ML13330A765
| ML13330A765 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 08/29/1979 |
| From: | Arenal A SOUTHERN CALIFORNIA EDISON CO. |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| NUDOCS 7910020616 | |
| Download: ML13330A765 (5) | |
Text
Southern California Edison Company P. 0. BOX 800 2244 WALNUT GROVE AVENUE ROSEmEAD, CALIFORNIA 91770 A.ARENAL TELEPHONE VICE PRESIDENT August 29, 1979 213-572-1476 U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region V 1990 North California Boulevard Suite 202, Walnut Creek Plaza Walnut Creek, California 94596 Attention:
Mr. R. H. Engelken, Director DOCKET No.
50-206 SAN ONOFRE UNIT 1
Dear Sir:
\\II IE BULLETIN 79-06C NUCLEAR INCIDENT AT THREE MILE ISLAND -
SUPPLEMENT Reference is made to your correspondence of July 26, 1979, forwarding the subject IE Bulletin.
Identified therein were actions required as a result of the incident at Three Mile Island.
Response to individual items specified in the Bulletin are listed below:
Short-Term Actions ITEM 1.
In the interim, until the design change required by the long-term actions of this Bulletin has been incorporated, institute the following actions at your facilities:
A.
Upon reactor trip and initiation of HPI caused by low reactor coolant system pressure, immediately trip all operating RCPs.
B.
Provide two licensed operators in the control room at all times during operation to accomplish this action and other immediate followup actions required during such an occurrence.
For facilities with dual control rooms, a total of three licensed operators in the dual control room at all times meets the requirements of this Bulletin.
7zo cos o &
U.S. Nuclear Regulatory Commission August 29, 1979 ITEM 1. Continued
RESPONSE
- a. Operating procedures have been revised to require trip of operating Reactor Coolant Pumps upon reactor trip and initiation of HPI caused by low reactor coolant system pressure.
- b. Two licensed operators are now in the control room at all times to accomplish this action and other immediate and followup actions required during such an occurrence.
ITEM 2.
Perform and submit a report of LOCA analyses for your plants for a range of small break sizes and a range of time lapses between reactor trip and pump trip.
For each pair of values of the parameters, determine the peak cladding temperature (PCT) which results.
The range of values for each parameter must be wide enough to assure that the maximum PCT or, if appropriate, the region containing PCTs greater than 2200 degrees F is identified.
RESPONSE
A series of Loss of Coolant Accident (LOCA) analyses for a range of break sizes and a range of time lapses between initiation of break and pump trip applicable to 2, 3, and 4 loop plants has been performed by the Westinghouse Owners Group.
A report summarizing the results of the analysis of delayed reactor coolant pump trip during small LOCA's will be submitted to Mr.
D. F. Ross by Mr.
Cordell Reed on August 31, 1979.
In the report, maximum PCT's for each break size considered and pump shutoff times have been provided.
The report concludes that if the reactor coolant pumps are tripped prior to the reactor coolant system pressure reaching 1250 psia, the resulting PCT's are Iess than or equal to those reported in the FSAR. In addition, it is shown that there is a finite range of break sizes and RCP trip times in all cases 10 minutes or later, which will result in PCT's in excess of 2200 0 F as calculated with conservative Appendix K models.
The conclusions of that study apply to all standard Westinghouse PWR's, where "standard," for this purpose, means recent vintage 2, 3 and 4 loop geometry configurations using a combination of safety injection accumulators and pumps to provide emergency core cooling.
The applicability of the results of that study to San Onofre Unit
- 1 has been evaluated.
Since the basic reactor coolant system geometry and plant operating conditions of San Onofre Unit #1 are similar to those of a "standard" Westinghouse plant, the trends observed in the study are expected to be similar for San Onofre Unit #1. That is, there will be an RCP trip time delay that could result in a higher PCT than that PCT calculated in the FSAR analysis.
That critical delay time and the magnitude of its
U.S. Nuclear Regulatory Commission August 29, 1979 impact on PCT, however, cannot be accurately determined at this time because of the analytical differences and the physical plant differences that exist between the generic study and San Onofre Unit 11.
However, the interim action specified by Item #1 above requires the immediate trip of all operating RCP's upon reactor trip and initiation of Safety Injection caused by low reactor coolant system pressure. This action, in essence, results in LOCA transient consequences which are bounded by the conservatism of San Onofre FSAR LOCA analyses since RCP trip in those analyses is assumed coincident with reactor trip. For all LOCA transients, reactor trip would take place on variable low pressure (1840 psig minimum setpoint) prior to initiation of safety injection (1685 psig setpoint).
However, the delay time between reactor trip and safety injection initiation would be negligible (a few seconds for the smallest breaks) for all LOCA's considered in the generic analysis.
Based on the above, we have concluded that automatic RCP trip coincident with safety injection initiation, as discussed further in our response to the long term action item below, is appropriate for San Onofre to provide assurance that the PCT's following all LOCA and non-LOCA transients remain within acceptable limits.
Therefore, further analyses to determine plant specific critical RCP trip delay time and the magnitude of its impact on PCT are not warranted.
ITEM 3.
Based on the analyses done under Item 2 above, develop new guidelines for operator action, for both LOCA and non-LOCA transients, that take into account the impact of RCP trip requirements. For Babcock &
Wilcox designed reactors, such guidelines should include appropriate requirements to fill the steam generators to a higher level, following RCP trip, to promote natural circulatio.flow.
RESPONSE
The Westinghouse Owners' Group has developed guidelines which were submitted to the NRC in Section 6 and Appendix A of WCAP 9600.
The evaluation provided as the response to Item 2 is consistent with the guidelines in WCAP 9600.
No changes to these guidelines are needed for both LOCA and non-LOCA transients.
ITEM 4.
Revise emergency procedures and train all licensed reactor operators and senior reactor-operators based on the guidelines developed under Item 3 above.
RESPONSE
The Owner's Group effort to revise emergency procedures covers many issues, including operation of the Reactor Coolant Pumps.
U.S. Nuclear Regulatory Commission August 29, 1979 The action taken in response to Item 1 is sufficient as an interim measure and no immediate need exists for changing our emergency procedures to include the tripping of the Reactor Coolant Pumps.
The expected schedule for revising the LOCA, steamline break and steam generator tube rupture emergency procedures is the following:
Mid-October:
Guidelines which have been reviewed by the NRC will be provided to each utility. Appropriate utility personnel associated with writing procedures will meet with the Owners' Group Subcommittee on Procedures and Westinghouse to provide the background for revising their emergency procedures.
1 to 2 months -- Plant specific procedures will be revised.
from Mid-October:
3 to 4 months -- Revised procedures will be implemented and from Mid-operators trained.
October Upon completion of the design change initiating automatic RCP trip on safety injection actuation, all applicable emergency operating instructions will be revised as appropriate and all licensed operators and senior operators will receive training regarding the details of the change and subsequent procedure revisions.
ITEM 5. Provide analyses and develop guidelines and procedures related to inadequate core cooling (as discussed in Section 2.1.9 of NUREG-0578 "TMI 2 Lessons Learned Task Force Status Report and Short-Term Recommendations") and define the conditions under which a restart of the RCP's should be attempted.
RESPONSE
Analyses related to inadequate core cooling and definition of conditions under which a restart of the RCP's should be attempted will be performed. Resolution of the requirements for the analyses and an acceptable schedule for providing the analyses and guidelines and procedures resulting from the analyses will be arrived at between the Westinghouse Owners' Group and the NRC staff.
Again, a standard Westinghouse PWR will be used in the study. The applicability of the results of these analyses to San Onofre Unit
- 1 will be evaluated. Any additional analyses required to develop plant specific guidelines and procedures will be identified and a schedule for completion will be prepared at that time.
0 U.S. Nuclear Regulatory Commission -
August 29, 1979 Long-Term Action ITEM 1. Propose and submit a design which will assure automatic tripping of the operating RCP's under all circumstances in which this action may be needed.
RESPONSE
A design change which will assure automatic tripping of any operating RCP upon initiation of a safety injection actuation signal is discussed below.
Spare safety injection initiation sequencer contacts (one per sequencer connected in parallel) and an auxiliary relay will be connected to the DC control power of the reactor coolant pumps.
Upon SIS actuation, the sequencer contacts will close and energize the auxiliary relay (connected in series with the contacts).
An auxiliary relay contact in each reactor coolant pump stop circuit will close, thereby energizing the circuit breaker internal trips.
Manually operated knife switches will be installed to permit on-line testing of the sequencers. The spare sequencer contacts which will be utilized are isolated such that failures in the new circuit will not affect the sequencers.
The above described design change will assure automatic tripping of the operating RCP's under all circumstances when this action is required.
We will review and document the above design change in accordance with the requirements of 10CFR50.59.
Installation., testing and implementation of the change will be accomplished during the first plant shutdown of sufficient duration-following review committee approval and receipt of material.
Updh implementation, we will delete the requirements for providing two licensed operators in the control room per your direction in Item 1 above.
Sincerely, A. Arenal cc:
Director, Office of Inspection and Enforcement Division of Reactor Operations Inspection