ML041060368
| ML041060368 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 04/15/2004 |
| From: | Shukla G NRC/NRR/DLPM/LPD4 |
| To: | Rueger G Pacific Gas & Electric Co |
| Shukla G, NRR/DLPM/LPDIV-2/415-8439 | |
| Shared Package | |
| ML041060372 | List: |
| References | |
| TAC MC2535 | |
| Download: ML041060368 (16) | |
Text
April 15, 2004 Mr. Gregory M. Rueger Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P. O. Box 3 Avila Beach, CA 93424
SUBJECT:
DIABLO CANYON POWER PLANT, UNIT 1 - ISSUANCE OF EXIGENT AMENDMENT RE: REVISION TO TECHNICAL SPECIFICATION 3.3.5, LOSS OF POWER DIESEL GENERATOR START INSTRUMENTATION (TAC NO. MC2535)
Dear Mr. Rueger:
The Commission has issued the enclosed exigent Amendment No. 165 to Facility Operating License No. DPR-80 for the Diablo Canyon Power Plant, Unit 1. The amendment consists of changes to the Technical Specifications (TS) in response to your application dated April 2, 2004, as superseded by application dated April 8, 2004.
The amendment revises the TS 3.3.5, "Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation," to allow performance of Surveillance Requirement (SR) 3.3.5.2 for the trip actuation device operational test, prior to first entry into MODE 4, by adding a note to the FREQUENCY column of SR 3.3.5.2 on a one-time basis.
A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commissions next regular biweekly Federal Register notice.
Sincerely,
/RA/
Girija S. Shukla, Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-275
Enclosures:
- 1. Amendment No. 165 to DPR-80
- 2. Safety Evaluation cc w/encls: See next page
April 15, 2004 Mr. Gregory M. Rueger Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P. O. Box 3 Avila Beach, CA 93424
SUBJECT:
DIABLO CANYON POWER PLANT, UNIT 1 - ISSUANCE OF EXIGENT AMENDMENT RE: REVISION TO TECHNICAL SPECIFICATION 3.3.5, LOSS OF POWER DIESEL GENERATOR START INSTRUMENTATION (TAC NO. MC2535)
Dear Mr. Rueger:
The Commission has issued the enclosed exigent Amendment No. 165 to Facility Operating License No. DPR-80 for the Diablo Canyon Power Plant, Unit 1. The amendment consists of changes to the Technical Specifications (TS) in response to your application dated April 2, 2004, as superseded by application dated April 8, 2004.
The amendment revises the TS 3.3.5, "Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation," to allow performance of Surveillance Requirement (SR) 3.3.5.2 for the trip actuation device operational test, prior to first entry into MODE 4, by adding a note to the FREQUENCY column of SR 3.3.5.2 on a one-time basis.
A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commissions next regular biweekly Federal Register notice.
Sincerely,
/RA/
Girija S. Shukla, Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-275
Enclosures:
- 1. Amendment No. 165 to DPR-80
- 2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
PUBLIC GHill (2) TBoyce RidsNrrDlpmPdiv (HBerkow)
PDIV-2 Reading RidsNrrPMGShukla JKnox RidsNrrLAEPeyton RidsOgcRp RidsNrrLADBaxley RJenkins RidsAcrsAcnwMailCenter RidsRgn4MailCenter (BJones)
TS: ML NRR-100 PKG: ML041060372 ACCESSION NO.: ML041060368 NRR-058 *No major changes to SE input OFFICE PDIV-2/PM PDIV-1/LA EEIB/SC* OGC Nlo PDIV-2/SC NAME GShukla DBaxley RJenkins MLItzkowitz SDembek DATE 4/14/04 4/14/04 4/12/04 4/14/04 4/14/04 DOCUMENT NAME: C:\ORPCheckout\FileNET\ML041060368.wpd OFFICIAL RECORD COPY
Diablo Canyon Power Plant, Unit 1 cc:
NRC Resident Inspector Richard F. Locke, Esq.
