ML18046A744
| ML18046A744 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/17/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| Shared Package | |
| ML18046A745 | List: |
| References | |
| TASK-03-01, TASK-03-07.B, TASK-03-12, TASK-06-02.D, TASK-06-03, TASK-06-07.B, TASK-09-03, TASK-10, TASK-3-1, TASK-3-12, TASK-3-7.B, TASK-6-2.D, TASK-6-3, TASK-6-7.B, TASK-9-3, TASK-RR LSO5-81-06-060, LSO5-81-6-60, NUDOCS 8106250282 | |
| Download: ML18046A744 (11) | |
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Docket No. 50-255.
LS05-81-06-060
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Mr. David P. Hoffman Nuclear Licensing A$1inist~ator Consumers Power Company 1945 W. Parnall Road Jackson, M1ch1g~n 49201
Dear Mr. Hoffman:
SUBJECT:
SYSTEMATIC EVALUATION PROGRAM (SEP) FOR PALISADES NUCLEAR POWER PLANT, UNIT 1 - EVALUATION REPORT ON TOPlCS VI-2.D AND VI-3 (DOCKET NO. 50-255)
Enclosed is a copy of our draft eva)uation of SEP Topics VI-2.D, "Mass. and
- Energy Re leas~ for Possible Pipe Brea~) Inside Containment", and.Vf-3, "Con-
- tainment P.ressure and Heat Removal Capability".
- This evaluation compares your **fac11 ity~ as described 1n Docket No. 50-255, with the criteria cur-rently used by the regulatory staff for 11cens1ng new fac11 it1es. Appendix A to our draft evaluation* is a draft Technical Evaluation Report from our
- :contractor, Lawrence L1~.rmore National Laboratory.
NRC comments are. ident-ified by pen and ink changes to *Appendix A:- The final Technical Eval~ation Report will be fon-1arded as soon as it is availabl~, however, the conclus-l.,r ions discussed in our draft evaluation are not expected to change.
Please inform us if your as.-bu11t facility differs from the licensing basis assum-
.ed in our assessment.*
Coimnent~ are requested within 30 *days *of the receipt of this letter so that they may be consi~ered 1n our final evaluation.
ih.1.s eva 1 uat1on w111 be a b~,~ i c input to the. integrated SC!fety assessment for your facility unless.yolj identify changes needed to reflect the as-buil~." conditions at your fac111ty.
This assess.ment may be revised 1n the future if your.~f~cil_ity design is change~ or if NRC criteria relating to *
.thh subject ar:e :modified before the integrated assessment is completed.
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Enclosure:
Draft SEP Topics V-I-2. D arid VI~3 ;
Sincerely,.
Dennis M. Crutchfield, Chief Operating Reactors Branch No.
Division of LicensinR
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UNITED ST ATES N.UCLEAR REGULATORY COMMIS~ION WASHINGTON, D. C. 20555 Mr. David P: Hoffman Nuclear Licensing Administrator Consumers Power Company 1945 W. Parnall Road Jackson; Michigan 49201
Dear Mr. Hoffman:
June 17, 1981
SUBJECT:
SYSTEMATIC EVALUATION. PROGRAM (SEP) FOR PALISADES NUCLEAR POWER PLANT, UNIT l - EVALUATION REPORT ON TOPICS v1~2.D AND VI-3 (DOCKET NO. 50-255)
Enclosed is a copy of our draft evaluation of SEP Topics VI-2.D, "Mass and Energy Release for Possible Pipe Break Inside Containment", and VI-3, "Con-tainment Pressure and Heat Removal Capability". This evaluation compares
- your facility~ as described in Docket-No. 50-255, with the criteria cur-
- rently.used by the *regulatory staff for licensing new facilities. Appendix A to our draft evaluation is a draft Technical Evaluation Report from bur contractor, Lawrence Li vermo.re Nati ona 1 Laboratory.
NRC comments are i dent-*
ified by pen and ink changes to Append.ix A.
