ML041950260
| ML041950260 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 07/02/2004 |
| From: | Shukla G NRC/NRR/DLPM/LPD4 |
| To: | Rueger G Pacific Gas & Electric Co |
| Shukla G, NRR/ADIP/NRLPO,415-8439 | |
| Shared Package | |
| ML041950300 | List: |
| References | |
| TAC MB6758, TAC MB6759 | |
| Download: ML041950260 (19) | |
Text
July 2, 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 NOS. 1 AND 2 - ISSUANCE OF AMENDMENT RE: CREDIT FOR AUTOMATIC ACTUATION OF PRESSURIZER POWER OPERATED RELIEF VALVES (TAC NOS. MB6758 AND MB6759)
Dear Mr. Rueger:
The Commission has issued the enclosed Amendment No. 171 to Facility Operating License No. DPR-80 and Amendment No. 172 to Facility Operating License No. DPR-82 for the Diablo Canyon Power Plant, Unit Nos. 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated September 24, 2002, and its supplements dated November 21, 2003, and March 9, 2004.
The amendments revise TS Section 3.4.11 to credit the automatic actuation of the pressurizer power operated relief valves for mitigating the plant transient of inadvertent actuation of the safety injection system.
The application also requested to revise TS 3.4.10, "Pressurizer Safety Valves," to allow pressurizer safety valve loop seal temperatures to be less than the lower design temperature during plant heatup and cooldown in Mode 3, and in Mode 4 when any reactor coolant system cold leg temperature is greater than the low temperature overpressure protection arming temperature specified in the pressure temperature limits report, provided at least one Class I PORV is available and capable of providing automatic pressure relief. Subsequently, this request was withdrawn by the letter dated March 9, 2004. The enclosed Notice of Partial Withdrawal of Application for Amendments to Facility Operating Licenses has been forwarded to the Office of the Federal Register for publication.
G. Rueger A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's 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 and 50-323
Enclosures:
- 1. Amendment No. 171 to DPR-80
- 2. Amendment No. 172 to DPR-82
- 3. Safety Evaluation
- 4. Notice of Partial Withdrawal cc w/encls: See next page
G. Rueger A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's 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 and 50-323
Enclosures:
- 1. Amendment No. 171 to DPR-80
- 2. Amendment No. 172 to DPR-82
- 3. Safety Evaluation
- 4. Notice of Partial Withdrawal cc w/encls: See next page DISTRIBUTION:
PUBLIC PDIV-2 Reading RidsNrrDlpmPdiv (HBerkow)
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TBoyce TS: ML041960100 NRR-100 PKG: ML041950300
- Safety Evaluation Input ACCESSION NO.: ML041950260 NRR-058 OFFICE PDIV-2 PDIV-2/PM PDIV-2/LA IROB/SC SRXB/SC EEIB/SC OGC PDIV-2/SC NAME GTaylor:mp GShukla EPeyton DTrimble JUhle EMarinos RHoefling SDembek DATE 6/16/04 6/16/04 6/15/04 1/20/04*
4/16/04 10/17/03*
6/23/04 6/29/04 DOCUMENT NAME: C:\\MyFiles\\Copies\\AMD MB6758-59.wpd OFFICIAL RECORD COPY
Diablo Canyon Power Plant, Units 1 and 2 cc:
NRC Resident Inspector Diablo Canyon Power Plant c/o U.S. Nuclear Regulatory Commission P.O. Box 369 Avila Beach, CA 93424 Sierra Club San Lucia Chapter c/o Henriette Groot 1000 Montecito Road Cayucos, CA 93430 Ms. Nancy Culver San Luis Obispo Mothers for Peace P.O. Box 164 Pismo Beach, CA 93448 Chairman San Luis Obispo County Board of Supervisors Room 370 County Government Center San Luis Obispo, CA 93408 Mr. Truman Burns Mr. Robert Kinosian California Public Utilities Commission 505 Van Ness, Room 4102 San Francisco, CA 94102 Diablo Canyon Independent Safety Committee ATTN: Robert R. Wellington, Esq.
Legal Counsel 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 Richard F. Locke, Esq.
