ML13317A714
| ML13317A714 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/27/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Dietch R Southern California Edison Co |
| References | |
| TASK-15-01, TASK-15-02, TASK-15-1, TASK-15-2, TASK-RR LSO5-81-10-050, LSO5-81-10-50, NUDOCS 8110300343 | |
| Download: ML13317A714 (15) | |
Text
October 27, 1981 Docket No. 50-206 LS05 10-050 Mr. R. Dietch, Vice President Nuclear Engineering and Operations 198 1 Southern California Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Rosemead, California 91770
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Dear Mr. Dietch:
SUBJECT:
SAN ONOFRE 1 - SEP TOPICS XV-1 AND XV-2 (SYSTEMS)
By letter dated July 1, 1981, you submitted safety assessment reports for the above topics. The staff has reviewed these assessments and our con clusions are presented in the enclosed safety evaluation reports, which complete these'topics for San Onofre 1.
These evaluations will be a basic inputs to the integrated assessment for your facility. The evaluations may be revised in the future if your facility design is changed or if NRC criteria relating to these topics are modified before the integrated assessment is completed.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosures:
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 October 27, 1981 Docket No. 50-206 LSO5-81-10-050 Mr. R. Dietch, Vice~-President Nuclear Engineering and Operations Southern California Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Rosemead, California 91770
Dear Mr. Dietch:
SUBJECT:
SAN ONOFRE 1 - SEP TOPICS XV-1 AND XV-2 (SYSTEMS)
By letter dated July 1, 1981, you submitted safety assessment reports for the above topics. The staff has reviewed these assessments and our con clusions are presented in the enclosed safety evaluation reports, which complete these topics for San Onofre 1.
These evaluations will be a basic input to the integrated assessment for your facility. The evaluations may be revised in the future if your facility design is changed or if NRC criteria relating to these topics are modified before the integrated assessment is completed.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors.Branch No,, 5 Division of Licensing
Enclosures:
As stiated' cc w/enclosures:
See next page
Mr. R. Dietch cc Charles R. Kocher, Assistant General Counsel James Beoletto, Esquire.
Southern Cali-fornia 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 P. 0. Box 1831 San Diego, California 92112 Resident Inspector/San Onofre NPS c/o U. S. NRC P. 0. Box 4329 San Clemente, California 92672 Mission Viejo Branch Library 24851 Chrisanta Drive Mission Viejo, California 92676 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
SYSTEMATIC EVALUATION PROGRAM TOPIC XV-1 SAN ONOFRE TOPIC XV-1:
DECREASE IN FEEDWATER TEMPERATURE, INCREASE IN FEEDWATER FLOW, INCREASE IN STEAM FLOW AND INADVERTENT OPENING OF A STEAM GENERATOR RELIEF OR SAFETY VALVE I.
INTRODUCTION These events involve an unplanned increase in heat removal by the secondary system which can cause a decrease in the temperature of the reactor coolant, an increase in reactor power due to the negative moderator temperature coef ficient and a decrease in the reactor coolant system steam generator pressure.
Depending on the magnitude of the temperature reduction the plant may stabilize in new operating conditions, or the overpower or variable low pressure pro tection may cause a reactor trip.
Each event description below has a separate section for evaluation and conclusions.
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 evaluation 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 determination of the margins of'safety during normal operations and transient conditions anticipated during the life of the facility.
The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water-coolant 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 opera tion, including the effects of anticipated operational occurrence.
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 opera tional occurrences.
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.
-2 III.
RELATED SAFETY TOPICS Various other SEP topics evaluate such items as the reactor protection' system. The effects of single. fa-ilures on safe shutdown capability are con sidered under Topic VII-3.
IV.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.1.1., 15.1.2, 15.1.3 and 15.1.4.
The evaluation includes review of the analysis for the event and identifica tion of the features in the pl'ant 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.
(Draft Standard ANS 58..8 used as guidance). Deviations from the criteria specified in the Standard Review Plan are identified.
V.
INDIVIDUAL EVENT REVIEW DECREASE IN FEED4ATER FLOV A. Evaluation The licensee has not presented a detailed analysis of the transient, but has identified the increase in feedwater flow transient as a more limiting bounding event.
The feedwater system consists of two trains, each having one high pressure heater and a set of low pressure heaters.
From the figure in Reference 1, which presents a heat balance of the turbine cycle, it can be seen that the high pressure heaters increase the enthalpy by 301.0 Btu/1b.
Each of these heat sources could be bypassed by opening a single valve in a bypass line.
The licensee has estimated the consequences of the loss of both high pressure heaters in Reference 1, but has not considered the possibility of the loss of the whole set of low pressure heaters.
However, from the plant energy balance, it is estimated that the increase in feedwater flow transient is also a bounding event for loss of all low pressure heaters in one train.
