ML110600678
| ML110600678 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/18/2011 |
| From: | Balwant Singal Plant Licensing Branch IV |
| To: | Reddemann M Energy Northwest |
| Wilkins, L E, NRR/DORL/LPL4, 415-1377 | |
| References | |
| TAC ME3713 | |
| Download: ML110600678 (15) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 March 18, 2011 Mr. Mark E. Reddemann Chief Executive Officer Energy Northwest P.O. Box 968 (Mail Drop 1023)
Richland, WA 99352-0968
SUBJECT:
COLUMBIA GENERATING STATION -ISSUANCE OF AMENDMENT RE:
CHANGE TO TECHNICAL SPECIFICATIONS RELATING TO TRAVERSING IN-CORE PROBE CONTAINMENT ISOLATION INSTRUMENTATION (TAC NO.
ME3713)
Dear Mr. Reddemann:
The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 220 to Facility Operating License No. NPF-21 for the Columbia Generating Station. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated March 29,2010, as supplemented by letter dated January 14, 2011.
The amendment revises TS 3.3.6.1, "Primary Containment Isolation Instrumentation," by deleting channel check Surveillance Requirement 3.3.6.1.1 from TS Table 3.3.6.1-1, "Primary Containment Isolation Instrumentation," for the traversing in-core probe isolation instrumentation.
A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely,
/
.~
Balwant K. Sing ai, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosures:
- 1. Amendment No. 220 to NPF-21
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 220 License No. NPF-21
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Energy Northwest (licensee), dated March 29, 2010, as supplemented by letter dated January 14, 2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's 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.
Enclosure 1
-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 Facility Operating License No. NPF-21 is hereby amended to read as follows:
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 220 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3. The license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Au-f~ ~~
Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License No. NPF-21 and Technical Specifications Date of Issuance: March 18, 2011
ATTACHMENT TO LICENSE AMENDMENT NO. 220 FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 Replace the following pages of the Facility Operating License No. NPF-21 and Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Facility Operating License REMOVE INSERT
-3 Technical Specification REMOVE INSERT 3.3.6.1-8 3.3.6.1-8
- 3 (3) 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; (4) Pursuant to the Act and 10 CFR Parts 30,40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
(6) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to store byproduct, source and special nuclear materials not intended for use at Columbia Generating Station. The materials shall be no more than 9 sealed neutron radiation sources designed for insertion into pressurized water reactors and no more than 40 sealed beta radiation sources designed for use in area radiation monitors. The total inventory shall not exceed 24 microcuries of strontium-90, 20 microcuries of uranium-235, 30 curies of plutonium-238, and 3 curies of americium-241.
C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3486 megawatts thermal). Items in Attachment 1 shall be completed as specified. Attachment 1 is hereby incorporated into this license.
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 220 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(a) For Surveillance Requirements (SRs) not previously performed by existing SRs or other plant tests, the requirement will be considered met on the implementation date and the next required test will be at the interval specified in the Technical Specifications as revised in Amendment No. 149.
Amendment No. 220
Pr i illd r'y Con trJ i nllH~n t Isolation lnstruillentar,ion 3.3.6.1
",[)!e 3.3. h. 1-) (page 4 of 4)
Pr1 Jr." ry Containment Isoldl;Ofl Ifl:; t r,lme ~ t d t IOn APPL reMILE CONDITIONS MonfS Cil ilEOUiRED RE'~REN(£D OThER r;HAN~tLS -ROM SPEClflEQ PER TRIP REOUIREO SURVi:ILLANCr ALLOWAiH E FIJ~C r I ON COND:TIONS SYSTE'~ Act I o~ C.l REQUIREMENTS VALUE
~. R"',R :,DC 51' tf,o' Iso1atiofl
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- c. Heilt fxc!'Jn<jer J I per room F SR 3. J. 6 ..1. J Area :ill L3.6. \.4 Temperature- High SR 3,3.6. I .6 Room ')0:' Ar'"d ( 140"F ROO1i 501 Ari!o .s. [60*F R'Jom 60':1 ArQa S. 150" Room 606 Ar!'4 ~ 140*'
2( d) SR C. ReactO" Ve$sc~ 3.4. ') J 3.3.n.!.l L 9.5 inches
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SR },).6.1.6
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- d. L *~8 l:1cneS I-iat!'r t evel Low 5R 3.3.(1, .4 low. Leve! 2 SR 3, J, 6. .6 IL Dry'",,, 11 F r,:!:;) ;)l_re \ .2.3 2 (i SR 3 j ,r,) , .: of.- i ! . liB p5;g rtlc;tl SR ) J 6 4 SR ),3 6.1.6 (dl Only one trip 5y5te~ regYirQd In HOCES 4 dnJ 5 wllh QHA Shutdown C0011n9 System integ"ity
!1ldWU ned.
