ML18153B148
| ML18153B148 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 11/10/1994 |
| From: | Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML18153B149 | List: |
| References | |
| 94-648, NUDOCS 9411220189 | |
| Download: ML18153B148 (10) | |
Text
{{#Wiki_filter:e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 November 10, 1994 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC. 20555 Gentlemen: VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 Serial No. NL&P/ETS: Docket Nos. License Nos. PROPOSED TECHNICAL SPECIFICATIONS CHANGE CLARIFICATION OF CHANNEL FUNCTIONAL TESTING FOR 94-648 RO 50-280 50-281 DPR-32 DPR-37 REACTOR PROTECTION AND ENGINEERED SAFEGUARD SYSTEMS Pursuant to 1 O CFR 50.90, the Virginia Electric and Power Company requests amendments, in the form of a change to the Technical Specifications, to Facility Operating License Nos. DPR-32 and DPR-37 for Surry Power Station Units 1 and 2. The proposed changes will clarify the surveillance requirements for Reactor Protection and Engineered Safeguard Systems instrumentation and actuation logic. A discussion of the proposed Technical Specifications change for Surry is provided in. The proposed Technical Specifications change is provided in. It has been determined that the proposed Technical Specifications change does not involve an unreviewed safety question as defined in 1 O CFR 50.59 or a significant hazards consideration as defined in 1 O CFR 50.92. The basis for our determination that the change does not involve a significant hazards consideration is provided in Attachment 3. The proposed Technical Specifications change has been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Management Safety Review Committee. Should you have ~my questions or require additional information, please contact us. Very truly yours, ~U,p~~ James P. O'Hanlon Senior Vice President - Nuclear Attachments
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e cc: 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 Commissioner Department of Radiological Health Room 104A 1500 East Main Street Richmond, Virginia 23219
COMMONWEAL TH OF VIRGINIA ) ) COUNTY OF HENRICO ) The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. P. O'Hanlon, who is Senior Vice President - Nuclear, of Virginia Electric and Power Company. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief. Acknowledged -before me this /()7J/ day of~ /J..u), 192!/:_. My Commission Expires: ~. 3/, 19:/1_. ~-{!k Notary bttc (SEAL) Discussion of Change Surry Power Station
Discussion of Changes Introduction A comprehensive review of the Technical Specifications surveillance program for Surry was performed to ensure that a test was being performed for each Technical Specifications surveillance line item in Table 4.1-1, "Minimum Frequencies for Check, Calibration and Test of Instrument Channels." The review included an evaluation of the technical adequacy of each test with respect to the Technical Specifications requirement. The review identified certain instrument and logic circuit tests for which the licensed plant design does not facilitate the complete testing of the circuit or actuation logic without using temporary modifications (e.g., temporary jumpers or lifting leads). Consistent with the approved licensing basis and industry testing standards, these instrument and logic circuits are tested to the extent practicable on a monthly basis and are fully tested on a refueling basis. To ensure the Technical Specifications accurately reflect the actual testing being performed and are consistent with the existing plant design and/or the original licensing basis, Virginia Electric and Power Company is proposing an administrative change to the Technical Specifications to identify the instrument channels and logic actuation circuits that cannot be fully tested in accordance with the definition of channel functional test.
Background
Surveillance testing is required during power operation to ensure system operability. To satisfy the need for testing without interrupting power operation, an overlapping testing scheme is used. Each test within the scheme adequately tests a portion of the protection systems without causing or preventing an actuation of a protective function. Completion of the overlapping testing scheme confirms that the protection and safeguards systems will perform as designed. The few exceptions to this testing approach are limited to components that would require the installation of jumpers or temporary modifications for testing and, therefore, are more subject to equipment damage or adversely impacting power operations due to safety system actuation. Instead, these components are tested when the plant is shutdown. The original licensed plant design for the Reactor Protection and Engineered Safeguards Systems uses relays for the actuation logic and to initiate the safety function (e.g., pump starts, valve repositioning, etc.). This design does not support complete functional testing of each actuation logic combination or the associated actuation relays at power without using temporary modifications. The Reactor Protection System is Page 1 of 5
e e capable of being tested from the analog channel through the final actuation device (the trip breaker). However, portions of the Engineered Safeguards System and the Auxiliary Feedwater System automatic actuation logic and actuation relays cannot be fully tested. As part of the initial licensing process, the Reactor Protection and Engineered Safeguards Systems were reviewed and approved by the NRC. The Reactor Protection System is similar to the Westinghouse three loop reference plant and, therefore, only the differences from the reference design were reviewed by the NRC. Since the Engineered Safeguards System initiating circuitry is of a different design than the reference, a detailed review of the complete initiating circuitry was also performed by the NRC. As noted in the NRC's February 25, 1972 Safety Evaluation Report for the initial license, no deficiencies were observed in the initiating circuitry of the reactor trip, safety injection, or emergency feedwater systems. However, the licensed design and testing capability of the safety injection actuation logic was not explicitly included in the original Technical Specifications issued on May 25, 1972. The Auxiliary Feedwater System was upgraded in accordance with the requirements of NUREG-0737, "Post TMI Requirements." The system upgrades, including the actuation logic and testing capabilities, were reviewed and approved by the NRC in a Safety Evaluation Report dated August 7, 1981. Auxiliary Feedwater System actuation circuitry testing requirements were incorporated into the Technical Specifications by License Amendments 72 and 73 issued by the NRC on September 29, 1981. Similar to the Safety Injection System actuation logic, the licensed design and testing capability of the Auxiliary Feedwater System actuation logic was not explicitly included in the Technical Specifications change. Specific Changes Each instrumentation channel or actuation logic in Technical Specifications Table 4.1-1, "Minimum Frequencies for Check, Calibration and Test of Instrument Channels" that cannot be fully tested in accordance with the existing definition of channel functional test without lifting leads or installing jumpers will be identified with a note in the remarks column. The following clarifications are provided: Item 19, Containment Sump Level - This line item is duplicated in TS Table 4.1-2 Items 9 and 1 O and is deleted from Table 4.1-1. Page 2 of 5
e e Item 26, Logic Channel Testing - Notes are included in the remarks column to
- identify the logic channels tested monthly and to what extent the logic is tested. These notes are:
- 1)
Reactor protection, safety injection, and the consequence limiting safeguards systems logic channels are tested per this line item.
- 2)
The master and slave relays are not included in the monthly logic channel test of the safety injection system. Item 32, Auxiliary Feedwater
- a.
Steam Generator Water Level Low-Low - A low-low level on two-out-of-three channels in two-out-of-three steam generators starts the turbine driven pump. The two-out-of-three channel low-low level logic matrix is tested monthly. However, the two-out-of-three steam generator logic (auto start) cannot be tested without implementing temporary modifications or actuating the pump. Therefore, a note is included in the remarks section to indicate that the auto start of the turbine driven pump is not included in the monthly test but is tested within 30 days prior to each startup. Each pump autostart test satisfies the surveillance requirements for any startup within 30 days of the completed test.
- b.
RCP Undervoltage - An undervoltage condition on two-out-of-three station service busses initiates a reactor trip and starts the turbine driven pump. This actuation logic and the actuation relay contacts for reactor trip are tested monthly as part of Item 8, "4KV Voltage and Frequency." However, the actuation relay contacts for auto start of the pump cannot be tested monthly without implementing temporary modifications or actuating the pump. Therefore, the monthly requirement is changed to Not Applicable. In addition, a note is included in the remarks section to indicate that the actuation logic and relays are tested within 30 days prior to each startup. Each pump autostart test satisfies the surveillance requirements for any startup within 30 days of the completed test. Item 37, Safety Injection Input from ESF - This item is being renamed to Safety Injection Input to RPS. This item tests the reactor trip signal generated by the Safety Injection System and is being identified as such. Page 3 of 5
e Safety Significance The*proposed change to clarify the surveillance requirements for the Reactor Protection and Engineered Safeguards Systems instrumentation and actuation logic is administrative in nature. This change does not modify the existing test or test methodology. The circuits will continue to be tested to the fullest extent practical in accordance with the existing design, consistent with the regulatory position of not installing jumpers or lifting any leads to perform testing at power. The proposed change to clarify the surveillance requirements for the Reactor Protection and Engineered Safeguards Systems instrumentation and actuation logic has no impact on the probability of an accident occurrence. The instrumentation and actuation logic will continue to be operated in the same manner. The actual test frequency is not changing. This proposed change is only a clarification to reconcile the Technical Specifications with the license and design basis of the Reactor Protection and Engineered Safeguards Systems instrumentation and actuation logic. Since surveillance testing is not changing, the proposed clarification to more accurately describe actual testing does not contribute to the probability of any previously analyzed accident. The Reactor Protection and Engineered Safeguards Systems instrumentation and actuation logic will be operated in the same manner and operability requirements are not being altered. Therefore, the consequences of any design basis accident are not being increased by the proposed changes to clarify the surveillance test requirements for the Reactor Protection and Engineered Safeguards Systems instrumentation and actuation logic. There are no plant modifications or changes in methods of plant operation introduced by this clarificat_ion of the monthly surveillance testing for the Reactor Protection and Engineered Safeguards Systems instrumentation and actuation logic. The plant is not being operated in a different manner due to the proposed changes. Therefore, no new accidents or accident precursors are generated by the proposed changes to clarify the surveillance test requirements. Clarification of the testing for the Reactor Protection and Engineered Safeguards Systems instrumentation and actuation logic surveillance requirements does not affect the margin of safety in that the operability requirements for these safety systems remain unchanged. The existing testing is performed in accordance with the plant design and licensing basis and provides adequate indication of the operability of the affected instrumentation or actuation logic. Furthermore, the Reactor Protection and Engineered Page 4 of 5
e Safeguards Systems instrumentation and actuation logic are fully tested on a refueling cycle bas*is which includes operation of each relay and end device. Therefore, the margin of safety is not altered by the proposed clarification of the testing for the Reactor Protection and Engineered Safeguards Systems instrumentation and actuation logic. Page 5 of 5
e Technical Specifications Change Surry Power Station}}