Diablo Canyon Power Plant Pacific Gas & Electric Company c/o U.S. Nuclear Regulatory Commission P.O. Box 7442 P.O. Box 369 San Francisco, CA 94120 Avila Beach, CA 93424 Mr. David H. Oatley, Vice President Sierra Club San Lucia Chapter and General Manager c/o Henriette Groot Diablo Canyon Power Plant 1000 Montecito Road P.O. Box 56 Cayucos, CA 93430 Avila Beach, CA 93424 Ms. Nancy Culver City Editor San Luis Obispo The Tribune Mothers for Peace 3825 South Higuera Street P.O. Box 164 P.O. Box 112 Pismo Beach, CA 93448 San Luis Obispo, CA 93406-0112 Chairman Mr. Ed Bailey, Radiation Program Director San Luis Obispo County Board of Radiologic Health Branch Supervisors State Department of Health Services Room 370 P.O. Box 942732 (MS 178)
County Government Center Sacramento, CA 94234-7320 San Luis Obispo, CA 93408 Mr. James D. Boyd, Commissioner Mr. Truman Burns California Energy Commission Mr. Robert Kinosian 1516 Ninth Street (MS 31)
California Public Utilities Commission Sacramento, CA 95814 505 Van Ness, Room 4102 San Francisco, CA 94102 Mr. James R. Becker, Vice President Diablo Canyon Operations Diablo Canyon Independent Safety and Station Director Committee Diablo Canyon Power Plant ATTN: Robert R. Wellington, Esq. P.O. Box 3 Legal Counsel Avila Beach, CA 93424 857 Cass Street, Suite D Monterey, CA 93940 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower & Pavillion 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064
PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON POWER PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 165 License No. DPR-80
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Pacific Gas and Electric Company (the licensee) dated April 2, 2004, as superseded by application dated April 8, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 165, are hereby incorporated in the license. Pacific Gas and Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 10 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Stephen Dembek, Chief, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 15, 2004
ATTACHMENT TO LICENSE AMENDMENT NO. 165 TO FACILITY OPERATING LICENSE NO. DPR-80 DOCKET NO. 50-275 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT 3.3-41 3.3-41 3.3-42 3.3-42
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 165 TO FACILITY OPERATING LICENSE NO. DPR-80 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT, UNIT 1 DOCKET NO. 50-275
1.0 INTRODUCTION
By application dated April 2, 2004 (ADAMS Accession No. ML040970311), as superseded by application dated April 8, 2004 (ADAMS Accession No. ML041040599), Pacific Gas and Electric Company (PG&E or the licensee) requested exigent changes to the Technical Specifications (TS) (Appendix A to Facility Operating License No. DPR-80) for the Diablo Canyon Power Plant (DCPP), Unit 1.
The amendment would revise the TS 3.3.5, "Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation," to allow performance of Surveillance Requirement (SR) 3.3.5.2 for the trip actuation device operational test (TADOT), prior to first entry into MODE 4, by adding a note to the FREQUENCY column of SR 3.3.5.2 on a one-time basis.
PG&E requested the license amendment on an exigent basis to allow a delay in the performance of the SR 3.3.5.2 testing, which has come due on DCPP Unit 1. Early performance of the SR 3.3.5.2 testing is not appropriate from an outage safety standpoint.
From an outage defense-in-depth standpoint, it is better to perform SR 3.3.5.2 later in the 12th Refueling Outage (1R12) as originally planned, when the desired redundancy of safety-related equipment is available. Approval to extend the due date for SR 3.3.5.2 will allow SR 3.3.5.2 testing to be performed in conjunction with load shed and auto sequencing testing for each DG and its associated vital bus at the end of the outage, prior to entry into Mode 4, as originally planned. This approval is requested to support planned maintenance on DG 1-1, which is the only Unit 1 DG that is in compliance with SR 3.3.5.2. Approval to extend SR 3.3.5.2, will allow declaring DG 1-3 operable to support planned maintenance, without first having to perform SR 3.3.5.2 testing.
2.0 REGULATORY EVALUATION
The NRC staff finds that the licensee in Section 6.0 of its April 8, 2004, submittal identified the applicable regulatory requirements. The regulatory requirements that the staff based its acceptance are provided below.
General Design Criterion (GDC)-17, Electric Power System, of Appendix A, General Design Criterion for Nuclear Power Plants, to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR) requires that nuclear power plants have an onsite electric power system and an offsite electric power system to permit the functioning of structures, systems, and components
important to safety. The safety function of each system (assuming the other system is not functioning) is to provide sufficient capacity and capability to assure that (1) fuel design limits and design conditions of the reactor coolant boundary are not exceeded as a result of anticipated operational occurrences (AOOs) and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents. The onsite electric power supplies (including the batteries) and the onsite electric distribution system is required to have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.
