ML13333A452
| ML13333A452 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/27/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | James Drake Southern California Edison Co |
| References | |
| TASK-17, TASK-RR NUDOCS 7912170189 | |
| Download: ML13333A452 (8) | |
Text
NRC PDR Local PDR ORB #2 Reading Docket No. 50-206 NRR Reading DGEisenhut RHVollmer DLZiemann HSmith WPO'Connor Mr. James H. Drake OELD Vice President I&E (3)
Southern California Edison Company DCrutchfield (2) 2244 Walnut Grove Avenue JRBuchanan Post Office Box 800 TERA Rosemead, California 91770 NOV 2 7 ACRS (16)
Dear Mr. Drake:
RE:
COMPLETION OF SEP TOPIC XV SAN ONOFRE UNIT 1 Your letter dated December 7, 1978, indicated that you have examined our araft evaluation of the subject topic dated October 29, 1978. You suggested editorial or corrective changes to the assessment to make it more accurately reflect your facility design. We have incorporated your suggested modifi cations in the enclosed assessment. With these modifications our review of SEP Topic XV-17 is complete and will be a basic input to the integrated assessment of your facility.
The subject assessment compares your facility design with the criteria currently used by the staff in licensing new facilities. This assessment may need to be re-examined if you modify your facility or if the criteria are changed before we complete our integrated assessment.
Sincerely, by "Origir&a h
6 Dennis Z emann, Chief Operating Reactors Branch #2 Division of Operating Reactors
Enclosure:
Completed SEP Topic XV-17 cc w/enclosure:
See next page OFFICE
..DOR:ORB #2 DOR:
I 'DOR:ORB #.
O F C
SURNAME
..PD..'u.nnor c fie d DLZiemann.
cATE 01/7M9 1
7.9.c0..OVENEN NG O1C 1979.289-369 NRC FORM 318 (9-76) NRCIV 02,10
'I-S.
GOVERrV.'EN1T PPINTING OF'FICE: 1979-289-369
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. James H. Drake 2 -
November 27, 1979 cc w/enclosure:
Charles R. Kocher, Assistant General Counsel Southern California Edison Company Post Office Box 300 Rosemead, California 91770 David R. Pigott Samuel B. Casey Chickering & Gregory Three Embarcadero Center Twenty-Third Floor San Francisco, California 94111 Jack E. Thomas Harry B. Stoehr San Diego Gas & Electric Company P. 0. Box 1831 San Diego, California 92112 U. S. Nuclear Regulatory Commission ATTN:
Robert J. Pate P. 0. Box 4167 San Clemente, California 92672 Mission Viejo Branch Library 24851 Chrisanta Drive Mission Viejo, California 92676 K M C, Inc.
ATTN:
Richard Schaffstall 1747 Pennsylvania Avenue, N. W.
Washington, D. C. 20006
COMPL
- November 27, 1979 San Onofre Unit 1 Topic XV-17 Radiological Consequences of Steam Generator Tube Failure (PWR)
The safety objective of this topic is to assure that the releases from this postulated event will not result in exposures in excess of the established guidelines.
The double ended severance of a steam generator tube is considered a limiting fault not expected to take place during the lifetime of the plant. Nevertheless, it is analyzed because the consequences of this postulated event could include the release of significant amounts of radioactive material. The significance of this accident, compared with 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 containment structure to the turbine and/or condenser, or if there is a concurrent loss of offsite power, to the environment through the safety and relief valves.
Based on analyses of the types of tube degradation that have been observed at the San Onofre Unit 1 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 San Onofre Unit 1 Technical Specifications, Section 4.16. In addition, Technical Specification 3.1.4 limits the allowable primary to secondary leakage
-2 to 0.3 gpm in any one steam generator. If thislimit is exceeded, shutdown procedures must be initiated.
An analysis of the radiological consequences of a steam generator tube failure at the San Onofre Unit.1 plant has been performed following the assumptions and procedures indicated in the S.R.P. 15.6.3, "Radiological Consequences of a Steam Generator Tube Failure (PWR)". The specific assumptions. made regarding the plant conditions prior to the postulated accidents and the expected systems responses are listed in Table XV-L.*
In particular, 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 prior to the accident the primary and secondary coolant activities were at the maximum levels allowed by the Technical Specifications 3.1.1 and 3.4.2.
The estimated 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.
On the basis of these results, we conclude that operation of the San Onofre Unit 1 Generating Station is safe with regard to a possible steam generator tube failure, and that the risk presented by this
- The system assumptions made in this review will be confirmed during the DBE reviews for this facility.
-3 postulated accident is similar to that of plants licensed under current criteria.
Since the plant design conforms to current licensing criteria, this completes the evaluation of this SEP topic.
TABLE XV-1 Assumptions Made in Analysis of the Radiological Consequences of Postulated Tube Failure, Main Steam Line Failure and Rod Ejection Accident
- 1. 103% of rated reactor power = T387 Mwth.
- 2. Loss of offsite power following the accident.
- 3. Primary coolant activity prior-to the accident of 1.iCi/g of Dose Equivalebt 1-131 and 100/E -Ci/g of noble gases.
- 4.
Iodine spiking factor of 500 after the accident.
- 5. Primary coolant activity of 60.uCi/g of Dose Equivalent 1-131 at.time of accident for cases assuming a previous iodine spike.
- 6. Secondary coolant activity prior to the accident of 0.1 4Ci/g Dose Equivalent 1-131.
- 7.
Iodine decontamination factor of 10 between water and steam.
- 8. Meteorological conditions corresponding to a 30 meter elevated release with fumigation and 1 9/sec wjnd speed at a distance of 283 meters (X/Q = 1.1 x 10- sec/m ).*
- 9. No additional fuel clad failures as a result of any of the accidents.
For the Steam Generator Tube Failure Accident
- 1. Failed steam generator is not isolated following the accident.
- 2. 50,000 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 half during the first 30 minutes).
- 3. All releases through the secondary side safety and relief valves.,
For the Main Steam Line Failure Accident
- 1. Total primary.to secondary leak rateof 1. gpm.
g-2 For the Control Rod Ejection Accident
- 1. All releases through the secondary side safety and relief valves.
- 2. Total primary to secondary leak rate of 1. gpm.
- As per Regulatory Guide 1.5, "Assumptions Used for Evaluating The Potential Radiological Consequences of a Steam Line Break Accident for Boiling _ater Reactors". The 0-2 hour X/Q for a ground release is 9.5 x 10. sec/m based on the site meteorological data. Use of this X/Q would result in a reduction of about 10% in the calculated offsite doses.
TABLE XV-2 ACCIDENT DOSES AT NEAREST SITE BOUNDARY 2-hour Dose 2-hour Whole.
to the Thyroid Body Dose (rem)
(rem)
Tube Failure Accident
- 55.
0.6 Tube Failure Accident with 150.
0.6 Previous Iodine Spike*
Steam Line Failure Accident
- 15.
0.01 Steam Line Failure Accident
- 26.
0.01 with Previous Iodine Spike*
Rod Ejection Accident 1.4 0.01 Rod Ejection Accident with 2.4 0.01 Prev ious Iodine Spike*
- 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.