ML051320220
| ML051320220 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/12/2005 |
| From: | Olshan L NRC/NRR/DLPM/LPD2 |
| To: | Rosalyn Jones Duke Energy Corp |
| Olshan L N, NRR/DLPM, 415-1419 | |
| Shared Package | |
| ML051450443 | List: |
| References | |
| TAC MC6376, TAC MC6377, TAC MC6378 | |
| Download: ML051320220 (16) | |
Text
July 12, 2005 Mr. Ronald A. Jones Vice President, Oconee Site Duke Energy Corporation 7800 Rochester Highway Seneca, SC 29672
SUBJECT:
OCONEE NUCLEAR STATION, UNITS 1, 2 AND 3 RE: ISSUANCE OF AMENDMENTS ELIMINATING REQUIREMENTS FOR POST ACCIDENT SAMPLING (TAC NOS. MC6376, MC6377, AND MC6378)
Dear Mr. Jones:
The Nuclear Regulatory Commission has issued the enclosed Amendment Nos. 346, 348, and 347 to Renewed Facility Operating Licenses DPR-38, DPR-47, and DPR-55, respectively, for the Oconee Nuclear Station, Units 1, 2, and 3. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated March 14, 2005.
The amendments delete TS 5.5.4, Post Accident Sampling.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Leonard N. Olshan, Sr. Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287
Enclosures:
- 1. Amendment No. 346 to DPR-38
- 2. Amendment No. 348 to DPR-47
- 3. Amendment No. 347 to DPR-55
- 4. Safety Evaluation cc w/encls: See next page
July 12, 2005 Mr. Ronald A. Jones Vice President, Oconee Site Duke Energy Corporation 7800 Rochester Highway Seneca, SC 29672
SUBJECT:
OCONEE NUCLEAR STATION, UNITS 1, 2 AND 3 RE: ISSUANCE OF AMENDMENTS ELIMINATING REQUIREMENTS FOR POST ACCIDENT SAMPLING (TAC NOS. MC6376, MC6377, AND MC6378)
Dear Mr. Jones:
The Nuclear Regulatory Commission has issued the enclosed Amendment Nos. 346, 348, and 347 to Renewed Facility Operating Licenses DPR-38, DPR-47, and DPR-55, respectively, for the Oconee Nuclear Station, Units 1, 2, and 3. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated March 14, 2005.
The amendments delete TS 5.5.4, Post Accident Sampling.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Leonard N. Olshan, Sr. Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287
Enclosures:
- 1. Amendment No. 346 to DPR-38
- 2. Amendment No. 348 to DPR-47
- 3. Amendment No. 347 to DPR-55
- 4. Safety Evaluation cc w/encls: See next page DISTRIBUTION: See next page Package Number: ML051450443 NRR-058 License Amendment Number: ML051320220 Tech Spec Number: ML OFFICE CLIIP LPM PDII-1/PM PDII-1/LA PDII-1/SC NAME WReckley LOlshan CHawes EMarinos DATE 5/ 6/ 05 6/ 6/ 05 6/ 6/ 05 6/ 6/ 05 OFFICIAL RECORD COPY
SUBJECT:
OCONEE NUCLEAR STATION, UNITS 1, 2 AND 3 - ISSUANCE OF AMENDMENTS RE: (MC6376, MC6377, AND MC6378)
Date: July 12, 2005 DISTRIBUTION:
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DUKE ENERGY CORPORATION DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 346 Renewed License No. DPR-38
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility) Renewed Facility Operating License No. DPR-38 filed by the Duke Energy Corporation (the licensee) dated March 14, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-38 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 346, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 180 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Evangelos C. Marinos, Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: July 12, 2005
DUKE ENERGY CORPORATION DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 348 Renewed License No. DPR-47
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility) Renewed Facility Operating License No. DPR-47 filed by the Duke Energy Corporation (the licensee) dated March 14, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-47 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 348, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 180 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Evangelos C. Marinos, Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: July 12, 2005
DUKE ENERGY CORPORATION DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 347 Renewed License No. DPR-55
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility) Renewed Facility Operating License No. DPR-55 filed by the Duke Energy Corporation (the licensee) dated March 14, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-55 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 347, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 180 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Evangelos C. Marinos, Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: July 12, 2005
ATTACHMENT TO LICENSE AMENDMENT NO. 346 RENEWED FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND TO LICENSE AMENDMENT NO. 348 RENEWED FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO. 347 RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 5.0-9 5.0-9
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 346 TO RENEWED FACILITY OPERATING LICENSE DPR-38 AMENDMENT NO. 348 TO RENEWED FACILITY OPERATING LICENSE DPR-47 AND AMENDMENT NO. 347 TO RENEWED FACILITY OPERATING LICENSE DPR-55 DUKE ENERGY CORPORATION OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287
1.0 INTRODUCTION
By letter dated March 14, 2005 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML050810436), Duke Energy Corporation (the licensee) submitted a request for changes to the Oconee Nuclear Station, Units 1, 2, and 3, Technical Specifications (TS). The requested changes would delete TS 5.5.4, Post Accident Sampling.
