ML18043A575
| ML18043A575 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/05/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Bixel D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TASK-15-02, TASK-15-12, TASK-15-17, TASK-15-2, TASK-RR 790308, NUDOCS 7903260283 | |
| Download: ML18043A575 (14) | |
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DKDavis Docket No. 50-255 MAR o* 5. 7979 Mr. David Bixel Nuclear Licensing Admini.strator Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201
. Deaf Mr. Bixel:
NRC*
_NRR Reading VStello DEisenhut RHVollmer
- OELD Ol&E (3)
DLZiemann RDSilver HSmith BGrimes
. TERA JRBuchanan ACRS (16)
Enclosed are copies of our. draft evaluations. of Systematic Evaluation Progr~am Topics XV-17 and XV-18, and a portion of the draft evaluation of Topic XV-12.
You are requested to examine the facts upon which the staff has* based its evaluations and respond either by confirming that the facts are correct, or by identifying any errors. If in error, p 1 ease supply correc*ted information for the docket.
~!e en<:ourage you
- to supply for the docket any other material related to these topics that might affect the staff's evaluation.
Your. respon_se within 30 days of* the date you receive this 1 etter is requested. If no response. is received within that time, we* will_
assume that you have no comments or corrections..
Enclosures:
Topics XV-17 XV-18 XV-12 cc w/enclosures:
See next page 7 9 0 3 2 6 0 2~ 3 -
OP'P'ICll:..
DATii:~
Sincerely,
{§riii\\nsi"il~-ne~~:rn 11 z;i<:ll-~""~
p,!;)p).l.l.ll )!.,,. :
Dennis L *. Ziemqnn, Chief
- Operating Reactors Branch' #2 Division of Operating Reactors
\\1r u.a. ~cvERNMBN~ PR1NT1NG OP'P'ICI!: t 111 - z1s
- 711
t*~r. David Bixel cc w/enclosures:
M. I. Miller, Esquire Isham, Lincoln & Beale Suite 4200 One First National Plaza Chicago, Illinois 60670 Mr. Paul A. Perry, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michi~an Avenue
- Jackson, Michigan 49201 Myron M. Cherry, Esquire Suite 4501 One IBM P 1 aza Chicago, Illinois 60611 Kalamazoo Public Library 315 South Rose Street Kalamazoo, Michigan 49006 K M C, Inc.
ATTN:
Jack McEwen 1747 Pennsylvania Avenue, N. W.
Suite 1050 Washington, D. C.
20006 Palisades Topic XV-17 Jtadiolog1ca1 Consequences of Steam &enerRtor lube failure (PWR)
The safety objective of this topic is to assure that the releases from this postulated event will not T"esu1t 1n exposures in excess of t.he
. established guidelines.
The double ended severance of a steam generator tube is considered a
- limiting fault not expected to take place during the 1ffetime of the plant. Wevertheless, it ts analyzed because the consequences of *this
~stula:ted event could include the release of significant amounts.-t>f radioactive material. The significance of this accident, as compared to a small loss-of-coolant accident, is due to the path created for the release of reactOr coolant via the secondary side of the steam generator, out of the reactor contai11nent structure to the turbine and/or condenser, or if there is a concurrent 1oss of offsite power, to the envirorrnent through.the -safety and relief valves.
The sudden complete failure of a steam generator tube as postulated for these calculations is not considered a* likely event.
Based on analyses of the types of tube degradation that have been observed at the Palisades steam generators the most likely event would be the gradual increase of the primary to secondary leakage over a time period.
To assure that the integrity of the steam generator tubes is maintained through-the life of the plant, periodic inspections are performed as specified in the Palisades Technical Specifications 3.4.5 *. In addition, Technical Specification 3.4.6 limits the allowab1e primary to secondary leakage to 0.6 gpm in any one steam generator.
An analysis of the radiological consequences of 1 steam generator tlJbe failure at the Palisades plant has been performed following the assll?lptions ~nd procttdures indicated in the S.R.P. 15.6.3, *Radiological "Consequences of a Steam Generator Tube Failure (PWR)~. The specific QSSLITlptions made regarding the plant conditions prior to the postulated
- accidents and the expected systems responses are listed in Table XV-1-.
~n particular, it has been,conservatively asst1ned that the.accident.is followe~ by a complete loss cf offsite power. Therefore, the plant is cooled down by releasing secondary steam to the envirorment through the.