Pacific Gas & Electric Company P.O. Box 7442 San Francisco, CA 94120 Mr. David H. Oatley, Vice President and General Manager Diablo Canyon Power Plant P.O. Box 56 Avila Beach, CA 93424 City Editor The Tribune 3825 South Higuera Street P.O. Box 112 San Luis Obispo, CA 93406-0112 Mr. Ed Bailey, Chief Radiologic Health Branch State Department of Health Services P.O. Box 997414 (MS 7610)
Sacramento, CA 95899-7414 Mr. James D. Boyd, Commissioner California Energy Commission 1516 Ninth Street (MS 31)
Sacramento, CA 95814 Mr. James R. Becker, Vice President Diablo Canyon Operations and Station Director Diablo Canyon Power Plant P.O. Box 3 Avila Beach, CA 93424
PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 171 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 September 24, 2002, and its supplements dated November 21, 2003, and March 9, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's 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.
171, 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 30 days from the date of issuance, or within 30 days following upgrade of the automatic actuation circuitry for the Class 1 power operated relief valves for each unit, whichever is later.
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: July 2, 2004
PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 172 License No. DPR-82 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Pacific Gas and Electric Company (the licensee) dated September 24, 2002, and its supplements dated November 21, 2003, and March 9, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's 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-82 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.
172, 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 30 days from the date of issuance, or within 30 days following upgrade of the automatic actuation circuitry for the Class 1 power operated relief valves for each unit, whichever is later.
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: July 2, 2004
ATTACHMENT TO LICENSE AMENDMENT NO. 171 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 172 TO FACILITY OPERATING LICENSE NO. DPR-82 DOCKET NOS. 50-275 AND 50-323 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 3.4-19 3.4-19 3.4-20 3.4-20 3.4-22 3.4-22
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 171 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 172 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323
1.0 INTRODUCTION
By application dated September 24, 2002, and its supplements dated November 21, 2003, and March 9, 2004, Pacific Gas and Electric Company (PG&E), requested changes to the Technical Specifications (TSs) (Appendix A to Facility Operating License Nos. DPR-80 and DPR-82) for the Diablo Canyon Power Plant (DCPP), Units 1 and 2.
The amendments revise TS Section 3.4.11 to credit the automatic actuation of the pressurizer power operated relief valves (PORVs) for mitigating the plant transient of inadvertent actuation of the safety injection (SI) system.
Specifically, the proposed changes would revise:
1.
TS 3.4.11, Condition A, to read, "One or more PORVs inoperable solely due to excessive seat leakage."
2.
TS 3.4.11, Condition B, to read, "One PORV inoperable for reasons other than excessive seat leakage."
3.
TS 3.4.11, Condition E, to read, "Two Class 1 PORVs inoperable for reasons other than excessive seat leakage."
4.
TS 3.4.11 - Two new surveillance requirements (SRs) will be added: SR 3.4.11.4 to read, "Perform a COT on each required Class 1 PORV, excluding actuation," with a frequency of 92 days; and SR 3.4.11.5, "Perform CHANNEL CALIBRATION for each required Class 1 PORV actuation channel," with a frequency of 24 months. A COT is a channel operational test.
Additionally, the license amendment request (LAR) included a request to revise TS 3.4.10, "Pressurizer Safety Valves," to allow pressurizer safety valve (PSV) loop seal temperatures to be less than the lower design temperature during plant heatup and cooldown in Mode 3, and in Mode 4 when any reactor coolant system cold leg temperature is greater than the low temperature overpressure protection arming temperature specified in the pressure temperature limits report, provided at least one Class I PORV is available and capable of providing automatic pressure relief. However, by letter dated March 9, 2004, the licensee withdrew the proposed changes to TS 3.4.10.
The supplemental letters dated November 21, 2003, and March 9, 2004 provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on December 24, 2002 (67 FR 78522).