The final Technical Evaluation Report will be forwarded.as soon as it is available, however, the conclus-ions discussed in o~r draft evaluation are not expected to change.
Please inform us if your:as-built facility differs from the licensing basis assum~
ed in our assessment.
Comments are requested \\vithin 30 days of the rec:e*ipt of this letter so that they may _be considered in our final evaluation.'.
This evaluation will be.a basic input to the integrated safety as*sessment for your facility unles.s *you identify changes needed to reflect the as-
. built conditions at your facility. This assessment may be revised in the_
future if your facility design is changed or if NRC criteria relating to *
- this subj~ct are modified before the integrated assessment is ~ompleted.
Enclosure:
. Draft SEP Topics VI-2.D and VI-3
~cw/enclosure:*
See next page Sincerely,
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"Jft.. t~M Dennis M. Crutc!ifield, Ch;jf Operating Reactors Branch No. 5 Division of Licensing
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4' Mr; David P. Hoffman cc M. *I. Miller, Esquire Isham~ Lincoln & Beale Suite 4200 One First NatTonal Plaza Chicago~ Illinois 60670
. Mr. Paul A. Perry, Secretary Consumers Power Corrpany 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Corrpany 212 West Michigan Avenue Jackson, Michigan 49201.*
Myron M. Cherry, Esquire suite 4501 One IBM Plaza Chicago, Illinois 60611 Ms. Mary.P. Sinclair Great Lakes Energy Alliance 5711. Summerset Ori ve Midland, Michigan. 48640 Kalamazoo Public Library 3*15 South Rose Street Kalamazoo, Michigan* 49006 Township Supervisor Covert TOn-nship Route l, Box lO * *,
Van Buren County, Michigan.. 49043*
Offic~ of the Governo~ (2)
Roa~ l ~*capitol Btiilding Lansing, Michigan 48913 Direct-Or, Criteria and Standards Division Office of Radiation Programs
. (ANR-460).
U. S. Envi ronrriental Protection
- . Agency Washington, D. c.
20460
- u. S. Environmental Protection Agency Federal Activities Branch
- Region V Office ATTN:
EIS COORDINATOR 230 South Dearborn Street Chicago, Illinois 60604.
Charles Bechhoefer, Esq., Chairman Atomic Safety and Licensing Board Panel U.
S~ Nuclear Regulatory Conr.iissidn Washington, D~ C.
20555 Dr~ George c. Anderson Department of Oceanography University of Washington Seattle, Washington 98195 Dr~ M. Stanley Livingston.
1005 Calle Largo Santa Fe, New Mexico 87501 Resident Insp~ctor c/o U. S. NRC P. O. Box 87 South Haven,.Michigan 49090 Palisades Plant ATTN:.
Robert W. Montross Plant Manag,er
. Covert, Michigan*
- 49043 William J.**scanlon, Esquire 2034 Pauline Boulevard Ann A~bor, Michigan 48103
SAFETY EVALUATION.REPORT ON CONTAINMENT PRESSURE AND HEAT REMOVAL CAPABILITY SEP* TOPIC. VI AND MASS AND ENERGY RELEASE FOR POSSIBLE PIPE BREA~
INSIDE CONTAINMENT,
FOR THE PALISADES NUCLEAR POWER PLANT DOCKET NO. 50-255 REGULATORY DOCKET. HlE COPY
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1
TABLE OF CONTENTS I.
Introductfon I I.
Review Criteria I II.
Related Safety Topics IV.
- Review Guidelines
- v.
Evaluation VI.
Conclusions
- VI I.
References
- Appendix A.
(Draft)
Containment Analysis and Evaluation for the Palisades Nuclear Power Plant, Unit 1
I.
Introduction The Palisades Nuclear Power Plant, Unit 1 began corrrnercial operations in 1971. Since then the staff 1s safety review criteria have changed.
As part of the systematic Evaluation Program (SEP), the containment pressure*
and heat. removal capability (Topic VI-3) and the mass and energy release for possible pipe break inside containment (Topic VI~2.D) have been re-
.* evaluated.