-safety and relief valves. In addition, it has been assllned that prior to the accident the primary and secondary coolant.activities were a~ the
. maximum levels allowed by the Technical Specifications 3.4.8 and 3. 7. l
- The estima~ed site boundary doses resulting from this postulated accident (see Table XV-2)-have been found to be within the. 10 CFR Part 100 guidelines as specified in the Acceptance Criteria for S.R.P. 15.6.3.
t On the basis of these results, we conclude that operation of the Palisades plant is safe with regard to a possible steam generator tube failure, and that the risk presented by.this
¥CStulated accident is similar to that of plants 1icensed under c~rre.nt criteria.
This completes the evaluation cf this SEP topic. Since the plant desi'gn
..conforms to current licensing criteria, no ad.ditional SE? review.is required.
Topic XV-17 (Palisades)
References
~ 3 -
- 1.
Palisades Plant~ PSAR.
Section 14.15
- 2.
Letter from D. P. Hoffman, Consumers Power Company to the Director. NRR.
June 26, 1978 "Palisades Plant ~Steam Generator Operating History Questionnaire Docket 50~255
. 3.
Consumers Power Company - Steam Generator Repatr Report for the Palisades Plant, Docket 50~255 January 1979.
- 4.
NUREG-75/087, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, U. S. Nuclear Regulatory Commission, September 1975, (All references are to Rev. 0 dated 11/24/75 unless otherwise indicated.)
Palisades Topic XV-18 Radiological.Consequences of Main Steam Line Failure.
Outside Contai'1nent The safety objective of this topic is to assure that the releases from this postulated event will not result 1n exposures 1n excess of the established guidelines.
The rupture of a ma.in steam 1;ne is considered a limiting fault not expected to take-place during the 11fet;me of the plant. 1'e.verthe1ess, it is postulated because its consequences could include the release of significant amounts of radioactive material. In particular, the failure of a steam line outside containment would result in the release of activity contained within the secondary system, in addition to ~pening a potential, albeit small path for the release of reactor coolant to the environment via postulated steam generator leaks.
An analysis of the radiological consequences of a matn ste~m ltne failure at the Palisades plant has been perfonried following the assumptions and procedures indicated in the Appendix to S.R.P. 15.1,5, "Radiological Consequence_s of Main Steam Line Failures Outside Containment (PWRL 11 The specific assumptions made regarding the plant conditfons prior to the postulated accident and the expected responses are list in Table XV-1.
In particular, it has been assumed that one steam generator is bio~~
dry.within 60 seconds following the accident, and that 1 gpm of re.act':lr
/.
coolant is released directly to the*environment during the first two hours. This is in accordance with Techntca1 Speci'fication 3,1,5 whi"ch limits the allowable steam generator primary to secondary leakage to 0.6 gpm in any one steam generator.
In addition, it has been assumed that prior to the acctden~ the primary and secondary coolant activities were at the maximum levels allowed by the Technical Specifications 3,4,8 and 3,7,1, An evaluation ~f thts accident in support of Amendment 31 to the provisional operating license in November 1977 concluded that no additional fuel clad failures would occur.
The estimated site boundary doses resulting from th_is postulated accident (see Table XV-2) have been found to be within the 10 CFR Part 100 guidelines as specified in the Acceptance Criteria for S. R. P. 15. l. 5.
On the basis of these results, we conclude that operation of the Palisades plant is safe with regard to a possible main steam line failur~, and that the risk presented by this postulated accident is similar to that of plants licensed under current criteria.
This completes the evaluation of this SEP.topic, Since the plant design conforms to current licensing criteria, no additional SEP review i~
required.
Topic XV-18 (Palisades)*
References
- l. Palisades Plant~ FSAR.
Sectton 14,14
- 2.
Letter from D. P. Hoffman, Consumers Power Company-to the Director, NRR.
June 26, 1978 "Palisades Plant - Steam Generator Operating Htstory Questionnatre,"
Docket 50-255
- 3.
Consumers Power Company " Steam Generator Repair Report for the Palisades Plant, Docket 50~255 January 1979
- 4.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 31 to Provisional Operating License No.
DPR~20. Palisades Plant. Section 4.4 Docket 50-255,
- 5.
NUREG-75/087, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, U. S. Nuclear Regulatory Commission, September 1975,
.(All references are to Rev. 0 dated 11/24/75 unless otherwise indicated.)
---~
Palisades Topic XV-12 Spectrum of Rod Ejection Accidents (PWR)
~ Radiologic~l Consequences The safety objective of this review is to assure that the releases from this postulated event will not result in exposures in excess of the established guidelines.
