ML14108A399

Issuance of Amendment Request to Relocate Flood Protection Technical Specification to the Technical Requirements Manual

ML14108A399
Person / Time
Site:	Hope Creek
Issue date:	12/18/2014
From:	Carleen Parker Plant Licensing Branch 1
To:	Joyce T Public Service Enterprise Group
Lamb J, NRR/DORL/LPLI-2
References
TAC MF2738
Download: ML14108A399 (13)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 18, 2014 Mr. Thomas Joyce President and Chief Nuclear Officer PSEG Nuclear LLC P.O. Box 236, N09 Hancocks Bridge, NJ 08038

SUBJECT:

HOPE CREEK GENERATING STATION -ISSUANCE OF AMENDMENT RE:

REQUEST TO RELOCATE FLOOD PROTECTION TECHNICAL SPECIFICATION TO THE TECHNICAL REQUIREMENTS MANUAL (TAC NO. MF2738)

Dear Mr. Joyce:

The Commission has issued the enclosed Amendment No. 196 to Renewed Facility Operating License No. NPF-57 for the Hope Creek Generating Station (Hope Creek). This amendment consists of changes to the technical specifications (TSs) in response to your application dated September 5, 2013.

The amendment relocates the operability and surveillance requirements (SRs) for flood protection from the Hope Creek TSs to the Hope Creek Technical Requirements Manual (TRM).

The affected TSs are: Flood Protection, TS 3.7.3 and SR 4.7.3. As part of the amendment, the operability and SRs for the flood protection will be relocated verbatim into the Hope Creek TRM.

A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely~

CJ .y~

Carleen ~ Parker, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-354

Enclosures:

1. Amendment No. 196 to Renewed License No. NPF-57

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 196 Renewed License No. NPF-57

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

A. The application for amendment filed by PSEG Nuclear LLC dated September 5, 2013, 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 set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No.

NPF-57 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 196, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. The license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULA TORY COMMISSION

'-y/_Y_'\ V\.../*~~-~ .. _____ )

Meena K. Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and the Technical Specifications Date of Issuance: becember 18, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 196 RENEWED FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following page of the Renewed Facility Operating License with the revised page.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert Page 3 Page 3 Replace the following pages of the 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 xiii xiii XX XX 3/4 7-9 3/4 7-9 3/4 7-10 3/4 7-10

reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) PSEG Nuclear LLC, 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 as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. Mechanical disassembly of the GE 14i isotope test assemblies containing Cobalt-60 is not considered separation.

(7) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I 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 PSEG Nuclear LLC is authorized to operate the facility at reactor core power levels not in excess of 3840 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 196, and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-57 Amendment No. 196

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7.3 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-9 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM . . . . . . . . . . . . . . . . . 3/4 7-11 3/4.7.5 DELETED 3/4.7.6 SEALED SOURCE CONTAMINATION . . . . . . . . . . . . . . . . . . . . * . . . . . . 3/4 7-19 3/4.7.7 MAIN TURBINE BYPASS SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-21 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. Sources-Operating . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-1 Table 4.8.1.1.2-1 Diesel Generator Test Schedule .... 3/4 8-10 A.C. Sources-Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-11 3/4.8.2 D.C. SOURCES D.C. Sources-Operating . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-12 Table 4.8.2.1-1 Battery Surveillance Requirements ... 3/4 8-15 D.C. Sources Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-17 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS Distribution- Operating . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-18 Distribution- Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-21 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Primary Containment Penetration Conductor Overcurrent Protective Devices . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-24 Table 3.8.4.1-1 Primary Containment Penetration Conductor Overcurrent Protective Devices...... 3/4 8-26 Motor Operated Valve Thermal Overload Protection (Bypassed) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-30 HOPE CREEK xiii Amendment No. 196

INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-1 3/4.7.2 CONTROL ROOM SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-1 3/4.7.3 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-1b 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM ........... . B 3/4 7-1c 3/4.7.5 DELETED 3/4.7.6 SEALED SOURCE CONTAMINATION . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-4 3/4.7.7 MAIN TURBINE BYPASS SYSTEM . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-4 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES and ONSITE POWER DISTRIBUTION SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES ......... . B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.2 INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.3 CONTROL ROD POSITION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.4 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.5 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.6 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.7 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.8 and 3/4.9.9 WATER LEVEL -REACTOR VESSEL and WATER LEVEL- SPENT FUEL STORAGE POOL ....... . B 3/4 9-2 3/4 . 9. 10 CONTROL ROD REMOVAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION .... B 3/4 9-2 HOPE CREEK XX Amendment No. 196

PLANT SYSTEMS 314.7.3 DELETED HOPE CREEK 3/4 7-9 Amendment No. 196

THIS PAGE INTENTIONALLY BLANK HOPE CREEK 3/4 7-10 Amendment No. 196

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 196 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-57 PSEG NUCLEAR LLC HOPE CREEK GENERATING STATION DOCKET NO. 50-354

1.0 INTRODUCTION

By letter dated September 5, 2013, 1 PSEG Nuclear LLC (PSEG or the licensee) submitted a request for changes to the Hope Creek Generating Station (Hope Creek) technical specifications (TSs).

The amendment removes the operability and surveillance requirements (SRs) for flood protection from the Hope Creek TSs and places them in the Hope Creek Technical Requirements Manual (TRM).

The affected TSs are: Flood Protection, TS 3.7.3 and SR 4.7.3. As part of the amendment, the operability and SRs for the flood protection will be relocated into the Hope Creek TRM.

2.0 REGULATORY EVALUATION

The U.S. Nuclear Regulatory Commission (NRC) staff used the following regulatory basis for its evaluation of the licensee's amendment request:

• The regulations in Title 10 of the Code of Federal Regulations (1 0 CFR) 50.36(c)(2) state, in part, that "[w]hen a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met ... "

• The regulations in 10 CFR 50.36(c)(2)(ii) state, "[a] technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

1 Agencywide Documents Access and Management System (ADAMS) Accession No. ML13249A242.

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety."

• The regulations in 10 CFR 50.36(c)(3) state, "surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."

• Regulatory Guide (RG) 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 2, issued December 1980, lists five types (Types A-E) of variables to help designers select the accident monitoring instrumentation and applicable criteria. Categories 1, 2, and 3 separate the type criteria into groups for a graded approach to requirements, depending on the importance to safety or the measurement of a specific variable.

• The NRC Staff's Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Commission's Interim Policy statement Criteria to Standard Technical Specifications, dated, May 9, 1988.

On July 22, 1993, the Commission issued its final policy statement indicating that satisfying the guidance in the policy statement also satisfies Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.36, "Technical Specifications" (58 FR 39132; July 22, 1993). The final policy statement described the safety benefits of the improved Standard Technical Specifications (STS) and encouraged licensees to use the improved STS as the basis for plant-specific TS amendments and for complete conversions to the improved STS. Further, the final policy statement gave guidance for evaluating the required scope of the licensee's improved TS and defined the guidance criteria for determining which of the limiting condition for operations (LCOs) and associated surveillances should remain in its improved TS. Using this approach, licensees should keep existing LCO requirements that fall within or satisfy any of the final policy statement criteria in the TS. Those LCO requirements that do not fall within or satisfy any of these criteria may be relocated to licensee-controlled documents. The Commission codified the four criteria in 10 CFR 50.36(c)(2)(ii) (60 FR 36953; July 19, 1995). NUREG-1443, "Standard Technical Specifications General Electric BWR/4 Plants," dated April2012, (ADAMS Accession

No. ML12104A192) was developed based on the criteria in the Final Policy Statement.

The proposed amendment would relocate TS 3.7.3 and SR 4.7.3 to the Hope Creek TRM. The TRM is described in Section 13.5, "Plant Procedures," of the Hope Creek Updated Final Safety Analysis Report (UFSAR). Specifically, UFSAR Section 13.5.4 reads as follows:

13.5.4 Technical Requirements Manual (TRM)

The Technical Requirements Manual (TRM[)] contains technical requirements and/or supporting information (e.g., tables and component lists) which were once contained in the HCGS [Hope Creek] Technical Specifications (TS) (i.e.,

Appendix A of the HCGS Facility Operating License). Removal of the TS and information is approved by the NRC through individual TS amendments. The TRM is intended to provide operational guidance and requirements for various plant conditions, actions, and testing similar to TS, however, these requirements are in accordance with licensing commitments. These changes add the TRM into the scope of procedures to be processed through the Station Qualified Reviewer (SQR) process and reviewed by PORC [Plant Operations Review Committee]. Future changes to the relocated requirements and supporting information are processed in accordance with the Quality Assurance Topical Report (QATR), and are subject to 10CFR50.59 review. All non-editorial changes are reviewed by PORC.

The TRM is comprised of an index, the individual specification[s] and bases. The manual is intended to provide a single location for the relocated TS items as a convenience for operations and other station personnel. The individual sections of the TRM contain the relocated licensing commitments, which are subject to the provisions of 10CFR50.59 described above, and are controlled in accordance with the applicable established procedure process.

Nuclear Energy Institute (NEI) guidance document NEI 98-03, Revision 1, "Guidelines for Updating Final Safety Analysis Reports" (ADAMS Accession No. ML003779028) lists the following methods of controlling the TRM on page 7 of Appendix A:

The TRM or other licensee controlled document is explicitly "incorporated by reference" into the UFSAR. Under this approach, the referenced document is subject to the change control requirements of 10 CFR 50.59 and the update/reporting requirements of 10 CFR 50.71(e), e.g., periodic submittal of change pages, etc.

The TRM or other licensee controlled document is treated in a manner consistent with procedures fully or partially described in the UFSAR. Under this approach, the referenced document is maintained on-site in accordance with licensee administrative processes, and changes are evaluated using 10 CFR 50.59.

Regulatory Guide (RG) 1.181, "Content of the Updated Final Safety Analysis Report in Accordance with 10 CFR 50.71(e)" dated September 1999 (ADAMS Accession No. ML992930009), states that Revision 1 of NEI 98-03 provides methods that are acceptable to the

NRC staff for complying with the provisions of 10 CFR 50.71 (e).

Based on the description in Hope Creek UFSAR Section 13.5.4, the Hope Creek TRM is treated in a manner consistent with procedures fully or partially described in the UFSAR. After the information in TS 3. 7.3 and SR 4. 7.3 is relocated to the TRM, any future changes to the flooding protection would be controlled by the provisions in 10 CFR 50.59.

3.0 TECHNICAL EVALUATION

3.1 Background The requirement for flood protection ensures that facility protection features are in place in the event of flood conditions. The limit of elevation 10.5 feet Mean Sea Level (MSL) [99.5 feet PSEG datum] is based on the elevation at which facility flood protection features provide protection to safety-related equipment.

3.2 Description of Proposed TS Change The proposed license amendment would remove the following items from the TS and relocate them to the TRM:

1. TS 3.7.3 on page 3/4 7-9. This specification reads:

Flood protection shall be provided for all safety related systems, components, and structures when the water level of the Delaware River reaches 6.0 feet Mean Sea Level (MSL) USGS datum (95.0 feet PSE&G datum) at the Service Water Intake Structure.

The associated APPLICABILITY is: "At all times."

The associated Actions follow:

a. With severe storm warnings from the National Weather Service which may impact Artificial Island in effect or with the water level at the service water intake structure above elevation 6.0 feet MSL USGS [United States Geological Survey] datum (95.0 feet PSE&G datum), initiate and complete:

1. The closing of all service water intake structure watertight perimeter flood doors identified in Table 3. 7. 3- 1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or declare affected service water system components inoperable and take the actions required by LCO 3. 7 .1.2;

-and-

2. The closing of all power block watertight perimeter flood doors identified in Table 3.7.3-1 within 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Once closed, all access through the doors shall be administratively controlled.

b. With the water level at the service water intake structure above elevation 10.5 feet MSL USGS datum (99.5 feet PSE&G datum), be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUT DOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. SR 4.7.3 on page 3/4 7-9. This specification reads:

The water level at the service water intake structure shall be determined to be within the limit by:

a. Measurement in accordance with the Surveillance Frequency Control Program when the water level is below elevation 6.0 MSL USGS datum (95.0 feet PSE&G datum), and

b. Measurement at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when severe storm warnings from the National Weather Service which may impact Artificial Island are in effect.

c. Measurement in accordance with the Surveillance Frequency Control Program when the water level is equal to or above elevation 6.0 MSL USGS datum (95.0 feet PSE&G datum).

3. Table 3.7.3-1, "Perimeter Flood Doors" on page 3/4 7-10.

Upon approval of the proposed changes, the operability and SRs for flood protection will be incorporated into the Hope Creek TRM.

The following associated changes will also be made to support relocation of TS 3/4.7.3:

1. The TS Index will be revised to delete the reference toTS 3/4.7.3 and Bases 3/4. 7.3.

2. TS Bases page B 3/4 7-1b will be changed to delete the description for 3/4.7.3, Flood Protection. This bases definition will also be incorporated verbatim into the Hope Creek TRM.

3.3 Evaluation of Proposed TS Changes While flood protection is important to ensure safety-related equipment is protected, it is not a detector or indicator of reactor coolant pressure boundary (RCPB) degradation. The TS for Leakage Detection Systems is an example of a requirement provided to monitor degradation of the RCPB. Therefore, flood protection does not meet Criterion 1 for inclusion in the TS.

Flood protection is not a process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or Transient analysis. This is consistent with the NRC Final Policy Statement, which provided that Criterion 2 analyses are contained in Chapters 6 and 15 of the Final Safety Analysis Report (FSAR). Flood events are contained in Chapters 2 and 3 of the Hope Creek Updated FSAR (UFSAR). Therefore, flood protection does not meet Criterion 2 for inclusion in the TS.

Flood protection is not part of the primary success path in the mitigation of a DBA or transient.

Therefore, flood protection does not meet Criterion 3 for inclusion in the TS.

Hope Creek operating experience has shown that site flooding (i.e., flooding above plant grade elevation) is a highly unlikely event. Hope Creek has never initiated a plant shutdown due to a high water level of 99.5 feet (PSEG Datum). Additionally, .the Hope Creek UFSAR highest historical high water was 97.5 feet (PSEG Datum), recorded in November 1950, which is 4 feet below plant grade.

The Hope Creek response dated July 31, 1997, to Generic Letter 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities- 10 CFR 50.54(f),"

Supplement 4, provided the IPEEE for severe accident vulnerabilities. The IPEEE found that the evaluation of "other" external events [i.e., high winds, floods, etc.] were screened out by compliance with Standard Review Plan (SRP) criteria or by demonstration that their predicted core damage frequency fell below the IPEEE screening criteria. In addition, the IPEEE found no known plant-unique external event that poses a significant threat of severe accidents. The NRC staff evaluation report, dated July 26, 1999, on the IPEEE submittal concluded that the IPEEE results are reasonable given the Hope Creek design, operation, and history. Based upon this risk data, flood protection is not a structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Therefore, flood protection does not meet Criterion 4 for inclusion in the TS.

Flood protection does not meet any of the four screening criteria in the Final Policy Statement and 10 CFR 50.36. This conclusion is supported by the absence of operability and SRs for flood protection in the improved standard technical specifications (ISTS) presented in NUREG-1433. Accordingly, this proposed change conforms to theISTS, and flood protection requirements can be established in a licensee-controlled document, the Hope Creek TRM.

Future changes to flood protection requirements in the TRM will be subject to the controls of 10 CFR 50.59.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 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, and changes surveillance requirements. 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 (79 FR 21299, dated April15, 2014). 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.

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.

Principal Contributor: Matt Hamm Date: becernber 18, 2014.

ML14108A399 OFFICE LPL 1-2/PM LPL 1-2/LA SBPB/BC STSB/BC OGC LPL 1-2/BC LPL 1-2/PM REIIiot BHarris NLO NAME Jlamb A Baxter GCasto MKhanna CParker w/comments DATE 06/06/14 05/21/14 6/25/14 06/23/14 12/04/14 12/15/14 12/18/14