The purpose* of this evaluation is to document the deviations from current safety cri ter-i a as they r~l ate to the contai nrnent pressure and heat re-moval capability and the mass/energy release for possible pipe break *inside containment.
Furtherm~re, independent analyses in accordance with current
- criteria were performed to determine the adequacy of the containment design
- basis (e.g., design pressure and temperature) and to provide input for Un-resolved Safety. Issue (USI) A-24, Qualification of Class lE Safety Related Equipment.
The significance of the identified deviations, and recommended corrective measures to improve safety, will be the subject of a subsequent, integrated assessment of the Palisades plant.
- 11.
Review Criteria The review criteria used in the current evaluation of SEP Topics VI-2.D and VI~3 for the Palisades plant are contained in the following documents:
- 1) 10 CFR Part 50, App~ndix A, General Design Criteria for Nuclear Power
- Plants:
a)
GDC 16.- Containmment design; b)
GDC 38 - Containment heat removal; and c)
GDC 50 - Containment design basis~
- 2) 10 CFR Section 50.46, 11Acceptance Criteria for Emergency Core Cooling System for Light Water Nuclear Power Reactors."
'l
- 3) 10 CFR Part 50, Appendix K, 11 ECCS Evaluation Models"
- 4)
NUREG 75/087, Standard Review Plan* for the Review of Safety Analys_is Reports for Nuclear Power Plants (SRP 6.2.1, Containment Functional -
Design).
III. Related Safety Topics The review areas identified below are not addressed in this report, but are related to the SEP topics of mass and energy release for possible pipe break-inside containment, and/or containment pressure and heat re-mova1 capability.
- 1. III-1,*Classification of Structures, Components and Systems (Seismic and Quality)
- 2.
III-78, Design Codes, Design Criteria, Load Combinations, and Reac-tor Cavity Design Criteria
- 3.
- 4.
IX-3, Station Service and Cooling Water Systems
- 5.
X, Auxiliary Feedwater System
- 6.
III-12, Environmental Qualification of Safety Related Equipment*
IV.
Review Guidelines General Design Criterion (GDC) 16 of Appendix A to 10 CFR Part 50 requires that a reactor containment and associated systems shall be provided to es-tablish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment de-sign conditions important to safety are no~ exceeded for as long as the postulated accident conditions require.
GDC 38 requires a containment heat
- removal system be provided whose system safety function shall be to reduce the containment pressure and temperature following any loss-of-coolant acci-dent (LOCA) and maintain them at an acceptably low level; furthermore, the
v
- system safety function shall be achievabl~ assuming a-single failure.
GDC 50 requires that the containment structure and the containment heat removal
. system shall be designed so that the structure can accorrrnbdate, with suffi-cient margin, the-calculated pressure and temperature conditions resulting from any LOCA. This margin as obtained from the conservative calculation of mass/energy release and the containment model is discussed in the Standard Review Plan (SRP} Section 6.2.1, Containment Functional Design.
The containment design basis includes the effects of stored and generated energy in the accident.
Calculations of the energy available for rele?se should be done in accordance with the requirements of 10 CFR Part 50, *sec-tion 50.46 and Appendix K, paragraph I.A, and the conservatism as specified in SRP 6.2.1.3. The mass and energy release to the containment from a LOCA should be considered in terms of blowdown, reflood; and post-reflood. The mass and energy release for postulated secondary system pipe ruptures should be caltulated in accordance with SRP 6.2.1.4. The review also includes the analysis of postulated single active failures of components in the se.condary system.
By reviewing the licensee's analysis, we identify deviations from the cur-rent criteria and we perform independent analyses, as required, to evaluate the significance of these deviations.
In our analyses, we use the best esti-mate method; i.e., by using actual plant design data, we obtain a best esti-mate, but still reasonably conservative, containment analysis.
The evaluation is completed by comparing the results with the containment design _basis.