Columbia Generating Station 3.3,6, [*8 Amendment No. 220 149,161,169,209
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 220 TO FACILITY OPERATING LICENSE NO. NPF-21 ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397
1.0 INTRODUCTION
By application dated March 29, 2010, as supplemented by letter dated January 14, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML100990162 and ML110250476, respectively), Energy Northwest (the licensee) requested changes to the Technical Specifications (TSs) (Appendix A to Facility Operating License No. NPF-21) for the Columbia Generating Station (CGS). The requested changes would delete channel check Surveillance Requirements (SRs) for the traversing in-core probe (TIP) isolation instrumentation associated with TS 3.3.6.1, "Primary Containment Isolation Instrumentation. "
Specifically, the licensee proposed to remove channel check SR 3.3.6.1.1 from TS Table 3.3.6.1-1, "Primary Containment Isolation Instrumentation," for Function 6.a, "Traversing In-core Probe Isolation - Reactor Vessel Water Level- Low Low, Level 2," and Function 6.b, "Traversing In-core Probe Isolation - Drywell Pressure - High."
The supplemental letter dated January 14, 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 June 1, 2010 (75 FR 30444).
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 NRC's regulatory requirements related to the content of the TSs are contained in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36, "Technical specifications,"
which requires that the TSs include items in the following specific categories: (1) safety limits, Enclosure 2
-2 limiting safety systems settings, and limiting control settings; (2) limiting conditions for operations (LCOs); (3) SRs; (4) design features; and (5) administrative controls.
The regulations in 10 CFR 50.36(c)(3) specify that, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."
General Design Criterion (GDC) 13, "Instrumentation and control," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," states the following:
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.
GDC 56, "Primary containment isolation," states the following:
Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:
(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic
- 3 isolation valves shall be designed to take the position that provides greater safety.
The NRC's guidance for the format and content of licensee TSs are found in NUREG-1433, "Standard Technical Specifications, General Electric Plants BWR/4," Revision 3.0 (STS).
3.0 TECHNICAL EVALUATION
3.1 Background The containment systems are designed to prevent and mitigate the release of fission products to the environment during and after a design-basis accident. If a fission product release to the environment does occur, the design of the containment is such that the exposure limits of 10 CFR Part 100, "Reactor Site Criteria," will not be exceeded. Containment isolation mechanisms such as valves serve as barriers between fluids inside and outside containment.
The licensee added channel check SRs to the TSs via Amendment No. 208 dated September 12, 2008 1* Amendment No. 208 incorporated a number of Technical Specification Task Force (TSTF) change travelers, including TSTF-306, Revision 2, "Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration" (TSTF-306). TSTF-306 allows the plant to address inoperable instrumentation related to the containment isolation valves (CIVs) by providing the ability to isolate the affected valves and allow for subsequent opening of the valves under administrative controls. Amendment 208 also added the TIP isolation instrumentation to the equipment that this specification would cover via the addition of instrumentation SRs for a channel check, channel functional test, channel calibration, and logic system functional test.
In its letter dated March 29,2010, the licensee stated, in part, that, Subsequent to the NRC approval of Amendment 208, Columbia identified that the installed TIP instrumentation does not provide the appropriate information to support channel check surveillance. Failure to comply with the required TS surveillance resulted in declaration of the TIP Isolation instrumentation inoperable and isolation of the containment penetration.
In order to utilize the TIP system to support calibration of the Local Power Range Monitors (LPRMs) the plant staff must remove and then restore equipment tag out paperwork as well as provide additional administrative controls while the TIP containment isolation valves are opened.
Since the Amendment No. 208 was implemented, the licensee has employed administrative controls to unisolate this penetration for the use of the TIP system, which presents an operational burden for the station.
Lyon, C. F., U.S. Nuclear Regulatory Commission, letter to J. V. Parrish, Energy Northwest, "Columbia Generating Station - Issuance of Amendment Re: Adoption of Approved Generic Technical Specification Changes Associated with Containment Isolation Valves (TAC No. MD6208)," dated September 12, 2008 (ADAMS Accession No. ML081900496).
-4 Therefore, the licensee proposed to remove the requirement for performance of the channel check surveillance for the TIP isolation instrumentation associated with TS 3.3.6.1.