B. Conclusion It is concluded that the decrease in feedwater temperature transient is bounded by the increase in feedwater flow transient.
-3 INCREASE IN FEEDWATER FLOW A. Evaluation An increase in feedwater flow can result from excessive opening of the feedwater control valve, overspeed of a feedwater pump, or starting a second feedwater pump. The increase in feedwater flow will cause more heat to be extracted by the steam generator. When colder water reaches the core, the power is increased through negative reactivity feedback.
The licensee has presented an analysis of increase in feedwater flow in Reference 2 and 3.
The largest increase in feedwater flow at full load would be caused by simultaneous opening of all three feedwater control valves. Feedwater flow is limited by the feedwater system heat-flow characteristics.
The licensees' analysis has conservatively assumed a step increase to 140% of full flow, which is beyond the actual limit of the feedwater system.
However, in order to limit the consequences of this transient the operator is required to manually trip the reactor within two minutes following the increase in feedwater.
As a second case, the licensee has analyzed a step increase from 43% to 103% feed flow when operating at a steady-state power of 51% of design.
A conservative low feedwater temperature is assumed for this case.
For this case the operator is required to manually trip the reactor within 40 seconds following this increase in feedwater.
All core parameters used in the analysis are more conservative than the actual parameters calculated for the current cycle. No automatic plant controls are assumed to function, but a manual reactor trip is assumed soon after the operator has been alerted by high level alarms on the steam generator. -The other assumptions made in the analysis are in conformance with the criteria of SRP Section 15.1.1.
The results of the analysis show that in both cases the primary system parameters smoothly approach their asymptotical values and do not reach the protective limits. *Thus there is no immediate concern of DNB or overp.ressure.
San Onofre 1 does not have a reactor trip on steam generator high level.
It is assumed that the transient is terminated by the operator who trips the reactor manually on steam generator high level alarm to avoid flooding of the steam lines.
In neither case does the time available for operator action comply with the criteria of draft standard ANS 58.8 which we have used as guidance for our safety evaluation (minimum 10 minutes). The licensee submittal does not indicate if the feedwater pumps are turned off automatically or if the operator has to trip the pumps after he has tripped the reactor.
-4 The consequences of not terminating the increase in feedwater flow transient.before the steam lines have flooded have not been evaluated.
i If the flooding would result in a steamline rupture the ensuing reactor coolant system overcooling would be more severe than during the worst case steamline break accident analyzed by the-licensee.
It is our position that the event should be reanalyzed using the ANS 58.8 guidance for operator action. The same issue (steam generator overfill) is also included in Task A-47, Safety Implications of,.
Control Systems, of the NRC program of Unresolved Safety Issues.
B. Conclusions The analysis of increase feedwater flow has been evaluated against the criteria of SRP 15.1.1.
We have concluded that the operator action times assumed in the analyses do not meet the criteria of ANS 58.8 and therefore credit for such actions should not be given. The increase in feedwater flow event should be reevaluated using a minimum time for operator action of 10 minutes. Since the issue of steam generator overfill due to an increase in feedwater flow is also included as part of a USI task, no immediate licensee action is required. The results of t-he staff review of USI Task A-47 should satisfactorily resolve our concerns regarding this event.
INCREASE IN STEAM FLOW A. An increase in steam flow may be initated by opening of the turbine con trol valves.
The plant response to steam flow increase depends on the control mode and the magnitude of the moderator reactivity coefficient.
In general, however, the core power tends to increase to a level matching the increased steam flow. In the automatic control mode, the power is increased by a combined effect of the control rod withdrawal and the feedback from moderator temperature.
In the manual control mode, it is increased as a result of the negative moderator temperature coefficient when the primary temperature and both primary and secondary pressures decrease.
The licensee has presented an analysis of an increase in steam flow in Reference 2.
The maximum thermal power level for the current operating conditions, evaluated in Reference 4, corresponds to the situation with the turbine control valves fully open.
This eliminates the possibility of a load increase above this power level.
Theevent analyzed in Reference 2 is a transient response to a sudden requirement for 30 percent more load by the turbine governor control while operating at 70 percent load.
An automatic reactor control and a slightly positive moderator coefficient are assumed which yield the most severe transient. Manual control or negative moderator coefficient would result in a more smooth transfer to annew equilibrium state.
-5 The results of the analysis show that the plant parameters do not reach the protection limits, set to protect the plant against DNB and over pressure.
B.
Conclusion The analysis of increase in steam flow has been evaluated against the criteria of SRP 15.1.1 and we have concluded that it is in conformance with the criteria.
INADVERTENT OPENING OF A STEAM GENERATO\\ RELIEF OR SAFETY VALVE A.
Evaluation An atmospheric dump valve may be inadvertently opened by the operator or may open due to a failure in the control system that opens the valve.
