ML18153B283
| ML18153B283 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/05/1993 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 93-049, 93-49, NUDOCS 9308110173 | |
| Download: ML18153B283 (8) | |
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VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 August 5, 1993 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 Serial No.
NO/ETS Docket Nos.
License Nos.
REGULATORY GUIDE 1. 97 VARIABLE QUALIFICATION STATUS93-049 R1 50-280 50-281 DPR-32 DPR-37 Virginia Electric and Power Company's letter dated January 31, 1984, detailed our response to Regulatory Guide (RG) 1.97, Revision 3, as required by NUREG-0737.
The N RC issued a Safety Evaluation Report on March 3, 1988, that concluded that the post accident monitoring instrumentation at Surry conforms to RG 1.97, Revision 3.
Since our 1984 submittal, a number of engineering and construction activities have taken place. A!so, an NRC inspection was conducted in the areas of our conformance with RG 1.97 requirements (NRC Inspection Report 50-280/281 /90-22 dated September 6, 1990). In response to these activities we have completed extensive re-reviews of the variables associated with the Regulatory Guide. These reviews have determined that the intent of the Regulatory Guide is being met at Surry Power Station.
These reviews also have determined that some additional exceptions to the Design and Qualification Criteria of RG 1.97 are required. contains the justification for these exceptions. Your review and approval of these exceptions is requested.
Additionally, the status of ongoing engineering projects to correct recorder and transmitter concerns for RG 1.97 variables is included in Attachment 2 for information.
-Should you have any additional questions or require additional information, please contact us.
Very truly yours,
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- w. L. Stewart Senior Vice President - Nuclear Attachments
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U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.
Suite 2900.
Atlanta, Georgia 30323 Mr. M. W. Branch NRC Senior Resident Inspector Surry Power Station Surry Power Station Units 1 and 2 Regulatory Guide 1.97 Variables Exceptions and Deletions
L Exceptions and Deletions 1.
Variable D-24 "Heat Removal by the Containment Fan Heat Removal System," Category 2.
Our letter, Serial number 053, dated January 31, 1984, identified the Containment Recirculation Cooler Fan subsystem as the Surry equivalent to the Containment Fan Heat Removal System as specified in Table 3 of Regulatory Guide 1.97. During our re-review of RG 1.97 variables in 1991, we determined that the containment licensing calculations for Surry Power Station do not rely upon the Containment Recirculation Cooler Fan subsystem for accident mitigation purposes. Rather, the Containment Spray and Recirculation Spray Systems are the systems required for mitigation of design bases accidents. The flows and temperatures for these systems are already being monitored as Variables D-23, "Containment Spray Flow", D-30, "Component Cooling Water Temperature to ESF System" and D-31, "Component Cooling Water Flow to ESF System". Since no credit is being taken for the Containment Recirculation Cooler Fan subsystem as a containment heat removal system, we are deleting D-24 as a Reg. Guide 1.97 variable.
- 2.
Variable A-8, "High Head Safety Injection Flow", Category 1.
Variable A-8, "High Head Safety Injection" was deleted as a Type "A" Variable.
This change in status was identified to the NRC in our letter Serial Number 87-323, dated August 12, 1987. The recent RG 1.97 re-review has identified the need to reinstate this variable as a Type "A" Variable.
The basis for reinstatement is on Reactor Coolant Pump trip criteria (operator manually trips RCPs) during a small break LOCA.
The manual action performed by the operator (RCP pump trip) is based upon HHSI flow indicated with the RCS sub-cooling margin less than 30 degrees. The manual action is not dependent upon any minimum or trended HHSI flow rates.
A Design Change Package (DCP) is currently in the process of development which will add a redundant flow indicator to the Control Board and re-configure the header flow transmitter power supplies. Addition of this variable will be in compliance with the guidance of RG 1.97 with one exception. The guidance for a Category 1 variable specifies a recorder be incorporated into the instrumentation for trending this variable. There are no accident mitigating procedures that require trended HHSI flow information, and the redundant flow indicators will provide the operator with the necessary system flow information.
Therefore, we are taking an exception to the requirement of RG 1.97, Criterion 6, "Display and Recording" regarding the recording/trending of Variable A-8.
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- 3.
Variable B-14, "Containment Isolation Valve Position," Category 1.
