ML13310B570
| ML13310B570 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 07/25/1984 |
| From: | Medford M SOUTHERN CALIFORNIA EDISON CO. |
| To: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| Shared Package | |
| ML13310B572 | List: |
| References | |
| NUDOCS 8407310378 | |
| Download: ML13310B570 (7) | |
Text
Southern California Edison Company [984 JL 25 PM 12: h E
P.
. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD. CALIFORNIA 91770 M.O. MEDFORD TELEPHONE MANAGER, NUCLEAR LICENSING 0
) 5721749 4all U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement L: 0 Region V 1450 Maria Lane, Suite 210 Walnut Creek, California 94596-5368 Attention: Mr. John B. Martin, Regional Administrator
Dear Sir:
Subject:
Docket No. 50-206 Component Cooling Water System Maintenance Outage San Onofre Nuclear Generating Station Unit 1
References:
A. Letter, K. P. Baskin, SCE, to R. H. Engelken, NRC, dated June 9, 1980 B. Letter, M. 0. Medford, SCE, to 3. B. Martin, NRC, dated April 6, 1984 C. Letter, T. W. Bishop, NRC, to M. 0. Medford, SCE, dated May 4, 1984 D. Letter, M. 0. Medford, SCE, to D. M. Crutchfield, NRC, dated June 15, 1984 In the Reference A letter, SCE committed to maintain redundant or diverse means of decay heat removal capability during all modes of operation. In the Reference B letter, it was requested that NRC concurrence be given to a temporary revision to this commitment for the purpose of performing common train maintenance on components directly involved with meeting those commitments. The Reference C letter provided NRC concurrence that, due to the low decay heat levels associated with the current protracted outage, the commitments could be temporarily revised. As indicated in the Reference B letter, additional review was taking place to evaluate other activities that would necessitate a common train outage of the Residual Heat Removal (RHR) auxiliary systems. Our review has determined that requirements to test certain relief valves of the Component Cooling Water System (CCWS) require an additional outage similar to that for RHR maintenance. The purpose of this letter is to provide information regarding the common train outage of the CCWS and to request NRC concurrence with an additional temporary revision to our previous commitments.
PRIGPDAI 8407310378 840725 PDR ADOCK-05000206 Jet~e yL ~
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Mr. J. B. Martin
-2 The need for the CCWS outage stems primarily from a requirement that the relief valves (RV's) on the system (RV's 775A through 7751) be tested. Though Reference D indicated that these valves are not required to be tested as part of the Inservice Testing Program, there remains a requirement of the State of California Safety Order to test these valves at a frequency to be determined by the owner. These valves do not now have the capability of being tested on-line; therefore, it will be necessary to remove the valves for testing and maintenance which requires disabling the CCWS.
Enclosure.1 to this letter provides a description of the intended activities, by maintenance order (MO), which impact common components of the CCWS. Also included is a description of the requirements of other systems that will be affected by the CCWS outage. Since the CCWS is the only load supplied by the Saltwater Cooling System, during the CCWS outage the Saltwater Cooling System will also be secured.
This request is similar to that of the Reference B letter in that the decay heat being produced in the reactor is minimal.
However, the plant configuration for the CCWS outage will vary somewhat from the configuration of the RHR outage in that the CCWS provides cooling not only to the RHR heat exchangers which were disabled, but also to the Reactor Coolant Pump (RCP)
Thermal Barrier Coils which were not disabled. During the test done to verify conditions expected to occur during the RHR outage and during the RHR outage itself, Mode 5 conditions could be maintained indefinitely through Reactor Coolant System (RCS) ambient losses with the RCS in a filled and vented condition, RCP cooling, and cooling from the steam generators filled to approximately 300 inches. The effectiveness of this scenario was described in the Reference B letter.
During the CCWS outage, the same mechanisms of heat removal will be available except that the CCW to the RCP Thermal Barrier Coils will be terminated. The effect on the RCS heatup rate and additional administrative controls to monitor the RCS temperature are discussed below.
The additional effects of securing the CCWS on the RCS heatup rate will be minor as noted during the recent RHR outage. During that outage, CCW was secured to the RCP Thermal Barrier Coils for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. At 9:00 p.m. on May 21, 1984, when CCW cooling to the RCP's was secured, the RCS temperature was about 154.9 degrees Fahrenheit (OF).
At 8:00 p.m. on May 22, 1984, when CCW cooling to the RCP's was restored, the RCS temperature was about 157.8 OF.
The heatup rate experienced during this period was approximately 3.0oF averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The RCS heatup rate, both before CCW was secured and after CCW was restored, averaged approximately 1.50F per day. Although the increased heatup rate will represent a departure from the results reported in the Reference B letter, the heatup rate will remain low enough to permit the required work while remaining in Mode 5 with the Saltwater Cooling and CCWS out of service.
Mr. 3. B. Martin
-3 As a result of the heatup rate that will be experienced during the CCWS outage, additional control measures will be implemented to assure that an RCS temperature of 190OF is not exceeded. A plot of RCS temperature will be maintained and the heatup rate will be calculated every four hours.
The performance of work will be regulated in such a manner that the capability of restoring CCWS cooling to appropriate components within any 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> period during the outage is maintained. Action to restore CCWS will be initiated to permit at least 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> between the time restoration is begun until the projected heatup rate would result in an RCS temperature of 190 0F. Should the CCWS have to be restored before the planned work is completed, we intend to cooldown the RCS, secure the CCWS, and resume work subject to the control measures stated above. This cycle will be repeated if necessary to complete the required testing and maintenance. It is anticipated that the duration of the CCWS outage to implement the desired testing and maintenance will not exceed two weeks, but in any case, delays due to unforeseen circumstances are not expected to result in an outage of more than 21 days. The CCWS outage will not be permitted to proceed for more than 21 days without prior NRC resident staff notification. Current plans call for the commencement of the CCWS outage in late August after Hot Functional testing has been completed.
With the above precautions in effect, along with the routine control measures used to regulate plant activities, there is adequate assurance that an RCS temperature of 190OF will not be exceeded during the CCWS outage.
Because the concurrence provided in the Reference C letter was only for a temporary revision of our commitments and indicated that following the RHR outage the plant be returned to a condition consistent with the bulletin response contained in the Reference A letter, it is requested that NRC concurrence be provided for an outage of the CCWS. It is understood that upon completion of the CCWS outage, plant configuration will be immediately restored to comply with the decay heat removal commitments made in the Reference A letter.
If you have any questions or desire additional information regarding this matter, please contact me.
Very truly yours, Enclosure cc:
D. M. Crutchfield (NRC Division of Licensing)
A. E. Chaffee (NRC Senior Resident Inspector, San Onofre Units 1, 2 and 3)
A. D'Angelo (NRC Resident Inspector, San Onofre Unit 1)
Enclosure i INFORMATION IN SUPPORT OF COMPONENT COOLING WATER SYSTEM OUTAGE SAN ONOFRE UNIT 1
- 1. Description of Maintenance Activities The Component Cooling Water System (CCWS) maintenance and testing activities on common components are described in Attachment 1 to this enclosure which includes Drawing Nos. 5178310, 5178311 and 5178312.
The sequence of maintenance and testing activities will be such that impact on common train equipment is minimized. Maintenance orders may be added or deleted to this list of activities, as required, to support the CCWS outage within the scope of this submittal.
- 2. Makeup Requirements During CCWS Outage A review of the primary system makeup requirements during the 24 days of the heatup test for Configuration #2.of Special Procedure SO1-SPE-681 has determined that no makeup was required.
This was due to the fact that the pressurizer level did not vary sufficiently to require makeup to be initiated. The reduction in RCS leakage was probably attributable to the fact that the major sources of leakage are located in systems which were isolated due to the test configuration (charging, letdown, RHR).
Since this will also be the configuration during the CCWS outage, it is expected that no makeup will be required. If makeup is required during the outage, the system configurations will support this requirement since the charging test pump or the North Charging Pump, using its associated lube oil air cooling system, will be operable though in a secured state.
- 3. Reactor Coolant System Boron Dilution The only system which could provide unborated water to the primary system and which is affected by the CCWS outage is the charging system. During normal Mode 5 operation, a dilution could be initiated by the failure (opening) of valve FCV-1102A which supplies unborated water from the primary plant makeup tank to the chemical blending device.
The chemical blending device provides suction to the charging pumps which discharge to either the normal charging path through FCV-1112 or to the RCP seal water injection path via FCV-1115 A, 8, C, D, E and F. Dilution is possible with a single failure of FCV-1102A during normal Mode 5 operation because charging and letdown are in service with FCV-1112 and/or FCV-1115 A, B, C, 0, E and F in an open position. These dilution paths will be eliminated during the CCWS outage because normal charging and letdown will be secured by closing FCV-1112 and FCV-1115 A, B, C, D, E and F.
The plant configuration during the CCWS outage will therefore decrease the probability of an accidental dilution of the primary system for the only paths which are affected by the outage.
-2
- 4.
Potential Nitrogen Accumulation in the Reactor Vessel Head By letter dated December 12, 1983 we provided LER No.83-006 concerning an incident involving the Reactor Coolant System (RCS) at San Onofre Unit
- 1. This incident resulted in the accumulation of nitrogen gas in the reactor vessel head which was caused by the use of increased nitrogen pressure in the volume control tank (VCT).
VCT pressure is used as a driving force for charging flow to the RCS during some Mode 5 operational configurations. Since the solubility of nitrogen in the reactor vessel head is less than in the VCT, nitrogen comes out of solution in the RCS and collects in the reactor vessel head.
Corrective measures have been implemented as discussed in the LER for normal Mode 5 operations.
This problem is not a factor for the CCWS outage since the charging path will be isolated as discussed in Item 3 above. With the charging path isolated there is no open connection between the source of nitrogen (the VCT) and the RCS.
- 5. Reactor Coolant System Sampling The Technical Specifications require sampling for boron concentration twice each week. A concern was expressed regarding the adequacy of the fluid sample as a representative portion of the entire system due to the fact that no forced flow was provided during the RHR outage. A review of the sampling results during the heatup test was conducted.
The results were discussed in Attachment 2-4 of the Reference B letter. As can be seen from the sample boron concentrations descirbed in that attachment, the absence of forced flow had a negligible effect on the sampling. It is therefore not expected that boron redistribution due to the absence of forced flow will take place and the samples taken will be representative of the boron concentration throughout the entire RCS.
In order to provide additional assurances regarding the adequacy of RCS sampling for gross activity and boron concentration during the CCWS outage, the following steps will be taken:
- a. Two sample points at diverse locations in the RCS will be used.
- b. In addition to sampling the RCS, the potential point of dilution will also be sampled.
This sample, which will only be analyzed for boron concentration, will be taken at the charging pumps.
- c. Sampling frequency will be as listed in Technical Specification 4.1.C.
- 6. Additional Cooling Requirements for Component Cooling Water System Not only does the CCWS provide cooling to the RHR heat exchangers for decay heat removal purposes, but it also provides cooling water to several auxiliary systems.
Those systems which would normally receive cooling during this outage are discussed below.
-3
- a. Spent Fuel Pool Heat Exchanger The spent fuel pool is cooled by circulating CCW through the spent fuel pool heat exchanger. There is one spent fuel pool heat exchanger and one pump (there is an additional pump which is not connected to any piping but is available should the need arise).
The heat production by the spent fuel currently in the pool is minimal.
A calculation has been performed that demonstrates that for low levels of decay heat (as presently exist in the spent fuel pool) adequate cooling is provided by evaporation from the pool surface.
- b. Reactor Coolant Pumps The CCWS provides cooling to the Reactor Coolant Pump Thermal Barrier Coils and Upper and Lower bearing oil coolers.
Since the Reactor Coolant Pumps will not be run during the outage, there will be no requirement for this cooling.
- c. Charging/Test Pumps Should any makeup requirements arise, either the test pump, which does not rely on the CCWS for cooling, or the North Charging Pump, whose installed lube oil air cooling system is capable of maintaining proper bearing temperature, can be used.
- d. General The remaining loads supplied by the CCWS will either be secured during the outage or are so small that the loss of CCW will not manifest itself in any significant effect.
ATTACHMENT 1 COMPONENT COOLING WATER SYSTEM COMMON TRAIN OUTAGE ITEM M.O. #
OESCRIPTION
- 1.
84051268 Repair leaking North CCW Pump discharge valve (CCW-326).
- 2.
84051989 Repair frozen sample valve (CCW-492).
- 3.
84063181 Repair leaking Center CCW Pump discharge valve (CCW-327).
- 4.
84063182 Repair leaking North CCW Pump miniflow valve (CCW-346).
- 5.
84050303 Repair leaking Reactor Coolant Pump (RCP) A upper vent valve (CCW-013).
- 6.
84051301 Repair leaking Center CCW Pump miniflow valve (CCW-347).
- 7.
300123 Repair common inlet valve to RCP thermal barriers (CCW-453).
- 8.
84042409 Install lead seal on CCW Surge Tank relief valve (RV-787).
- 9.
84060489 Test Relief Valve 775A
- 10.
84064900 Test Relief Valve 7758
- 11.
84060491 Test Relief Valve 775C
- 12.
84060571 Test Relief Valve 775D
- 13.
84060575 Test Relief Valve 775E
- 14.
84060576 Test Relief Valve 775F
- 15.
84060577 Test Relief Valve 775G
- 16.
84060580 Test Relief Valve 775H
- 17.
84060581 Test Relief Valve 7751 4042u