ML100621169
| ML100621169 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 03/31/2010 |
| From: | Feintuch K Plant Licensing Branch III |
| To: | Costanzo C Florida Power & Light Group |
| Feintuch K, NRR/DORL/LPL3-1, 415-3079 | |
| References | |
| TAC ME0873, TAC ME0874 | |
| Download: ML100621169 (16) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 31, -2010 Mr. Christopher R. Costanzo Vice President Duane Arnold Energy Center 3277 DAEC Road Palo,IA 52324-9785
SUBJECT:
DUANE ARNOLD ENERGY CENTER - ISSUANCE OF AMENDMENT PERTAINING TO CHANGES TO TECHNICAL SPECIFICATIONS 5.5.12 (PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM) AND 3.6.1.3 (PRIMARY CONTAINMENT ISOLATION VALUES), ASSOCIATED WITH THE MAIN STEAM ISOLATION VALVE (MSIV) LEAKAGE TESTING REQUIREMENTS (TAC NOS. ME0873 AND ME0874)
Dear Mr. Costanzo:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 276 to Facility Operating License No. DPR-49 for the Duane Arnold Energy Center (DAEC, the licensee). This amendment consists of a change to the Technical Specifications (TS) in response to your application dated March 4, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090680040).
The letter dated March 4, 2009 was sent by the licensee under the company name FPL Energy Duane Arnold, LLC. Effective April 16, 2009, the legal name of the company became NextEra Energy Duane Arnold, LLC, according to a letter from FPL Energy dated March 24, 2009. In a letter dated April 17, 2009, the licensee requested a license amendment to change its name on the license to NextEra Energy Duane Arnold, LLC. In a letter dated November 13, 2009, the NRC issued Amendment No. 275 to Facility Operating License No. DPR-49, which changed the name of the licensee to NextEra Energy Duane Arnold, LLC.
Amendment No. 276 changes the Duane Arnold Energy Center (Duane Arnold) Technical Specifications (TS) Section 5.5.12 (Primary Containment Leakage Rate Testing Program) to reflect the exemption to specific requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix J, which was granted by the I\\IRC in a letter dated March 29,2010 (ADAMS Accession No. ML100501057). Specifically, the exemption permits the licensee to exclude main steam isolation valve (MSIV) leakage from the overall integrated leakage rate test measurements required by Section III.A of Appendix J, Option B.
This change eliminated the double counting of the leakages of interest, once as part of the actual containment leakage and again as part of the main steam pathway leakage used in dose calculations.
In addition, Amendment NO.276 also changes TS Section 3.6.1.3 (Primary Containment Isolation Valves) to remove the repair criterion for MSIVs that fail their as-found leakage rate acceptance criterion located in the current Surveillance Requirement 3.6.1.3.9.
C. Costanzo
- 2 A copy of the Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, j(~;J~J(
Karl Feintuch, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-331
Enclosures:
- 1. Amendment No. 276 to License No. DPR-49
- 2. Safety Evaluation cc w/encls: Distribution via ListServ
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY DUANE ARNOLD, LLC DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 276 License No. DPR-49
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by FPL Energy Duane Arnold, LLC dated March 4, 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 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 Facility Operating License No. DPR-49 is hereby amended to read as follows:
- 2 (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, are hereby incorporated in the license. NextEra Energy Duane Arnold, LLC, 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 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION r-¥£-r Ro ert Pascarelli, 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: March 31,2010
ATTACHMENT TO LICENSE AMENDMENT NO. 276 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace the following page of Renewed Facility Operating License DPR-49 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 3
3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain a marginal line indicating the area of change.
REMOVE INSERT 3.6-15 3.6-15 5.0-17 5.0-17 5.0-18 5.0-18
- 3 2.8.(2) NextEra Energy Duane Arnold, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time 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 as of June 1992 and as supplemented by letters dated March 26,1993, and November 17,2000.
2.8.(3) NextEra Energy Duane Arnold, 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; 2.8.(4) NextEra Energy Duane Arnold, 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 radioactive apparatus components; 2.8.(5) NextEra Energy Duane Arnold, LLC, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This 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:
Maximum Power Level 2.C.(1) NextEra Energy Duane Arnold, LLC is authorized to operate the Duane Arnold Energy Center at steady state reactor core power levels not in excess of 1912 megawatts (thermal).
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, are hereby incorporated in the license. NextEra Energy Duane Arnold, LLC shall operate the facility in accordance with the Technical Specifications.
PCIVs 3.6.1.3 SR 3.6.1.3.9 Verify leakage rate through each MSIV is
- 5: 100 seth and that the combined maximum pathway leakage rate for all four main steam lines is ~ 200 seth when tested at ~ 24 psig.
FREQUENCY In accordance with the Primary Containment Leakage Rate Testing Program DAEC 3.6-15 Amendment No. 276
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)
- 2.
Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;
- 3.
Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
- 4.
Other appropriate limitations and remedial or compensatory actions.
- b.
A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
- 1.
A required system redundant to system(s) supported by the inoperable support system is also inoperable; or
- 2.
A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or
- 3.
A required system redundant to support system(s) for the supported systems (1) and (2) above is also inoperable.
- c.
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.
5.5.12 Primary Containment Leakage Rate Testing Program
- a.
A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.
- b.
This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak Test Program," dated September 1995, as modified by the following exceptions to NEI 94-01, Rev. 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J":
- 1.
The first Type A test after the September 1993 Type A test shall be performed no later than September 2008.
(continued)
DAEC 5.0-17 Amendment No. 276
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)
- 2.
Exemption from Section liLA of 10 CFR Part 50, Appendix J, Option B, to allow the contribution from Main Steam pathway leakage to be excluded from the overall integrated leakage rate from Type A tests.
- 3.
Exemption from Section III.B of 10 CFR Part 50, Appendix J, Option B, to allow the contribution from Main Steam pathway leakage to be excluded from the sum of the leakage rates from Type B and Type C tests.
- c.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 45.7 psig.
- d.
The maximum allowable primary containment leakage rate, La, at Pa, shall be 2.0% of primary containment air weight per day.
- e.
Leakage Rate acceptance criteria are:
- 1.
Primary Containment leakage rate acceptance criterion is ::; 1.0 La. During the first startup following testing 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 the Type A tests; and
- 2.
The air lock testing acceptance criterion is overall air lock leakage rate::; 0.05 La when tested at 2 Pa.
- f.
The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.
5.5.13 Control Building Envelope Habitability Program A Control Building Envelope (CBE) Habitability Program shall be established and implemented to ensure that CBE habitability is maintained such that, with an OPERABLE Standby Filter Unit System, CBE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CBE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:
- a.
The definition of the CBE and the CBE boundary.
- b.
Requirements for maintaining the CBE boundary in its design condition including configuration control and preventive maintenance.
(continued)
DAEC 5.0-18 Amendment No. 276
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 276 TO FACILITY OPERATING LICENSE NO. DPR-49 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331
1.0 INTRODUCTION
By application dated March 4, 2009, FPL Energy Duane Arnold, LLC (the licensee, later renamed NextEra Energy Duane Arnold by Amendment No. 275 dated November 13, 2009, to the Facility Operating License DPR-49) requested: (1) an exemption for the Duane Arnold Energy Center (Duane Arnold) from certain requirements of Appendix J to 10 CFR Part 50; (2) a license amendment to Duane Arnold Energy Center (Duane Arnold) Technical Specifications (TS) Section 5.5.12 (Primary Containment Leakage Rate Testing Program) for Duane Arnold that would reflect the exemption to specific requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix J; and (3) a license amendment (not related to the exemption) to TS Section 3.6.1.3 (Primary Containment Isolation Valves) to remove the repair criterion for main steam isolation valves that fail their as-found leakage rate acceptance criterion found in current Surveillance Requirement (SR) 3.6.1.3.9.
This Safety Evaluation addresses all three actions requested by the licensee. The Commission's evaluation of Duane Arnold's exemption request pursuant to 10 CFR 50.12 is contained in a letter to the licensee dated March 29, 2010.
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations (10 CFR), Section 50.54(0), requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR Part 50. Appendix J specifies the leakage test requirements, schedules, and acceptance criteria for tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. Option B of Appendix J is titled "Performance-Based Requirements." Option B,Section II I.A., requires that the overall integrated leakage rate must not exceed the allowable leakage rate (La) with margin, as specified in the TS. The overall integrated leakage rate, as specified in the 10 CFR Part 50, Appendix J definitions, includes the contribution from main steam isolation valve (MSIV) leakage. The licensee is requesting a permanent exemption from Option B, Section 1I1.A.,
requirements to permit exclusion of MSIV leakage from the overall integrated leakage rate test measurement.
- 2 Option B,Section III.B., requires the sum of the leakage rates of all Type B and Type C local leakage rate tests to be less than the performance criterion (La) with margin, as specified in the TS. The licensee is also requesting exemption from this requirement, to permit exclusion of the MSIV leakage rates from the sum of the Type B and Type C test leakage rates.
Pursuant to 10 CFR 50.12, the Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of 10 CFR Part 50 when (1) the exemptions are authorized by law, will not present an undue risk to public health and safety, and are consistent with the common defense and security, and (2) when special circumstances are present. The regulation lists valid types of special circumstances. For example, special circumstances are present whenever, according to 10 CFR Part 50.12(a)(2)(ii), "Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule...."
1a CFR Part 50 Appendix J, Option B,Section V.B.3., requires that the regulatory guide or other implementation document used by a licensee to develop a performance-based leakage testing program must be included, by general reference, in the plant TS. Duane Arnold TS 5.5.12 requires that leakage rate testing be performed as required by 10 CFR Part 50, Appendix J, Option B, as modified by the approved exemption, and in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, with one exception listed in the TS. This RG endorses, with certain exceptions, Nuclear Energy Institute (NEI) report NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 26, 1995.
Section 182a of the Atomic Energy Act (Act) requires applicants for nuclear power plant operating licenses to include TS as part of the license. These TS are derived from the plant safety analyses.
10 CFR 50.36 contains the requirements for the content of TS. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation:
(1) safety limits (SLs), limiting safety system settings (LSSSs), and limiting control settings; (2) limiting conditions for operation (LCOs); (3) Surveillance Requirements (SRs); (4) design features; and (5) administrative controls.
10 CFR 50.36 does not specify each particular requirement to be included in a plant's TSs, nor does it specify the format of a plant's TS. Rather, the NRC publishes generic guidance on TS format and content. The NRC published a set of Standard Technical Specifications (STS) in NUREG-1433. Revision 3 "Standard Technical Specifications, General Electric Plants, BWR/4."
The STS are a guide to what a plant's TS should contain with regard to format and content. The STS are not requirements in a regulatory sense, but licensees adopting portions of the improved STS to existing technical specifications should adopt all related requirements, as applicable, to achieve a high degree of standardization and consistency.
The NRC staff reviewed the proposed changes for compliance with 10 CFR 50.36 and agreement with the precedent as established in STS. In general, licensees cannot justify technical specification changes solely on the basis of adopting the model STS. To ensure this, the NRC staff makes a determination that proposed changes maintain adequate safety.
- 3 The licensee is proposing to add, in TS 5.5.12, two more exceptions from the guidelines of RG 1.163 and NEI 94-01, which would implement the exemption that the licensee is requesting; The licensee is also proposing a change to the leakage rate test acceptance criteria in TS 3.6.1.3, to remove the repair criterion for MSIVs that fail their as-found leakage rate acceptance criterion located in current Surveillance Requirement (SR) 3.6.1.3.9. These changes are consistent with the Improved Standard TS (ISTS), NUREG-1433.
10 CFR 50.65 (Maintenance Rule), establishes requirements governing the performance of systems, structures, and components (SSCs) that meet certain functional criteria within the licensing basis. MSIVs meet these functional criteria under the design and licensing basis of Duane Arnold. Accordingly, performance goals have been established for the MSIVs within the Duane Arnold Maintenance Rule program, which include as-found leakage limits commensurate with the TS values.
3.0 TECHNICAL EVALUATION
3.1 10 CFR 50, Appendix J Exemption Request In letters dated July 31, 2001 and April 16, 2001, from the Nuclear Regulatory Commission (NRC) to the licensee, the NRC approved the use of the Alternative Source Term (AST) in the calculations of the radiological dose consequences of design basis accidents (DBA) for Duane Arnold. In the NRC staff's safety evaluation, dated July 31,2001, the NRC staff accepted that main steam pathway leakage is treated separately from the remainder of the assumed leakage from primary containment in the LOCA analysis.
The MSIV leakage effluent has a different pathway to the environment, when compared to a typical containment penetration. It is not directed into the secondary containment and filtered through the standby gas treatment system as is other containment leakage. Instead, the main steam leakage is collected and treated via an alternative leakage treatment (ALT) path having different mitigation characteristics.
In performing accident analyses, it is appropriate to group various leakage effluents according to the treatment they receive before being released to the environment (e.g., from main steam pathways). The proposed exemption would more appropriately permit ALT pathway leakage to be independently grouped with its unique leakage limits. In this manner, the Duane Arnold containment leakage testing program will be made more consistent with the limiting assumptions used in the associated accident consequence analyses.
The licensee has analyzed the MSIV and main steam pathway leakage separately from the overall containment integrated leakage, local leakage across pressure retaining, leakage limiting boundaries, and containment isolation valve leakage in its dose consequence analyses.
Specifically, the AST DBA analyses use the main steam piping, main steam drain lines, and main condenser as an alternate means for MSIV and main steam pathway leakage treatment.
Based on the foregoing, the separation of the main steam pathways from the other containment leakage pathways is warranted because a separate radiological consequence term has been provided for these pathways. The revised design basis radiological consequences analyses address these pathways as individual factors, exclusive of the primary containment leakage.
- 4 Therefore, the NRC staff finds the proposed exemption from Appendix J, to separate MSIV leakage from other containment leakqge, to be acceptable.
The NRC staff's evaluation of Duane Arnold's 10 CFR 50.12 exemption request is contained in the March 29, 2010, letter to the licensee.
3.2 Changes to TS 5.5.12, Primary Containment Leakage Rate Testing The licensee proposes to revise TS 5.5.12, "Primary Containment Leakage Rate Testing Program," to include the 10 CFR 50.12 exemption to 10 CFR Part 50, Appendix J discussed above, which exclude the main steam pathway leakage contribution to both the overall containment leakage for Type A tests and the total leakage of individual components tested as part of Type Band C tests. These changes are included with the other previously-approved exception to the guidelines contained in NEI 94-01, Rev. O. The staff finds these changes to be acceptable because they conform to the requested exemption.
In addition, TS 5.5.12 is being re-formatted into sub-sections and sub-parts, in order to better accommodate the addition of these new exceptions. This new format is consistent with the current version of the ISTS (NUREG-1433, Rev. 3.1). The staff considers these formatting changes to be editorial in nature and acceptable.
3.3 Changes to TS 3.6.1.3, Primary Containment Isolation Valves The licensee proposes to make a change to SR 3.6.1.3.9, which will remove the current as-left repair criterion that is invoked whenever a MSIV fails its individual local leakage rate test as found acceptance criterion. Specifically, SR 3.6.1.3.9 states any MSIV that fails the as-found leakage rate criterion of ~ (less than or equal to) 100 standard cubic feet per hour (scfh) must be repaired to an as-left leakage rate criterion of ~ (less than or equal to) 11.5 scfh. Neither the pertinent regulations, 10 CFR Part 50, Appendix J, Option B, nor the associated guidance documents (RG 1.163) specify an as-left repair criterion for MSIVs that fail their individual as found leakage limit and thus, no exemption pursuant to 10 CFR 50.12 is required for this proposed change.
The current as-left leakage limit (repair criterion) for MSIVs was added to the Duane Arnold TS as a prudent measure as Duane Arnold was among the first plants to adopt the BWROG report (NEDC-31858P, Revision 2) for increasing MSIV leakage limits and eliminating the MSIV Leakage Control System in 1994-95, when industry operating experience with MSIV leakage was not extensive. Since that time, performance has improved significantly and later plants have been approved by the NRC to increase their allowable MSIV (or main steam pathway) leakage using the BWROG report, but without specifying a repair criterion within the TS. The licensee states that this operating experience is also true for Duane Arnold and its Appendix J Program records indicate that no MSIV has failed an as-found leakqge test in the last 10 years.
In addition, the licensee indicates that failure of a MSIV to pass an as-found leakage test would constitute a "Maintenance Rule Functional Failure" in the Duane Arnold Maintenance Rule Program and would be entered into the Corrective Action Program for resolution. It is the NRC
- 5 staff's determination that it is not necessary to have such a prescriptive repair criterion in the TS to achieve the underlying purpose of proper maintenance and overall reliable performance of the MSIVs.
Furthermore, since the associated analyses of radiological release consequences from DBAs do not use this as-left criterion as an input to those calculations, specifying a repair criterion within the TS is overly conservative, is not required by 10 CFR 50.36(c)(3), and is not consistent with the applicable Improved Standard Technical Specifications (ISTS), NUREG-1433.
Therefore, the NRC staff finds the proposed change, to remove the repair criterion for MSIVs that fail their as-found leakage rate acceptance criterion found in current SR 3.6.1.3.9, to be acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Iowa State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
S 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 issued on June 30,2009 (74 FR 31324). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). In addition, the amendment changes recordkeeping, reporting, or administrative procedures. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). 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.
With respect to Duane Arnold's request for exemption from certain requirements in 10 CFR Part 50, Appendix J, Option B, the commission prepared an environmental assessment, and determined pursuant to 10 CFR 51.32, that granting of the exemption would not have a significant effect on the quality of the human environment (75 FR 13318)
6.0 CONCLUSION
Based on the forgoing, the NRC staff finds: (1) the proposed exemption from 10 CFR Part 50, Appendix J, (2) the associated change to TS 5.5.12, and (3) the unassociated change to TS 3.6.1.3, to be acceptable.
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
- 6 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: March 31,2010
C. Costanzo
- 2 A copy of the Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, lRAI Karl Feintuch, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-331
Enclosures:
- 1. Amendment No. 276 to License No. DPR-49
- 2. Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION:
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