NRC Staff note the recommenda-tions that the Committee made with regard to T'!A's Hartsville Plant.
He recommended that the time to make modifications in nuclear power plants is at the Construction Permit stage, not the Operating License stage.
R. Mattson said that one of the lessons learned is that the NRC Staff can no longer stall on such matters.
(For a current listing of the currently identified areas for investigation for the Lessons Learned Task Force, see Apcendix XI.) 1623 157
MINtf*ES CF THE 230TH ACRS VIETI %
June 14-16,1979 2.
C:erations Lessons Learned J. Milhoan, NRC Staff, identified the areas relating to person-nel and operator procedures and training for study by the Lessons Learned Cperations Subgroup (see Acpendix XII).
He said that one of the objectives is to assure that, in the future, operators are trained adeouately to cope with unantici-pated transients.
S. Levine, NRC Staff, cointed out that one of the reasons that the mI-2 accident went on for such a lorg time was that there were no operators present who wre knowledgeacle of both the engineering and the systems involved. A menber suggested that one of the difficulties encountered may be that there is too much direction to detail emanating from the NRC Staff, and that face the reality that the people who operate a plant one must must be provided adequate latitixie so that they can properly operate that plant.
Another member suggested that in making its determinations regardirq recanmendations for operation, the NRC Staff also look at the near-accidents that have occurred, such as at Brown's Ferry, Oyster Creek, and Davis Besse, as well as mI-2.
3.
Svstems and Ecuinnent Lessons Learned R. Tedesco, NRC Staff, itemized a preliminary list of reactor systems and equipent that are being reviewed in the light of the WI-2 accident (see Appendix XIII).
F.
Metrocolitan Edison's (Met. Ed.) Presentation 1.
Introduction R. C.. Arnold, General Public Utilities (GPJ), discussed the total ~ experience and nuclear experience of ~its employees (see Appendix XIV).
2.
Possible Generic Imcrevements R.
- Keaten, "et. Ed., discussed ;:cssible generic improvenents for the NI-2 plant, including operational instrumentation, diagnostic equipnent, cceputer capability, and equipnent design (see Acpendix XV). He noted that the lists of items he presen-ted are neither canplete nor are the reccamendations firm. He also roted that improvements wre being planned for the canput-er system 4en the accident took place. 1623 158
MINtfrES CF TFS 2321 ACFS VIETING June 14-16, 1979 3.
0:eration of the Unit J. Herbein, Met. Ed., noted that shift checklists and log keeping are maintained throtshout the plant for a variety of support systems and equipment, including many non-safety-related syctems.
Sese logs are used to monitor ecuipnent perfonnance as wil as plant status for indications and warning of off-normal conditions, possible ecuipnent malfunctions, and changes in plant status (see Appendix XVI). Some of the plant eculpnent parameters are recuired by technical specifications to ce read and documented. Rese procedures have been reviewed by the plant on-site review committee and are accroved by the station superintendent.
The shift relief and log entries procedure recuires the shift foreman and control room operator to keep records relative to the hourly log, the control room log, and the shift foreman's log. This procedure provides guidance for individuals relieving the shift, who are required to become familiar with operations in progress, any special instruction that has been left to log duty personnel, and plant status.
Operators are required to acknowledge and have an understanding and awareness of changes in plant status by signing the control rocm log prior to assuming the shift duty.
Both the shif t foreman and the control room operator keep turnover notes relative to specific actions that have occurred or will occur on their shift.
Sese notes, passed from shift to shift, enable the operators to focus at watch relief on-the- -- - - --
other-than-normal conditions. Copies of these notes are passed
~
daily to the station manager for his review. In the future, this process will be formali::ed to a greater degree.
Met. Ed. is considering the following additions to the check-lists and logs:
e a critical valve and couponent checklists added to the-shift and daily check procedure, e
a former control room operator turnover checklise and/or status board that will precisely delineate the status of such items as the major equipment out of service, primary and secondary system paraneter abnormalities, system tests currently in progress, instrumenta tion out-of-service, electrical and mechanical maintenance in progress, off-normal indication for positions in key monitoring and control systems such as the ICS,and the RPS, and
_a_
1623 159
.a.
MI'HJrES CF W.E 230TH ACRS MEETING June 14-16, 1979 the location of caution tags, as wil as the status of key 4
core reactivity and heat transfer parameters, and the status of significant alarms.
To increase capabilities, %t. Ed., in the future, will assign a degreed engineer on snift during plant operations.
J. Herbein said that currently monthly computer summaries of all nuclear plant LERs are received by the training depar ment.
reviewd, and those that are pertinent, are included in lec-tures that are part of the operator's requalification program.
In the future, the summary provided by the NRC will be reviewed by the licensing group, and appropriate ERs will be sent to engineerim and/or the licensing group.
Following the review of the ERs, both the engineering and the licensing groups will forward reccmmendations to the training manager for incorpora-tion into the operator training program.
On the initial review, high priority items ERs will h forwarded to the plant operations review ccmmittee and the unit superintendent fo r-their information and action. We licensing group will manage the task tracking system and will keep a record of the actions taken by the engineering, licensing, and training groups. Past experience has indicated that the ccmplete GRs, rather than stnmaries, are needed to adequately access the significance.
Mr. Ckrent requested t$at Met. Ed. identify the publication
'ahich lists the ERs that Met. Ed. uses to identify trouble areas.
J. Herbein discussed proposed changes to emergency administra-tive surveillance and operating procedures.
With respect to emergency procedures, an objectives section will be added to indicate an overall direction to the operator.
tese proce-dures will also be made to reflect the NRC Sulletin items that wer e issued following the TMI-2 accident.
In addition, a multiple plant parameter philosophy will be used in energency procedures so that a reactor's coolant systems conditions can be judged.
Ebether, a separate natural circulation procedure will be provided.
With regard to administrative procedures, shift turnover checklists will reflect safety features, ccmpo-nent depressurization accidents, small break LOCAs, and Unit I system change modifications.
1623 160 MINtRES CF ':"4E 230TH ACRS v.EE'"!NG June 14-16, 1979 J. Herbein said that the training program will consist of three parts:
e Se operators will receive 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> of instruction at a 3 & W simulator.
Cperators will participate in a proctored and evaluated classroom training program at mI consisting of 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br /> of instruction, including lectures on such topics as heat transfer and fluid dynamics, reactor coolant system elevations and manometer effects, the m I-2 transient, B I emergency plan and procedures, simulated instrument failure drills in the Unit I control room, instructions in the proper use of the new direct NRC phone lines to Bethesda and King of Prussia, PA, procedures for surveil-lance and corrective maintenance, and procedures for the sign, switch, and valve alignment sheets to restore emergency equipnent to the normal lineup following main-tenance, and Each licensed ocerator will receive a ecccany written
~
~
and oral examination, in addition to both oral and written examinations originating with the NRC.
W.
Zew, v t.
Ed., the shift supervisor at mI-2 at the time e
of the accident, discussed with the Committee the conditions at the plant during the accident, and his perceptions of ahat was going on during the accident.
J. Herbein noted that the B & W site manager has been at the mI site since the construction phase of Unit I, and has acted as the liason between Met. Ed. and the B & W home office in Lynchburg. He has supported Met. Ed. in preparaton of operating
. procedures and their review, initial start-up and test coordin-ation, physics testing, refueling outage preparatier., and has helped in activity coordination and scheduling. B roughout the day of the DI-2 accident, he participated in technical discus-sicns, and concurred with c:xnmand teau decisions.
He also established ccmnunications betaeen the NI site and B & N Lynchburg, VA.
4.
Kev Decisions Processes R.
C. Arnold, GPJ, used a ost-DI-2 organization chart to outline the current decisons making process for the clean-up and recovery of the stricken unit (see Acpendix XVI). 1623 161
'C;U:'E3 CF SE 230TH ACES MEE"T!G June 14-16, 1979 In answe.r to a question regarding recommended changes to handle post-accident problems, R. C. Arnold stggested that the reactor heat removal system should be inside containment rather than leading to the auxiliary building, and that it might be helpful is one of the steam generators had auxiliary piping so that it could be availacle to be used in the manner in hich it is now being used at 'IMI-2 following pipi:q modifications.
5.
Additienal Questiens a.
What were the initiatino events leadine to water hammer and clant trio?
W.
Zewe noted that the initiating event of the TMI-2 accident was the closure of the condensate polisher valves on the condensate feedwater system, and starving the suction to the condensate booster punps and also to the feedwater pumps, which resulted in their automatic trip and an ensuing autcmatic trip of the turbine generator, which led to the high pressure trip of the reactor.
Folicwirq the valve closure there was water present in the control.
air system. Me water cane into the system as the result of a check-valve failure in the instrument air line.
The instrument air was being used, alorg with high pressure sluicing water, to " fluff" a clog of demineralizer resins.
,21s water is at approximately 50 pounds higher pressure than the control air. It is believed that the water hammer was caused by the closure of the valves.
C. Michelson, ACRS Consultant, suggested that this problem might be eliminated if the licensee were to eliminate the use of control air for service purposes.
b.
What eroblem in the centrol room ventilation caused the need for resoirators in TMI-l?
R.
Duciel, Met.
Ed., said that the Unit 2 control room ventilation system does rot provide automated controls.
Once the system is placed in an emergency line-up, it remains so until manually changed.
On the morning of the accident, this system was placed in an emergency line-co early in the transient, and should have provided fo r outside make-up air.
While the flow rate of the make-up air was not measured, Met. Ed. believes that the control room was not at positive pressure continually throtqh the day, and as a result, there wre, frem time to time, introductions of air-borne radioactivity into the control rocm.
Bis occurred in both units.
Se primary 1623 162
MOU"ES CF T9E 23C*H ACRS MEETrJG June 14-15,1979 reason for de introduction of air-corne radioactivity can be tied to 2e stagnant meteorological condi ticr.3 that existed.
".avels of some of the particulate isotopes reached 10-100 times the maximum pemissible concentration.
He said that he believes there was a significant safety facto r if the operators wore respiratory equignent, but that the control rocm could not have been rendered uninha-bitable. However, no analysis has been made.
c.
Natural Circulation R.
Keaten discussed the conditions for loss of natural circulation at T*4I-2 operating in its current mode of natural circulation heat removal (see Appendix XVII).
He said that no analysis has been made for the con:fition the muld exist when the power-operated relief valve is open.
d.
PCRV Block Valve Ooeration R.
Keaten discussed the block valve operations for the power-operated relief valve for 4-5 hours and 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after plant trip. He said that an inspection of the data obtained from the "MI-2 accident infers that the block valve was open during the first 4-5 hours of the accident.
The operating staff believes that the valve was closed during the majority of. this time, but there is no ecmmer data to directly indicate the actual conditions.
