The U.S. Nuclear Regulatory Commission (the Commission) is considering issuance of an amendment to Facility Operating License No. DPR-59, issued to Entergy Nuclear Operations, Inc. (ENO or the licensee) for operation of the James A. FitzPatrick Nuclear Power Plant, (FitzPatrick), located in Oswego County, New York.
The initial notice of consideration of issuance of amendment to facility operating license and opportunity for hearing was originally published in the Federal Register (64 FR 60854) on November 8, 1999, and corrected in the Federal Register (64 FR 69574) on December 13, 1999. The information included in the supplemental letters indicates the original notice, that included 13 proposed beyond-scope issues (BSls) to the improved Technical Specifications (ITS) conversion, needs to be expanded and revised to include a total of thirty one BSls and requires re-notice in the Federal Register. This notice supercedes the previous notice.
The proposed amendment, requested by the Power Authority of the State of New York, the former licensee, in a letter dated March 31, 1999, as supplemented by letters dated May 20, June 1, July 14, October 14, 1999, February 11, April 4, April 13, June 30, July 31, September 12, September 13, and October 23, 2000, represents a full conversion from the current Technical Specifications (CTS) to a set of ITS based on NUREG-1433, “Standard Technical Specifications (STS) for General Electric Plants, BWR/4” Revision 1, dated April 1995. On November 21, 2000, the Power Authority of the State of New York's (PASNY's) ownership interest in FitzPatrick was transferred to Entergy Nuclear FitzPatrick, LLC, to possess and use FitzPatrick and to Entergy Nuclear Operations (ENO), Inc. to possess, use and operate FitzPatrick. By letter dated January 26, 2001, ENO requested that the NRC continue to review and act on all requests before the Commission which had been submitted by PASNY before the transfer. ENO has supplemented the original application with supplements by letter dated February 7, February 20, May 31 and August 6, 2001. NUREG-1433 has been developed by the Commission's staff through working groups composed of both NRC staff members and industry representatives, and has been endorsed by the staff as part of an industry-wide initiative to standardize and improve the Technical Specifications (TSs) for nuclear power plants. As part of this submittal, the licensee has applied the criteria contained in the Commission's “Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors (Final Policy Statement),” published in the Federal Register on July 22, 1993 (58 FR 39132), to the CTS and using NUREG-1433 as a basis, proposed an ITS for FitzPatrick. The criteria in the Final Policy Statement was subsequently added to 10 CFR 50.36, “Technical Specifications,” in a rule change that was published in the Federal Register on July 19, 1995 (60 FR 36953) and became effective on August 18, 1995.
The licensee has categorized the proposed changes to the CTS into four general groupings. These groupings are characterized as administrative changes, relocated changes, more restrictive changes and less restrictive changes.
Administrative changes are those that involve restructuring, renumbering, rewording interpretation and complex rearranging of requirements and other Start Printed Page 59496changes not affecting technical content or substantially revising an operating requirement. The reformatting, renumbering and rewording process reflects the attributes of NUREG-1433 and does not involve technical changes to the CTS. The proposed changes include: (a) Providing the appropriate numbers, etc., for NUREG-1433 bracketed information (information that must be supplied on a plant-specific basis, and which may change from plant to plant), (b) identifying plant-specific wording for system names, etc., and (c) changing NUREG-1433 section wording to conform to existing licensee practices. Such changes are administrative in nature and do not impact initiators of analyzed events or assumed mitigation of accident or transient events.
Relocated changes are those involving relocation of requirements and surveillances for structures, systems, components, or variables that do not meet the criteria for inclusion in TSs. Relocated changes are those CTS requirements that do not satisfy or fall within any of the four criteria specified in the 10 CFR 50.36(c)(2)(ii) and may be relocated to appropriate licensee-controlled documents.
The licensee's application of the screening criteria is described in the attachment of the licensee's March 31, 1999, submittal, which is entitled, “Application of NRC Selection Criteria to James A. FitzPatrick Nuclear Power Plant Technical Specifications” (Split Report) in Volume 1 of the submittal. The affected structures, systems, components or variables are not assumed to be initiators of analyzed events and are not assumed to mitigate accident or transient events. The requirements and surveillances for these affected structures, systems, components, or variables will be relocated from the TSs to administratively controlled documents such as the quality assurance program, the final safety analysis report (FSAR), the ITS BASES, the Technical Requirements Manual (TRM) that is incorporated by reference in the FSAR, the Core Operating Limits Report (COLR), the Offsite Dose Calculation Manual (ODCM), the Inservice Testing (IST) Program, or other licensee-controlled documents. Changes made to these documents will be made pursuant to 10 CFR 50.59 or other appropriate control mechanisms, and may be made without prior NRC review and approval. In addition the affected structures, systems, components, or variables are addressed in existing surveillance procedures that are also subject to 10 CFR 50.59. These proposed changes will not impose or eliminate any requirements.
More restrictive changes are those involving more stringent requirements compared to the CTS for operation of the facility. These more stringent requirements do not result in operation that will alter assumptions relative to the mitigation of an accident or transient event. The more restrictive requirements will not alter the operation of process variables, structures, systems, and components described in the safety analyses. For each requirement in the STS that is more restrictive than the CTS that the licensee proposes to adopt in the ITS, the licensee has provided an explanation as to why it has concluded that adopting the more restrictive requirement is desirable to ensure safe operation of the facility because of specific design features of the plant.
Less restrictive changes are those where CTS requirements are relaxed or eliminated, or new plant operational flexibility is provided. The more significant “less restrictive” requirements are justified on a case-by-case basis. When requirements have been shown to provide little or no safety benefit, their removal from the TS may be appropriate. In most cases, relaxations previously granted to individual plants on a plant-specific basis were the result of (a) generic NRC actions, (b) new NRC staff positions that have evolved from technological advancements and operating experience, or (c) resolution of the Owners Groups' comments on the Improved standard Technical Specifications. Generic relaxations contained in NUREG-1433 were reviewed by the staff and found to be acceptable because they are consistent with current licensing practices and NRC regulations. The licensee's design is being reviewed to determine if the specific design basis and licensing basis are consistent with the technical basis for the model requirements in NUREG-1433, thus providing a basis for the ITS, or if relaxation of the requirements in the CTS is warranted based on the justification provided by the licensee.
These administrative, relocated, more restrictive, and less restrictive changes to the requirements of the CTS do not result in operations that will alter assumptions relative to mitigation of an analyzed accident or transient event.
In addition to the proposed changes solely involving the conversion, there are also changes proposed that are different to the requirements in both the CTS and the Standard Technical Specifications (STS) NUREG-1433. These proposed beyond-scope issues to the its conversion are as follows:
1. ITS 184.108.40.206, Reactor Protection System (RPS) Instrumentation Function 5, reactor scram on main steam isolation valve (MSIV) closure. The trip setting valve was changed from less than or equal to 10 percent (in the CTS) to less than or equal to 14 percent in the ITS.
2. ITS 220.127.116.11 changed the CTS allowable values for turbine stop valve closure, the turbine control valve fast closure and the EHC oil pressure low functions setpoints based on recent setpoint calculations.
3. ITS 18.104.22.168, Suppression Pool Water Temperature is modified by footnote (c), which states : “A channel requires 15 to 16 RTDs to be OPERABLE.” This results in a CTS change and a deviation from the STS.
4. ITS 22.214.171.124 changes the CTS and ISTS channel configuration from 2 channels per trip system to 4 channels in one trip system.
5. ITS 126.96.36.199 changed the CTS allowable values for CS pump flow, LPCI pressure, LPCI pump flow, HPCI vessel water level high and HPCI pump discharge flow low based on recent setpoint calculations.
6. ITS 188.8.131.52, Automatic Depressurization System (ADS) initiation timer and the containment Spray (CS) and Low-Pressure Coolant Injection (LPCI) pump start timer values were changed from the CTS and the ISTS and tolerances relaxed to allow the extension of calibration Frequency to 24 months in the ITS.
7. ITS 184.108.40.206 changed CTS Table 3.2-2 Item 9, Reactor Low Pressure, LPCI and Core Spray Injection Valve Open Permissive of >450 psig to >410 psig in ITS Table 220.127.116.11-1 Functions 1.c and 2.c.
8. ITS 18.104.22.168 changed CTS Table 3.3-2, Item 5, Reactor Low Level Containment spray Interlock trip level setting of >~0.0 inch to >~1.0 inch in ITS Table 22.214.171.124-1.
9. ITS 126.96.36.199 changed the trip setpoint Allowable Values in CTS Table 3.2-2 for the Core Spray Pump Start Timer (item 11), the RHR LPCI Pump Start Timer (item 12, and the Auto Blowdown Timer (item 13) in ITS Table 188.8.131.52-1 Functions 1.d, 2.f, 4.b and 5.b to reflect values corresponding to a 6-month to 24-month reduction in calibration Frequency.
10. ITS 184.108.40.206 changed the trip setpoint Allowable Values in CTS Table 3.2-1 for the suppression Chamber High Level (item 13) in ITS Table 220.127.116.11-1 Function 3.e to 14.5 inches which is <~6 inches above normal level.
11. ITS 18.104.22.168 changed the CTS Table 3.2-2 trip level setting for Item 24, Reactor Low-pressure from 285 to 335 Start Printed Page 59497psig to >~300 psig in ITS Table 22.214.171.124 Function 2.d.
12. ITS 126.96.36.199 changed the Allowable Values in CTS Table 3.2-1 for the HPCI Turbine steam Line High Flow to reflect values corresponding to 160 to 161 inches of water differential pressure (dp) in ITS TABLE 188.8.131.52-1 Function 3.a.
13. ITS 184.108.40.206 changed the trip setpoint Allowable Value “HPCI/Reactor Core Isolation cooling (RCIC) Steam Line Low Pressure” in ITS Table 220.127.116.11-1 Function 3.b and 4.b to reflect values corresponding to >60 and <~90 for HPCI and >61 and <~90 for RCIC.
14. ITS 18.104.22.168 changed the CTS allowable values of setpoint temperatures for the RWCU, HPCI, and RCIC.
15. ITS 22.214.171.124 changed the CTS allowable values for the setpoints for main steam line flow high, main steam tunnel area temperature high, HPCI steam line flow high, HPCI turbine exhaust diaphragm pressure high, HPCI steam line penetration (drywell entrance) area temperature high, HPCI steam line torus room area temperature high, HPCI equipment area temperature high, RHR heat exchanger A area temperature high, reactor building (RB) southwest area of elevation 272 feet temperature high, RCIC steam line flow high, RCIC steam supply line pressure low, RCIC turbine exhaust diaphragm pressure high RCIC steam supply line pressure low, RCIC turbine exhaust diaphragm pressure high, RCIC steam line steam line penetration (drywell entrance) area temperature high, RCIC steam line torus room area temperature high, RCIC equipment area temperature high, RWCU suction line penetration area temperature high, RWCU heat exchanger room area temperature high, RWCU pump area temperature high (Pumps A and B), and SDC reactor pressure high to be consistent with support setpoint calculations.
16. ITS 126.96.36.199 changed the LCO section of the Bases consistent with the changes made to accommodate RAI 188.8.131.52-1.
17. ITS 184.108.40.206 safety analysis section of the Bases has been changed to be consistent with changes made as a result of RAI 3.3.1-1.SI
18. ITS 220.127.116.11 changed the Trip Level Settings for Loss of Offsite Power (LOP) instrumentation listed in CTS Table 3.2.-2 to new ITS Allowable Values listed in ITS Table 18.104.22.168-1.
19. ITS 22.214.171.124 changed CTS 4.9.G.3 setpoint or Allowable Value of >~108V to >109.9V in its SR 126.96.36.199.3.
20. ITS 3.4.7 added an RHR Shutdown Cooling-Hot Shutdown specification to the ITS SPECIFICATION based on current licensing basis been restored to operable status within 30 days. ITS 188.8.131.52 ACTION B specifies initiating action in accordance with ITS 5.5.6 which relates to reporting requirements.
21. ITS 3.4.9, Reactor Coolant System (RCS) Pressure/Temperature (P/T) Limits in CTS were changed to add a new alternate criteria in ITS to allow idle recirculating pump (loop) start if the operating loop is greater than 40 percent flow or if the idle loop is less than 40% flow for less than or equal to 30 minutes.
22. ITS 3.5.1 and ITS 3.5.2, Emergency Core Cooling System (ECCS)-Operating and Shutdown, High-Pressure Coolant Injection (HPCI) and Residual Heat Removal (RHR) LPCI pump flow rates in CTS were reduced to SAFER/GESTR-Loss-of-Coolant Accident (LOCA) flow rates in the ITS.
23. ITS 3.5.3 adds an additional requirement to ITS SR 184.108.40.206 that requires the performance of the surveillance “Once each startup prior to exceeding 25% RTP.”
24. ITS 3.5.3 divides the existing CTS 4.5.E.1.d surveillance requirement that “RCIC delivers at least 400 gpm against a system head corresponding to a reactor vessel pressure of 1195 psig to 150 psig” into two separate Surveillance Requirements: ITS SR 220.127.116.11 and ITS SR 18.104.22.168.
25. ITS 22.214.171.124 deletes the CTS 4.7.A.1 requirement to inspect the interior surface of the drywell and suppression chamber above the water line every 24 months based on the inspection being required by the primary containment leakage rate testing program 3 times in 10 years.
26. ITS SR 126.96.36.199.1 changes the note in the ISTS markup that LPCI and Core Spray air operated testable check valve leakage test failure does not result in an ITS SR 188.8.131.52.1 failure.
27. ITS 184.108.40.206 changed CTS LPCI and CS testable check valve testing per Primary Containment Leakage Rate Testing (PCLRT) program (twice every 24 months).
28. ITS SR 220.127.116.11.1, SR 18.104.22.168.2, and B 22.214.171.124 changes the frequency of performing a functional test of each required vacuum breaker from 31 days as indicated in the ISTS to a new schedule in accordance with the IST Program which is 92 days.
29. ITS SR 126.96.36.199.2 was changed to add the word “required” to make it clearer that the SR is only applicable to the single RHR pump in a subsystem rather than both pumps in a subsystem that are provided by design.
30. ITS 3.8.1, AC Sources—Operating, Condition D for two reserve circuits inoperable in CTS was changed to add new interim power reduction to less than or equal to 45 percent with a 36-hour Completion Time in the ITS.
31. ITS 3.8.4, DC Sources—Operating (in CTS) was changed to allow 8 hours to restore one inoperable source in the ITS.
Before issuance of the proposed license amendments, the Commission will have made findings required by the Atomic Energy Act of 1954, as amended (the Act) and the commission's regulations.
By December 28, 2001, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.714, which is available at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland, or electronically on the Internet at the NRC Web site http://www.nrc.gov/NRC/CFR/index.html. If there are problems in accessing the document, contact the Public Document Room Reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to firstname.lastname@example.org. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following factors: (1) The nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the Start Printed Page 59498petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include alist of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. Petitioner must provide sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross-examine witnesses.
A request for a hearing and petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemakings and Adjudications Staff, or may be delivered to the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to Mr. David E. Blabey, attorney for the licensee, 1633 Broadway, New York, New York 10019.
Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for hearing will not be entertained absent a determination by the commission, the presiding officer or the presiding Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of the factors specified in 10CAR 2.714(a)(1)(l)-(v) and 2.714(d).
If a request for a hearing is received, the Commission's staff may issue the amendment after it completes its technical review and prior to the completion of any required hearing if it publishes a further notice for public comment of its proposed finding of no significant hazards consideration in accordance with 10 CFR 50.91 and 50.92. For further details with respect to the proposed action, see the licensee's application dated March 31, 1999, as supplemented by letters dated May 20, June 1, July 14, October 14, 1999, February 11, April 4, April 13, June 30, July 31, September 12, September 13, October 23, 2000, February 7, February 20, May 31, and August 6, 2001. Documents may be examined, and/or copied for a fee, at the NRC's Public Document room, located at One White Flint North, 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. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by email to email@example.com.Start Signature
For the Nuclear Regulatory Commission.
Dated at Rockville, Maryland, this 21st day of November 2001.
Guy S. Vissing,
Project Manager, Section 1, Project Directorate I, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 01-29585 Filed 11-27-01; 8:45 am]
BILLING CODE 7590-01-P