ML19351A180
| ML19351A180 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/22/1981 |
| From: | Paulson W Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-06, TASK-3-6, TASK-RR NUDOCS 8106260198 | |
| Download: ML19351A180 (12) | |
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o, IJNITED STATES
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- o NUCLEAR REGULATORY COMMISSION h
WASHINGTON, D. C. 20555 8
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.....,o June 22,1981
- Docket No. 50-155 LICENSEE: _ CONSUMERS' POWER' COMPANY (CPC)
FACILITY: Big Rock Point Plant
SUBJECT:
SUMMARY
OF MEETING HELD WITH CONSUMERS ' POWER COMPANY TO.
CISCUSS' SEISMIC DESIGN CONSIDERATIONS (SEP TOPIC III-6).
FOR THE BIG ROCK POINT PLANT-i On _May,21,1981 a meeting was held in Bethesda', Maryland with representatives of Consumers Power Company.- The purpose of the meeting was. to (1) discuss thei licensee's SEP seismic reeviluation status. (2)~ discuss the NRC letter dated
- (pril 24,1981 (seismic reevaluation' program plan scope and sckdule), and -
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- 3) clarify questions raisEJ by the NRC contractor (Lawrence Livermore l
Laboratory). ' A list of attendees is enclosed-(Enclosure 1). Handouts pro-vided by' the:NRC staff are also enclosed (Enclosure _2).
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' An overview of_ the SEP seismic reevaluation status was presented by Consumers Power Company representatives. Thef' stated that the Probabilistic Risk l
- Assessment (PRA) of the Big Rock Point Plant which has been ' submitted to the p
NRC for review shows that seismic risk is not important. The licensee representatives ' requested that we review the seismic related portion of the
' risk assessment. They stated that they will supplement the PRA with a letter
' report as to why they conclude that they don't have to proceed with seismic analyses of piping and equipment. To date, seismic analyses have been com-pleted for structures, underground facilities ~and the main cooling loop.
Other piping, including the main steam line piping, has not been done.
The NRC staff noted that the D' Appolonia report prepared for the licensee has not been' docketed. The staff indicated that the report.is needed by
- July 15,1981 The lictisee stated that if their review of the report i
could not be completed by that date, they would sumarize the methods used l
and provide -them to the NRC.
With regard to shutdown capability, the NRC staff indicated that the licensee
'should show that integrity of the reactor coolant pressure bou'ndary will be maintained during a seismic event up to and including the SSE, that at least l
one system that can withstand-a seismic event is available under non-accident i,
conditions' to bring the primary coolant down to 212*F, and that the failure of other systems would not affect the integrity of the reactor coolant boundary.
In addition, the ECCS should be demonstrated by analysis to be able to with-stand the SSE. With regard to the spent fuel pool, the licensee shou'd E
demonstrate (1) the structural integrity of the spent fuel structure a,1d racks, and (2) ability to provide makeup water to the pool following the SSE.
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-- June 22,1981
.- The staff cuationed -theilicensee that in generating the in-structure
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response spectra using the site: specific spectra, the.use of only one time history may not be. adequate because of inadequate frequency con-tent,-amplitude and dJration. - The licensee should justify the adequacy of the time-history or histories used in the analysis.
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In'sumary, the l'i_censee -reiterated that because of the results of the PRA,..they feel that the risk from a seismic event at Big Rock Point. is low. Further, based on the PRA and the D' Appolonia report, they con--
clude they do not have to do any additional. seismic analyses of the piping.
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t,/d.V~,[@g j lid:'tv Walter A. Paulson, Project Manager Operating Reactors Branch #5 Division of Licensing
Enclosures:
As stated cc w/ enclosures:
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3-June 22,1981 CC:
- Mr. Paul A. Perry, Secretary U. S. Environmental Protection Co.isumers Power Company
- Agency 212 West Michigan Avenue Federal Activities Branch
. Jackson, Michigan 49201 Region V Office ATTN: EIS C00RDINATOR Judd L. Bacon,1 squire 230 South Dearborn Street Consumers Power Conpany Chicago, Illinois 60604 212 West Michigan Avenue Jackson, Michigan 49201 Herbert Grossman, Esq., Chairman Atomic Safety and Licensing Board Joseph Gallo, Esquire U. S. Nuclear Regulatory Commission
- Isham, Lincoln & Beale Washington, D. C.
20555
.1120 Connecticut Avenue Room 325 Dr. 0 scar H. Paris Washington, D. C.
20036 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Peter W. Steketee, Esquire Washington, D. C.
20555 505 Peoples Building Grand Rapids, Michigan 495,03 Mr. Frederick J. Shon Atonic Safety and Licensing Board Alan S. Rosenthal, Esq., Chairman U. S. Nuclear Regulatory Commission Atomic Safety & Licensing Appeal Board Washington, D. C.
20555 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Big Rock Point Nuclear Power Plant ATTN: Mr. C. J. Hartman Mr. John O'Neill, II Plant Superintendent Route 2, Box 44 Charlevoix, Michigan 49720
- Maple City, Michigan 49664 Christa-Maria Charlevoix Public Library Route 2 Box 108C 107 Clinton Street Charlevoix, Michigan 49720 Charlevoix, Michigan William J. Scanlon, Esquire Chairman 2034 Pauline Boulevard
~ County Board of Supervisors Ann Arbor, Michigan 48103 Charlevoix County Charlevcix, Michigan 49720 Resident Inspector Big Rock Point Plant Office of the Governor (2) c/o U.S. NRC Room 1 - Capitol Building RR #3, Box 600 Lansing, Michigan 48913 Charlevoix, Michigan 49720 Herbert Semmel Mr. Jim E. Mills Council for Christa Maria, et al.
Route 2, Box 108C Urban Law Institute Charlevoix, Michigan 49720 Antioch School of Law 263316th Street, NW Thoras S. Moore Washington, D. C.
20460 Atomic Safety & Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.
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.Dr. ' dohn H. Buck.
Atomic: Safety and Licensing Appeal l Board U..S. Nuclear _ Regulatory Coranission Washington,-D.
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- 20555,
.Ms.~ JoAnn Bier
' 204 Clinton Street Charlevoix, Michigan ~.49720 Mr. David P.:Hoffman.
Nuclear Licensing Administrator Consumers Power _ Company-1945 W. Parnall' Road Jackson, Michigan. 49201 A
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GUIDELINES AND PROCEDURES FOR THE REVIEW 0F SEISMIC QUALIFICAT10N OF SEP GROUP 11 PLANTS (San Onofre 1, Lacrosse, Big Rock Point, Yankee Rowe, Haddam Neck, and Dresden 1)
I.
BACKGROUND In order to determine the margin of safety of-the selected eleven op.erating.
- nuclear power plants relative to those designed under current standards, criteria, and procedures, and to define the nature and extent of retrofitting to bring these plants -to acceptable levels of capability if they are not alrecdy.
at such levels, the Office of Nuclear Reactor Regulation (NRR) has been proceed-ing with Phase II of the Systenutic Evaluation Program (SEP) since October 1977 through the review of 137 safety topics developed in Phase I of the SEP. The seismic design considerations, Topics II-4.A B, and C, III-6, III-11, and IX-1 are among the 137 safety topics to be reviewed.
_II.
SCOPE AND OBJECTIVE-The objective of Topic III-6 (including Topics 111-11 and IX-1) is to review and evaluate the seismic resistance of facilities.
As a minimum, the seismic program should provide for an evaluation of:
1.
The integrity' of the reactor coolant pressure boundary, 2.
The capability of systems and components to shutdown the reactor and maintain
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it in a safe shutdown condition (i.e., Topic VII-3),
3.
...a capability of systems and components necessary to mitigate the consequences of accidents (i.e., Topics XV on design basis events) including fuel storage (TopicIX-1),and 4.
The integrity of structures containing the Items 1, 2 and 3 above.
III. GENERAL CRITERIA AND REFERENCES The bases used by the staff for the review and evaluation will be the following:
1.
NUREG/CR-0098, " Development of Criteria for Seismic Review of Selected Nuclear Power Plants," N. M. Newmark and W. J. Hall, May 1978.
l 2.
The Final Ground Response Spectra from the NRC SEP Site Specific Spectra l
Project. The interim spectra was forwarded by the August 4,1980 NRC.
letter to the Group II owners.
j 3.
"SSRT Guidelines for SEP Soil-Structure Interaction Review," N. M. Newmark, December 8,1980.
4.
Standard Review Plan, Sections 2.5, 3.7, 3.8, 3.9, and 3.10.
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Regulatory Guides' 1.26, 1.29, 1.60, 1.61, 1.92, 1.100, and 1.122.
6.
For mechanical and olectrical equipment not covered by NUREG/CR-C098 or Regulatory Guide 1.61, a damping value of up to 4 percent can be used for qualification by analysis. A higher damping value nay be used if sufficient justification is provided and found acceptable.
7.
In general, Items 1, 2 and 3. should be used as a group in cases where
- the criteria in Items 1, 2 and 3 differ from those in the Standard Review Plan or Regulatory Guides.
IV. GENERAL PROCEDURES A) : Licensees are to implement a program to perform the required analysis and to submit a Safety Analysis Report (SAR) addressing the analysis program, scope of analysis performed and the results of the analysis.
In accordance with 10 CFR 50.54(f) of the Commission Regulation, a letter was issued on August 4,1980 to Group II plants licensees requesting that the licensee:
1.
submit details of a seismic evaluation program plan addressing the scope of review, evaluation criteria and a schedule for conslation; and
-2.
provide justification for continued operation in the interim until the program is complete.
The program plan should be acceptable to the NRC Staff.
B) When the' analysis is completed, the licensee will submit the Safety Analysis Report for staff review.
In addition to the detailed descriptions of the methods and procedures used, and the results obtained, a summary of the evaluations of structures, systens and components should be provided in the SAR.
The licensee should provide the following information in tabular form for each equipment item:
1.. Method of qualification used:
(a) Analysis or test or a combination of test end analysis (indicate the company that prepared the report, the reference report number and -
date of the publication).
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If by test, describe whether it was a single or multi-frequency test and whether input was single axis or multi-axis.
(c) If by analysis, describe whether static or dynamic, single or nultiple-axis analysis was used. Provide. natural frequency (or frequencies) and tha danping value used in the analysis.
2.
Inf cate whether the equipment has met the qualifiation requirements or whether codification is required.
3.
Indicate the system in which the equipment is located and whether the equipment is required for:
a)_ hot stand-by b
cold shutdown c
both d
neither 4.
Location of system or conponent, i.e., building and elevation.
5.
Seismic input (indicate the spectra or acceleration used).
6.
Indicate whether it is within the scope of NSSS or B0P.
V.
STAFF REVIEW PROCEDURES A? Based on the licensee's SAR and the pertinent reports referenced therein, the staff and available contractor or consultants will conduct the revic::;. The staff review procedure will be different from that of Group I plants in that the review of Group II plants will be similar to an Operating License review.
B) The staff will conduct a site review of the qualification methods, procedures, and results for a list of selected safety related structures, systems, and conponents and their supporting structures. The intention is also to observe the field conditions and installations of structures, systems, and conponents, based on which judgments can be made as to the validity of the mathematical modeling employed in the evaluation program, with respect to the configurations and the boundary conditions.
C)
In the evaluation of structures, the staff will review at least the following:
1.
Adequacy of site ground response spectra ar.d synthetic time histories input to structures for SSE.
2.
Adequacy of structural mathematical model.
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Nornal, seismic and seismic relAiza
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x'u <ocoinations, stresses, and deformations.
5.
Adequacy of ' floor response spectra.
6.
Relative motions which might influence piping entRrtry hebadings; or spanning between buildings, tilt, or interactPer ("?etts.
D)
In the evaluation of systems, components, and their vvyyortir the staff will review at least the following:
1..
Adequacy of the inputs to each system or compcNert ander $5E loading.
2.
Adequacy of the analytical model and assumptions used to,4imulate the field installation conditions.
3.
Structural integrity of the systems, components, and their supports.
E).In conducting the reviews, depending on the circumstances..the staff nay want to perform some independent avaluations such as confirmatory analyses or consultant's view to enhance the overall review evaluations. From, among the list of selected safety related structures, system., and componnets to be reviewed, the staff will conduct an independent confirmatory analyses for the following:
1.
Containment building and other buildings _ deemed desirable.
2.
Some piping systens.
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- 3. ' Some mechanical and electrical equipment.
F) After each site visit, the staff will issue a trip report, which will
. identify findings, conclusions and follow-up items.
G) At the conclusion of review, the staff will issue a Safety Evaluation i
. Report (SER) for the Topic III-6.
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PL"ifliU i).iiit DIC ROCK POINT
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No specific information was given on the codes and allowables that will be used in the structural analysis.
B.
Information was not presented on the type of structural analysis that will be performed on the various structures.
(i.e., finite elet 9t, stick model, etc.).
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No information was presented on how directional structural response due
.to scismic excitation will be combined.
D.
Information was not presented pertaining to structural damping, computer code use, coil interaction modeling, extraction of significant modes, relative displacements and seismic component input.
PIPING 4
A.
Computer modeling methodology such as eccentric mass techniques, mass distribution, support flexibility and selected. spectra has not been presented.
B.
No information was presented on damping' values, support analysis, and relative displacement.
i C.
No specific information was presented on the codes and allowables that l:
will be used in the piping analysis.
i D.
No information was given on what computer codes are to be used in the analysis.
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l MECHANICAL & ELECTRICAL EQUIPMENT A.
Provide a list of equipment associated with the systems to be included in the program.
l B.
Provide criteria, methods and procedures to be used for equip-l ment qualifications.
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EcView of Seismic Recvaluation Program Plan Big Rock Point Nuclear Power Plant May 14, 1981 The material presented by licensee is sparse. No sp2cific questions and chficien::ics can be observed other than those already imluded in the previous review coments, Items 2,4,5, and 7, (April 24,1981).
Structures and primary ccolant icop.
o "Scicatic Safety Margin Evaluaticn - BRP NFP", by D'Appolonia Vol. I - General guidelines Sections 1-2 Executive Sumary and report organizaticn.
Sections 3-7_
Methcds of Analysis, descriptions of site and
. structures.
Sectico 8 Sumary of Vol. II Vol. X (details of analysis)
Sectico 9 Sumary of results and conclusions Vol. II-VoL X Details of analysis for each structure o
It seems that Secticns 3-7 of Vol. I covers most of the topics called for in the requested seismic evaluaticn program plan (for structures).
o There is a conflict in statement regarding whether or not to consider live load in the Licensee's October 10, 1980 letter and in Secticn 9.3, Vol I of D'Appolonia report. It is also not clear whether thermal effect is considered in the primary coolant loop stress analysis (Secticn 9.2.1)'.
Systems and Ccupcnents o
General statements about complying with SRP, RG and NURE/CR-0098 are not acceptable since these guidelines are scnetimes quite general and can be subjected to various interpretation for a specific plant. SRP and NURM/CR-0090 sre review guidelines intended mainly for the use of reviewers.
o What the program plan (for systems and cmponents) calls for is I
material similar to Section 3-7, Vol. I of D'Applcnia's report such as the detailed descripticns of analysis methods and procedures, design codes, ard cmputer programs. To state this in another l
manner, relevant topics included in regular PSAR Sections 3.7, 3.8, 3.9 and 3.10 should be included in the seisnic evaluaticn program plan.
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1 MEETING
SUMMARY
DISTRIBUTION Docket NRC'PDR Local PDR ORB Reading J. 01shinski S. Varga T. Ippolito R. Clark J. Stolz D. Crutchfield HSmith W. Paulson OELD OI&E(3)
ACRS (10)
- CCTEftAs, E. Adensam P. Y. Chen K. Herring R. Hermann W. Russell f' k.lk 4$ h.e w
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JUN 251981* T TS. u.um.Y
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