ML020020206
| ML020020206 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 02/06/2002 |
| From: | Milano P NRC/NRR/DLPM/LPD1 |
| To: | Kansler M Entergy Nuclear Operations |
| Milano, P , NRR/DLPM, 415-1457 | |
| Shared Package | |
| ML020390455 | List: |
| References | |
| TAC MB3355 | |
| Download: ML020020206 (13) | |
Text
February 6, 2002 Mr. Michael Kansler Senior Vice President and Chief Operating Officer Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
SUBJECT:
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 - ISSUANCE OF AMENDMENT RE: ELIMINATION OF REQUIREMENT FOR A POST ACCIDENT SAMPLING SYSTEM (PASS) (TAC NO. MB3355)
Dear Mr. Kansler:
The Commission has issued the enclosed Amendment No. 210 to Facility Operating License No. DPR-64 for the Indian Point Nuclear Generating Unit No. 3. The amendment consists of changes to the Technical Specifications (TSs) in response to your application transmitted by letter dated October 23, 2001.
The amendment deletes TS 5.5.3, Post Accident Sampling and thereby eliminates the requirements to have and maintain the PASS.
A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.
Sincerely,
/RA/
Patrick D. Milano, Sr. Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-286
Enclosures:
- 1. Amendment No. 210 to DPR-64
- 2. Safety Evaluation cc w/encls: See next page
February 6, 2002 Mr. Michael Kansler Senior Vice President and Chief Operating Officer Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
SUBJECT:
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 - ISSUANCE OF AMENDMENT RE: ELIMINATION OF REQUIREMENT FOR A POST ACCIDENT SAMPLING SYSTEM (PASS) (TAC NO. MB3355)
Dear Mr. Kansler:
The Commission has issued the enclosed Amendment No. 210 to Facility Operating License No. DPR-64 for the Indian Point Nuclear Generating Unit No. 3. The amendment consists of changes to the Technical Specifications (TSs) in response to your application transmitted by letter dated October 23, 2001.
The amendment deletes TS 5.5.3, Post Accident Sampling and thereby eliminates the requirements to have and maintain the PASS.
A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.
Sincerely,
/RA/
Patrick D. Milano, Sr. Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-286
Enclosures:
- 1. Amendment No. 210 to DPR-64
- 2. Safety Evaluation cc w/encls: See next page PUBLIC PDI R/F JMunday SLittle PMilano OGC GHill (2) WBeckner JLamb WReckley ACRS BPlatchek, RI Package: ML020390455 Accession Number: ML020020206 Amendment 210: ML020380574 OFFICE PDIII-1\PM PDI-1\PM PDI-1\LA PDI-1\ASC NAME JLamb PMilano SLittle JMunday DATE 2/4/02 2/4/02 2/4/02 2/5/02 Official Record Copy
DATED: February 6, 2002 AMENDMENT NO. 210 TO FACILITY OPERATING LICENSE NO. DPR-64 INDIAN POINT UNIT 3 PUBLIC PDI R/F JMunday SLittle PMilano OGC GHill (2)
WBeckner JLamb WReckley ACRS BPlatchek, RI cc: Plant Service list
ENTERGY NUCLEAR OPERATIONS, INC.
DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 210 License No. DPR-64
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Entergy Nuclear Operations, Inc. (the licensee) dated October 23, 2001, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and 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-64 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 210, 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 the date of its issuance and shall be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Joel Munday, Acting Chief,Section I Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: February 6, 2002
ATTACHMENT TO LICENSE AMENDMENT NO. 210 FACILITY OPERATING LICENSE NO. DPR-64 DOCKET NO. 50-286 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Page Insert Page 5.0-9 5.0-9
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 210 TO FACILITY OPERATING LICENSE NO. DPR-64 ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286
1.0 INTRODUCTION
By letter dated October 23, 2001, Entergy Nuclear Operations, Inc. (the licensee) submitted a request for changes to the Indian Point Nuclear Generating Unit No. 3 (IP3) Technical Specifications (TSs). The requested changes would delete TS 5.5.3, Post Accident Sampling and thereby eliminate the requirements to have and maintain the Post Accident Sampling System (PASS) at IP3.
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 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 TSs of the operating plants and was included in the initial TSs for plants licensed during the 1980s and 90s. Additional expectations regarding PASS capabilities were included in Regulatory Guide 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 power 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 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 reports submitted by the Combustion Engineering Owners Group (CEOG) and the Westinghouse Owners Group (WOG) 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 evaluations (SEs) for the two topical reports, the NRC staff has reviewed the available sources of information for use by decision-makers 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 groups 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 BACKGROUND
In a letter dated May 5, 1999 (as supplemented by letter dated April 14, 2000), the CEOG submitted the Topical Report CE NPSD-1157, Revision 1, Technical Justification for the Elimination of the Post-Accident Sampling System From the Plant Design and Licensing Bases for CEOG Utilities. A similar proposal was submitted on October 26, 1998 (as supplemented by letters dated April 28, 1999, April 10 and May 22, 2000), by the WOG in its topical report WCAP-14986, Post Accident Sampling System Requirements: A Technical Basis. The reports provided evaluations of the information obtained from PASS samples to determine the contribution of the information to plant safety and accident recovery. The reports considered the progression and consequences of core damage accidents and assessed the accident progression with respect to plant abnormal and emergency operating procedures, severe accident management guidance, and emergency plans. The reports provided the owners groups technical justifications for the elimination for the various PASS sampling requirements.