3.2 Proposed Changes The licensee's proposed TS change would remove channel check SR 3.3.6.1.1 from TS Table 3.3.6.1-1, "Primary Containment Isolation Instrumentation," for Function 6.a, "Traversing In-Core Probe Isolation - Reactor Vessel Water Level - Low Low, Level 2," and Function 6.b, "Traversing In-Core Probe Isolation - Drywell Pressure - High."
The CGS TSs define "channel check" as follows:
A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
Since CGS's TIP isolation instrumentation provides no means to perform channel checks (i.e.,
no indicators), in its request for additional information (RAI) dated December 9, 2010 (ADAMS Accession No. ML103370113), the NRC staff asked the licensee to provide justification or compensatory measures in lieu of a channel check to ensure the operability and SRs for the TIP isolation instrumentation are met. The licensee provided the required information in its letter dated January 14, 2011. The NRC staff reviewed this removal of channel check SRs based on the reliability of the instrumentation, safety significance, controls and monitoring, and the function and operability of the TIP within the Neutron Monitoring System.
3.3 NRC Staff Evaluation 3.3.1 Reliability of the Instrumentation In its letter dated January 14, 2011, in response to the NRC staff's RAI dated December 9, 2010, regarding ensuring instrumentation operability, the licensee stated, in part, that, The TIP system CIV isolation instrumentation has an extended history of reliable operation. A review of the channel functional and calibration history of the TIP CIV instrumentation indicates that there has not been any drift, failures, or trends of as-found conditions that have, or would have exceeded allowable values during the past five years of operation. In addition, for the four instruments that provide inputs into the TIP isolation instrumentation, there have been only five instances where adjustments due to being outside of as-found tolerances were required during the 85 calibrations in the past five years.
This result translates to an average of just over one adjustment per instrument every 5 years.
Therefore, the NRC staff concludes that the reliability of the TIP CIV instrumentation is acceptable.
-5 3.3.2 Safety Significance In its letter dated January 14, 2011, the licensee stated, in part, that, The TIP system is infrequently used and the majority of the CIVs are closed when the system is not in use. The TIP system CIVs consist of inline ball valves and shear valves for each of the five TIP subsystems (A through E) that penetrate containment. When the system is not in use, the ball valves are maintained in a closed position. Normal operation of the TIP system is performed when calibration of the Local Power Range Monitors (LPRMs) is required. This occurs on a periodicity related to core exposure and correlates to the surveillance scheduled at approximately 42 day intervals when the plant is running at full power. Use of the TIP system generally takes less than one shift at each surveillance interval to gather TIP detector data. As described in the Energy Northwest's submittal, the TIP system CIV isolation design is such that manually initiated shear valves may be utilized to provide the containment isolation function for those instances in which the ball valves cannot be isolated However, it should be noted that there is a common 3/8 inch purge line that penetrates the containment. This purge line branches to each of the five TIP indexer mechanisms inside containment. The TIP purge line contains a check valve and a globe valve. The TIP purge line CIV globe valve is maintained in the open position, even when the system is not in use, to maintain a dry air source, typically nitrogen, to preclude any potential for moisture intrusion into the system.
If the TIP system is not used, the safety significance of the system is reduced since the TIP system is not normally used and the TIP CIVs are closed. Therefore, the NRC staff concludes that the safety Significance justification of the CGS TIP CIV instrumentation is acceptable.
3.3.3 Controls and Monitoring When a containment isolation function actuates, the TIP CIVs will automatically close if not already positioned closed. For a typical system configuration when the TIP system is not in use, this would effectively result in only closing the TIP purge line globe valve, as the ball valves would already be in the closed position. If containment isolation is required while the TIP system is in use and the TIP ball valve cannot be isolated, the operator can control the shear valve by a manually operated key lock switch. The operator action will cut the TIP cable and close the TIP guide tube. The control room also has controls available to close the TIP purge line globe valve if needed.
In its letter dated January 14, 2011, the licensee stated, in part, that, There are other indications available to the control room personnel for the TIP system and related TIP CIVs. These indications include the status of the TIP CIVs' position, shear valve circuit integrity, and whether or not an isolation signal is present. The channel check that cannot be physically performed applies
-6 specifically to the TIP isolation instrumentation functions of reactor vessel water level - low low and drywell pressure - high. The TIP CIV instrumentation includes valve position indication lights that are verified on a monthly basis as required by Technical Specification (TS) Surveillance Requirement (SR) 3.3.3.1.1.
As part of the TIP operating procedure, the status of the TIP CIV isolation instrumentation is verified. After the requisite information is collected to support the LPRM calibration, the CIV ball valves associated with each TIP subsystem are closed. This position is verified at suspension of TIP system operations via plant procedures.
