ML13317A535

Forwards Final Topic Evaluation Concluding That Facility Design Meets Criteria of SEP Topic XV-5, Loss of Normal Feedwater Flow

ML13317A535
Person / Time
Site:	San Onofre
Issue date:	03/14/1983
From:	Paulson W Office of Nuclear Reactor Regulation
To:	Dietch R Southern California Edison Co
References
TAC-44652, TASK-15-05, TASK-15-5, TASK-RR LSO5-83-03-019, LSO5-83-3-19, NUDOCS 8303160140
Download: ML13317A535 (6)

Text

March 14, 1983 Docket No. 50-206 LS05-83 019 Mr. R. Dietch, Vice President Nuclear Engineering and Operations Southern California Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Rosemead, California 91770

SUBJECT:

SAN ONOFRE 1 -

SEP TOPIC XV-5, LOSS OF NORMAL FEEDWATER FLOW By letter dated February 1, 1982, the staff issued a safety evaluation report for the above topic. Your letter of February 23, 1983, provided additional information. Enclosed is the staff's final topic evaluation which concludes that your facility design meets current criteria with respect to this topic.

This evaluation will be a basic input to the integrated assessment for your facility. The evaluation may be revised in the future if your facility design is changed or if NRC criteria relating to this topic is modified before the integrated assessment is completed.

Sincerely, Original signed by/

Walter Paulson, Project Manager Operating Reactors Branch #5 Division of Licensing

Enclosure:

As stated cc w/enclosure:

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Mr. R. Dietch, Vice Presin't Docket No. 50-206 Nuclear Engineering and 0Wations San Onofre 1 Southern California Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Rosemead, California 91770 cc Charles R. Kocher, Assistant General Counsel James Beoletto, Esquire Southern California Edison Company Post Office Box 800 Rosemead, California 91770 David R. Pigott Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 Harry B. Stoehr San Diego Gas & Electric Company Post Office Box 1831 San Diego, California 92112 Resident Inspector/ San Onofre NPS c/o U.S. Nuclear Regulatory Commission Post Office Box 4329 San Clemente, California 92672 Mayor City of San Clemente San Clemente, California 92672 Chairman Board of Supervisors County of San Diego San Diego, California 92101 California Department of'Health ATTN: Chief, Environmental Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 95814 U.S. Environmental Protection Agency Region IX Office ATTN: Regional Radiation Representative 215 Freemont Street San Francisco, California 94111 Robert H. Engelken, Regional Administrator U.S. Nuclear Regulatory Commission, Region V 1450 Maria Lane Walnut Creek, California 94596

SYSTEMATIC EVALUATION PROGRAM TOPIC XV-5 SAN ONOFRE TOPIC:

XV-5, LOSS OF NORMAL FEEDWATER FLOW I.

INTRODUCTION A loss of normal feedwater flow could be caused by main feed pump failure, feed control valve malfunctions or a loss of offsite power. Loss of feedwater flow would result in decrease of steam generator water level, increase in primary system pressure and temperature, and reduction in the secondary system capability to remove the heat generated in the reactor core. To mitigate this accident condition, the auxiliary feedwater system (AFWS) would be used to transport the minimum required feedwater flow to the steam generator(s) to remove the decay heat after reactor scram. If the heat removal capability were not restored in a timely manner, fuel damage could possibly occur.

II. REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evalua tion of the design and performance of structures, systems and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, including determina tion of the margins of safety during normal operations and transient conditions anticipated during the life of the facility.

Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radioactivity.

The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water-cooled reactors.

GDC 10 "Reactor Design" requires that the core and associated coolant, control and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during normal operation, including the effects of anticipated operational occurrences.

GDC 15 "Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated operational occurrences.

-2 GDC 26 "Reactivity Control System Redundancy and Capability" requires that the reactivity control systems be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded.

III.

RELATED SAFETY TOPICS Various other SEP topics evaluate features of the reactor protection system. The effects of single failures on safe shutdown capability are considered under Topic VII-3.

IV. REVIEW GUIDELINES The review is conducted in accordance with SRP 15.2.7. The evaluation includes review of the analysis for the event and identification of the features in the plant that mitigate the consequences of the event as well as the ability of these systems to function as required. The extent to which operator action is required is also evaluated.

V.

EVALUATION Reactor protection is provided by (a) trip on high pressurizer water level (b).trip on steam flow-feedwater flow mismatch (c) automatic initiation of the AFWS on low steam generator water level, thus restoring the heat removal capability of the secondary system.

In Reference 1, the licensee submitted information on minimum flow require ments of the AFWS for plant transients. The analysis assumptions are described in References 1 and 2. The AFWS consists of 1 motor driven pump with capacity of 235 gpm and 1 turbine driven pump of 300 gpm capacity.

This flow capacity is sufficient to remove the reactor core decay heat after reactor scram following a loss of main feedwater coincident with the loss of offsite power. The results also indicate that a minimum heat transfer area can be maintained in the steam'generator which receives auxiliary feedwater flow. This prevents primary system overpressurization and DNB in the reactor core. The pressurizer does not fill during the transient.

In Reference 3, the staff concluded that the licensee had responded accept ably to Enclosure 2 of the staff's November 15, 1979 evaluation of auxiliary feedwater system requirements (see NUREG-0645, Vol. II).

This enclosure concerned the design basis events considered in establishing auxiliary feedwater system flow and the analysis, assumptions and acceptance criteria used.

-3 VI.

CONCLUSIONS As part of the SEP review for San Onofre 1, we have reviewed the licensee's analysis of the loss of normal feedwater event and have concluded that the initial conditions are acceptable and the peak pressures and DNBR reached during the transients are within the acceptable limits of the SRP section 15.2.7.

-4 REFERENCES

1. Letter from K. P. Baskin to D. M. Crutchfield dated November 18, 1981, "Automatic Initiation of the AFWS - Amendment No. 97."

2. Letter from K. P. Baskin to D. M. Crutchfield dated March 6, 1981.

3. Letter from D. M. Crutchfield to R. Dietch, dated October 22, 1982.

4. Letter from R. W. Krieger to D. M. Crutchfield dated February 23, 1983.