Electric power from the transmission network to the onsite electric distribution system is required to be supplied by two physically independent circuits designed and located so as to minimize the likelihood of their simultaneous failure. Each of these circuits is required to be designed to be available in sufficient time following a loss of all onsite alternating current (ac) power supplies and the other offsite electric power circuit, to assure that fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits is required to be available within a few seconds following an accident to assure that core cooling, containment integrity, and other vital safety functions are maintained. In addition, GDC-17 requires provisions to minimize the probability of losing electric power from the remaining electric power supplies as the result of LOP from the unit, the offsite transmission network, or the onsite power supplies.
The regulation at 10 CFR 50.36, Technical Specifications, requires the TS to be derived from the analyses and evaluation included in the safety analysis report. A TS limiting conditions for operation (LCOs) is required to be established for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design-basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The SRs are included as part of the TS, to assure that the necessary quality of systems and components are maintained and LCOs will be met. When an LCO is not met, either due to component failure or maintenance outage, action is required within a specified time by the TS to restore required equipment to an operable condition. This specified time to take action is referred to as the allowed outage time (AOT) or completion time (CT). The AOT or CT is a temporary relaxation of operability for required equipment, which provides a limited time to fix components and return required equipment to an operable status. Establishing this limited time to fix components is based, primarily, on the reliability of remaining required operable equipment during the short time period of a CT, commensurate with reliability when all required equipment is operable.
DCPP Design Considerations Appendix 3.1A to the DCPP Final Safety Analysis Report Updated (FSARU) states that the Diablo Canyon Power Plant Units 1 and 2 designs conform to GDC-17 (dated 1971), instead of GDC-39, dated 1967. This is an exception to the commitments to GDCs dated 1967.
The DCPP electric power system design includes offsite and onsite electric power systems to permit functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) is to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design
conditions of the reactor coolant boundary are not exceeded as a result of AOOs and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.
The DCPP offsite system includes two physically independent offsite power supply circuits between the offsite transmission network and the onsite electric distribution system that are designed and located to minimize to the extent practical the likelihood of their simultaneous failure under operating, postulated accident and environmental conditions. Each offsite circuit is designed to be available in sufficient time following the loss of all onsite ac standby power supplies and the other offsite circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. In addition, one of these two offsite circuits is designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.
The onsite electric power supplies and distribution systems are designed with sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure. The DCPP onsite system design includes, in part, three independent and redundant ac system divisions each with its associated onsite ac standby DG power supply and distribution system (train or load group).
As part of DCPPs design, LOP instrumentation, DG start relays, and a system for starting loads in sequenced steps are installed to support the availability of sufficient capacity and capability of offsite and onsite power supplies and their associated load groups. The LOP instrumentation consists of two under-voltage sensing schemes installed as part of each of the three independent and redundant load groups. The first scheme detects loss of voltage and the second scheme detects degraded voltage conditions. The first scheme consists of a separate pair of under-voltage protection relays on each of the three safety buses. The relays have a two-out-of-two logic arrangement for each bus to prevent inadvertent tripping of operating loads during a loss of voltage either from a single failure in the potential circuit or from human error.
The second scheme consists of two relays and timers for each bus having a two-out-of-two logic arrangement.
The LOP instrumentation at DCPP trips the offsite power supply breaker to the load group from the normal and standby offsite supplies. Tripping the supply breakers isolates (and thus protects) the load group from the degraded and transient voltage conditions that may exist on the offsite power supply during a loss of offsite power (or degraded power) event. This protection pursuant with the requirements of GDC-17, (a) minimizes the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the LOP generated by the nuclear power unit, the LOP from the transmission network, or the LOP from the onsite electric power supplies and (b) thus supports the availability of sufficient capacity and capability of the load group when needed.
It has been the staffs position that LOP (and degraded power) schemes include coincident actuation logic. Coincident actuation logic precludes spurious trips thus eliminating unnecessary challenge to offsite and onsite power sources and safety systems. This coincidence logic, pursuant with the requirements of GDC-17, minimizes the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the LOP generated by the nuclear power unit, the LOP from the transmission network, or the LOP
from the onsite electric power supplies. The DCPP actuation logic design of two-out-of-two channels tripping includes coincident actuation logic, meets this staff position, eliminates challenge to offsite and onsite power sources, and therefore minimizes the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the LOP from the transmission network.
Each DG is automatically started by a signal generated by an undervoltage (UV) relay connected to (or which monitors the voltage on) the load groups 4.16kV bus. This automatic start supports the availability of sufficient capacity and capability of onsite power supplies and their associated load group.
The DCPP design includes a system for automatically sequencing loads. At DCPP, when voltage is reestablished to the load group from either the offsite or onsite power source, safety loads are connected to the bus in sequence steps. This sequencing supports the availability of sufficient capacity and capability of the offsite and onsite power supplies and its associated load group.
The LOP instrumentation, DG start relays, and a system for starting loads in sequenced steps support the availability of sufficient capacity and capability of offsite and onsite power supplies and their associated load groups. The design of the electric power system (which includes the availability of support from LOP instrumentation, DG start relays, and a system for starting loads in sequenced steps) meets the requirements of GDC-17.
3.0 TECHNICAL EVALUATION
The NRC staff has reviewed the licensee's technical and regulatory analyses in support of its proposed license amendment which are described in Section 5.0 of the licensee's April 8, 2004, submittal.
Surveillance requirement 3.3.5.2 requires the performance of a TADOT. The TADOT (which is a test that checks the trip devices that provide actuation signals directly) requires a test that checks that the DG start relay (one per safety bus) will generate a LOP signal within the specified under voltage conditions. The TADOT also requires a test that checks that the second level of UV relays and timers (two relays and timers per bus) each generate a degraded voltage signal within the specified under voltage conditions and times.
The licensee stated that these relays and timers do not experience time-dependent drift and that a review of surveillance, maintenance and operational history of these devices indicate that there were not time-dependent failures or problems in meeting the TS test requirements. In addition, there were no recurring failures observed in historical data.
The proposed extension of the allowed interval for performing the TADOT does not impact the design basis requirements of the offsite or onsite electric power systems. The design will continue to provide sufficient capacity and capability, pursuant with the requirements of GDC 17, to ensure the availability of necessary power to safety systems when needed. The reliability of these relays and timers is such that their capability to open or close (or to perform their required safety function) when needed is expected to be available to perform their required safety function if needed for a loss of offsite power event. If the relays should fail during a loss of offsite power event, the safety analysis for current operating mode of the plant (refueling)
indicates that sufficient time will be available to manually start and load the diesel generator as needed to meet safety requirements.
Based on the licensees findings and conclusions, the reliability of these relays and timers, the additional time permitted by the safety analysis to manually start and load the diesel generator, and the change having no impact on design basis requirements, the staff concludes the proposed change is acceptable.
Conclusion Therefore, based on the above review, the staff concludes that the effect on safety of extending the surveillance interval is small and that the proposed extension does not impact the design basis requirements of electrical systems. The reliability of these relays and timers and the additional time permitted by the safety analysis to manually start and load the diesel generator provides additional assurance that the design with the extended surveillance interval continues to meet the requirements of GDC 17. The proposed extension of surveillance interval, prior to first entry into Mode 4, is therefore considered acceptable.
4.0 EXIGENT CIRCUMSTANCES
The Commissions regulations, 10 CFR 50.91, contain provisions for issuance of amendments when the usual 30-day public comment period cannot be met. One type of special exception is an exigency. An exigency is a case where the staff and licensee need to act promptly. In this case, the licensee cannot perform a test on the schedule required by TS SR 3.3.5.2. Pursuant to 10 CFR 50.91(a)(6), the licensee requested the proposed amendment on an exigent basis.
Under such circumstances, the Commission notifies the public in one of two ways: (1) by issuing a Federal Register notice providing an opportunity for hearing and allowing at least two weeks for prior public comments, or (2) by issuing a press release discussing the proposed changes, using local media. In this case, the Commission used the second approach and published a public notice in the local newspaper San Luis Obispo Tribune on April 13 and 14, 2004.
Testing to satisfy SR 3.3.5.2 is normally performed at the end of each refueling outage, in conjunction with testing the load shed and auto sequencing functions for each DG and its associated vital bus. Performance of SR 3.3.5.2 requires de-energizing the vital bus for each DG. This testing is done prior to entering Mode 4 to assure that all components associated with the vital buses are available for testing.
The 18-month frequency for SR 3.3.5.2, and the interval of 1.25 times the frequency allowed by SR 3.0.2, normally provides sufficient time to schedule the tests that satisfy SR 3.3.5.2 at the end of each refueling outage without it becoming overdue. However, following the Unit 1 eleventh Refueling Outage, Unit 1 experienced problems with the main generator, which resulted in a 52-day forced Mode 5 outage. This outage affected core burnup for the cycle, which subsequently moved the start date for the current Unit 1 outage 1R12 to a later date, and made SR 3.3.5.2 due in the beginning of 1R12.
For Unit 1 in its present condition (core offloaded), SR 3.3.5.2 does not have to be current as long as there is no movement of irradiated fuel assemblies. However, to allow scheduled fuel reconstitution activities to proceed as planned, a performance of the TADOT for SR 3.3.5.2 for Bus H and DG 1-1 was conducted during the weekend of April 3-4, 2004, using a special test, with increased management oversight. Completion of this test allowed making DG 1-1 operable so that fuel reconstitution activities could proceed.
SR 3.3.5.2 testing for Bus F prior to restoration of Bus G has been determined to be unacceptable based on the outage risk assessment results. In the case of Bus F testing, the only available auxiliary saltwater (ASW) pump would have to be tripped and restarted twice to complete the test. The other ASW pump is powered from 4-kV Bus G, which is cleared for maintenance at this time, per the outage safety schedule. The ASW system provides cooling to the spent fuel pool cooling system via the component cooling water system. With the core offloaded to the spent fuel pool, interrupting ASW flow at this time places the unit in an unacceptable outage risk assessment condition. Currently, 4-kV vital Bus G is out of service for maintenance, making its associated DG 1-2 unavailable for testing. To provide a plant configuration with the desired redundancy of safety-related equipment (defense-in-depth) without further disruption of the outage plan, it is advisable to retain the existing schedule for performance of the SR 3.3.5.2 TADOTs for Buses F and G later in the outage. This will require extending the frequency of SR 3.3.5.2 as requested in this license amendment request.
Sufficient testing has been performed to maintain DG 1-1 operable in Modes 5 and 6, to allow continued fuel movement in the fuel handling building. However, the licensee will not be able to load the core until extension of the due date for SR 3.3.5.2 is approved, or the test has been completed to support mode 6 activities. At least two DGs are required to provide emergency backup power for two source range channels for core reload. Performance of the surveillance now to meet the existing surveillance frequency requirements will result in de-energizing a vital 4-kV bus when less redundant equipment is available. This results in an undesirable interruption of ASW cooling for the fuel in the spent fuel pool. Approval to extend SR 3.3.5.2 will allow declaring DG 1-3 operable at that time to support planned maintenance, without first having to perform SR 3.3.5.2 testing.
Therefore, early performance of the SR 3.3.5.2 test is not appropriate from an outage defense-in-depth standpoint, and it is safer to perform SR 3.3.5.2, prior to first entry into Mode 4, when the desired redundancy of safety-related equipment is available.
There were no public comments in response to the notice published in the San Luis Obispo Tribune on April 13 and 14, 2004.
5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION The Commissions regulations in 10 CFR 50.92 state that the Commission may make a final determination that a license amendment involves no significant hazards considerations if operation of the facility in accordance with the amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in the margin of safety.
Operation of the DCPP Unit 1 in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident evaluated. Extending the frequency of SR 3.3.5.2 on a one-time basis for Unit 1 does not involve an accident initiator.
The SR only verifies that UV relays operate (change state) when a UV condition occurs on a 4-kV vital bus. Other SRs performed in Mode 1 are used to verify relay setpoints and characteristics that are subject to time-dependent changes. Therefore, the proposed change will not increase the probability of any accident previously evaluated. Also, the proposed change has no effect on the radiological consequences of any accident previously evaluated because it will not delay or prevent any plant equipment from performing its design function.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Operation of the DCPP Unit 1 in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change does not involve or create an accident initiator. It only extends the surveillance interval for verifying operability of UV relay sequencing. Other SRs are used to verify relay setpoints and characteristics that are subject to time-dependent changes.
Therefore, the proposed change does not create the possibility of a new or different accident from any accident previously evaluated.
Operation of the DCPP Unit 1 in accordance with the amendment will not involve a significant reduction in the margin of safety. The proposed change only involves extending a SR frequency that has no time-dependent characteristics. The SR itself is not changed. Other SRs are used to verify relay setpoints and characteristics that are subject to time-dependent changes. Extension of this SR frequency has no effect on any margin of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based upon the above considerations, the staff concludes that the amendment meets the three criteria of 10 CFR 50.92. Therefore, the staff has made a final determination that the proposed amendment does not involve a significant hazards consideration.
6.0 STATE CONSULTATION
In accordance with the Commissions regulations, the California State official was notified of the proposed issuance of the amendment. The State official had no comments.
7.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has made a final finding that the amendment involves no significant hazards consideration. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
8.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: J. Knox Date: April 15, 2004