ML101880229

Issuance of License Amendment 298 Regarding Modification of the Tech Specs Requirements for Testing of the Shutdown Cooling System Isolation, Reactor Pressure-High

ML101880229
Person / Time
Site:	FitzPatrick
Issue date:	07/21/2010
From:	Bhalchandra Vaidya Plant Licensing Branch 1
To:	Fitzpatrick J Entergy Nuclear Operations
vaidya B, NRR/Dorl/lpl1-1, 415-3308
References
TAC ME1819
Download: ML101880229 (13)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 July 21, 2010 Vice President, Operations Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT RE: MODIFICATION OF THE TECHNICAL SPECIFICATIONS REQUIREMENTS FOR TESTING OF THE SHUTDOWN COOLING SYSTEM ISOLATION, REACTOR PRESSURE - HIGH (TAC NO. ME1819)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 298 to Renewed Facility Operating License No. DPR-59 for the James A. FitzPatrick Nuclear Power Plant (JAFI\IPP). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated July 31, 2009, as supplemented by letters dated March 5, 2010, and June 17, 2010.

The amendment revised the JAFNPP TS Surveillance Requirements (SR) for testing of the Residual Heat Removal System Shutdown Cooling (SDC) mode Containment Isolation, Reactor Pressure - High Function by replacing the current requirement to perform TS SR 3.3.6.1.3, Perform Channel Calibration, with TS SR 3.3.6.1.1 Perform Channel Check, SR 3.3.6.1.2, Perform Channel Functional Test, SR 3.3.6.1.4, Calibrate the Trip Units, and SR 3.3.6.1.5, Perform Channel Calibration. These changes are to support a proposed plant modification to increase the reliability of SDC isolation logic by changing the source of the reactor high pressure input signal.

V. P. Operations -2 A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, Bhalchandra K. Vaidya, 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. 298 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 AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 298 Renewed Facility Operating License No. DPR-59

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

A. The application for amendment by Entergy Nuclear Operations, Inc. (the licensee) dated July 31,2009, as supplemented by letters dated March 5, 2010, and June 17, 2010, 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:

-2 (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 298, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION l~i.~

Nancy L. Salgado, 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: July 21, 2010

ATTACHMENT TO LICENSE AMENDMENT NO.298 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 3.3.6.1-10 3.3.6.1-10

-3 (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 I: 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 N029sare 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 298

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 5 of 6)

Primary Containment Isolation Instrumentation APPLICABLE REQUIRED CONDITIONS SURVEILLANCE ALLOWABLE MODES OR CHANNELS REFERENCED REQUIREMENTS VALUE OTHER PER TRIP FROM SPECIFIED SYSTEM REQUIRED FUNCTION CONDITIONS ACTION C.l

5. Reactor Water Cleanup (RWCU)

System Isolation

a. RWCU Suction Line 1.2.3 1 F SR 3.3.6.1.3 <:: 144°F Penetration Area SR 3.3.6.1.7 Temperature High

b. RWCU Pump Area 1.2.3 1 per F SR 3.3.6.1.3 <:: 165°F for Pump Temperature - High room SR 3.3.6.1.7 Room A and <::

175°F for Pump Room B

c. RWCU Heat 1,2,3 1 F SR 3.3.6.1.3 <;155°F Exchanger Room Area SR 3.3.6.1.7 Temperature High

d. SLC System Initiation 1,2 2(d) I SR 3.3.6.1.7 NA

e. Reactor Vessel Water Level 1,2.3 2 F SR 3.3.6.1.1 ~ 177 inches

- Low (Level 3) SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7

f. Drywell Pressure High 1.2,3 2 F SR 3.3.6.1.1 <:: 2.7 psig SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7

6. Shutdown Cooling System Isolation

a. Reactor Pressure High 1,2.3 1 F SR 3.3.6.1.1 <:: 74 psig SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7

b. Reactor Vessel Water Level 3.4.5 2(e) SR 3.3.6.1.1 ~ 177 inches

- Low (Level 3) SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7 (continued)

(d) SLC System Initiation only inputs into one of the two trip systems and only isolates one valve in the RWCU suction and return line.

(e) Only one trip system required in MODES 4 and 5 when RHR Shutdown Cooling System integrity maintained.

JAFNPP 3.3.6.1-10 Amendment 298

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 298 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-59 ENTERGY NUCLEAR OPERATIONS, INC.

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

1.0 INTRODUCTION

By letter dated July 31,2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092160401), as supplemented by letters dated March 5, 2010 (ADAMS Accession Nos. ML100750522 and ML100750523), and June 17,2010 (ADAMS Accession No. ML101680559), Entergy Nuclear Operations, Inc. (the licensee) submitted a request for changes to the James A. FitzPatrick Nuclear Power Plant (JAFNPP) Technical Specifications (TS). The supplements dated March 5 and June 17,2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination.

The proposed change would revise the JAFNPP TS Surveillance Requirements (SR) for testing of the Residual Heat Removal (RHR) System Shutdown Cooling (SOC) mode Containment Isolation, Reactor Pressure - High Function by replacing the current requirement to perform TS SR 3.3.6.1.3, Perform Channel Calibration, with TS SR 3.3.6.1.1 Perform Channel Check, SR 3.3.6.1.2, Perform Channel Functional Test, SR 3.3.6.1.4, Calibrate the Trip Units, and SR 3.3.6.1.5, Perform Channel Calibration. These changes are to support a proposed plant modification to increase the reliability of SOC isolation logic by changing the source of the reactor high pressure input signal.

The JAFNPP's current design uses pressure switches to provide reactor pressure input to the SOC isolation logic. However, plant experience has shown that the pressure switches can cause spurious inadvertent system isolations because their locations in the system are susceptible to hydraulic pressure transients during system startup. To avoid the system sensitivity to hydraulic pressure transients, the plant modification will replace the pressure switch inputs with inputs from existing reactor steam dome pressure transmitters. The modification will also provide the steam dome pressure analog signal and bistable functions with transmitters and the analog transmitter trip system (ATIS).

The proposed amendment will replace the current 92-day channel calibration SR for the pressure switches with the SRs associated with the existing reactor steam dome pressure

-2 transmitters that will be modified to supply the reactor pressure input to the SDC isolation logic.

The proposed change replaces the current requirement to perform TS SR 3.3.6.1.3, "Perform Channel Calibration," with a requirement to perform TS SR 3.3.6.1.1, "Perform Channel Check,"

TS SR 3.3.6.1.2, "Perform Channel Functional Test," TS SR 3.3.6.1.4, "Calibrate the Trip Units,"

and TS SR 3.3.6.1.5, "Perform Channel Calibration."

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 (10 CFR) Part 50, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971, with the rule effective on May 21, 1971. In accordance with a U.S. Nuclear Regulatory Commission (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 UFSAR 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 NRC staff considered the following regulatory requirements in its review of the license amendment request:

• 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities,"

establishes the fundamental regulatory requirements for the domestic licensing of nuclear production and utilization facilities. Specifically, Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 provides, in part, the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety.

• General Design Criterion (GDC) 13, "Instrumentation and Control," requires that instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.

Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

-3

• GDC 20, "Protective System Functions," requires the protection system to be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences, and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

• 10 CFR 50.36(c)(3), "Surveillance requirements," requires that SRs 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 (LCOs) will be met.

3.0 TECHNICAL EVALUATION

JAFNPP's TS Table 3.3.6.1-1, "Function 6a, Shutdown Cooling System Isolation, Reactor Pressure-High," lists each primary containment isolation function, applicable modes, required channels per trip system, conditions referenced from required action, SRs, and allowable value.

Currently, TS Table 3.3.6.1-1 contains the following SRs:

• SR 3.3.6.1.3, "Perform Channel Calibration," at a frequency of 92 days

• SR 3.3.6.1.7, "Perform Logic System Functional Test," at a frequency of 24 months The licensee imposed SR 3.3.6.1.3 based on the source of the reactor pressure input to the isolation logic. Because the source of the reactor pressure input will be changed from the current pressure switches to existing pressure transmitters associated with the ATTS, the licensee proposed to revise the required surveillances to reflect the SRs for the reactor steam dome pressure transmitters per this license amendment request.

The proposed change would replace SR 3.3.6.1.3 for the subject function with the following SRs associated with the transmitters:

• SR 3.3.6.1.1, "Perform Channel Check," at a frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

• SR 3.3.6.1.2, "Perform Channel Functional Test," at a frequency of 92 days

• SR 3.3.6.1.4, "Calibrate the Trip Units," at a frequency of 184 days

• SR 3.3.6.1.5, "Perform Channel Calibration," at a frequency of 24 months The proposed change does not affect the existing logic system function test requirement, SR 3.3.6.1.7.

The licensee will not revise the TS bases associated with the SRs because it will use existing SRs in the changes. However, the licensee will revise the TS bases for primary containment isolation instrumentation function 6a to reflect the changes in the source of the reactor pressure inputs to the isolation logic.

In response to the NRC staff's request for additional information (RAI), the licensee provided the calculation of total loop uncertainty for the new design, including the uncertainties of existing reactor steam dome pressure transmitters, master trip unit (MTU), and slave trip unit (STU).

- 4 The applicanfs responses to the RAI indicate that the steam dome pressure transmitters are located outside of primary containment on the 300-foot elevation of the reactor building. The RAI responses also state that the SDC system high reactor-pressure isolation function is not credited in any accident analysis and is not required to mitigate the consequences of an accident. Therefore, the worst-case environmental conditions that the transmitters will be exposed to when the instruments are required to operate are the same as normal environmental conditions. The MTUs and STUs that make up the ATIS are physically located in the relay room. This room is below the control room and has a heating, ventilation, and air conditioning system that controls the environmental conditions. Therefore, the worst-case environmental conditions that the MTUs and STUs will be exposed to when the instruments are required to operate are the same as normal environmental conditions.

On the basis of normal environmental conditions, the licensee calculated the total loop uncertainty and provided the field trip set point (FTS), allowable value, as-found zone, and as left tolerance for the instrument channel. Because of the significant increase in the instrument loop span of the reactor steam dome pressure transmitters (1,200 pound-force per square inch gauge (psig)), there is a corresponding increase in instrument uncertainty. This increased uncertainty requires a change of FTS from 70 psig to 59 psig, and a lower allowable value in the technical requirements manual from 50 psig to 41 psig. However, the TS upper allowable value will remain the same at 74 psig. In addition, the licensee stated that the ATTS has a proven history of reliability and accuracy at JAF and confirmed that an FTS of 59 psig is conservative and will adequately protect analytical value.

The licensee will periodically test instrument channels in accordance with JAPs Surveillance Test Program. The licensee stated that conduct of surveillance test procedures in the field is controlled via station procedures, including Surveillance Test Program and Fundamentals of Maintenance. Test personnel are required to notify supervision of any parameter response that is different than expected and to immediately notify the operation shift manager or control room supervisor in the event that acceptance criteria are not satisfied.

The licensee will review and ensure that surveillance acceptance criteria were satisfied and "as found' and "as-left'values are within their required tolerances followinq a surveillance test. In the event that surveillance acceptance criteria were not met, procedures in the Surveillance Test Program and Corrective Action Program require the test personnel to initiate and document the failure in a condition report. If a condition report documents that an instrument failed to meet the acceptance criteria and could not be restored to within tolerance, the instrument would be declared INOPERABLE and the appropriate condition associated with the applicable LCO would be entered. Once an instrument has been declared INOPERABLE, the limiting condition for operation cannot be exited until the instrument meets all SRs to establish OPERABILITY. In cases where the instrument can be adjusted to within the tolerance, the instrument would not be declared INOPERABLE; however, the potential impact on past OPERABILITY would be evaluated as part of the Corrective Action Program.

After reviewing the total loop uncertainty calculation and the submitted information, the NRC staff finds that the proposed reactor steam dome pressure transmitter with the ATIS has larger uncertainty than that of the original pressure switch instrument. However, the licensee has provided an acceptable justification for the mild operation conditions, submitted the setpoint calculation, and confirmed that the revised FTS is still conservative and will adequately protect analytical value.

-5 Therefore, the NRC staff finds that the proposed change of the TS SR and the corresponding changes to the TS bases are acceptable.

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 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 (74 FR 51239 dated October 6,2009). 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) 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: P. Chung Date: July 21, 2010

V. P. Operations -2 A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, IRA!

Bhalchandra K. Vaidya, 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. 298 to DPR-59

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

PUBLIC LPL1-1 R/F RidsNrrDorlLPL 1-1 RidsOGCMailCenter RidsNrrDirsltsb RidsNrrDeEicb RidsNrrDssSrsb RidsAcrsAcnwMailCenter PChung, NRRlEICB RidsNrrPMBVaidya(paper copy) RidsNrrLASLittle (paper copy) MGray, RI ADAMS Accession No. ML101880229 (* No substantial change from SE Input Memo OFFICE LPL1-1\PM LPL1-1\LA NRRlEICB/BC(*) NRRlITSB/BC NRRlSRSB/BC NAME B. K. Vaidva SUttle WKemper REIliott AUlses DATE 7/08/10 7/08/10 06/30/10 7/08/10 7/12/10 OFFICE OGC LPL1-1\BC LPL1-1\PM NAME BHarris NSalgado B. K. Vaidya DATE 7/19/10 7/19/10 7/19/10 Official Record Copy