An analysis of the radiological consequences of a postulatd control rod ejection accident has been performed following the assumptio*ns and procedures indicated in Regulatory Guide 1,77 and the Appendix to S.R.P~ 15.4.8, "Radiological Consequences of Control Rod Ejection Accident (PWR)."
The specific assumptions made regarding the plant conditions prior to the postulated accident and the expected res~onses are listed in Table XV-1.
In p~rticular, it has been conservatively assumed that the accident is followed by a complete loss of offsite power.
Therefore,. the plant is cooled down by releasing secondary steam to the environment through the safety and relief valves.
In addition,-it has been assumed that 0,3%
of the rods suffer clad damage and 0.1% of the rods have at least incipient centerline melti_ng as a result of the accident, This is in accordance with the assumptions in Section 14.16 of the FSAR.
The estimated site boundary doses resulting from this postula:ed accident (see Table XV-2) have been found to be within the 10 CFK ~art 100 -
guidelines as specified in the Acceptance Criteria for S.~.?. 15,4.8 The estimated doses at the low population zone (LPZ) are lower.
On the basis of these results, we conclude that operation of the
.Palisades plant is safe with regard to a possible control rod ejection, and that the risk presented by this postulated accident is similar to that of plants licensed under current criteria.
Topic XV-12.
(Palisades)
References Jo e
.. 3..
- l. Palisades Plant - FSAR.
Section 14,16
- 2.
Letter from D. P. Hoffman, Consumers Power Company, to the Director, NRR.
June 26, 1978 "Palisades Plant -
Steam Generator Operating History Quest1onnaire.M Docket 50-.255
- 3.
Consumers Power Company - Steam Generator Repair Report for the Palisades Plant, Docket 50~255 January 1979,
- 4.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 31 to Provisional Operating License No. DPR-20.
Palisades Plant Docket No.
50~255 November 11 1977.
- 5.
NUREG-75/087, Standard Review Plan for the Review of Safety
- Analysis Reports for Nuclear Power Plants, LWR Edition, U. S. Nuclear Regulatory Commission, September 1975, (All references are to Rev. 0 dated 11/24/75 unless otherwise indicated.)
TABLE XV-1 Assumptions Made in Analysis of the Radiological Consequences of Postulated Tube Failure, and Main Steam Line Failure
- l. Reactor power= 2650 Mwth.
- 2.
Loss of offsite power following the accident.
- 3.
Primary coolant activity £rior to the accident of l.µCi/g of Dose Equivalent I-131 and 100/E µCi/g of noble gases.
- 4.
Iodine spiking factor of 500 after the accident.
- 5.
Primary coolant activity of 40.µCi/g of Dose Equivalent I-131 at time of accident for cases assuming a previous iodine spike.
- 6.
Secondar~ coolant activity prior to the accident of 0.1 µCi/g Dose Equivalent I-131.
7.* Iodine decontamination factor of 10 between water and steam.~
- 8. 0-2 hour X/Q for gro~nd rel~ase at exclusion area boundary boundary = 3.4 x 10-sec/m For the Steam Generator Tube Failure Accident
- 1. Failed steam generator is not isolated during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the accident.
- 2.
98,000 lb. of primary coolant leak to the secondary side of the failed steam generator through the failed tube during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (one h~lf during the first 30 minutes},
- 3. All rel~ases through the secondary side safety and relief valves.
- 4.
No additional fuel clad failures as a result of the accident.-
For the Main Steam Line Failure Accident
- 1. Total primary to secondary leak rate of l. gpm.
2.* No additional fuel clad failures as a result of the accident..
fa e For the Control Rod Ejection Accident
- 1. Total primary to secondary leak rate of 1. gpm.
- 2.
0.3% of rods suffer clad damage.
- 3. o.l % of rods have at least incipient center line melting.
TABLE XY-2 ACCIDENT DOSES AT NEAREST SITE BOUNDARY 2-hour Dose 2-hour Whole to.the Thyroid (rem)
.Body Dose..
(rem)
Tube Failure Acci9ent
- 12.
0,4 Tube Failure Accident 60, 0,4 with Previous Iodine Spike*
Steam Line Fa i 1 ure
- 1. 7
< 0. 01 Accident Steam Line Failure 2.6
< 0. 01 Accident with Previous Iodine Spike*
Rod Ejection Accident**
Case 1 3.6 0.05 Case 2 1.0
< 0. 01
- For this accident sequence it is assumed that an iodine spike was initiated some time before the accident resulting in. the highest coolant activity allowed by the Technical Specifications.
- Case 1 assumes a 11 rel eases through the secondary site sa*fety and relief valves.
Case 2 assumes all releases through the containment.