In the aftermath of the accident at Three Mile Island (TMI), Unit 2, the Nuclear Regulatory Commission (NRC) imposed requirements on licensees for commercial nuclear power plants to install and maintain the capability to obtain and analyze post-accident samples of the reactor coolant and containment atmosphere. The desired capabilities of the post-accident aampling System (PASS) were described in NUREG-0737, Clarification of TMI Action Plan Requirements. The NRC issued orders to licensees with plants operating at the time of the TMI accident to confirm the installation of PASS capabilities (generally as they had been described in NUREG-0737). A requirement for PASS and related administrative controls was added to the TS of the operating plants and was included in the initial TS for plants licensed during the 1980s and 90s. Additional expectations regarding PASS capabilities were included in Regulatory Guide (RG) 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident.
Significant improvements have been achieved since the TMI accident in the areas of understanding risks associated with nuclear plant operations and developing better strategies for managing the response to potentially severe accidents at nuclear plants. Recent insights about plant risks and alternate severe accident assessment tools have led the NRC staff to conclude that some TMI action plan items can be revised without reducing the ability of licensees to respond to severe accidents. The NRCs efforts to oversee the risks associated with nuclear technology more effectively and to eliminate undue regulatory costs to licensees and the public have prompted the NRC to consider eliminating the requirements for PASS in TS and other parts of the licensing bases of operating reactors.
The NRC staff has completed its review of the topical report submitted by the Babcock and Wilcox (B&W) Owners Group (BWOG) that proposed the elimination of PASS. The justifications for the proposed elimination of PASS requirements center on evaluations of the various radiological and chemical sampling and their potential usefulness in responding to a severe reactor accident or making decisions regarding actions to protect the public from possible releases of radioactive materials. As explained in more detail in the NRC staffs safety evaluation (SE) for the topical report, the NRC staff has reviewed the available sources of information for use by decisionmakers in developing protective action recommendations and assessing core damage. Based on this review, the NRC staff found that the information provided by PASS is either unnecessary or is effectively provided by other indications of process parameters or measurement of radiation levels. The NRC staff agrees, therefore, with the owners group that licensees can remove the TS requirements for PASS, revise (as necessary) other elements of the licensing bases, and pursue possible design changes to alter or remove existing PASS equipment.
2.0 REGULATORY EVALUATION
In its letter dated June 25, 2001, the BWOG submitted for the NRC staff's review topical report BAW-2387, Justification for the Elimination of the Post-Accident Sampling System (PASS) from the Licensing Basis of Babcock and Wilcox-Designed Plants. The NRC staff's SE for the BWOG topical report is dated November 14, 2002 (ADAMS Accession No. ML022560119).
The BWOG proposed elimination of the PASS requirements from the standard TSs by submitting change traveler TSTF-442.
The ways in which the requirements and recommendations for PASS were incorporated into the licensing bases of commercial nuclear power plants varied as a function of when plants were licensed. Plants that were operating at the time of the TMI accident are likely to have been the subject of confirmatory orders that imposed the PASS functions described in NUREG-0737 as obligations. The issuance of plant-specific amendments to adopt this change, which would remove PASS and related administrative controls from the TS, would also supersede the PASS-specific requirements imposed by post-TMI confirmatory orders.
The NRC staff prepared this model SE relating to the elimination of requirements on post-accident sampling for B&W plants and solicited public comments in the Federal Register on March 3, 2003 (68 FR 10052) in accordance with the consolidated line item improvement process (CLIIP). The use of the CLIIP in this matter is intended to help the NRC to efficiently process amendments that propose to remove the PASS requirements from the TS. Licensees of nuclear power reactors to which this model apply were informed by a notice in the Federal Register on May 13, 2003 (68 FR 25664) that they could request amendments conforming to the model and, in such requests, should confirm the applicability of the SE to their reactors and provide the requested plant-specific verifications and commitments.
3.0 TECHNICAL EVALUATION
The technical evaluations for the elimination of PASS sampling requirements are provided in the SE dated November 14, 2002, for BWOG topical report BAW-2387. As described in its SE for the topical report, the NRC staff finds that the post-accident sampling requirements for the following may be eliminated for B&W plants:
- 1. Reactor coolant dissolved gases.
- 5. Reactor coolant pH.
- 6. Reactor coolant boron.
- 7. Reactor coolant conductivity.
- 8. Radionuclides in the reactor coolant.
- 9. Containment atmosphere hydrogen.
- 10. Containment atmosphere oxygen.
- 11. Radionuclides in the containment atmosphere.
- 12. Radionuclides in the containment sump.
- 13. Containment sump pH.
PASS sampling of the above 15 parameters is specified in NUREG-0737 and RG 1.97. The sampling of the parameters are either not required to manage an accident and recover plant conditions, or not necessary due to redundancy in sampling capabilities. Based upon the detailed justifications provided in topical report BAW-2387 and its associated SE of November 14, 2002, the NRC staff concludes that the proposal to eliminate PASS sampling of the above parameters is acceptable. The licensee noted in its application that the PASS may continue to be used to collect boron samples under some plant conditions. As stated in the NRC staffs SE approving BAW-2387, the removal of PASS-related requirements from the TS does not preclude licensees from keeping the PASS in the plant and using the system as long as it does not adversely affect safety-related systems.
The NRC staff concludes that sampling of radionuclides is not required to support emergency response decisionmaking during the initial phases of an accident because the information provided by PASS is either unnecessary or is effectively provided by other indications of process parameters or measurement of radiation levels. Therefore, it is not necessary to have dedicated equipment to obtain this sample in a prompt manner.
The NRC staff does, however, believe that there could be significant benefits to having information about the radioisotopes existing post-accident in order to address public concerns and plan for long-term recovery operations. As stated in the SE for the topical report, the NRC staff has found that licensees could satisfy this function by developing contingency plans to describe existing sampling capabilities and what actions (e.g., assembling temporary shielding) may be necessary to obtain and analyze highly radioactive samples from the reactor coolant system (RCS), containment sump, and containment atmosphere. The use of the contingency plans for obtaining samples would depend on the plant conditions and the need for information by the decisionmakers responsible for responding to the accident (see Section 4.0 below).
In addition, the NRC staff considers radioisotope sampling information to be useful in classifying certain types of events (such as a reactivity excursion or mechanical damage) that could cause fuel damage without having an indication of a loss of reactor coolant inventory.
However, the NRC staff agrees with the topical reports contentions that other indicators of failed fuel, such as radiation monitors, can be correlated to the degree of failed fuel.
In lieu of the information that would have been obtained from PASS, the NRC staff believes that licensees should maintain or develop the capability to monitor radioactive iodines that have been released to offsite environs. This information would be useful for decisionmakers trying to assess a release of and limit the publics exposure to radioactive materials.
The NRC staff believes that the changes related to the elimination of PASS that are described in the topical report, related SE, and this proposed change to the TS are unlikely to result in a decrease in the effectiveness of a licensees emergency plan. Each licensee, however, must evaluate possible changes to its emergency plan in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.54(q) to determine if the change decreases the effectiveness of its site-specific plan. Evaluations and reporting of changes to emergency plans should be performed in accordance with applicable regulations and procedures.
The NRC staff notes that containment hydrogen concentration monitors are required by 10 CFR 50.44 and are relied upon to meet the data reporting requirements of 10 CFR Part 50, Appendix E, Section VI.2.a.(ii)(3). The NRC staff concludes that these hydrogen monitors provide an adequate capability for monitoring containment hydrogen concentration during the early phases of an accident. The NRC staff sees value in maintaining the capability to obtain grab samples for complementing the information from the hydrogen monitors in the long term (i.e., by confirming the indications from the monitors and providing hydrogen measurements for concentrations outside the range of the monitors). The licensees contingency plan for obtaining highly radioactive samples will include sampling of the containment atmosphere and may, if deemed necessary and practical by the appropriate decisionmakers, be used to supplement the hydrogen monitors.
4.0 VERIFICATIONS AND COMMITMENTS As requested by the NRC staff in the notice of availability for this TS improvement, the licensee has addressed the following plant-specific verifications and commitments.
4.1 Each licensee should verify that it has, and make a regulatory commitment to maintain (or make a regulatory commitment to develop and maintain), contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, suppression pool, and containment atmosphere.
The licensee has made a regulatory commitment to develop and maintain contingency plans for obtaining and analyzing highly radioactive samples from the RCS, containment sump, and containment atmosphere. The contingency plans will be contained within appropriate plant procedures. The licensee will implement this regulatory commitment as part of the implementation of this license amendment.
4.2 Each licensee should verify that it has, and make a regulatory commitment to maintain (or make a regulatory commitment to develop and maintain), a capability for classifying fuel damage events at the Alert level threshold (typically this is 300 FCi/ml dose equivalent iodine). This capability may utilize the normal sampling system and/or correlations of radiation readings to radioisotope concentrations in the reactor coolant.
The licensee has established the capability for classifying fuel damage events at the Alert level threshold. The licensee has committed to maintain the capability for the Alert classification within the emergency plan and applicable implementing procedures. The licensee has implemented this commitment.
4.3 Each licensee should verify that it has, and make a regulatory commitment to maintain (or make a regulatory commitment to develop and maintain), the offsite capability to monitor radioactive iodines.
The licensee has verified that it has the ability to assess radioactive iodines released to offsite environs. The licensee has committed to maintain the capability for monitoring iodines within the emergency plan and applicable implementing procedures. The licensee has implemented this commitment.
The NRC staff finds that reasonable controls for the implementation of, and for subsequent evaluation of, proposed changes pertaining to the above regulatory commitments are provided by the licensees administrative processes, including its commitment management program.
Should the licensee choose to incorporate a regulatory commitment into the emergency plan, final safety analysis report, or other document with established regulatory controls, the associated regulations would define the appropriate change control and reporting requirements.
The NRC staff has determined that the commitments do not warrant the creation of regulatory requirements, which would require prior NRC approval of subsequent changes. The NRC staff has agreed that NEI 99-04, Revision 0, "Guidelines for Managing NRC Commitment Changes,"
provides reasonable guidance for the control of regulatory commitments made to the NRC staff.
(See Regulatory Issue Summary 2000-17, Managing Regulatory Commitments Made by Power Reactor Licensees to the NRC Staff, dated September 21, 2000 [ADAMS Accession No. ML003741774].) The commitments should be controlled in accordance with the industry guidance or comparable criteria employed by a specific licensee. The NRC staff may choose to verify the implementation and maintenance of these commitments in a future inspection or audit.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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 (70 FR 24649). 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.
7.0 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 Contributor: W. Reckley Date: July 12, 2005