ML093420970
| ML093420970 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/01/2010 |
| From: | Kalyanam N Plant Licensing Branch IV |
| To: | Entergy Operations |
| Kalyanam N, NRR/DORL/LPL4, 415-1480 | |
| References | |
| TAC ME0660 | |
| Download: ML093420970 (13) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 1, 2010 Vice President, Operations Arkansas Nuclear One Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802
SUBJECT:
ARKANSAS NUCLEAR ONE, UNIT NO.1 - ISSUANCE OF AMENDMENT RE:
TECHNICAL SPECIFICATION CHANGE TO ADOPT LOW PRESSURE AIR LOCK SEAL TEST (TAC NO. ME0660)
Dear Sir or Madam:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 242 to Renewed Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit NO.1 (ANO-1). The amendment consists of changes to the Technical Specifications (TSs) in response to Entergy Operations, Inc.'s (the licensee) application dated February 16, 2009.
The change modifies TS 5.5.16, "Reactor Building Leakage Rate Testing Program," which currently contains reactor building leak rate criteria for overall Type A, B, and C testing, but does not specify criteria for Type B air lock leakage testing. The licensee has proposed to add criteria for overall air lock leakage testing and adopt a low pressure test method relevant to the air lock door seals, consistent with NUREG-1430, Revision 3.1, "Standard Technical Specifications (STS) for Babcock & Wilcox Plants."
A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, N. Kaly Kalyanam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-313
Enclosures:
- 1. Amendment No. 242 to DPR-51
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY OPERATIONS, INC.
DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT NO.1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 242 Renewed License No. DPR-51
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee), dated February 16, 2009, 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 license 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
- 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-51 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 242, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMIVIISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License No. DPR-51 and Technical Specifications Date of Issuance: February 1, 2010
ATTACHMENT TO LICENSE AMENDMENT NO. 242 RENEWED FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313 Replace the following pages of the Renewed Facility Operating License No. DPR-51 and 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.
Operating License REMOVE INSERT 3
3 Technical Specifications REMOVE INSERT 5.0-18 5.0-18
- 3 (5)
EOI, 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 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; (6)
EOI, 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 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; 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 EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 242, are hereby incorporated in the renewed license.
EOI shall operate the facility in accordance with the Technical Specifications.
(3)
Safety Analysis Report The licensee's SAR supplement submitted pursuant to 10 CFR 54.21 (d),
as revised on March 14,2001, describes certain future inspection activities to be completed before the period of extended operation. The licensee shall complete these activities no later than May 20, 2014.
(4)
Physical Protection EOI shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Arkansas Nuclear One Physical Security Plan, Training and Qualifications Plan, and Safeguards Contingency Plan," as submitted on May 4,2006.
Renewed License No. DPR-51 Amendment No. 242 Revised by letter dated July 18, 2007
Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.16 Reactor Building Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the reactor building 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 contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, except that the next Type A test performed after the April 16, 1992 Type A test shall be performed no later than April 15, 2007.
In addition, the reactor building purge supply and exhaust isolation valves shall be leakage rate tested once prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days.
The peak calculated reactor building internal pressure for the design basis loss of coolant accident, Pa, is 54 psig.
The maximum allowable reactor building leakage rate, La, shall be 0.20% of containment air weight per day at Pa.
Leakage rate acceptance criteria are:
- a.
Reactor Building leakage rate acceptance criteria is ::;; 1.0 La. During the first unit startup following each test performed in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and < 0.75 La for Type A tests.
- b.
Air lock testing acceptance criteria are:
- 1.
Overall air lock leakage rate is :s; 0.05 La when tested at ~ Pa;
- 2.
For each door, leakage rate is :s; 0.01 La when tested at ~ 10 psig.
The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Reactor Building Leakage Rate Testing Program.
The provisions of SR 3.0.3 are applicable to the Reactor Building Leakage Rate Testing Program.
ANO-1 5.0-18 Amendment No. 24a,~,~,~,242
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 242 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS, INC.
ARKANSAS NUCLEAR ONE, UNIT NO.1 DOCKET NO. 50-313
1.0 INTRODUCTION
By application dated February 16, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090480125), Entergy Operations, Inc. (the licensee),
requested changes to the Technical Specifications (TSs) for Arkansas Nuclear One, Unit NO.1 (ANO-1).
The amendment revises TS 5.5.16, "Reactor Building Leakage Rate Testing Program," which currently contains reactor building leak rate criteria for overall Type A, B, and C testing, but does not specify criteria for Type B air lock leakage testing. The proposed changes would revise the criteria for overall air lock leakage testing and adopt a low pressure test method relevant to the air lock door seals. The proposed change is consistent with NUREG-1430, Revision 3, "Standard Technical Specifications (STS) for Babcock & Wilcox Plants." Additionally, many domestic commercial nuclear plants, including Arkansas Nuclear One, Unit 2 (ANO-2), contain both the criteria for the overall air lock leak test and the low pressure test method associated with the door seals.
2.0 REGULATORY EVALUATION
Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The TSs ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The U.S. Nuclear Regulatory Commission's (NRC) regulatory requirements related to the content of the TSs are contained in Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR 50.36) that requires that the TSs include items in the following categories:
(1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. However, the rule does not specify the particular requirements to be included in a plant's TSs.
- 2 Pursuant to 10 CFR 50.54(0), plants are required to comply with the requirements set forth 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors." Three major test types are described in Appendix J. Type A tests refer to overall leak rate testing of the containment building structure. Type Band C tests are often referred to as "local" leak rate tests. Type B tests apply to containment boundaries other than valves including doors and hatches with resilient seals or gaskets. Type C tests apply to containment isolation valve leakage.
General Design Criteria (GDC) 16, 50, 51, 52, and 53 govern the requirements for the reactor containment structure. These requirements include measures for leak rate testing of the structure as a whole, and for leak testing of components that penetrate the structure.
Criterion 16--Containment design. 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.
Criterion 50--Containment design basis. The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident.
Criterion 51--Fracture prevention of containment pressure boundary. The reactor containment boundary shall be designed with sufficient margin to assure that under operating, maintenance, testing, and postulated accident conditions (1) its ferritic materials behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.
Criterion 52--Capability for containment leakage rate testing. The reactor containment and other equipment which may be sUbjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure.
Criterion 53--Provisions for containment testing and inspection. The reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leaktightness of penetrations which have resilient seals and expansion bellows.
NRC Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program,"
endorses the use of American National Standards Institute/American Nuclear Society
- 3 (ANSI/ANS) 56.8-1994, Containment System Leakage Testing Requirements, to meet the requirements of Appendix J. RG 1.163 also endorses Nuclear Energy Institute (NEI) 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J."
The NRC staff reviewed the proposed changes for compliance with 10 CFR 50.36 and agreement with the precedent as established in NUREG-1430, Revision 3, "Standard Technical Specifications, Babcock & Wilcox Plants," dated June 2004. In general, licensees cannot justify TS changes solely on the basis of adopting the Standard Technical Specification (STS) model. Licensees may revise the TSs to adopt the improved STS format and content, provided that a plant specific review supports a finding of continued adequate safety because: (1) the change is editorial, administrative, or provides clarification (Le., no requirements are materially altered); (2) the change is more restrictive than the licensee's current requirement; or (3) the change is less restrictive than the licensee's current requirement, but nonetheless still affords adequate assurance of safety when judged against current regulatory standards.
3.0 TECHNICAL EVALUATION
3.1 Proposed TS Change The current TS 5.5.16 states, in part, that:
Reactor Building leakage rate acceptance criteria is :s; 1.0La. During the first unit startup following each test performed in accordance with this program, the leakage rate acceptance criteria are < 0.60Lafor the Type B and Type C tests and < O. 75Lafor Type A tests.
This TS Section would be modified to provide editorial spacing between the numerical value and maximum allowable reactor building leakage rate (La).
The proposed TS 5.5.16 requirement for reactor building leak rate acceptance criteria will be sub-bulleted as part "a." and a new part "b." will be added to describe air lock testing acceptance criteria. Both part "a." and part "b." will be under a new sub-header and, consistent with the NUREG-1430, Revision 3, STS for Babcock & Wilcox Plants, will read as follows:
Leakage rate acceptance criteria are:
- a.
Reactor BUilding leakage rate acceptance criteria is ~ 1.0 La. During the first unit startup following each test performed in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and < 0.75 La for Type A tests.
- b.
Air lock testing acceptance criteria are:
- 1.
Overall air lock leakage rate is s 0.05 La when tested at ~ Pa;
- 2.
For each door, leakage rate is s 0.01 La when tested at ~ 10 psig.
- 4 Pais the peak calculated reactor building internal pressure for the design basis loss-of-coolant accident.
The proposed change describes both the overall local leak test criteria and criteria for the air lock door seals, consistent with NUREG-1430, Revision 3.
3.2 Background
As stated in ANSI/ANS 56.8-1994, Section 3.3.2, Type Band C tests shall be conducted at a differential pressure not less than Pa, except on air lock door seals, which may have a lower pressure specified in the plant licensing basis.
As stated in NEI 94-01, Revision 0, Section 10.2.2.1, "Test Interval,"
Containment airlock(s) shall be tested at an internal pressure of not less than
[Pacl prior to a preoperational Type A test. Subsequent periodic tests shall be performed at a frequency of at least once per 30 months. Containment airlocks tests should be performed in accordance with ANSI/ANS-56.8-1994.