At 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after trip, the operators were deliberately intending to open the block valve as a method of depressuri::ing the primary system so that the decay heat removal system could be operated. The shift supervisor gave instructions to the operator to open that valve while the supervisor was watching the building pressure indicator, and that when the
~
block valve was opened, the pressure spike in the reactor containment building was obtained.
e.
Hydrecen Bubble Calculations R. Keaten briefly discussed de reduction of the hydrogen bucble volume in the reactor coolant system during the later stages of the ':MI-2 accident (see Appendix XVIII).
He noted that no correction was made for the presence of steam in the bubble. d"he initial calculations were made for a temperature of 280 F and a pressure of 1000 psi, which approximated the reactor conditions at the time the bubble was first noticed. 1623 163
\\
v.INtJ* 3 CF *"4E 23021 ACF.S MIETING June 14-16, 1979 G.
Babcock and Wilcox's (3 & W) oresentation 1.
Chronolocv of Events and 3 E W Sucecrt for TMI-2 J.
- 9. Taylor, 3 5 W, discussed de chronology of events and 3 & N actions with regard to the NI-2 accident, an organiza-tion chart showing the relationships of the offices involved in the supcort, the level of 3 & W support chronologically, and the tasks in which 3 & W was involved (see Appendix XX).
He discussed the ex.munications during de accident, and noted the difficulties that were encountered oecause of busy telephone lines.
2.
Actiens Taken to Reduce the Procability of de T."I-2 Initiatine Secuence E. A. Womack, 3 & W, discussed the actions taken by 3 & W to reduce the probability of a recurrence of a mI-2 type initiat-ing sequence.
He noted that 3 & W has issued a bulletin regardirq high pressure system injection operation, and the conditions for maintenance of that operation.
mis bulletirr was subsecuently sent to all pressurized water reactor plants by the NRC.
He discussed also the second NRC bulletin, which focused on the challenges to the pressurizer pilot-operated relief valve which is used both in 3 & W plants and other plants, but used differently in the 3 & W control system. He said that B & W believes these actions have significantly contributed to mitigating the passibility that such an event will occur again.
He said that B & W is continuing to learn, and expects to continue to learn, more about the events that took place at TMI-2.
He focused on the operation of the pilot-operated relief valve, showing part of the event tree that described the situation in B & W plants prior to April 21, 1979.
He discussed the expected system behavior with all systems wrking normally, the loss of main feedwater e/ent analysis carried out for the FSAR, the system pressure tran-sient followirg the loss of main feedwater, and c:xnpared de aforesaid descriptions with those of system behavior expected '
after the operating changes wre made on April 21, 1979 (see Appendix XXI).
3.
Small--3reak LCCA Phenomenoloav
- 3. M. Dunn, B & W, discussed the work done recently by 3 & W on small-break WCA phenomenology (see Appendix XXII). 1623 164
M_*W"ES OF THE 230TH ACRS MEETING June 14-16,1979 During this discussion, Mr. Plesset requested dat the NRC Staff provide infonnation during the June 19-20 ICCS Succmit-tee V.eeting regarding de ptential mechanical forces trat can be developed by injecting cold water into steam, steam /wa ter mixtures, and hot pipes.
4.
Small-3reak LOCA 0;erational Guidelines E. R. Kane, 9 & W, discussed the small-break T.CCA operational guidelines (see Acpendix XXIII).
C. Michelson suggested eat B & W also consider the maximum high pressure injection effect if a small break occurs during a semi-cooled or start-up condition.
He also suggested that B & W address the problem of a small-break LOCA while the reactor is in a shut-down cooling mode.
5.
Transient Excerience Review and Failure Modes and Effects Analysis E.
A. Womack discussed B & W's transient experience review
~
and failure modes and effects analysis (MEA) including the' scope for a reliability analysis of the integrated control system (ICS), the fo rmat for FIA reporting, progess made in the MEA system, a list of power-operated relief valve actua-tions for various anticipated transients, a list of reactor trips, a list of challenges to the reactor protection system, ICS hardware failures, an analysis of 246 reactor trips, ICS control response, failures of inputs to the ICS, and abnormal transient operation guidelines (see Appendix XXIV).
6.
Other B & W Actions to Date D. H. Roy, B & W, discussed other B & W actions since the mI-2 accident, includity accident prevention, accident mitiga-tion, and accident recovery (see Appendix XXV).
He also discussed B & W actions regarding,TRS recamnendations.
W. Lipinski, ACRS Consultant, suggested that there is a need to consider the criteria for the primary system boundary.
D. H. Roy said dat B & W is currently receivitxg and evaluating all LERS originating with PWRs.
He said the company has enhanced its sensitivity to problens since the mI-2 accident. 1623 165
MI' ires IF ~"iE 220C3 ACRS MEE"T;G June 14-15,1979 Curing this discussion,.vr.
Sisss recuested that the ';RC Staff provide information regarding the cacacity of the ecer-gency cowr supply for dual plants with swing diesels to supply power for EECS to both units, simultaneously, in the event of loss of off-site power.
III.
Meeting on Millstone Nuclear Power Station Unit 2 (Power Increase)
(Gpen to Puolic)
[ Note:
Elpidio G.
Igne was the Cesignated Federal Employee fo r this portion of the meeting.]
A.
Subconnittee Recort Mr. Etherington, Succorctittee O. airman, reviewed the background for the Licensee's request to increase the power level of the Millstone Nuclear Station Unit 2 from 2560 to 2700 MWt (see Appendix XXVI). He stressed that the issue here is that, *4111e the site Td containment have been reviewed for 2700 MWt, the EC~S analjses performed for the FSAR were carried out for a power ~ ~~-
level of 2560 MWt. However, subsecuent a~CS analyses have been-performed at 2700 MWt.