An exception to the following RG 1.97 Criteria is requested for the specific containment isolation valve position indicator listed below:
1.
Criterion 1, "Equipment Qualification"
- 2.
Criterion 2, "Redundancy"
- 3.
Criterion 3, "Power Source" These containment isolation valves are either closed or locked closed during accident conditions. The operation of the associated systems are not designed for or relied upon to mitigate the consequences of any accident. The isolation valves themselves will remain subject to the appropriate 1 O CFR 50 Appendix J tests. The exceptions to the RG criteria are for the valve position indications only and will not affect the containment integrity requirements of the valves. Based upon this justification, we are requesting exceptions to Design and Qualification Criteria 1, 2, and 3 for the valve position indications for the valves identified in the following tables.
The valves listed are Unit 1 specific, however the exceptions are also requested for the corresponding Unit 2 valves.
Containment Purge
- Valve VS-MOV-1 OOA VS-MOV-1008 VS-MOV-100C VS-MOV-1 OOD VS-MOV-101 VS-MOV-102 Loop Fill Header Valve CH-FCV-1160 Status During Accident (1)
Closed (breaker locked open)
Locked Closed Closed (breaker locked open)
Locked Closed Locked Closed Locked Closed Status During Accident Closed Note: This valve does not use a open/closed limit switch for valve position indication. The Control Room operator is provided with a percent open indicator instead.
Containment Vacuum Air Ejector Suction Valve Status During Accident (2)
CV-HCV-100 Locked Closed
Variable B-14, "Containment Isolation Valve Position,". (continued)
RHR to RWST Cross connect Valve RH-MOV-100 Status During Accident (3)
Locked Closed (1)
These valves are also required to be locked closed with RCS temperature above 200°F, per Technical Specification 3.8.A.2. These valves are also deenergized.
(2)
This valve is required to be locked shut when RCS temperature is above 200°F, per Technical Specification 3.8.A.3.
(3)
The motor operator has been removed and the valve is locked closed above 200°F.
e Surry Power Station Units 1 and 2 Regulatory Guide 1.97 Variables Status of Engineering Project
Status of Engineering Projects
- 1.
HHSI Flow Indication Addition.
This design change will provide the redundant flow indicator for safety injection flow in the Control Room, as noted in item number 2 of Attachment 1.
These packages are being developed and will be implemented by the end of each refueling outage prior to operating cycle 13 and 14 for Surry Units 2 and 1, respectively. These outages are currently scheduled for 1994 and 1995 for Surry Units 2 and 1, respectively.
- 2.
Inside Recirculation Spray Pump Transmitter Replacement In our letter of November 29, 1990, we reported that although Recirculation Spray (RS) flow should have been included in the scope of Item D-23 along with the Containment Spray flow, it had not been included in this variable. The letter went on to say that despite this omission, adding RS flow instrumentation would not provide any useful information to the operator that could not be already derived from existing Control Room instrumentation (e.g., RS pump motor current, RS pump discharge pressure, and containment temperature and pressure.).
Subsequent reevaluation has determined that the existing instrumentation for one of these variables (RS pump discharge pressure) is not qualified for post accident conditions inside containment and requires replacement and relocation. A design change is being developed to relocate the discharge pressure transmitters above the containment submergence level and also to make the transmitters environmentally qualified.
These packages are being developed and will be implemented by the end of each refueling outage prior to operating cycle 13 and 14 for Surry Units 2 and 1, respectively. These outages are currently scheduled for 1994 and 1995 for Surry Units 2 and 1, respectively.
- 3.
Equipment Identification Labeling of the Category A, B, and C variables, in accordance with RG 1.97 Criterion 8, "Equipment Identification," was being completed concurrent with the Surry Control Room Design Review (CRDR) project. To date, appropriate labeling of control room instruments for RG 1.97 variables used in the EOPs has bee completed.
Subsequently, reassessment of the remaining CRDR corrective actions has been performed and concluded that, as future labeling discrepancies are identified, they will be addressed by the station operations labeling program and/or the Nuclear Design Change Program. Therefore, the remaining labeling of RG 1.97 variables for control room equipment will be accomplished by one of these programs. The remaining Category A, B, and C variables (instruments) are scheduled to be labeled by the end of 1993.