ML13333A495
| ML13333A495 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 01/29/1980 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | James Drake Southern California Edison Co |
| References | |
| TASK-15-17, TASK-RR NUDOCS 8002140433 | |
| Download: ML13333A495 (8) | |
Text
to Docket No. 50-206 2
Mr. James H. Drake Vice President Southern California Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Rosemead, California 91770 TEMA
Dear Mr. Drake:
a ps)
RE:
COMPLETION OF SEP TOPIC XV-17 Radiolog u
2 es of Steam Generator Tube Failure (PWR)
Your letter dated December 7, 1979, indicated that you have examined our draft evaluation of the subject topic dated November 27, 1979. 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 nay need to be re-examined if you modify your facility or if the criteria are changed before we complete our integrated assessment.
Sincerely, Dennis L. Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors
Enclosure:
Completed SEP Topic XV-17 cc w/enclosure:
See next page SURNAME K....
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0 1 0..
DATE NRC FORM 318 (9-76. NRCM 0240 U.S.GOV ENT PRINTING OFFICE: 1979-289-369 January 29, 1980 cc Charles R, Kocher, Assistant Director, Technical Assessment Genere& Counsel Division Southern California Edison Company Office of Radiation Programs Post Office Box 800 (AW-459)
Rosemead, California 91770 U. S. Environmental Protection Age ncy David R. Pigott Crystal -Mall 12 Samuel B. Casey Arlington, Virginia 20460 Chickering & Gregory Three Embarcadero Center U. S. Environmental Protection Twenty-Third Floor Agency San Francisco, California 94111 Region IX Office ATTN: EIS COORDINATOR Jack E. Thomas 215 Freemont Street Harry B. Stoehr San Francisco, California 94111 San Diego Gas & Electric Company P. 0. Box 1831 K M C, Inc.
San Diego, California 92112 ATTN:
Richard E. Schaffstall 1747 Pennsylvania Avenue, N. W.
Resident Inspector Suite 1050 c/o U. S. NRC Washington, D. C. 20006 P. 0. Box 3550 San Onofre, California 92672 Mission Viejo Branch Library 24851 Chrisanta Drive Mission Viejo, California 92676 Mayor City of San Clemente San ClementeS, California 92672 Chai rman Board of Supervisors County of San Diego San Diego, California 92101 California Department of Health ATTN:
Chief, EnvironEental Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 9ES314
ComPl Revised-December 31, 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 postulatea 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 0
to 0.3 gpm in any one steam. gener itor.
If this limit is exceeded, shutdown procedures must be initiated.
An analysis of the radiological conse-quenfes of a'steam generator tube failure at the San Onofre Unit 1 plant has been performed following the assumptions and proced.ires indicated in the S.R.P. 15.6.3, Radiological Consequences of a Stean Generator Tube Failure (PWR)".
The specific assumptions made regarding the plant conditions prior to the pcstulated accidents and the expected systems responses are listed in Table XV-1.*
In particular, it has been conservatively assumed that the a::ident 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 additicn, it has been assumed that prior to the accident the primary and secondary coolant activities were at the maximu 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 XY-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 7ade in this review will be confirmed durinc the DBE reviews for this 'acility.
e 0
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 a 1387 Mwth.
- 2. Loss of offsite power following the accident.
- 3. Primary coolant activity prior to the accident of l.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 I alsec wind speed at a distance of 282 meters (X/Q = 9.5 x 10-4 sec/m 3).
- 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.
-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.
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 47.5 0.5 Tube Failure Accident with 129.5 0.5 Previous Iodine Spike*
Steam Line Failure Accident 13.0 0.01 Steam Line Failure Accident 22.5 0.01 with Previous Iodine Spike*
Rod Ejection Accident 1.2 0.01 Rod Ejection Accident with 2.1 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.