3.
Status of the NRC Staff Review E. L. Conner, Jr., NRC Staff, discussed the safety reviews for both Millstone 2 and Calvert Cliffs, similar plants, incitxiing the FSAR reviews, the ACRS review, envirormental review, the NRC Staff actions regarding power level increases over the past 3 years, actions regarding the mresolved issues identified in the ACRS recort of June 11, 1974 on Millstone 2, the original license conditions, system changes for operation at the increased powe level, the systems requiring re-analysis, the Licensee Event Reports statistics for 1978, the abnormal occurrences at Mill-stone 2, and the occupational doses, (see Acpendix XX'nI).
C.
Licensee Presentation R.
Harris, Northeast Public Utilities, described the Millstone Plant and its site, its licensing and operating history, an overview of the recuest for a power increase, the cycle 3 core design, power increase methodolcgy charges, transient and acci-dent analyses, cycle 3 power increase modifications, snielding arrangements, and gamm surveys (see Acperdix XX'nII). 1623 166
NES OF ~"i2 230TH ACES MErNG June 14-15, 1979 3.
Committee Action
?.e Committee voiced no objection to the proposed increase in power level for the Millstone Nuclear Power Station Unit 2 from 2560 to 2700 MWt and instructed the ACRS Executive Director to so notify the NRC Executive Director for Cperations (see Appendix IsCV).
IV.
Meetinc with NRC Staff on NUR G-0531, "Investication and Evaluation of Stress Corrosion Cracxinc in Picinc of Licht Water Reactor Plants, Fecruarv,1979" (Open to Puclic)
(Note:
Elpidio G. Igne was the Cesignated Federal Employee for this portion of the meeting.]
W. Ha:1eton, NRC Staff, discussed the history of previous NRC Staff stress corrosion cracking studies, and the current study leading to NUREG-0 531, Investication and Evaluation of Stress Corrosion Cracking in Picino et Lignt Water Reactor Plants, February, 1979.
He discussed the acoge of tne st0dy, the participants, summaries of German and Japanese experience, the causes of intergranular stress corrosion cracking, the major conclusions reached, recommendations-of the study group, and NRC Staff actions and follow-on plans (see Appendix XXIX).
L.
C. Shao, NRC Staff, said that within 2 years the results of the rffice of Nuclear Regulatory Research studies to address the : robles c.c stress corrosion cracking, pipe leaks, and breaks should be available.
V.
Meeting with NRC Staf f on Recent Oceratinc Exoerien.:, Licensinc Actions, Generic Matters, and Future Acenda (Open to Public)
[ Note:
Gary R.
Quittschrelcer was the Designated Federal Employee for this portion of the meeting.]
A.
Duane Arnold: Reclacement of Recirculation No:cles 1.
Summary of Allecations a.
Welder Qualific-' ions C. Williams, NRC Staff, said that princically there wre t*e series of allegations regarding wldi:q of pipe at Duane Arnold.
?e first involved alleged ron-conformance with respect to welder cualification documentation con-trol. Were were instances were the welder's cualifica-tions could not be established clearly. Se IE investiga-tion demonstrated that, in each of those instances, adequate related documentation resolved any seriout, 1623 167
MDtfrIS CF T9E 230TH ACRS VIITING
' June 14-16, 1979
- problem, i.e., no circumstances wre found wnere a welder actually had failed to meet the qualification recuire-cents of ASME Sec. 9.
In association with the repa irs, there was an additional requirement that any welder involved in this work must demonstrate his proficency in a simulated restricted environment. me recirculation inlet piping, in some instances, is bounded by suoports and other equignent, so that a welder had only about 15-16" clearance. The additional requirement was enat he weld at least IS" of a similar joint configuration in that envi-ronment.
In many instances the record was not clear whether he had estaclished this requirement.
b.
Radiocrachic Film Ouality C. Williams said that the second major category of allega-tions involved alleged misinterpretations of radiographic film quality relating to the examination of welds 2 and 6.
Bree radiographic inspectors, wno were contracted. by
~~
the licensee, alleged that they had previously rejected radiographs during their end-process accumulation data, and ultimately had found that the licensee, in further considerations, had accepted these radiographs.
Two inspectors from Region III reviewed the subject radio-graphy after the fact of acceptance by the Licensee, and concluded that in several instances the welds did not, in their best judgement, meet the requirements of AS4E, Sec.
3, paragraph N34424.
IE engaged consultants to review the wo rk, and although there is not complete agreement with
~
~
the IE inspection, there is substantial agreement between the consultant's interpretation of the rejectable condi-tions of certain of the welds, Region III's interpreta-tions, and those of the alleger.
We Licensee has main-tained that the radiographs, though having certain anom-alles, are not questionable to the extent that they should be rejected. These issues were not resolved, and the
^
matter was passed to IE Headquarters for resolution.
In answer to a question, C. Williams said that IE does not believe that there is a question of bad faith, but rather a difference in Judgenent as to what is acceptable under the code.
2.
Resolution V. Noonan, NRC Staff, said that his group focused on the safety of the welds from a standpoint of stresses, rather than 1623 168
- CHJrIS CF :"dE 230T*4 MRS VIE"'ING June 14-16, 1979 de acceptability of the welds with respect to codes.
In all cases, the safety of the wlds was show. to wat the curren:
??AR allowables. He also discussed the weld designs, and :he mcdifications of the designs of the safe ends to eliminate crevices (see Appendix XXX).