Evaluation Based on a review of the existing docket of Palisades; two areas were identi-fied as deviating from the current o-iteria. First, in the LOCA analysis as
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described in the Palisades Final Safety*A~alysis Report (FSAR), the licen-see did not consider the reflood phase *or energy released from the secondary system to the containment through the steam generator.
If the reflood phase were considered, the Palisades plant would meet the provisions of SRP 6.2.1.3 and 10 CFR Part 50, Appendix K.
Second, in the main steam line break (MSLB) analysis as presented in the licensee's submittal dated January 21, 1980 (Reference 1), the single failure of a main steam isolation valve (MSIV) was not considered.
We find that the MSIV failure can be shown to be the worst single failure since it would allow both steam generators to blow down.
Be-cause the Palisades plant MS!Vs are designed as check valves, they close only in the forl'lard flow direction.
In the event of a MSLB upstream of the MSIV, the fluid from the broken loop steam generator, and also from the unbroken loop steam generator could be released to the containment.
This blowdown of two steam generators could produce peak containment pressures and tempera-tures that are beyond design limits. Therefore, the licensee's analysis which did not consider the MSIV single failure represents a deviation from the provisions of SRP 6.2.1.4.
To assess the significance of these two deviations, our consultant, Law-r.ence Livermore National Laboratory (LLNL) performed an independent analy-sis which is presented -in draft Appendix A of this report.
Mass and energy release rates utilized in the analysis were calculated using both RELAP-4 MOD 6 and MOD 7 in accordance with current criteria. Calculation of the post-acci dent containment pressure and temperature was done using CONTEMPT-LT /028.
Several* cases were analyzed for both the primary and secondary system analy-ses.
For the primary system pipe breaks, the calculated transient reflects a post-accident peak containment pressure from 66 to 69.7 psia and a temperature
from 273°F to 325°F, depending on the particular assumptions made. *The con-tainment design pressure and temperature for Palisades are 55 psig (70 psia) and 283°F.. There is~ therefore, a range of 0% to 5% margin between the cal-culated post-accident containment pressure for primary system breaks as identified in draft Appendix A and the design pressure.
Furthermore, because of the thermal capacity of the containment structure, the containment design temperature is not exceeded even though the containment atmosphere temperature exceeds 283°F.
The containment response to a secondary system pipe break is also given in draft.Appendix A.
This analysis was for a MSLB which considered the blow-down of both steam generators resulting from the postulated single failure of a MSIV.
The calculated transient reflects a post-accident containment pres-sure of 97 to 107 psia and a temperature of 420°F to 465°F.
The peak calcu-lated containment pressure exceeds the design va1ue by a substantial amount~
Analyses have also been performed assuming a design change which would prevent the blowdown of both steam generators.
The results show that the calculated peak pressure is 68.5 psia, which is 1.2 psi below the design value.
The cal-cul ated peak containment temperature is 413° F.
Because the therma 1 capacity of the containment structure, the containment design temperature is not ex-ceeded.
VI.
Conclusions
- We have identified the deviations of the Palisades plant from current licens-ing criteria in Section V, above.
From the independent containment analyses
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report~d in draft Appendix A,_ we conclude that the Palisades containment de-sign pressure can meet our current criteria, if design changes were made so
- .that the single failure of a MSIV can be accommodated.
The short duration containment atmosphere over-temperature conditions during a MSLB (Figure 4.8T) and LOCA (Figure 3.llT} should be considered as an input to the environmental qualification of Class lE safety related equipment, USI A-24.
VII. References
- 1. Letter from Roger W.
H~ston of Consumers Power Company to.Dennis L. Zie-mann of NRR, Nuclear Regulatory Commission,
Subject:
Automatic Initiation of Auxiliary Feedwater System at Palisades Plant, Docket I-lo. 50-255, Li-cense DRP-20, dated January21, 1980 *
. 2. Letter from D. P. Hoffman of Consumers Power Company to D. Crutchfield of NRC,
Subject:
Proposed Technical Specifications Change Related to Containment Spray Initiation Time, Docket No-. 50-255, License DPR-20, dated November 24, 1980 *