A steam generator safety valve may be opened only as a result of a valve failure. The consenuences of an inadvertent opening of either valve are the same as for a small break in the steam line: a reduction of the reactor coolant temperature and pressure and insertion of reactivity because of a negative moderator temperature coefficient.
If the reactor is at power during the inadvertent opening of a relief or safety valve, a.
reactor trip is caused by overpower protection.
Continued cooling to hot shutdown conditions actuates the safety injection system due to coincident low pressurizer pressure and low pressurizer level signals.
The core is ultimately shut down by the boric acid. injec tion delivered by the safety injection system.
The licensee has analyzed inadvertent opening of a steam generator relief or safety valve several times in the past.
Results of the most recent analysis are presented in Reference 4.
Inadvertent opening of a valve, equivalent to a steam leak of 152 lb/sec at 920 psia, has been analyzed assuming initial hot shutdown conditions.
This causes a larger and more rapid cooling than an event starting from power operation, because the stored energy in the reactor coolant system is smaller and the water inventory and pressure in the steam generator are higher.
The method and the assumptions used in the analysis are in conformance with the criteria of SRP Section 15.1.1.
The core parameters are more conserva tive than the actual parameters calculated for the current cycle.
The results of the analysis show that the reactor does not become critical during the transient. Thus there is no concern of DNB or overpressure.
-6 B. Conclusion The analysis of an inadvertent opening of a steam generator relief or safety valve has been.,evaluated against the criteria-of SRP 15.1.1 and we have concluded that it is in-conformance with the criteria.
VI.
TOPIC CONCLUSIONS Each of the events covered by this topic have been reviewed and we have concluded that the analyses are in conformance with SRP criteria with the following exception:
The increase in feedwater flow event should be reevaluated using a minimum time for operator action of 10 minutes. Since the issue of steam generator-overfill due to an increase in feedwater flow is also included as part of a USI task, no immediate licensee action is required.
The results of the staff review'of USI Task A-47 should satisfactorily resolve our concerns regarding SEP Topic XV-1.
REFERENCES
- 1. Letter, W. C. Moody (SCE) to D. M. Crutchfield (NRC)
Subject:
Design Basis Event Reviews, Systematic Evaluation Program, San-Onofre Nuclear Generating Station Unit 1, July 1, 1981.
- 2. San Onofre Nuclear Generating Station Unit 1, Part II Final Safety Analysfs, 1970.
- 3. San Onofre Nuclear Generating Station Unit 1, Part I Operating History and Verification of Design Objectives, Appendix A, 1970.
- 4. Reload Safety Evaluation Cycle 8, Revision 1, San Onofre Nuclear Generating Station Unit 1, October 1980.
SYSTEMATIC EVALUATION PROGRAM TOPIC XV-2 SAN ONOFRE TOPIC:
XV-2, EVALUATION OF STEAMLINE BREAK INSIDE/OUTSIDE CONTAINMENT I.- INTRODUCTION The steam release arising from a rupture of a main steam line would result in an initial increase in steam flow, which decreases during the accident as the steam pressure falls. The increased energy removal from the reactor coolant system causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an increase ih -reactor power and in a reduction of core shutdown margin.
The systems providing the necessary protection against the coisequences of a steamline break are:
- 1) reactor trip from nuclear overpower or from low pressurizer pressure,
- 2) safety injection system actuation from low pressurizer pressure or high containment pressure, and
- 3) isolation of the main feedwater lines in conjunction with receipt of the safety injection signal.
Following the reactor trip there is still a possibility that the core will be come critical and return to power. The core is ultimately shut'down by the boric acid injection delivered by the safety injection system.
II. REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for an operating license provide an analysis and evaluation of the design and performance of structures, systems and components of the facility with the objective of assess ing tfhe risk to public health and safety resulting from operation of the facility. The steam line break is one of the postulated accidents used to evaluate the adequacy of these structures, systems and components with respect to the public health and safety.
The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 27, "Combined Reactivity Control System Capability" requires that the reactivity control systems, in conjunction with poison addition by the emergency core cooling system, has the capability to reliably control reactivity changes to assure that under postulated accident conditions, and with appropriate margin for stuck rods the capability to cool the core is maintained.
-2 GDC 28, "Reactivity Limits" requires that the reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, itssupport structures or other-reactor pressure vessel internals to impair -significantly the capability to cool the core.
GDC 28 specifically requires that these postulated reactivity accidents include consideration of the steam line break accident.
III.
RELATED TOPICS SEP Topics III-5.A and III-5.B consider the dynamic effects of the pipe break on safety related equipment.
SEP Topics VI-2.D and VI-3 consider the ability of containment and the con tainment heat removal systems to mitigate the temperature/pressure transient.
Other SEP topics address such items as ESF initiation, emergency power supplies, and containment isolation.
IV. REVIEW GUIDELINES The review was conducted in accordance with SRP 15.1.5.
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. Deviations from the criteria specified in the Standard Review Plan are identified.
V.
EVALUATION In comparison with most Westinghouse-designed plants, San Onofre has certain specific features which influence the course of steamline break accidents:
- 1) no safety injection actuation signals are derived directly from secondary side parameters,
- 2) the steam outlets of all three steam generators are connected to a single steam header inside the reactor containment and cannot be isolated separately,
- 3) the steam is led out of the containment in two steam lines which are not isolated automatically during the accident, and
- 4) normal feedwater pumps, initially filled with unborated water, are used as safety injection pumps.
-3 The initial steamline break analysis was presented by the licensee in Ref erence 1. Since then the plant has been backfitted to improve its protection against loss of coolant accidents and steamline breaks. The most important modifications are:
- 1) new diesel generators-,capable to supply electrical power for the whole safety injection system,
- 2) reduced delays in valve alignment for safety injection mode,.resulting in earlier boron injection to reactor coolant loops, and
- 3) increased boron concentration in emergency coolant.
A steamline break reanalysis taking into account the plant modifications is presented in Reference 2. The most recent analysis, incorporating the appli cable standard Westinghouse steamline rupture analysis assumptions and the cur rent core data is presented in Reference 3.
The most recent analysis, presented in.Reference 3, covers four different cases of the double-ended rupture of the largest steam line. The cases are a rupture inside the flow restrictor and a rupture outside of it, each analyzed both with and without offsite power available.
In the analysis the initial plant conditions are assumed to be end-of-life core conditions, and the plant at hot shutdown.
These conditions normally max imize the consequences of the accident in Westinghouse designs. A steamline break at full power has been analyzed in Reference 1 and shown to be less severe than the corresponding case at hot shutdown.
In addition to double-ended breaks a small break equivalent to.about 10 percent of full load steam flow has been analyzed in Reference 3 and a small break equivalent to about 40 percent of full load steam flow has been analyzed in Reference 1.
The analysis presented in Reference 4 is performed using standard Westinghouse steamline rupture analysis methods and conservative current cycle physics para meters. An end of life shutdown margin of 1.9 percent Apwas assumed while the calculated shutdown margin is 2.3 percent Ap.
The methods and the assumptions used in the analysis are in conformance with the acceptance criteria of SRP Section 15.1.5.
The analyses presented in Reference 1 can be regarded as supporting sensitivity studies which point out the most conservative combination of accident assumptions.
-4 The results of the analysis show that the worst case is a break outside of the flow restrictor with offsite power available. The peak core thermal power during the transient is 43.1lpercent of full power. The minimum value of the DNBR is greater than-the 1.30 limit-and no DNB is expiected to occur.
The results of the analysis, assuming a double-ended break at full power or a small break at hot shutdown co-nditions, do not indicate a return to criticality.
The pressure in the reactor coolant and main steam systems is maintained well below the design pressure in all of-the cases analyzed.
The results of the analysis for the worst case indicate further that after about 200 seconds the reactor coolant system has stabilized at a pressure of 1200 psi and a temperature of 2500 F. All steam generators are fed by the auxiliary feedwater system and all three experience blow down through the break.
This simultaneous blow down of all steam generators can cause a pressurized thermal shock to the reactor pressure vessel.
The effects of thermal shocks are currently analyzed as a generic issue by Materials Engineering Branch and San Onofre 1 is one of the plants to be evaluated.
It is our position that the acceptance of steamline break analysis is pending until the evaluation by MTEB has been completed.
The response of the containment to steamline break is not evaluated in this connection, but it is a separate issue under Topic VI-2.D.
The radiological consequences of a steamline break accident have been previously evaluated and found acceptable under Topic XV-2 (dose), formerly XV-18.
VI.
CONCLUSIONS As part of the SEP review for San Onofre 1, we have evaluated the licensee's analysis of the steamline break accident (Ref. 1 and 3), against the criteria of SRP Section 15.1.5. Based on this evaluation we have concluded that the protection of the reactor pressure vessel against the pressurized thermal shock, caused by the accident, needs further consideration. This is a generic issue being pursued independently of the SEP.
REFERENCES
- 1. San Onofre Nuclear Generating Station Unit 1 Part II Final Safety Analysis, 1970.
- 2. Letter from K. P. Baskin to K. R. Goller-dated December 30,- 1976,-enclosure titled "Steamline Accident Reanalysis, San Onofre Nuclear Generating Station, Unit l."
- 3. Reload Safety Evaluation San Onofre Nuclear Generating Station Unit 1 Cycle 8 Revision 1, October 1980.