3.
D. C. Cook: Crackinc in Feedwater (Carbcn-Steel) Pices E. Jo rdan, NRC Staff, reported On cracks in carbon-steel piping of the feedwater lines of D. C. Cook Unit 2.
me plant was shut down because of unidentified leakage of 3 gym.
Inspection of the source of the leakage revealed that there were through-wall, circumferential cracks in 2 feedwater lines near the no::le welds. Non-destructive testing in other Unit 2 nozzles identified cracking in these nozzles also. Unit 1 was shut down for refuel-ing; inspection found that all 4 no::les in the feedwater system of Unit 1 also had cracks.
Testing was by both ultrasonic and radiographic means.
Unit I has operated for approximately 4 years, and Unit 2 for 1 year.
E. Jordan described the piping and w lds involved, the cracks discovered, and also identified the plants for which feedwatar nozzles had been inspected since this matter ws identified on May 25, 1979 (see Appendix XXXI).
He noted that all the piping involved is Class 2 piping, radiogra@ed at time of installation, be not subject to periodic, in-service inspection.
E. Jordan said that the Licensee is making repairs, including replacement of the elbow, rebuilding wall thickness, and grinding and cleaning the interior surface of the nozzle. me cause of the problen has not been identified. The Licensee and the vendor, W, have perfoccad stress analyses which do not clearly indicate that the material was over-stressed. There are some evidences of fatigue.
The piping on the Unit 2 steam generators is being instrumented with strain guages and accelerometers to measure differential motion to characteri e the fatigue aspects.
The Licensee has ccomitted to reinspect the refueling outage.
E.
Jordan mted also that San Cnofre has cracks in 3 nozzles.
C.
Inadecuacies in the Desicn of Nuclear Power Plant Picino to Withstand Seismicallv-Induced Loacs W. Russell, NRC Staff, discussed the problen in seisnic analysis which led to the NRC Staff's shutting down of S nuclear gewer plants pending reanalysis and possible modifications to meet codes. Be specifics of the problem addressed the methodology of,b
}bh
'4DIV"ES CF ~'HE 230TH AC:S MEE-'ING Cune 14-16, 1979 algebraic str'r.ation and intramedal restenses.
Se code involved in de original analysis was Sheck II, a proprietary code of Stone and Wecster.
"he n ?ety concern was that systems that could both cause an accident aed/or mitigate an accident wre affected. Be specific concern s.es that the use of an algebraic method without a time sequence could cancel loads in the analysis.
W.
Russell discussed de NRC Staff's order to show cause, the status of the piping reanalysis as of June 14, 1979, the informa-tion requested by the NRC Staff, and ce response of operators of reactors, both those shut-down and others (see Acpendix :CCCI).
He noted that at Surry, the licensee is using soil-structure interaction techniques in its current analyses.
D.
Beaver Vallev 1: Failure of Steam Du== Valve to Close D.
A. Beckman, NRC Staff, reported on an incident occuring at Beaver Valley 1, on January 13, 1979, during ',hich a steam dump valve failed to close (see Appendix X.OCIII).
Mr. Sender offered the opinion that freezing had a part in the event, and recommended that the NRC Staff should require all plants, located in areas where freezing may be a problem, to qualify its equipnent for low temperature environments.
In ans er to a question, A.
Schwencer, NRC Staff, nid dat no safety issue has been identified in this matter.
E.
Future Acenda te Committee agreed on a future agenda for AC"-iS reviews (see Appendix II).
With respect to the Committee's proposed report on Palo Verde, R.
Baer, NRC Staff, said that he does.not believe dat this report is on the critical path for issuance of a Construction Permit.
He said that at this time the NRC Staff resources are limited for case work because of the TMI-2 accident, and that NRC Staff members are not available now to address the ocen items.
Con-struction is not planned to start until January,1981.
In answer to a question, R. Baer said that currently the NRC Staff is dealing with applications for seven Cperating Licenses, six of them already having been reviewed by the Committee, fo r which the NRC Staff is currently writing supplanents dealing with the matters raised by the ?tI-2 accident.
He also said t".at the NRC Staff has not begun to write supplements for Constrtrtion Permits yet. 1623 170
MINU~IS OF S2E 230TH ACRS '4Sr:NG June 14-16, 1979 Mr. Ckrent recommended that the Committee review the NRC Staff's resolutions of 7.!-2 issues.
He suggested that this could be done either on a plant-by-plant basis, or as a generic review.
Mr. Fraley recxmended that the Committee write a recort on the lessons learned frcm the TMI-2 accident after it reviews the NRC Staff proposals on these matters. He suggested that the Committee require supplementary staff re;mrts on the 7.I-2 issues for both CLs and cps.
D. Xarner, Ari::ena Public Service Company, offered his company's opinion,that as replicate plants, the Palo Verde plants are unicue. He said that the ccmpany plans to file an FSAR for all 5 units in October, 1979, in preparation for the NRC Staff's CL review for Palo Verde Unit 1.
Any modifications made for Units 1-3 would be incorporated in Units 4 and 5.
He said that the Ccmmittee would have an opportunity to review these mcdifications for Unit 4 and 5 at the time of the FSAR/SER review. He said also that the empany has resolved the open items in accordance with NRC Staff wishes so that the NRC Staff input could be reduced.
Mr. Ckrent noted his opposition to the proposal that the Committee review :tcdifications to Palo Verde Units 4 and 5 during the CL review of Palo Verde Units 1-3.
F.
Subcor.Tnittee Activities A schedule of future subc::r::mittee activities was distributed to members (see Appendix III).