Door seals are not required to be tested when containment integrity is not required, however they must be tested prior to reestablishing containment integrity. Door seals shall be tested at [Pae], or at a pressure stated in the plant Technical Specifications.
In its letter dated February 16, 2009, the licensee stated that:
Entergy maintains a reactor containment leakage test program that meets the requirements stated above. However, Entergy has not yet adopted the low pressure air lock door seal test method described in the STS as permitted by ANSI/ANS 56.8-1994 (Reference 4) and NEI 94-01, Rev[ision] 0 (Reference 5).
Therefore, the proposed change will revise the current Reactor Building Leakage Rate Testing Program of ANO-1 TS 5.5.16 to provide criteria for both the 30 month pressure test at Pa and the low pressure test for door seals.
3.3
NRC Staff Evaluation
In its application dated February 16, 2009, the licensee stated that the information provided in its letter dated April 25, 1996, with respect to similar changes requested for the ANO-2 air lock
- 5 doors and approved in Amendment No. 175, dated September 26, 1996 (ADAMS Accession No. ML021560122), is also applicable to ANO-1. The licensee states in its letter dated February 16, 2009:
In response to an NRC Request for Additional Information (RAI), Entergy stated in letter dated April 25, 1996 (Reference 2):
TS 4.6.1.3.1.a is intended to detect degradation of the airlock door seals following the opening and closing of the airlock doors. The proposed air lock door seal leakage rate limit is 1% of the total containment leakage assumed in accident analysis.
TS 4.6.1.3.1.b requires the performance of an overall airlock leakage test.
This test has a limit of 5% of the total containment leakage assumed in the accident analysis. This test is a Type B test and the results are added to the total of all Type Band C tests. The total of the Type Band C tests must be less than 60% of the total allowed containment leakage. This ensures that the leakage through the airlock will not exceed the leakage assumed in the existing analyses for all evaluated accidents.
The licensee further stated that while Type B testing is applicable to ANO-1, the ANO-1 TSs do not currently contain the criteria for the test and, therefore, the licensee is proposing to include it in the new part "b.1" ofTS 5.5.16.
The change proposed by new part "b.2" of TS 5.5.16 is intended to detect degradation of the air lock door seals following the opening and closing of the air lock doors. The proposed air lock door seal leakage rate limit is 1 percent of the total containment leakage assumed in accident analysis. The door seal test is intended to be a gross test to verify that the door seals were not damaged during the opening and closing cycles. The test does not replace the required overall barrel leakage test. As discussed in the licensee's letter dated April 4, 1995, requesting a similar change for ANO-2, the air lock vendor (Trentec), recommended that a test pressure of 3 pounds per square inch gauge (psig) is sufficient to meet this test requirement. The licensee has selected a test pressure of 10 psig.
The NRC has determined that licensee's proposed changes meet the requirements of 10 CFR 50.54(0) and 10 CFR 50, Appendix J and are consistent with NUREG-1430, Revision 3, STS for Babcock & Wilcox plants and, therefore, are acceptable. In addition, the staff concludes that the licensee has provided adequate justification to support the requested changes and reasonable assurance that ANO-1 will be able to comply with the regulatory requirements and, therefore, meets 10 CFR 50.36. Therefore, the NRC staff concludes that the proposed TS changes are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment. The State official had no comments.
- 6
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. 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 published in the Federal Register on April 21, 2009 (74 FR 18253). 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: Brian Lee Date: February 1, 2010
February 1, 2010 Vice President, Operations Arkansas Nuclear One Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802
SUBJECT:
ARKANSAS NUCLEAR ONE, UNIT NO.1 - ISSUANCE OF AMENDMENT RE:
TECHNICAL SPECIFICATION CHANGE TO ADOPT LOW PRESSURE AIR LOCK SEAL TEST (TAC NO. ME0660)
Dear Sir or Madam:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 242 to Renewed Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit NO.1 (ANO-1). The amendment consists of changes to the Technical Specifications (TSs) in response to Entergy Operations, Inc.'s (the licensee) application dated February 16, 2009.
The change modifies TS 5.5.16, "Reactor Building Leakage Rate Testing Program," which currently contains reactor building leak rate criteria for overall Type A, B, and C testing, but does not specify criteria for Type B air lock leakage testing. The licensee has proposed to add criteria for overall air lock leakage testing and adopt a low pressure test method relevant to the air lock door seals, consistent with NUREG-1430, Revision 3.1, "Standard Technical Specifications (STS) for Babcock &Wilcox Plants."
A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, IRA!
N. Kaly Kalyanam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-313
Enclosures:
- 1. Amendment No. 242 to DPR-51
- 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
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