The specific samples and the NRC staffs findings are described in the following evaluation.
The NRC staff prepared this model SE relating to the elimination of requirements on post accident sampling and solicited public comment (65 FR 49271) 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 TSs. Licensees of nuclear power reactors to which this model apply were informed (65 FR 65018) that they could request amendments confirming the applicability of the SE to their reactors and providing the requested plant-specific verifications and commitments.
3.0 EVALUATION The technical evaluations for the elimination of PASS sampling requirements are provided in the SEs dated May 16, 2000, for the CEOG Topical Report CE NPSD-1157 and June 14, 2000, for the WOG Topical Report WCAP-14986. The NRC staffs SEs approving the topical reports are located in the NRCs Agencywide Documents Access and Management System (ADAMS)
(Accession Numbers ML003715250 for CE NPSD-1157 and ML003723268 for WCAP-14986).
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 TSs, supersede the PASS specific requirements imposed by post-TMI confirmatory orders.
As described in its SEs for the topical reports, the NRC staff finds that the following PASS sampling requirements may be eliminated for plants of Combustion Engineering and Westinghouse designs:
- 1. reactor coolant dissolved gases
- 4. reactor coolant pH
- 7. reactor coolant conductivity
- 8. reactor coolant radionuclides
- 9. containment atmosphere hydrogen concentration
- 10. containment oxygen
- 11. containment atmosphere radionuclides
- 12. containment sump pH
- 15. containment sump radionuclides The NRC staff agrees that sampling of radionuclides is not required to support emergency response decision making 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 radionuclides existing post-accident in order to address public concerns and plan for long-term recovery operations. As stated in the SEs for the topical reports, 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. (See item 4.1 under Licensee Verifications and Commitments.) These contingency plans must be available to be used by a licensee during an accident; however, these contingency plans do not have to be carried out in emergency plan drills or exercises. The contingency plans for obtaining samples from the RCS, containment sump, and containment atmosphere may also enable a licensee to derive information on parameters such as hydrogen concentrations in containment and boron concentration and pH of water in the containment sump. The NRC staff considers the sampling of the containment sump to be potentially useful in confirming calculations of pH and boron concentrations and confirming that potentially unaccounted for acid sources have been sufficiently neutralized. The use of the contingency plans for obtaining samples would depend on the plant conditions and the need for information by the decision-makers responsible for responding to the accident.
In addition, the NRC staff considers radionuclide 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 overheating on core exit thermocouples. However, the NRC staff agrees with the topical reports contentions that other indicators of failed fuel, such as letdown radiation monitors (or normal sampling system), can be correlated to the degree of failed fuel. (See item 4.2 under Licensee Verifications and Commitments.)
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. Although this capability may not be needed to support the immediate protective action recommendations during an accident, the information would be useful for decision makers trying to limit the publics ingestion of radioactive materials. (See item 4.3 under Licensee Verifications and Commitments.)
The NRC staff believes that the changes related to the elimination of PASS that are described in the topical reports, related SEs and this proposed change to 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 10 CFR 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 redundant, safety-grade, containment hydrogen concentration monitors are required by 10 CFR 50.44(b)(1), are addressed in NUREG-0737 Item II.F.1 and Regulatory Guide 1.97, and are relied upon to meet the data reporting requirements of 10 CFR Part 50, Appendix E, Section VI.2.a.(i)(4). The NRC staff concludes that during the early phases of an accident, the safety-grade hydrogen monitors provide an adequate capability for monitoring containment hydrogen concentration. 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). As previously mentioned, the licensees contingency plan (see item 4.1) for obtaining highly radioactive samples will include sampling of the containment atmosphere and may, if deemed necessary and practical by the appropriate decision-makers, be used to supplement the safety-related 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, containment sump, and containment atmosphere.
The licensee has verified that it has contingency plans for obtaining and analyzing highly radioactive samples from the RCS, containment sump, and containment atmosphere. The licensee has committed to maintain the contingency plans within its Chemistry Procedures.
The licensee has implemented this commitment.
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 .Ci/ml dose equivalent iodine). This capability may utilize the normal sampling system and/or correlations of sampling or letdown line dose rates to coolant concentrations.
The licensee has verified that it has a capability for classifying fuel damage events at the Alert level threshold. The licensee has committed to maintain the capability for the Alert classification within its Emergency Plan Implementing Procedures (EPIPs). 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 capability to monitor radioactive iodines that have been released to offsite environs.
The licensee has verified that it has the capability to monitor radioactive iodines that have been released to offsite environs. The licensee has committed to maintain the capability for monitoring iodines within its EPIPs. The licensee has implemented this commitment.
The NRC staff finds that reasonable controls for the implementation 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 Nuclear Energy Institute (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.) 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 New York State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement 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 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 previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (66 FR 64293). 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.
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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: J. Lamb Date: February 6, 2002
Indian Point Nuclear Generating Unit No. 3 cc:
Mr. Jerry Yelverton Mr. Harry P. Salmon, Jr.
Chief Executive Officer Director of Oversight Entergy Operations Entergy Nuclear Operations, Inc.
1340 Echelon Parkway 440 Hamilton Avenue Jackson, MS 39213 White Plains, NY 10601 Mr. Robert J. Barrett Mr. James Comiotes Vice President - Operations Director, Nuclear Safety Assurance Entergy Nuclear Operations, Inc. Entergy Nuclear Operations, Inc.
Indian Point Nuclear Generating Unit 3 Indian Point Nuclear Generating Unit 3 295 Broadway, Suite 3 295 Broadway, Suite 3 P. O. Box 308 P.O. Box 308 Buchanan, NY 10511-0308 Buchanan, NY 10511-0308 Mr. Dan Pace Mr. John Donnelly Vice President Engineering Licensing Manager Entergy Nuclear Operations, Inc. Entergy Nuclear Operations, Inc.
440 Hamilton Avenue Indian Point Nuclear Generating Unit 3 White Plains, NY 10601 295 Broadway, Suite 3 P.O. Box 308 Mr. James Knubel Buchanan, NY 10511-0308 Vice President Operations Support Entergy Nuclear Operations, Inc. Mr. John McCann 440 Hamilton Avenue Manager, Licensing and Regulatory Affairs White Plains, NY 10601 Entergy Nuclear Operations, Inc.
Indian Point Nuclear Generating Unit 2 Mr. Joseph DeRoy 295 Broadway, Suite 1 General Manager Operations P. O. Box 249 Entergy Nuclear Operations, Inc. Buchanan, NY 10511-0249 Indian Point Nuclear Generating Unit 3 295 Broadway, Suite 3 Resident Inspectors Office P. O. Box 308 U.S. Nuclear Regulatory Commission Buchanan, NY 10511-0308 295 Broadway, Suite 3 P.O. Box 337 Mr. John Kelly Buchanan, NY 10511-0337 Director - Licensing Entergy Nuclear Operations, Inc. Regional Administrator, Region I 440 Hamilton Avenue U.S. Nuclear Regulatory Commission White Plains, NY 10601 475 Allendale Road King of Prussia, PA 19406 Ms. Charlene Fiason Licensing Mr. John M. Fulton Entergy Nuclear Operations, Inc. Assistant General Counsel 440 Hamilton Avenue Entergy Nuclear Operations, Inc.
White Plains, NY 10601 440 Hamilton Avenue White Plains, NY 10601
Indian Point Nuclear Generating Unit No. 3 cc:
Ms. Stacey Lousteau Mr. Ronald Schwartz Treasury Department SRC Consultant Entergy Services, Inc. 64 Walnut Drive 639 Loyola Avenue Spring Lake Heights, NJ 07762 Mail Stop: L-ENT-15E New Orleans, LA 70113 Mr. Ronald Schwartz SRC Consultant Mr. William M. Flynn, President 64 Walnut Drive New York State Energy, Research, and Spring Lake Heights, NJ 07762 Development Authority Corporate Plaza West Mr. Ronald J. Toole 286 Washington Avenue Extension SRC Consultant Albany, NY 12203-6399 Toole Insight 605 West Horner Street Mr. J. Spath, Program Director Ebensburg, PA 15931 New York State Energy, Research, and Development Authority Mr. Charles W. Hehl Corporate Plaza West SRC Consultant 286 Washington Avenue Extension Charles Hehl, Inc.
Albany, NY 12203-6399 1486 Matthew Lane Pottstown, PA 19465 Mr. Paul Eddy Electric Division Mr. Alex Matthiessen New York State Department Executive Director of Public Service Riverkeeper, Inc.
3 Empire State Plaza, 10th Floor 25 Wing & Wing Albany, NY 12223 Garrison, NY 10524 Mr. Charles Donaldson, Esquire Mr. Paul Leventhal Assistant Attorney General The Nuclear Control Institute New York Department of Law 1000 Connecticut Avenue NW 120 Broadway Suite 410 New York, NY 10271 Washington, DC, 20036 Mayor, Village of Buchanan Mr. Karl Copeland 236 Tate Avenue Pace Environmental Litigation Clinic Buchanan, NY 10511 78 No. Broadway White Plains, NY 10603 Mr. Ray Albanese Executive Chair Jim Riccio Four County Nuclear Safety Committee Greenpeace Westchester County Fire Training Center 702 H Street, NW 4 Dana Road Suite 300 Valhalla, NY 10592 Washington, DC 20001