ML15014A277

Issuance of Amendment Application to Revise Technical Specifications for Technical Specification Low Pressure Safety Limit

ML15014A277
Person / Time
Site:	FitzPatrick
Issue date:	02/09/2015
From:	Pickett D Plant Licensing Branch 1
To:	Entergy Nuclear Operations
Pickett D
References
TAC MF2897
Download: ML15014A277 (14)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 9, 2015

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT- ISSUANCE OF AMENDMENT RE: APPLICATION TO REVISE TECHNICAL SPECIFICATIONS FOR TECHNICAL SPECIFICATION LOW PRESSURE SAFETY LIMIT (TAC NO. MF2897)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 309 to Renewed Facility Operating License No. DPR-59 for the James A.

FitzPatrick Nuclear Power Plant. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated October 8, 2013, as supplemented by a letter dated November 18, 2014.

The amendment modifies the TSs to reduce the reactor pressure associated with the Reactor Core Safety Limit from 785 psig to 685 psig in TS 2.1.1.1 and TS 2.1.1.2. This change addresses the potential to not meet the pressure/thermal power/minimum critical power ratio TS safety limit during a pressure regulator failure-maximum demand (open) (PRFO) transient. The PRFO transient was reported by General Electric Nuclear Energy as a notification pursuant to Title 10 of the Code of Federal Regulations, Part 21, "Reporting of Defects and Noncompliance."

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosures:

1. Amendment No. 309 to DPR-59

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR FITZPATRICK. LLC ENTERGY NUCLEAR OPERATIONS. INC.

DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 309 Renewed Facility Operating License No. DPR-59

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the James A. FitzPatrick Nuclear Power Plant (the facility) Renewed Facility Operating License No. DPR-59 filed by Entergy Nuclear Operations, Inc. (the licensee) dated October 8, 2013, as supplemented on November 18, 2014, 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 Renewed Facility Operating License No. DPR-59 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 309, are hereby incorporated in the renewed operating 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 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: February 9, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 309 RENEWED FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Replace the following page of the License 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 3 3 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 2.0-1 2.0-1

(4) ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use, at any time, any byproduct, source and special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus, components or tools.

(5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1:

Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level ENO is authorized to operate the facility at steady state reactor core power levels not in excess of 2536 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 309, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Fire Protection ENO shall implement and maintain in effect all provisions of the approved fire protections program as described in the Final Safety Analysis Report for the facility and as approved in the SER dated November 20, 1972; the SER Supplement No. 1 dated February 1, 1973; the SER Supplement No. 2 dated October 4, 1974; the SER dated August 1, 1979; the SER Supplement dated October 3, 1980; the SER Supplement dated February 13, 1981; the NRC Letter dated February 24, 1981; Technical Specification Amendments 34 (dated January 31, 1978), 80 (dated May 22, 1984), 134 (dated July 19, 1989), 135 (dated September 5, 1989), 142 (dated October 23, 1989), 164 (dated August 10, 1990), 176 (dated January 16, 1992), 177 (dated February 10, 1992), 186 (dated February 19, 1993), 190 (dated June 29, 1993), 191 (dated July 7, 1993), 206 (dated February 28, 1994) and 214 (dated June 27, 1994); and NRC Exemptions and associated safety evaluations dated April 26, 1983, July 1, 1983, January 11, 1985, April 30, 1986, September 15, 1986 and September 10, 1992 subject to the following provision:

Amendment 309 Renewed License No. DPR-59

Sls 2.0 2.0 SAFETY LIMITS {Sls) 2.1 Sls 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure< 685 psig or core flow

< 10% rated core flow:

THERMAL POWER shall be~ 25% RTP.

2.1.1.2 With the reactor steam dome pressure~ 685 psig and core flow

~ 10% rated core flow:

MCPR shall be~ 1.10 for two recirculation loop operation or

~ 1.13 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be~ 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all Sls; and 2.2.2 Insert all insertable control rods.

JAFNPP 2.0-1 Amendment 309

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 309 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-59 ENTERGY NUCLEAR FITZPATRICK, LLC AND ENTERGY NUCLEAR OPERATIONS. INC.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333

1.0 INTRODUCTION

By letter dated October 8, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13282A559) [Reference 1], as supplemented by a letter dated November 18, 2014 (ADAMS Accession No. ML14322B022) [Reference 2], Entergy Nuclear Operations, Inc. (Entergy, the licensee) requested an amendment to revise the Technical Specifications (TSs) for the James A. FitzPatrick Nuclear Power Plant (JAFNPP). The proposed change would reduce the reactor steam dome pressure specified within Reactor Core Safety Limits Specification 2.1.1, in the TS. This change would resolve a concern reported by General Electric Nuclear Energy as a notification pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) Part 21, "Reporting of Defects and Noncompliance," regarding the potential to momentarily violate Reactor Core Safety Limit 2.1.1.1 and 2.1.1.2 during a pressure regulator failure maximum demand (open) (PRFO) transient.

The supplemental letter dated November 18, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on July 8, 2014 (79 FR 38589).

2.0 REGULATORY EVALUATION

The following explains the use of general design criteria (GDC) for JAFNPP. The construction permit for JAFNPP was issued by the Atomic Energy Commission (AEC) on May 20, 1970, and the operating license was issued on October 17, 1974. The plant design criteria for the construction phase are listed in the Updated Final Safety Analysis Report (UFSAR)

Chapter 1.5, "Principal Design Criteria." The AEC published the final rule that added Appendix A to Title 10 of the Code of Federal Regulations (1 0 CFR) Part 50, "General Design Criteria for Nuclear Power Plants," in the Federal Register on February 20, 1971 (36 FR 3255),

with the rule effective on May 21, 1971. In accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, "SECY-92-223- Resolution of Deviations Identified During the Systematic Evaluation Program," dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971, which includes JAFNPP. However, the JAFNPP UFSAR, Chapter 16.6, "Conformance to AEC Design Criteria," evaluates JAFNPP against the 10 CFR Part 50, Appendix A GDC. Also, the initial AEC safety evaluation of JAFNPP, dated November 20, 1972, Chapter 14.0, stated, "Based on our evaluation of the design and design criteria for the James A. FitzPatrick Nuclear Power Plant, we conclude that there is reasonable assurance that the intent of the General Design Criteria for Nuclear Power Plants, published in the Federal Register on May 21, 1971 as Appendix A to 10 CFR part 50, will be met." Therefore, the NRC staff reviews amendments to the JAFNPP license using the 10 CFR Part 50, Appendix A GDC unless there are specific criteria identified in the UFSAR.

The regulatory requirements and guidance documents that the staff considered in its review of the proposed amendment included the following items.

Section 50.36 of 10 CFR, "Technical specifications," [Reference 6] provides the regulatory requirements for the content required in the TSs. As stated in 10 CFR 50.36(c)(1 )(i)(A):

Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be resumed until authorized by the Commission.

Compliance with the fuel licensing criteria of 10 CFR Part 50 Appendix A, GDC 10, "Reactor design," [Reference 7] is achieved by preventing the violation of fuel design limits. GDC 10 states:

The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits [SAFDLs] are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," (SRP) [Reference 8] provides guidance on, among other things, the acceptability of the reactivity control systems, the reactor core, and fuel system design. Specifically, Section 4.2, "Fuel System Design," specifies all fuel damage criteria for evaluation and whether fuel designs meet the SAFDLs. Section 4.4, "Thermal and Hydraulic Design," provides guidance on the review of thermal-hydraulic design in meeting the requirement of GDC 10 and the fuel design criteria established in SRP Section 4.2. It states that the critical power ratio (CPR) is to be established such that at least 99.9 percent of fuel rods

in the core would not be expected to experience departure from nucleate boiling or onset of transition boiling (OTB) during normal operation or anticipated operational occurrences (AOOs).

3.0 TECHNICAL EVALUATION

In 2005, General Electric (GE) submitted a 10 CFR Part 21 notification, SC05-03, "Potential to Exceed Low Pressure Technical Specification Safety Limit" to the NRC [Reference 3]. The Part 21 notification discussed how applying newer computer analysis codes to a PRFO transient could result in a condition where the reactor steam dome pressure could momentarily decrease below 785 pounds per square inch gauge (psig), while rated thermal power (RTP) was above the plant-specific thermal power limit specified in TS 2.1.1.1, which would result in a violation of the Reactor Core Safety Limit. Initially, the Boiling Water Reactor Owners' Group (BWROG) attempted to resolve the Part 21 issue. On July 18, 2006, the Technical Specifications Task Force (TSTF) and the BWROG submitted an Improved Standard Technical Specifications Change Traveler TSTF-495, Revision 0, "Bases Change to Address GE Part 21 SC05-03"

[Reference 4], proposing a modification to the "Applicable Safety Analysis" portion of the Reactor Core Safety Limit TS Bases (B 2.1.1 ). This change proposed to clarify that the safety limit was considered not to apply to momentary depressurization transients. In the NRC safety evaluation (SE) input for TSTF-495, dated August 14, 2007 [Reference 5], the staff stated that although the technical arguments presented in TSTF-495 had merit, the proposed change was not acceptable because it would set a precedent that could lead to erosion of safety margins protected by Safety Limits. The staff further stated that from a regulatory standpoint, the proposed change to the TS Bases was also not acceptable. Consequently, in April 2012, the BWROG discontinued the effort to resolve the issue generically and recommended that plants lower their Low Pressure Safety Limit to meet the lower range of their critical power correlation on a plant-specific basis. As a result, the licensee submitted this license amendment request to lower the lower-bound pressure limit for JAFNPP from 785 psig to 685 psig.

Some advanced fuel designs have an NRC approved critical power correlation with a lower-bound pressure significantly below the 785 psig reactor steam dome pressure specified in TS Reactor Core Safety Limits 2.1.1.1 and 2.1.1.2. The licensee proposes to utilize this fact and reduce the reactor steam dome pressure consistent with the approved lower-bound pressure for the critical power correlation for the GNF2 (Global Nuclear Fuel) fuel currently comprising the JAFNPP core. The GNF2 fuel utilizes the GEXL 17 critical power correlations with an approved pressure range that encompasses the licensee's proposed lower-bound pressure limit of 685 psig. Revising the reactor steam dome pressure specified in Reactor Core Safety Limits (SLs) 2.1.1.1 and 2.1.1.2 from 785 psig to 685 psig would resolve the 10 CFR Part 21 condition concerning the potential to violate a safety limit during a PRFO transient.

PROPOSED TS CHANGES The following change is proposed to TS SL 2.1.1, "Reactor Core SLs":

Current TS 2.1.1.1:

With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:

THERMAL POWER shall be s 25% RTP.

Proposed TS 2.1. 1.1:

With the reactor steam dome pressure < 685 psig or core flow < 10% rated core flow:

THERMAL POWER shall be ~ 25% RTP.

Current TS 2.1.1.2:

With the reactor steam dome pressure =::: 785 psig and core flow =::: 10% rated core flow:

MCPR [minimum critical power ratio] shall be =::: 1.10 for two recirculation loop operation or =::: 1.13 for single recirculation loop operation.

Proposed TS 2.1.1.2:

With the reactor steam dome pressure =::: 685 psig and core flow =::: 10% rated core flow:

MCPR shall be =::: 1.10 for two recirculation loop operation or =::: 1.13 for single recirculation loop operation.

In its October 8, 2013 submittal [Reference 1], the licensee stated that the GNF2 fuel type is currently in use in the JAFNPP reactor core. As stated earlier, GNF2 fuel type has an approved pressure range for its critical power correlations that encompass the licensee's proposed lower-bound pressure limit of 685 psig.

Each fuel vendor has developed critical power correlations valid over specified pressure and flow ranges (mass flow rates) that are approved by the NRC. These critical power correlations have become increasingly fuel design dependent as advanced fuel designs have evolved. This has resulted in an extension of the NRC approved pressure range to lower pressures as additional test data became available to demonstrate the validity of revised or new correlations for performance of critical power calculations. The critical power correlations for some advanced fuel designs have received NRC approval down to a lower pressure than those approved previously. The lower-bound of the extended pressure ranges for these advanced fuel designs can be used to establish a lower reactor steam dome pressure than the 785 psig value currently specified in JAFNPP Reactor Core Safety Limits 2.1.1.1 and 2.1.1.2. The licensee proposes to utilize the fact that the GNF2 fuel comprising the JAFNPP core utilizes critical power correlation that have an approved pressure range including the proposed lower safety limit of 685 psig. Therefore, a wider pressure range is available for transients to demonstrate compliance with MCPR limits. Thus, the proposed change offers a greater pressure margin for a PRFO transient than what is currently available.

In the 10 CFR Part 21 notification [Reference 3], GE concluded that fuel cladding integrity is not threatened during the PRFO because the CPR increases during depressurization such that the initial CPR is the limiting CPR condition during the entire transient, and the conditions that exceed the low pressure TS safety limit exist for only a few seconds. Nonetheless, GE considered the PRFO to be a known AOO that could contribute to exceeding a safety limit.

While this condition was not considered to involve an actual safety hazard, the potential for violating a Reactor Core Safety Limit had been identified, and restoration to comply with the safety limit is required for the PRFO event. As a result, the licensee proposes to revise the

reactor steam dome pressure TS Safety Limit consistent with the NRC approved pressure range of critical power correlations for the current JAFNPP fuel design. Lowering the reactor steam dome pressure specification in this fashion would provide margin to ensure that Reactor Core Safety Limit 2.1.1.1 is not violated and would resolve the 10 CFR Part 21 issue.

The licensee confirmed [Reference 2] that the JAFNPP Cycle 22 core consists of one fuel type, namely 560 GNF2 fuel assemblies, and that there is no lead test assembly (LTA) in the core.

The NRC staff reviewed the licensee's submittal [Reference 1], supplemental information provided by the licensee in response to the staff's questions [Reference 2], and related documentation (e.g., the JAFNPP TS and Updated Final Safety Analysis Report (UFSAR), the GE 10 CFR Part 21 notification, the proposed TSTF-495, and related staff SEs). The staff concluded that reactor depressurization transients, such as PRFOs, are non-limiting for fuel cladding integrity and that the proposed change in TS 2.1.1.2 will have no negative impact on the MCPR core operating limits. Although this condition does not involve an actual safety hazard, the potential for violating a TS Reactor Core Safety Limit was identified by the 10 CFR Part 21 notification and restoration to comply with the safety limit was required; hence, the licensee proposed this amendment in order to address this issue.

The NRC staff determined that revising the Reactor Core Safety Limits 2.1.1.1 and 2.1.1.2 reactor steam dome pressures from 785 to 685 psig resolves the 10 CFR Part 21 notification issue. TS Safety Limits are specified to ensure that SAFDLs are not exceeded during steady state operation, normal operational transients, and AOOs. The Reactor Core Safety Limits are set such that fuel cladding integrity is maintained and no significant fuel damage is calculated to occur due to OTB if the Safety Limits are not exceeded.

The NRC staff confirmed that the GEXL 17 correlation for GNF2 fuel was approved for use per NEDE-24011-PA, "General Electric Standard Application for Reactor Fuel (GESTAR II)"

[Reference 9]. The pressure range over which the GEXL 17 correlation is valid for performance of critical power calculations includes the proposed lower safety limit. The reactor steam dome pressure of 685 psig is established from the lower bound pressure (700.0 pounds per square inch absolute (psia)- 14.7 psia =685.3 psig;;:;; 685 psig). As a result, the staff concluded that the proposed change to the Reactor Core Safety Limits continues to ensure that a valid CPR is maintained for the AOOs described in the UFSAR, including the PRFO transient, and that the proposed value of 685 psig for the reactor steam dome pressure would not result in a violation of Reactor Core Safety Limit 2.1.1.1 during a PRFO transient. Furthermore, the proposed change will continue to provide protection during startup conditions to ensure that operation at less than 685 psig or less than 10% core flow, while greater than 25 percent RTP, would not occur. Since this approach follows, and is consistent with, the way the reactor steam dome pressure has been established and valid CPR values will be maintained, it is a safe and appropriate method to address the 10 CFR Part 21 notification and, therefore, is acceptable. If JAFNPP transitions to different fuel designs, the licensee will need to reevaluate the GE 10 CFR Part 21 notification to determine whether NRC approval is required to change the TS Safety Limits.

The NRC staff evaluated the proposed changes against the applicable regulatory requirements and acceptance criteria. The NRC staff concludes that as long as the core pressure and flow are within the range of validity of the specified CPR correlation (i.e., GEXL 17 correlations for JAFNPP), the proposed reactor steam dome pressure changes to Reactor Core Safety Limits

2.1.1.1 and 2.1.1.2 will continue to ensure that 99.9 percent of the fuel rods in the core are not expected to experience OTB. This satisfies the regulatory requirements regarding acceptable fuel design limits and continues to assure that the underlying criteria of the safety limit are met consistent with GDC 10 and 10 CFR 50.36(c)(1)(i)(A). Therefore, the proposed amendment is acceptable. The NRC staff further concludes that there is reasonable assurance that the health and safety of the public, following approval of this TS change, will be unaffected.

4.0 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.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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 as published in the Federal Register on July 8, 2014 (79 FR 38589). 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.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that 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.

7.0 REFERENCES

1) Letter from L.M. Coyle (Entergy) to USNRC, "Application to Revise Technical Specifications for Technical Specification Low Pressure Safety Limit, James A FitzPatrick Nuclear Power Plant," October 8, 2013, ADAMS Accession No. ML13282A559.

2) Letter from L.M. Coyle (Entergy) to USNRC, "Response to Request for Additional Information (RAI) for Low Pressure Safety Limit License Amendment (TAC No.

MF2897), James A FitzPatrick Nuclear Power Plant," November 18, 2014, ADAMS Accession No. ML143228022.

3) GE letter to the NRC, GENE SC05-03, "Potential to Exceed Low Pressure Technical Specification Safety Limit," Reportable condition pursuant to 10 CFR Part 21, dated March 29, 2005.

4) Technical Specifications Task Force letter (TSTF-06-20) transmitting TSTF-495, Revision 0, "Bases Change to Address GE Part 21 SC05-03," dated July 18, 2006, ADAMS Accession No. ML061990227.

5) Safety Evaluation Input For TSTF-495, Revision 0, "Bases Change To Address GE Part 21 SC05-03," (TAC No. MD2672), dated August 14, 2007, ADAMS Accession No. ML072280007.

6) U.S. Code of Federal Regulations, "Domestic Licensing of Production and Utilization Facilities -Technical specifications," Title 10 of the Code of Federal Regulations," Part 50, Section 36 (1 0 CFR 50.36).

7) U.S. Code of Federal Regulations, "Domestic Licensing of Production and Utilization Facilities - General Design Criteria for Nuclear Power Plants," Title 10 of the Code of Federal Regulations," Part 50, Appendix A (1 0 CFR 50, Appendix A).

8) U.S. Nuclear Regulatory Commission, i*standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," (NUREG-0800, Formerly issued as NUREG-75/087).

9) Memorandum from A J. Mendiola (NRC) to S. L. Rosenberg (NRC), "Staff Findings Regarding Supplemental Information Pertaining to the Compliance of the GNF2 Fuel Design to the General Electric Standard Application for Reactor Fuel (GESTAR II),"

August 4, 2009, ADAMS Accession No. ML092080499.

10) U.S. Nuclear Regulatory Commission, "Report to Congress on Abnormal Occurrences, Fiscal Year 2006," NUREG-0090, Vol. 29, April2007, ADAMS Accession No. ML071080195.

Principal Contributor: M. M. Razzaque, DSS/SRXB Date: February 9, 2015

ML15014A277 *b>Y memo OFFICE LPLI-1\PM LPLI-1\PM LPLI-1\LA DSS\SRXB DSS\STSB OGC LPLI-1\BC NAME DRender DPickett KGoldstein CJackson* REIIiot JWachutka BBeasley DATE 01/20/2015 1/22/2015 01/20/2015 12/19/2014 2/02/2015 1/26/2015 2/09/2015