ML13333A439
| ML13333A439 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/08/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | James Drake Southern California Edison Co |
| References | |
| TASK-15-02, TASK-15-2, TASK-RR NUDOCS 7911270521 | |
| Download: ML13333A439 (7) | |
Text
4.
0 NRC PDR Local PDR ORB #2 Reading NRR Reading Docket No. 50-206 DGEisenhut RHVollmer DLZiemann HSmith PWO'Connor OELD Mr. James H. Drake I&E (3)
Vice President DCrutchfield (2)
Southern California Edison Company JRBuchanan 2244 Walnut Grove Avenue TERA Post Office Box 800 ACRS (16)
Rosemead, California 91770
Dear Mr. Drake:
1979 RE:
COMPLETION OF SEP TOPIC XV SAN ONOFRE UNIT 1 Your letter dated December 7, 1978, indicated that you have examined our draft 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-18 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, Original signea ft, 11..:
Zc un Dennis L. Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors
Enclosure:
Completed SEP Topic XV-18 cc wienclosure:
See next page DOR:ORB #2 DOR*DOR:0RB#2 O F F IC E,.
2..
PWO'Connor:ah! DCrutch ld' DLZiemann S U R N A M E O...............................
1DAT 7/
/79
//1
/79 11 79 RD A T E C......................
?7)N C
2t GovE PN-EN' onlNTING OFFICE, 1979-289-369
Mr. James November 8, 1979 cc w/enclosure:
Charles R. Kocher, Assistant General Counsel Southern California Edison Company Post Office Box 800 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.
Suite 1050 Washington, D. C. 20006
Complete - November 8, 1979L San Onofre Unit 1 Topic XV-18 Radiological Consequences of Main Steam Line Failure Outside Containment 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 rupture of a main steam line is considered a limiting fault not expected to take place during the lifetime of the plant.
Nevertheless, 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 opening 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 main steam line failure at the San Onofre 1 plant has been performed following the assumptions and procedures indicated in the Appendix to S.R.P. 15.1.5, "Radiological Consequences of Main Steam Line Failures Outside Containment (PWR)." The specific assumptions made regarding the plant conditions prior to the postulated accident and the expected responses are listed in Table XV-1.
V
-2 In particular, it has been assumed that the three steam generators are blown dry inmediately following the accident, and that 1 gpm of reactor coolant is released directly to the environment during the first two hours. This is in accordance with Technical Specification 3.1.4 which limits the allowable steam generator primary to secondary leakage to 0.3 gpm in any one steam generator.
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. An evaluation of this.
accident for the Cycle 6 Reload in March 1977 concluded that no addition al fuel clad failures would occur.
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. 15.1.5.
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 main steam line failure, and that the risk presented by this 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 = 1387 Mwth.
- 2. Loss of offsite power following the accident.
- 3. Primary coolant activity prior to the accident of 1.uCi/g of Dose Equivalent 1-131 and 100/E uCi/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 uCi/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-/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 rate~of 1. gpm.
TABLE XV-2 ACCIDENT DOSES AT NEAREST SITE BOUNDARY 2-hour Dose 2-hour Whole to the Thyro-id 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 Previous 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.
E.m 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 Water Rsactors". The (-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.