2.0 REGULATORY EVALUATION
The staff finds that the licensee in Section 5.0 of its submittal identified the applicable regulatory requirements. Equipment malfunctions or operator errors could cause unplanned increases in reactor coolant inventory by inadvertent actuation of the SI system. Depending on the boron concentration and temperature of the injected water and the responses of the automatic control systems, a change of power level may result, and without adequate controls, could lead to fuel damage or overpressure of the reactor coolant system (RCS). Reactor protection and safety systems are actuated to mitigate these events. The staff review covers the sequence of events, the analytical model used for analyses, the values of parameters used in the analytical model, and the results of the transient analyses. The NRCs acceptance criteria are based on General Design Criteria (GDC) 10, 15, and 26 to ensure that the specified acceptable fuel design limits are not exceeded, the reactor coolant pressure boundary will not be breached, and the reliable control of reactivity changes during normal operation, including anticipated operational occurrences. Specific review criteria are contained in Section 15.5.1-2 of the Standard Review Plan (SRP).
Other documents reviewed by the staff in regard to the PORV actuation circuit upgrade, the adequacy of the proposed operator actions, and the associated operator action times, include:
IEEE 279, "Criteria for Protection Systems for Nuclear Power Generating Stations,"
(10 CFR 50.55 a(h)) requirements.
ANSI/ANS 58.8 (1994), "Time Response Design Criteria for Safety-Related Operator Actions," August 23, 1994.
NRC Information Notice 97-78 (IN 97-78), "Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Response Times,"
October 23, 1997.
3.0 TECHNICAL EVALUATION
The staff has reviewed the licensee's technical and regulatory analyses in support of its proposed license amendments which are described in Sections 4.0 and 5.0 of the licensee's submittal. The detailed evaluation below will support the conclusion 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 Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
3.1 Modifications to PORV Actuation Circuit The PORVs are air-operated and controlled by solenoid valves that are energized to open and spring to close with an automatic actuation to provide various safety-related functions. As such, the PORVs are Class I equipment and its actuation circuit should be comprised of Class 1E components and Class 1E power sources. In DCPPs current design, each PORV is powered from a separate vital bus backed up by station batteries. However, in the current design, part of the automatic control circuit for the Class I PORV is Class II. These Class II components are the pressure channel selector switch and Hagan Model 118 bistable comparator module in the control racks to actuate the PORV auxiliary relays. To allow credit to be taken for the Class I PORV, the licensee proposed to upgrade the automatic actuation circuit of the PORV by removing the Class II components and providing actuation of the PORV auxiliary relays directly from the Eagle 21 process protection system (the Eagle 21 partial trip (EPT) card outputs will actuate the PORV actuation and interlock relays). The EPT cards will be reconfigured to support an energize-to-trip function.
Eagle 21 is a new Westinghouse microprocessor based Class 1E digital process protection system which replaces the older analog protection and control process instrumentation system.
The Eagle 21 system is part of the reactor protection system (RPS), which includes the reactor trip functions and the engineered safety features actuation functions. The input signals to the Eagle 21 system include temperature, pressure, level, and flow measurement. This system also accepts analog voltage or current inputs from other nuclear process systems. The protection channel independence is maintained in the same way as the old system, such that a single failure of any one of the redundant channels cannot affect the other channels. Features of the Eagle 21 equipment include the following:
Automatic surveillance testing capability; Self calibration to reduce/eliminate rack drift and simplify calibration procedures; and Self diagnostic capability to reduce trouble shooting time.
The staff found the Eagle 21 system acceptable for its safety-related functions in the staff's safety evaluation (SE) dated May 16, 1990. The staff's evaluation of the licensees submittal found that the upgrade of the automatic actuation circuit meets the IEEE 279 requirements, and thus, ensures the Class 1 PORVs ability to mitigate the consequences of the spurious operation of the SI system at power events. The staff, therefore, found the proposed modifications to the PORV actuation circuit acceptable.
3.2 Inadvertent Actuation of SI System Two separate inadvertent actuations of the SI system at power cases are analyzed to ensure
.that the RCS pressure limits are not exceeded and that the DNBR limits are met. The analyses are discussed in Section 15.2.15 of the licensee's Final Safety Analysis Report (FSAR) Update. The current analysis that evaluates RCS overpressure and pressurizer overfill takes credit for operation of the PSVs to relieve the RCS overpressure condition. No credit is taken for automatic operation of the PORVs since part of the automatic actuation circuitry is not designed to safety grade standards, and therefore the PORVs cannot be used for mitigation of a design basis event.
In the application, the licensee has proposed modification of the automatic actuation circuits of the PORVs to safety grade standards so that the automatic actuation function of the PORVs can be credited for mitigation of an inadvertent actuation of the SI system without challenges to the PSVs (see Section 3.1 of this SE). The licensee has provided the results of its new analyses of the inadvertent actuation of the SI system with automatic actuation of the PORVs for relieving RCS pressure. The event was analyzed using a RETRAN computer code model of DCPP. The computer code has been verified for applicability to DCPP with all the restrictions and conditions of the staffs approval of the code satisfied. The licensee has used conservative assumptions in this analysis to maximize the effects of pressurizer overfill and RCS pressurization. The major assumptions are discussed below:
1.
The initial pressurizer pressure is assumed to be 2190 psia, which is 60 psi lower than nominal value. This lower RCS pressure results in increased SI flow during the transient and maximizes the challenges to the PSVs or PORVs.
2.
Pressurizer heaters remain on during the transient to maximize the pressurizer water volume.
3.
A turbine trip coincident with a reactor trip at the time of event initiation with limited decay heat removal capabilities maximizes the magnitude of the RCS pressure and temperature increase.
4.
Two trains of SI pumps are assumed to provide the maximum flow against RCS pressure. The refueling water storage tank water temperature is assumed to be 35F to maximize the SI flow density and mass injection rate.
Consistent with plant emergency operating procedure (EOP) steps, the analysis assumes operator actions to: (1) ensure the operability of the PORVs by opening their block valves; (2) terminate SI flow; (3) establish RCS letdown flow; and (4) stabilize the pressurizer level and terminate the event. The analysis assumes 11 minutes to make PORVs available and 30 minutes to terminate the event. The staff concludes that these assumed operator action times are appropriate, as described in Sections 3.4 and 3.5 of this SE.
The results of the licensees analysis of an inadvertent actuation of the SI shows that similar to the results in the current analysis, the pressurizer will become solid during the transient.
However, since the PORVs are assumed operable for relieving water, the PSVs are not challenged. Instead, the PORVs will be cycled for water release during the transient.
Depending on the scenario assumed, the PORVs may experience up to 93 cycles during this event. The backup nitrogen accumulators are designed to support 150 PORV cycles which ensures the operation of PORVs for this event. Since the PORVs are capable of providing pressure control during the event, the peak RCS pressure is kept below the PSV setpoint which is well below the maximum allowable limit of 110 percent of the system design pressure.
Therefore, the staff finds the results of the analysis acceptable.
The inadvertent actuation of the SI system at power case that verifies that the DNBR limits are met already assumes normal operation of the pressurizer PORVs and sprays. This assumption conservatively minimizes the RCS pressure during the transient and results in more limiting DNBR values. Therefore, the minimum DNBR case remains bounding and re-analysis is not required.
3.3 Technical Specifications Changes The licensee proposed changes to TS 3.4.11 regarding the PORVs. The proposed changes include the following:
1.
Action A - The condition for this action is currently worded as "one or more PORVs inoperable and capable of being manually cycled." The licensee's proposed new words are "one or more PORVs inoperable solely due to excessive seat leakage." The modified words will reflect the new requirement of ensuring that the automatic function of the PORVs is operable as well as the manual function of the PORVs as previously required. Based on the above discussion, the staff considers that this proposed change is acceptable.
2.
Action B - The condition for this action is currently worded as "one PORV inoperable and not capable of being manually cycled." The licensee's proposed new words are "one PORV inoperable for reasons other than excessive seat leakage." The modified words will reflect the new requirement of ensuring that the automatic function of the PORVs is operable as well as the manual function of the PORVs as previously required.
Based on the above discussion, the staff considers that this proposed change is acceptable.
3.
Action E - The condition for this action is currently worded as "two class I PORVs inoperable and not capable of being manually cycled." The licensee's proposed new words are "two class I PORV inoperable for reasons other than excessive seat leakage." The modified words will reflect the new requirement of ensuring that the automatic function of the PORVs is operable as well as the manual function of the PORVs as previously required. Based on the above discussion, the staff considers that this proposed change is acceptable.
4.
SR 3.4.11.4 and 3.4.11.5 - The licensee proposes to add two SRs to perform a COT on each required Class I PORV, excluding actuation, every 92 days and perform channel calibration for each required Class I PORV actuation channel every 24 months. Based on the above discussion, these proposed SRs would ensure operability for the automatic function of the PORVs and therefore, are acceptable.
3.4 Operator Actions The staff has determined that the proposed operator actions contained in the licensees amendment request are in compliance with IN 97-78. Contained within IN 97-78 are nine criteria which the NRC uses, in part, to evaluate the crediting of operator actions.
1.
Specific operator actions required: As presented in the LAR, the operator actions and action times to be credited for the spurious operation of the SI system at power are:
(1)
Within 9 minutes of event initiation, stop the positive displacement charging pump; (2)
Within 11 minutes of event initiation, make a PORV available - check/open a PORV block valve; (3)
Within 14 minutes of event initiation, identify the SI as unnecessary, and stop all but one centrifugal charging pump. In one additional minute, throttle charging flow; (4)
Within 21 minutes of event initiation, restore instrument air to containment; and (5)
Within 26 minutes of event initiation, establish RCS letdown and terminate the event.
Although this may appear to be a significant number of actions, the current licensing basis (CLB) already includes actions (1), (3), (4), and (5) to terminate the event. Only action (2) is new, and this is a simple action requiring the checking of valve position indications and/or manipulation of one main control room switch. Note that the PORV block valves and their position indication are electrically powered from safety-grade supplies.
The only other difference is that the LAR allows a longer time, 26 minutes, to complete all of these actions, compared to the 16 minutes allowed in the CLB. The action times will be discussed more fully in the next section concerning ANSI/ANS 58.8.
- 2.
Potentially harsh or inhospitable environmental conditions expected: Actions (1) through (5) above are all performed in the main control room.
- 3.
Discussion of ingress/egress paths taken to accomplish functions: This criterion is not applicable, because the operator actions are all performed in the main control room.
- 4.
Procedural guidance for required actions: Actions (1) through (5) above are all procedurally directed by DCPPs EOPs, specifically Procedure E-0, "Reactor Trip or Safety Injection," and Procedure E-1.1, "SI Termination."
- 5.
Operator training necessary to carry out actions including operator qualifications required: The licensee has committed to performing specific training on the importance of performing the required operator actions with respect to the mitigation of the spurious operation of the SI system at power, and ensuring that pressurizer overfill does not occur. Operator performance during a spurious SI at power event on the plant simulator has already been evaluated, including evaluating operator action times and preventing pressurizer overfill. Evaluating operators on this event will continue in the future as a part of licensed operator training. If the operator action times are not conservatively bounded by the times for actions (1) through (5) listed above, then an evaluation will be performed to demonstrate acceptable results with respect to preventing pressurizer overfill, or the operator crews will be retrained on performing the appropriate actions within acceptable time frames.
- 6.
Additional support personnel and/or equipment required to carry out actions: No additional support personnel or equipment are required. Actions (1) through (5) can be readily carried out by the normal level of main control room staffing.
7.
Information required to determine if operator action is required including qualified instrumentation used to diagnose the situation and verify that action has been taken:
All of the instrumentation and indications required by the operators to determine if action is required, to diagnose the situation, and to verify the success of the actions is located in the main control room. The required instrumentation and indications are electrically powered from safety-grade supplies.
- 8.
Ability to recover from credible errors in performing the actions and expected time required to make the recovery: With the exception of action (5), to establish RCS letdown, all of the other actions require the simple manipulation of one or two control room switches, and verification of the action by checking one or two control room indications. Recovery time from any sort of credible error is expected to be very short and well within the 26 minutes allowed to terminate the event.
With regard to action (5), establishing RCS letdown, this action does require approximately 8 switch manipulations and checking approximately 10 indications.
However, establishing RCS letdown is directed by a series of 11 specific and straightforward steps contained in Appendix R of EOP E-1.1. Recovery time from any sort of credible error on an individual step is expected to be very short, and even if more than one error occurs, it is expected that recovery will still occur such that letdown will be established within 26 minutes.
- 9.
Consideration of risk significance of actions: The risk significance of the actions is judged to be low, given that there is a very high degree of confidence that DCPP operators will correctly take the necessary actions in a timely fashion.
3.5 Operator Action Times The staff has determined that the proposed operator action times contained in the LAR are in accordance with the guidance contained in ANSI/ANS 58.8. Specifically, for a plant condition 2 event, such as the spurious operation of the SI system at power, ANSI/ANS 58.8 establishes:
A minimum of 5 minutes for operators to verify automatic responses, observe plant parameters, and plan subsequent actions before any manipulation occurs.
A minimum of 2 minutes to perform each manipulation.
Applying these two criteria to the applicable actions and plant procedures (E-0 and E-1.1), the action times presented in the LAR were found to be reasonable. In addition, the licensee has tested the response times in the simulator of 12 crews of licensed operators to a spurious SI at power event. The results of the time testing were presented in Calculation No. STA-154, and indicated that the LAR proposed action times were bounded by the average crew times plus one standard deviation. This time testing data provided additional evidence that the LAR proposed action times were reasonable.
3.6 Conclusion Based on the above evaluations, the staff finds that the proposed modification of the automatic actuation circuits of the PORVs and the changes to TS 3.4.11 regarding the PORVs to ensure the operability of the automatic actuation function are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change 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 change surveillance requirements. The NRC staff has determined that the amendments involve 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 previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding (67 FR 78522). Accordingly, the amendments meet 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 amendments.
6.0 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 Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: I. Ahmed C. Liang D. Muller Date: July 2, 2004
7590-01-P UNITED STATES NUCLEAR REGULATORY COMMISSION PACIFIC GAS AND ELECTRIC COMPANY DOCKET NOS. 50-275 AND 50-323 NOTICE OF PARTIAL WITHDRAWAL OF APPLICATION FOR AMENDMENTS TO FACILITY OPERATING LICENSES The U.S. Nuclear Regulatory Commission (the Commission) has granted the request of Pacific Gas and Electric Company (the licensee) to partially withdraw its September 24, 2002, application for proposed amendment to Facility Operating License Nos. DPR-80 and DPR-82 for the Diablo Canyon Power Plant, Unit Nos. 1 and 2, located in San Luis Obispo County, California.
A portion of the proposed amendments would have revised Technical Specification 3.4.10, "Pressurizer Safety Valves," to allow pressurizer safety valve (PSV) loop seal temperature to be less than the lower design temperature during plant heatup and cooldown in Mode 3, and in Mode 4 when any reactor coolant system cold leg temperature is greater than the low temperature overpressure protection arming temperature specified in the pressure temperature limits report, provided at least one Class 1 power operated relief valve is available and capable of providing automatic pressure relief. The loop seal revision was intended to allow gradual stabilization of the loop seal temperatures during plant heatups and cooldowns, and avoid having to partially drain the loop seals to establish the minimum design PSV inlet temperature.
The Commission had previously issued a Notice of Consideration of Issuance of Amendment published in the Federal Register on December 24, 2002 (67 FR 78522).
However, by letter dated March 9, 2004, the licensee withdrew that portion of the proposed change.
For further details with respect to this action, see the application for amendment dated September 24, 2002, and its supplement dated November 21, 2003, and the licensee's letter dated March 9, 2004, which withdrew a portion of the application for license amendment.
Documents may be examined, and/or copied for a fee, at the NRCs Public Document Room (PDR), located at One White Flint North, Public File Area O1 F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible electronically from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams/html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC PDR Reference staff by telephone at 1-800-397-4209, or 301-415-4737 or by email to pdr@nrc.gov.
Dated at Rockville, Maryland, this 2nd day of July 2004.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Girjia Shukla, Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation