Text

Issuance of Amendment Revise Calculated Peak Containment Internal Pressure
	ML113220370
Person / Time
Site:	Palisades
Issue date:	01/19/2012
From:	Mahesh Chawla; Plant Licensing Branch III
To:	; Entergy Nuclear Operations
	Chawla M, NRR/DORL, 415-8371
References
	TAC ME6875
	Download: ML113220370 (10)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 19, 2012 Vice President, Operations Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043-9530

SUBJECT:

PALISADES NUCLEAR PLANT - ISSUANCE OF AMENDMENT RE: REVISE CALCULATED PEAK CONTAINMENT INTERNAL PRESSURE (TAC NO. ME6875)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 244 to Renewed Facility Operating license No. DPR-20 for the Palisades Nuclear Plant. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated August 16, 2011, supplemented by letter dated October 6, 2011.

The amendment revises TS Section 5.5.14, "Containment Leak Rate Testing Program" to increase the value of the calculated peak containment internal pressure, P a, from 53 pounds per square inch gauge (psig) to 54.2 psig. This increase is due to an increase in the calculated mass and energy release during the blowdown phase of the design basis loss-of-coolant accident.

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

Sincerely, Mahesh L. Chawla, Project Manager Plant licensing Branch 111-1 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket No. 50-255

Enclosures:

1. Amendment No. 244 to DPR-20

2. Safety Evaluation cc w/encls:

Distribution via listServ

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

DOCKET NO. 50-255 PALISADES NUCLEAR PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 244 License No. DPR-20

1.

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

A.

The application for amendment by Entergy Nuclear Operations, Inc. (the licensee),

dated August 16, 2011, as supplemented by letter dated October 6,2011, 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; 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.

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2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment and Paragraph 2.C. (2) of Renewed Facility Operating license No. DPR-20 is hereby amended to read as follows:

The Technical Specifications contained in Appendix A, as revised through Amendment No. 244, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Shawn A. Williams, Acting Chief Plant licensing Branch 111-1 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating license and Technical Specifications Date of Issuance:

January 19, 2012

ATTACHMENT TO LICENSE AMENDMENT NO. 244 RENEWED FACILITY OPERATING LICENSE NO. DPR-20 DOCKET NO. 50-255 Replace the following page of the Renewed Facility Operating License No. DPR-20 with the attached revised page. The changed area is identified by a marginal line.

REMOVE INSERT Page 3 Page 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 a marginal line indicating the area of change.

REMOVE INSERT 5.0-18 5.0-18

- 3 (1)

Pursuant to Section 104b of the Act, as amended, and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," (a) ENP to possess and use, and (b) ENO to possess, use and operate, the facility as a utilization facility at the designated location in Van Buren County, Michigan, in accordance with the procedures and limitation set forth in this license; (2)

ENO, pursuant to the Act and 10 CFR Parts 40 and 70, to receive, possess, and use source and special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3)

ENO, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use byproduct, source, and special nuclear material as sealed sources for reactor startup, reactor instrumentation, radiation monitoring equipment calibration, and fission detectors in amounts as required; (4) ENO, 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 for sample analysis or instrument calibration, or associated with radioactive apparatus or components; and (5) ENO, 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 operations of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act; 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)

ENO is authorized to operate the facility at steady-state reactor core power levels not in excess of 2565.4 Megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) The Technical Specifications contained in Appendix A, as revised through Amendment No. 244, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) ENO shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the SERs dated 09/01178,03/19/80,02/10/81, 05/26/83, 07/12/85, 01/29/86, 12/03/87, and 05/19/89 and subject to the following provisions:

Renewed License No. DPR-20 Amendment No. 244

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Functions Determination Program (SFDP) (continued)

c.

A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program. the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.14 Containment Leak Rate Testing Program

a.

A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50. Appendix J, Option B. as modified by approved exemptions. This program shall be in accordance with the guidelines of Regulatory Guide 1.163, "Performance-Based Containment Leakage-Test Program." dated September 1995, except that the next Type A test performed after the May 3, 2001, Type A test shall be performed no later than August 3,2012. as modified by the following exceptions:

1.

Leakage rate testing is not necessary after opening the Emergency Escape Air Lock doors for post-test restoration or posHest adjustment of the air lock door seals. However, a seal contact check shall be performed instead.

Emergency Escape Airlock door opening. solely for the purpose of strong back removal and performance of the seal contact check.

does not necessitate additional pressure testing.

2.

Leakage rate testing at Pais not necessary after adjustment of the Personnel Air Lock door seals. However, a between-the-seals test shall be performed at ;:::10 psig instead.

3.

Leakage rate testing frequency for the Containment 4 inch purge exhaust valves, the 8 inch purge exhaust valves, and the 12 inch air room supply valves may be extended up to 60 months based on component performance.

b.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pal is 54.2 psig. The containment design pressure is 55 psig.

c.

The maximum allowable containment leakage rate, La, at Pa, shall be Palisades Nuclear Plant 5.0-18 Amendment No. 489, 4Q4, ~)

~,24G,244

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 244 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-20 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT DOCKET NO. 50-255

1.0 INTRODUCTION

By letter dated August 16,2011 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML112280430), and as supplemented by the letter dated October 6, 2011 (ADAMS Accession Number ML112790383), Entergy Nuclear Operations, Inc.

(the licensee) requested U.S. Nuclear Regulatory Commission (NRC) review and approve the proposed license amendment request.

The amendment revises Technical Specification (TS) Section 5.5.14, "Containment Leak Rate Testing Program" to increase the value of the calculated peak containment internal pressure from 53 pounds per square inch guage (psig) to 54.2 psig. This increase is due to an increase in the calculated mass and energy release during the blowdown phase of the design basis loss of-coolant accident (LOCA). The increase in the predicted mass and energy release is due to the correction of an error in the calculation of the current value of Pa. The regulations at 10 CFR Part 50 Appendix J Option B define P a as the calculated peak containment internal pressure related to the design basis LOCA as specified in the TS and specifies the requirements for containment leakage rate testing.

The supplement dated October 6, 2011, 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 November 15, 2011, (76 FR 707703).

2.0 REGULATORY EVALUATION

No regulation specifically addresses the determination of the mass and energy release into the containment following a postulated design basis accident. The regulations at Title 10 of the Code of Federal Regulations (10 CFR) Part 50 Appendix A, General Design Criteria 16 and 50 address the capability of the containment to withstand the containment pressure resulting from a postulated design basis LOCA.

Enclosure

- 2 General Design Criterion 16, Containment design, specifies that the reactor containment and associated systems shall be provided to establish an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require, General Design Criterion 50, Containment design basis, specifies that the reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed to accommodate, without exceeding (with sufficient margin) the design leakage rate resulting from a design basis LOCA The regulations at 10 CFR Part 50, Appendix J, Option B define Pa as the calculated peak containment internal pressure related to the design basis LOCA as specified in the TS and specifies the requirements for containment leakage rate testing. The requirements of Palisades Nuclear Plant TS 5.5.14, "Containment Leak Rate Testing Program" provide more detailed leakage rate testing requirements.

3.0 TECHNICAL EVALUATION

Westinghouse provided the licensee with the mass and energy release calculations that are the basis for the current value of Pa. The licensee used these mass and energy release values to calculate Pa using the GOTHIC computer code. The licensee's August 16, 2011, letter states that Westinghouse identified a non-conservative LOCA mass and energy release input for the containment response analysis.

In response to a request for additional information, the licensee stated in an October 6, 2011, letter to the NRC that the non-conservative input to the mass and energy release affected the calculation of the blowdown portion of the LOCA calculated with the CEFLASH4A computer code.

The current TS value of Pa is 53 psig. The corrected value of Pa is 54.2 psig. The revised value is less than the containment design pressure which is 55 psig. The containment design pressure contains sufficient margin to ensure containment integrity, In response to a staff request for additional information, the licensee stated that upon approval of this license amendment, the Palisades' Appendix J Type A test procedures will be revised to specify that Pa is 54.2 psig in Section 5.5.14, "Containment Leakage Rate Testing Program", of the Palisades Technical Specifications.

Section 5.5.14 specifies that containment leakage rate testing shall be in accordance with the guidelines of Regulatory Guide 1.163 Revision 0, "Performance-based Containment Leakage Testing Program," dated September 1995. Regulatory Guide 1.163 in turn references NEI 94-01, Revision 0, dated July 21, 1995, "Industry Guideline for Implementing Performance Based Option of 10 CFR 50 Appendix J," (ADAMS Accession No. ML11327A025) prepared by the Nuclear Energy Institute. This document provides methods acceptable to the NRC staff for complying with the provisions of Option B in Appendix J to 10 CFR Part 50. NE194-01 references American National Standards Institute (ANSI) 56.8-1994, "Containment System

- 3 Leakage Testing Requirements," (ADAMS Accession No. ML11327A024) for detailed descriptions of the technical methods and techniques for performing Type A, B, and C tests.

ANSI 56.8-1994 specifies that Type A testing should be performed at a pressure not less than 0.96 times the calculated peak containment pressure (Pa) and not more than containment design pressure. The licensee's October 6, 2011, letter stated that Type A test procedures will be revised to incorporate the new 0.96 times Pa value.

The licensee also stated in the October 6, 2011, letter that Appendix J Type Band C test procedures will not require revision upon approval of this license amendment request. ANSI 56.8-1994 Section 3.3.2 requires that Type Band C testing be performed at a pressure not less than P a and not more than 1.1 times P a. Palisades Nuclear Plant site procedures for Type B and C testing require that testing be performed within a range of pressures that, with the revised Pa, will continue to be within the range of pressures required by ANSI 56.8-1994. Therefore, the Type Band C test procedures will not require revision upon approval of the license amendment.

The NRC staff agrees with the licensee's interpretation of compliance with the requirements of TS 5.5.14.

4.0

SUMMARY

The NRC staff has reviewed and assessed the information provided by the licensee in the August 16, 2011, license amendment request and the supplemental October 6, 2011, letter.

The staff finds the license amendment request to be acceptable in that it complies with GDC 16 and 50, 10 CFR Part 50 Appendix J and the implementation of the proposed change complies with the requirements of Palisades Nuclear Plant TS 5.5.14.

5.0 STATE CONSULTATION

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

6.0 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 or change the 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 (76 FR 70773). 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.

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7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Richard Lobel Date: