ML020040143
| ML020040143 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 11/05/2002 |
| From: | Robert Fretz NRC/NRR/DLPM/LPD1 |
| To: | Keiser H Public Service Enterprise Group |
| Fretz, R, NRR/DLPM, 415-1324 | |
| References | |
| TAC MB2815, TAC MB2816 | |
| Download: ML020040143 (17) | |
Text
November 5, 2002 Mr. Harold W. Keiser Chief Nuclear Officer & President PSEG Nuclear LLC - X04 Post Office Box 236 Hancocks Bridge, NJ 08038
SUBJECT:
SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS RE: ELIMINATION OF REQUIREMENTS FOR POST-ACCIDENT SAMPLING (TAC NOS. MB2815 AND MB2816)
Dear Mr. Keiser:
The Commission has issued the enclosed Amendment Nos. 254 and 235 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station, Unit Nos. 1 and 2. These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated August 17, 2001, as supplemented on August 12, 2002.
These amendments delete TS Section 6.8.4.e, Post-Accident Sampling," for Salem Nuclear Generating Station, Unit Nos. 1 and 2, and License Condition 2.C.25, "Post-Accident Sampling," for Unit 2, and thereby eliminate the requirements to have and maintain the post-accident sampling program.
A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Robert J. Fretz, Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311
Enclosures:
- 1. Amendment No. 254 to License No. DPR-70
- 2. Amendment No. 235 to License No. DPR-75
- 3. Safety Evaluation cc w/encls: See next page
- See previous concurrence OFFICE CLIIP/LPM*
PDI-2/PM PDI-2/LA OGC*
PDI-2/SC(A)
NAME WReckley RFretz TLClark AHodgdon REnnis for JAndersen DATE 09/12/01 09/11/02 9/11/02 04/04/02 9/11/02
PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 254 License No. DPR-70 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the PSEG Nuclear LLC, and the Exelon Generation Company, LLC (the licensees) dated August 17, 2001, as supplemented on August 12, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 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-70 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 254, 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 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA REnnis for/
James W. Andersen, Acting Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: November 5, 2002
ATTACHMENT TO LICENSE AMENDMENT NO. 254 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 Replace the following page of the Appendix A, Technical Specifications, with the attached revised page as indicated. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Pages Insert Pages 6-19 6-19
PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 235 License No. DPR-75 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the PSEG Nuclear LLC, and the Exelon Generation Company, LLC (the licensees) dated August 17, 2001, as supplemented on August 12, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 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-75 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 235, 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 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA REnnis for/
James W. Andersen, Acting Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Attachments: Changes to the Technical Specifications and License Date of Issuance: November 5, 2002
ATTACHMENT TO LICENSE AMENDMENT NO. 235 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Replace the following pages of Facility Operating License No. DPR-75 and 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 License, Page 18 License, Page 18 6-19 6-19
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 254 AND 235 TO FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311
1.0 INTRODUCTION
By letter dated August 17, 2001, as supplemented on August 12, 2002, PSEG Nuclear LLC (PSEG or the licensee) submitted a request for changes to the Salem Nuclear Generating Station, Unit Nos. 1 and 2 (Salem), Technical Specifications (TSs). The requested changes would delete TS Section 6.8.4.e, Post-Accident Sampling," for Salem, and License Condition 2.C.25, "Post-Accident Sampling," for Salem Unit No. 2, and thereby eliminate the requirements to have and maintain a post-accident sampling program. The August 12, 2002, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
In the aftermath of the accident at Three Mile Island (TMI), Unit 2, the U.S. 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 Sampling 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 TSs of the operating plants and was included in the initial TSs for plants licensed during the 1980s and 1990s. 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 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 reduce unnecessary regulatory burdens on licensees have prompted the NRC to consider eliminating the requirements for PASS in the TSs and other parts of the licensing bases of operating reactors.
The 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 staffs safety evaluations for the two topical reports, the 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 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 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 October 26, 1998 (as supplemented by letters dated April 28, 1999, April 10 and May 22, 2000), the WOG submitted the topical report WCAP-14986, Post Accident Sampling System Requirements: A Technical Basis. The report provided evaluations of the information obtained from PASS samples to determine the contribution of the information to plant safety and accident recovery. The report 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 report provided the owners group's technical justification for the elimination for the various PASS sampling requirements. The specific samples and the staffs findings are described in the following evaluation.
The NRC staff prepared a model safety evaluation (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 evaluation for the elimination of PASS sampling requirements for Westinghouse nuclear steam supply system plants is provided in the safety evaluation dated June 14, 2000, for the WOG topical report WCAP-14986. The NRC staffs safety evaluation approving the topical report is located in the NRCs Agencywide Documents Access and Management System (ADAMS) (Accession Number ML003723268).
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 the proposed removal of PASS and related administrative controls from the TSs supersedes the PASS specific requirements imposed by post-TMI confirmatory orders.
As described in its safety evaluation for the topical report, the staff finds that the following PASS sampling requirements may be eliminated for plants of 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 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 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 safety evaluation for the topical report, the 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 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 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 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 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 Verifications and Commitments.)
In lieu of the information that would have been obtained from PASS, the 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 Verifications and Commitments.)
The staff believes that the changes related to the elimination of PASS that are described in the topical report, related safety evaluation and this proposed change to the TSs are unlikely to result in a decrease in the effectiveness of the licensees emergency plan. The 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. The licensee should perform the appropriate evaluations and report changes to its emergency plan in accordance with applicable regulations and procedures.
The 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 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 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.
The elimination of PASS requirements involves the deletion of Condition 2.C.25.d in the Unit 2 operating license. The change is included in the licensees application to revise the TS in order to take advantage of the CLIIP. The staff has reviewed the change and agrees that the revisions are necessary due to the removal of the TS section on PASS. The changes do not revise technical requirements beyond those reviewed by the NRC staff in connection with the supporting topical reports or the preparation of the TS improvement incorporated into the CLIIP.
4.0 VERIFICATIONS AND COMMITMENTS As requested by the staff in the notice of availability for this CLIIP topic, 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 committed to develop and maintain contingency plans for obtaining and analyzing highly radioactive samples from the RCS, containment sump, and containment atmosphere. By letter dated August 12, 2002, the licensee stated that it has developed procedures to obtain and analyze highly radioactive samples of reactor coolant, containment sump and containment atmosphere. PSEG also stated in its August 12, 2002, letter that this commitment would be fully implemented within 90 days following issuance of this 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
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 the capability for classifying fuel damage events at the Alert level threshold. The licensee has committed to maintain the capability for the Alert classification in plant implementing procedures. PSEG stated in its August 12, 2002, letter that this commitment would be fully implemented within 90 days following issuance of this amendment.
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 in plant implementing procedures. PSEG stated in its August 12, 2002, letter that this commitment would be fully implemented within 90 days following issuance of this amendment.
The NRC staff finds that the licensees administrative processes, including its commitment management program, provide reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitments. 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 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). The commitments should be controlled in accordance with the industry guidance or comparable criteria employed by a specific licensee. The 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 Commissions regulations, the New Jersey State official was notified of the proposed issuance of the amendments. In a letter dated January 23, 2002, the State official provided the following comments:
We noted, that in Attachment 1 to the August 17th letter, PSEG commits to develop and maintain contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, containment sump and containment atmosphere. In order to better understand the options available to PSEG to meet this commitment, we held a conference call with PSEG Chemistry and Licensing personnel on January 22, 2002. Based on this conversation, it is our understanding that post accident sampling will be performed, but sampling locations and potential dose rates to sampling personnel are not yet known.
It is our expectation, that following an accident, the radionuclide mix be characterized within a reasonable length of time, say within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, by whatever means available. This mix must be defined prior to consideration of protective action recommendations being developed for ingestion pathway issues. We believe that the federal Department of Energy with their role in post accident radiation surveys and assessment would have similar needs.
Our concern is that the NRC has established no acceptance criteria associated with the development of these contingency plans. Our discussion with PSEG personnel revealed that for other plants the time to define the mix of radionuclides available for release was as long as 7 days. Clearly, this would not meet our needs.
The January 23, 2002, letter from the State of New Jersey, Department of Environmental Protection (NJ-DEP) reiterated similar comments provided by NJ-DEP in a January 10, 2000, letter responding to the NRC staffs solicitation of public comments during the review of topical report WCAP-14986:
We understand that a new scheme for classifying emergency conditions has been developed and some power plants are using them. This scheme does not rely on the results of analysis of samples of plant fluids. With the elimination of the need for determining air samples for core damage assessment, they [the WOG, et al.] believe that the radionuclide determination in PASS is completely eliminated.
While this may be more effective in quickly determining emergency classification levels, it may not be as effective in determining short and longer-term protective actions. Off-site officials need to have a mechanism to know not only inferred source term, but also actual source term over time, and the stability and potential hazards that remain in the coolant.
Both magnitude and mix of radionuclides is important once the reactor is considered stable and recoverable. As part of the licensees plant recovery determination, a series of PASS samples should be taken. The inventory derived from these samples would support the data collected from other areas and could provided a clearer understanding of the release fraction. This inventory and source term data would enhance the confidence of public health officials in their evaluation of subsequent protective actions in the ingestion phase.
As a public health official and offsite responder whose role as a decision-maker in making protective action recommendations for the governor, the elimination of the PASS could have a detrimental effect on public actions. An assurance from the licensee of the total inventory of radionuclides for potential release needs to be understood. This understanding of what potential hazards are contained in the reactor has to become important as some phase of an emergency and is certainly necessary for public assurance.
Therefore, we recommend that the PASS not be completely eliminated and in fact be used as a tool to understand plant stability and more accurately portray core inventory and release fractions.
In addition, NJ-DEP officials further clarified in a telephone conference call with the NRC staff on February 14, 2002, that its primary concern was that adequate measures would be available to obtain radionuclide samples once the PASS was removed from service, and that NJ-DEP needed to understand what alternatives measures would be available for sampling.
Subsequent discussions were held between NJ-DEP and the NRC staff on July 31 and November 4, 2002. During the November 4, 2002, conference call, the NJ-DEP representative also expressed a concern about efforts currently underway by the staff that would allow licensees to change the safety classification of the hydrogen monitors from safety-related to non-safety related. This issue was raised in the context that the NRC could be removing equipment that may be important to respond to potential security-related scenarios in the post-September 11, 2001, environment. The staff noted that it would be carefully considering the equipment and instrumentation needed to respond to scenarios associated with possible changes to the design-basis threat resulting from the NRCs comprehensive review of security and safeguards programs. If the ongoing research and security review recommends any changes to the facility or plant equipment, the NRC will take appropriate action.
The NRC staff understands NJ-DEPs responsibility to provide appropriate protective action recommendations (PARs) to the governor and other officials, and that all organizations involved in responding to an accident would need to have the best information available prior to making decisions affecting public health and safety. The staff further recognizes that the ultimate decision to order the evacuation of citizens surrounding the plant or to recommend the prophylactic use of KI tablets must be based on timely and accurate information, and that the public needs to have confidence in the decisions made by public officials. By letter dated September 30, 2002, the staff provided NJ-DEP a summary of its technical basis for approving PSEGs request for changes to the Salem TSs. A copy of this letter is available for inspection at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland, or from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov/reading-rm.html (the Public Electronic Reading Room) by referencing ADAMS Accession No. ML022470413.
6.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to the 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 (66 FR 55022). 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: November 5, 2002
PSEG Nuclear LLC Salem Nuclear Generating Station, Unit Nos. 1 and 2 cc:
Mr. Elbert C. Simpson Senior Vice President &
Chief Administrative Officer PSEG Nuclear - N19 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Mark B. Bezilla Vice President - Technical Support PSEG Nuclear - N10 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. David F. Garchow Vice President - Operations PSEG Nuclear - X10 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Gabor Salamon Manager - Nuclear Safety and Licensing PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Jeffrie J. Keenan, Esquire PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Ms. R. A. Kankus Joint Owner Affairs PECO Energy Company Nuclear Group Headquarters KSA1-E 200 Exelon Way Kennett Square, PA 19348 Lower Alloways Creek Township c/o Mary O. Henderson, Clerk Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 Dr. Jill Lipoti, Asst. Director Radiation Protection Programs NJ Department of Environmental Protection and Energy CN 415 Trenton, NJ 08625-0415 Richard Hartung Electric Service Evaluation Board of Regulatory Commissioners 2 Gateway Center, Tenth Floor Newark, NJ 07102 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Senior Resident Inspector Salem Nuclear Generating Station U.S. Nuclear Regulatory Commission Drawer 0509 Hancocks Bridge, NJ 08038 Mr. Norm Cohen The UNPLUG Salem Campaign 21 Barr Avenue Linwood, NJ 08221