VI.
Executive Sessions (Cpen to Public)
(Note:
James M. Jacobs was the Designated Federal Employee for this portion of the meeting.]
A.
Subcomittee Reoorts 1.
Reculatorv Activities The Committee concurred in the regulatory position of Regula-tory Guide 1.9 (Rev. 2), Selection Desicn and Oualification of Diesel Generator Units Used as Stancbv (On-site) Electric Power Systems at Nuclear Power Plants.
. 1623 171
MD'JI'ES CF '"HE 230""'. ACRS MEETING June 14-16, 1979 2.'
Meeting with Reoresentatves of Jacanese Reculatory Acencies Messrs. Tawroski and Plesset reported the highlights of their meeting, in Japan on May 21-25, 1979, with representa-tives of Jacanese regulatory agencies, including the Committee on the Examination of Reactor Safety. Discussions centered en reactor safety policy and practice, and the cooperative Emergency Core Cooling Systems (ECCS) Progran.
3.
Reliability and Probabilistic Assessment Mr. Kerr, Subcommittee Chairman, noted that in January the Commissioners asked the NRC Staff to prepare detailed guidance on the use of reliability and probabilistic assessment in the regulatory process.
'he NRC Staff provided a draft report on this subject, that did not provide much guidance.
Mr.
Kerr was of the opinion that the RES interpretation of the Commissioner's request does not coincide with that of the
~
Committee. He said that the Subconmittee will be hapoy to review a detailed guidance document wh'en it is received.
4.
LER Subcommittee Recort Mr..%eller, Subecmmittee Chairman, identified the general areas that have been involved in the Subc::mmittee's review of the LERs generated during the period 1976-1978. He said that the Sube:mmittee plans to have a draft report available for---
Committee consideration at the 231st ACRS Meeting (July).
He said that the Subc:mmittee and its consultants have tried to critique the significant IIRs and to evaluate the corrective actions that have been taken by the utilities and/or the NRC Staff.
R.
F.
Fraley suggested that the possible use of LERs to identify precursors of potential accidents should be investi-gated.
B.
TMI-2 Imolications
~- ~~
Members discussed the ':MI-2 accident, identifying those areas of concern that should be included in the Ccemittee's next report on the implications of the accident.
Items identified included hydrogen formation, hydrogen explosions, hydrogen-oxygen recombi-nation, steam-water explosions, steam-fuel cladding interactions, core melt implications, and analytical sttriies to consider scenar-ios beyond Class 3 accidents. 1623 172
.~
- cures OF THE 230TH ACRS MEE-'ING June 14-16, 1979 Mr. Ckrent offered to precare a draft recort for consideration at the 231st ACRS Meeting (July).
Mr. Kerr suggested the discussions of this type should be sched-uled for Thursday and Friday sessions of the meeting, not Satur-day.
C.
Action on Plants Reviewed hv ACRS but for which no CLs'Have Been Issued The Committee ag reed to recuest that the NRC Staff brief the Committee regarding additional recuirements that de NRC will impose on those plants for which the Committee has written favor-able CL reports but for which Cperating Licenses have not been issued (see Appendix XXXIV).
D.
Secuevah Nuclear Power Plant Units 1 and 2 The Committee agreed to defer further action on its review of the application for an Operating License for the Sequoyah Nuclear Power Plant Units 1 and 2 cntil the NRC Staff makes known its plans (via a SER sucplement) regarding this plant (see Appendix XXXIV).
E.
Palo Verde Nuclear Generatina Station Units 4 and 5 The Committee agreed to defer further action on its review of the application for a Construction Permit for the Palo Verde Nuclear Generating Station Units 4 and 5 until the NRC Staff makes known its plans regarding this plant (see Appendix XXXIV).
F.
Three Mile Island Nuclear Station Unit 1: Return to Power The Committee expects that the NRC Staff will keep it info rmed regarding the Applicant's proposals and the NRC Staff's intentions regarding the conditions that must be satisfied prior to the return to operation of Three Mile Island Nuclear Station Unit 1.
'Ihe Committee agreed to review this matter with special attention to the interactions between Unit 1 and the recovery operations at Unit 2. "'his review is scheduled for the 232nd ACRS Meeting (August) (see Acpendix XXXIV).
G.
TMI Imolications Subcommittee Assianment The Committee agreed that the ':MI Implications Subcommittee should study SRC Nuclear Power Plant Siting Practices in the near future, but did not set a date for the sttxdy.
Included in the study will be the interrelations between siting and Class 9 accidents, and consideration of realistic analyses for greater than Class 8 accidents. 1623 173
"INLTSS OF THI 23C':"i ACRS MES*n:G June 14-16, 1979 H.
Testimony 5efore TVI-2 Investicatory Groucs The Chairman noted the possibility that the Committee may be asked to testify before one or more of the groups investigating de 24I-2 accident. It was the consensus of the Committee dat it would be appropriate to consider the underlying causes and factors leading to this accident. The Committee agreed to devote part of the 231st ACRS Meeting (July) to consideration of the underlying factors relating to the "MI-2 accident.
I.
Subco:nnittee Press Conferences "he Committee agreed dat Subcommittee Chairmen should not sched-ule press conferences following subc::mmittee meetings.
It was proposed that comments to the press by Subcommittee Chairmen et. al. should be limited to factual information concerning de Ccmmittee's review process.
'Ihe Chairman brifly discussed a proposed letter (see Appendik XXXV) regarding the scope of interviews, sceeches, etc. by ACRS Members.
J.
ACRS As a Research " User" The Committee deferred action on a proposed request that-the--
ACRS be considered a " user" of RSR information and agreed to discuss at the 231st ACRS Meeting (July) the NRC Staff require-ments for an identified user in order to obtain authori::ation for research.
K.
NRC Research Procram Mid-Year Recort Mr. Siess noted that this propsed report is being prepared at the request of the Commissioners to be used as a basis for the Fiscal Year 1981 research budget request.
July 1 will be the deadline for information or c:rments regarding this report.
The repart will be discussed, finali::ed, and transmitted both orally and in writing to the Commissioners during the 231st ACRS Meeting.
The Committee requested that the ACRS Staff obtain H.
Cento n's comments on the proposed supplement to the Fiscal Year 1980 research budget for technical assistance prograns related to the TMI-2 accident.
Comments on the proposed supplement will be incit.xded in the abcVe report. 1623 IN
.M21'rIS CF THE 230TH ACRS MEETING June 14-15,1979 L.
Pice Crack Studies It was agreed that the Committee will give further consideration to a letter to the NRC recommending reexamination of primary system piping integrity, taking into account recent nuclear plant experience (e.g. stress corrosion cracki.z, etc.).
Mr. Ckrent was requested to prepare a draft of this proposed repart for consider-ation during the 231st ACRS Meeting (July).
M.
TMI-2 Bulletins and Orders Subcormittee Se Committee agreed that a new Subcxctittee should ce named to consider the responses of vendors / utilities to II Sulletins and NRC Crders resulting from the NI-2 accident; Mr. Mathis, Chairman, and Messrs. Sender, Etherington, Lawreski, Plesset, and Shewaon have been assigned.
The ACRS Staff was requested to obtain the NRC Staff's " report"
_ ~
on auxiliary feedwater system reliability for nuclear power plants.
N.
Use of Consultants The Chairman informed the Committee that its consultant, C.
Michelson, can devote up to 1/2 of his time to mI-2 matters for the next several months.
It ws the consensus of the Committee that formal, final reports from its W.I-2 consultants need not be required, but that written excents on appropriate matters identified by the consultants, should be prepared.
~ ~ ~ ~ -
O.
Scoce of ACRS Activities me Committee offered no ccmments regarding R.
F. Fraley's memo of June 12, 1979,.. Follow-Uo Items for ACRS Consideration (see Appendix XXXVI).
Mr. Siess noted his concern that the Committee may be getting outside its expertise when opinions of consultants have to be depended upon entirely. -
1623 175
ACRS Recorts and Letters 1.
Summary Comoarison of Stainless Steel and Zircalov Fuel Rod Cladding te Committee approved a reply to Commissioner Gilinsky's letter regarding a ccmparison of the merits of stainless steel vs. Zircaloy for nuclear fuel cladding (see Appendix XXXVII).
2.
ACRS Action on Procesed Revisions of Reculatorv Guides te Committee approved a memorandum to the Executive Director for Cperations informing him that the Committee concurs in the regulatory position of Regulatory Guide 1.9 (Rev. 2), Selec-tion, Design, and Qualificatoion of Diesel-Generator Units Used as Standbv (On-Site) Electric Power Systems at Nuclear Power Plants (see Appendix X:CWIII).
3.
Letter to D. L. Basdekas he Committee prepared an acknowledgement to a letter received from D. L. Basdekas, NRC Staff, regardity the safety imolica-tions of the MI-2 accident regarding reactor control systems (see Appendix XXXIX). Bis matter was referred to the ACRS Power and Electrical Subcommittte for follow-up in considering the implications of the mI-2 accident.
The 230th ACRS Meeting was adjourned at 4:00 p.m., Saturday, June 16, 1979.
h-a
~ \\623 \\,ib
APENDIXES 'ID UIE ph 230TH ACRS MEETIE JUNE 14-16, 1979 1T7 1623 2+0-1623 -145
230th ACRS Meeting Meeting Dates:
June 14-16,1979 APPENDIX I ATTENDEES ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Max W. Carbon, Chairman Milton S. Plesset, Vice-Chairman Myer Bender Harold Etherington William Kerr Stephen Lawroski J. Carson Mark William M. Mathis Dade W. Moeller David Okrent Jeremiah J. Ray Paul G. Shewmon Chester P. Siess ACRS STAFF Raymond F. Fraley, Executive Director Marvin C. Gaske, Assistant Executive Director James M. Jacobs, Technical Secretary Herman Aldennan Andrew L. Bates David E. Bessette John Bickel Paul A. Boehnert Sam Duraiswamy Elpidio G. Igne David H. Johnson Morton W. Libarkin Richard K. Major Thomas G. McCreless John C. McKinley Robert E. McKinney Ragnwald Muller Gary R. Quittschreiber Jean A. Robinette Richard P. Savio Peter Tam Dwight W. Underhill Hugh E. Voress
}h23 h Harold Walker Gary Young l
A/
1623 M6
230th ACRS Meeting Meeting Dates:
June 14-16,1979 CONSULTANTS I. Catton C. Michelson W. Lipinski I
I 1623 W M
1623 e;2
NRC ATTENDEES c
230TH ACRS MEETING Thursday, June 14, 1979 Div. of Project Manaaement Div. of Operatino Reactors Robert Baer R. H. Vollmer Walter Hauss P. S. Kapo Fred Allenspach D. C. Iann John F. Stolz T. A. Kevevn I. A. Peltier W. S. Hazelton D. O'Reilly C. Y. Cheng C. Heltemes, Jr.
V. S. Noonan T. H. Liu J. R. Fair Div. of Systems Safety J. Martore R. A. Hermann T. M. Novak T. A. Kevevn R. P. Denise R. A. Hermann B. W. Sleren M. L. Boyle Nuclear Reactor Regulation R. Mattson R. Tedesco R. A. Caora D. Ross M. W. Hodaes D. Skovholt Standards Develocment Management & Proaram Analysis E. C. Wenzinger C. E. Shortt Research Office of Inscection and Enforcement R. D15alvo H. S. Wong E. L. Jord C. Williams, Region III SCSB J. J. Burns, Region 6 X! Taboada S&P W. T. Russell J. M. Grant
, 5?L 1 6 2 3 1"413' A2 1623 243
APPLICANT ATTENDEES 230TH ACRS MEETING c
yune 14,1979 Northeast Utilities APS R. M. Kac1cn F. Farrell A. Gelin R. D. Hart D. B. Kauman P. V. Gurnex E. Randolph Foster S
/t 1623-gy i623 1l.;
\\
PUBLIC ATTENDEES 230TH ACRS MEETING c
,1,,,,,14, 1979 J. East, VEPC0 A. R. Rozy, NUSCO, 423 Baldwin Av., Meriden - Conn F. L. Carpentino, Cembustion Engineering, Durham, Conn R. T. Harris, NUSCO, Weathersfield, Conn.
R. R. Mills, Scmcustion Engr., Bloomfield, CT Mr. Layer, BBR, 7735 Old Georgetown Rd., Bethesda, MD N. Shirley, GE, Gaithersburg, MD A. Kimmins, Wash. Public Power Supply Systems, Richland, WA J. B. Hoch, Pacific Gas & Electric Co., San Francisco, CA F. Stetson, AIF, Rockville, MD E. Fuller, S. Levy, Inc., Campbell, CA R. Adamson, McGraw-Hill, Skillman, NJ Kunihiro, Ota, KEPCO,1725 K St., NW Wash., DC M. H. Furbush, PG&E,14190 Amherst St., Los Altos Hills, CA Gautman Sen, Public Service Electric & Gas, Newark, NJ G. Adamantiades, EPRI, Wash., DC K. Tortino, Lowenstein-Newman, Wash., DC W. H. House, II, Bechtel Pcwer Corp., Frederick, MD R. Borsum, B&W, Derwood, MD G. A. Blanc, PG&E,11513 Falls Road, Potomac, MD N. M. Johnson, S&W, Boston, MA P. Seiffer, Dave, Purcell, Jefferson, NJ
~
P. J. Kochis, Bechtel Power Corp.,12242 Erhison Rd., Ellicott City, MD Ed. Fuler, S. Levy, Inc., Campbell, CA J. East, VEPCO, Richmond, VA A. L. Millet, Ottawa News Service, Wash., DC 20003 W. Williams, Jr., S.C. Public Service Auth.
Osmund W. Dixon, S.C. Elec & Gas Co.
B. Beatright, Pickard, Lowe & Garrick J. E. McEwen, Stafco M. An. Stafco, Inc.
James Blocm 1623 215 iso 1623 &
45
NRC STAFF ATTENDEES 230TH ACRS MEETING June 15,1979 Div. of Project Manacement Div. of Operatina Reactors R. Baer D. Allison F. Williams S. MacKay S. Varga D. Wigginton I&E D. Beckman, Region I
/d3 1623 6 1623 I G
" Y, g
F.
IftVITED ATTEftDEES 230TH ACRS MTG.
c June 15,1979 General Public Utilities Metropolitan Edison Service Comoany E. Wallace G. Miller R. Reaten R. Dubrit R. Arnold W. Zlu H. Dieckamp J. Hertium Public Service Electric & Gas B&W G. Seu J. H. Taylor D. H. Roy SCPSA E. A. Womack W. Williams, Jr.
B. Durm J. J. Cudlin E. R. Kane
/P/
1623 W2' 1623 C:7 h-7
~
PUBLIC ATTENDEES 230TH ACRS MTG.
c June 15,1979 A. Yuspeh, Shaw, Pittman, et al S. Wynkoop, McGraw Hill, Arlington, VA J. Maffre, AIF K. Layer, BBR, 7735 Old Georgetown Road, Bethesda, MD Kunihiro, Ota, KEPCO,1725 K St, NW, Wash., DC A. Kimmins, WPPSS, Richland, WA Hiroyoshi Hamada, The Tbkyo Electric Power, Wash., DC R. Adamson, McGraw Hill, Skillman, NJ R. G. Cockrell, WPPSS, Richland, WA D. B. Karner, Arizona Public Service Co., Phoenix AR M. H. Schwartl, Pickard, Lowe & Garrick, 1200 18th St., NW, Suite 612, Wash., DC 20036 Bob Adamson, McGraw-Hill, Skillman, NJ D. Harbrecht, Pittsbrugh, Press, Wash., DC B. Montgomery, Bechtel June 16,1979 R. H. Leyse, EPRI, Rockville, MD 1 6 2 3 ;? '. 3 j,S )
w lf
. APPENDIX II ACRS FUTURE AGENDA ACRS MEETING TYPE OF REACTOR SER ISSUE PROJECT REVIEW VENDOR DATE
~
Jul y Bailly Generating Sta.
Filing Design Auoust Three Mile Island 1 Re start September None October tbne November Shoreham OL GE 10/1/79 LaSalle 1&2 OL GE 10/1/79 1623 W p3 7'e 4
,q