Based on the above, the NRC staff concludes that the licensee can provide proper controls and monitoring of the TIP system.
3.3.4 Neutron Monitoring System The function and operation of the TIP within the Neutron Monitoring System is described in the licensee's Final Safety Analysis Report (FSAR), Section 7.7.1.6, "Neutron Monitoring System Traversing In-Core Probe." The CGS FSAR Section 7.7.1.6.1, "Function," states that, Flux readings along the axial length of the core are obtained by fully inserting the traversing ion chamber into one of the calibration guide tubes, then taking data as the chamber is withdrawn. The analog data is available for driving a recorder and for use by the process computer. One traversing ion chamber and its associated drive mechanism is provided for each group of seven to nine fixed in-core assemblies.
The CGS FSAR Section 7.7.1.6.2, "Operation," states, in part, that.
The [TIP] system allows calibration of LPRM signals by correlating TIP signals to LPRM signals as the TIP is positioned in various radial and axial locations in the core.
A valve system is provided with a valve on each guide tube entering the drywell.
A ball valve and a cable shearing valve are mounted in the guide tubing just outside the drywell. The ball valves are closed except when the TIP is in operation. They maintain the leaktightness integrity of the drywell. A valve is also provided for a nitrogen gas purge line to the indexing mechanisms. A guide tube ball valve opens only when the TIP is being inserted. The shear valve is used only if containment isolation is required and the ball valve cannot be isolated. The shear valve, which is controlled by a manually operated key lock switch, can cut the cable and close off the guide tube. The shear valves are actuated by explosive squibs.
-7 In its letter dated March 29, 2010, the licensee stated:
The TIP system uses a small bore penetration, and its isolation in a design basis event is via the manually operated shear valves. The ability to manually isolate the TIP system by either the normal isolation valve or the shear valve would be unaffected by inoperable instrumentation. Therefore, the option to isolate the penetration and to continue plant operation was provided. In order to implement this allowance, a separate isolation instrumentation function is used for the TIP system.
The reactor vessel water level-low low, level 2 isolation function receives input from two reactor vessel water level channels and the drywell pressure - high isolation function receives input from two drywell pressure channels. These channels provide input to two logic trip circuits grouped in one-out-of-two logic. Each of these trip circuits is connected in one-out-of-two taken twice that a low low, level 2 (C) or drywell pressure - high (C) input and a low low, level 2 (D) or drywell pressure - high (D) input will initiate an isolation of the TIP valves.
TSTF-306 proposed surveillances to satisfy the requirements of 10 CFR 50.36(c)(3), which included a channel check, channel calibration, channel functional test, and logic system functional test. The channel check surveillance helps to satisfy the requirements of 10 CFR 50.36(c)(3) in part by providing detection of a gross channel failure and confirmation that the instrumentation continued to operate properly between each channel calibration. The NRC staff concludes that with the removal of the channel check surveillance from the CGS TS, the instrumentation that supports the equipment is maintained since the remaining SRs would still be in place, which is consistent for compliance with the requirements of 10 CFR 50.36(c)(3).
The NRC has previously approved exclusion of channel check for the TIP isolation instrumentation related to drywell pressure for Susquehanna Steam Electric Station, Units 1 and 2, via Amendment Nos. 213 and 1882 , when TSTF-306 was incorporated. When the NRC approved CGS's conversion to the improved STS in 19973 , a number of systems required deviations from the STS for channel check SRs on instrumentation since CGS's design did not include indications to support performance of a channel check. The proposed TS change is consistent in approach to the deviations from the STS that have been previously approved by the NRC for CGS.
Based on the above, the NRC staff concludes that the proposed changes provide reasonable assurance that the function of the TIP CIV instrumentation is appropriately monitored between channel functional tests. The NRC staff also concluded that the licensee's SRs adopted under TSTF-306 ensure that the removal of the channel check surveillance remains in compliance 2 Guzman, R. V., U.S. Nuclear Regulatory Commission, letter to Bryce L. Shriver, PPL Susquehanna, LLC, "Susquehanna Steam Electric Station, Units 1 and 2 - Issuance of Amendment Re: Intermittent Opening of Isolated Flow Paths and TIP Isolation (TAC Nos. MB6665 and MB6666), dated June 5,2003 (ADAMS Accession No. ML031560495).
3 Colburn, T G., U.S. Nuclear Regulatory Commission, letter to J. V. Parrish, Washing Public Power Supply System, "Issuance of Amendment for the Washington Public Power Supply System Nuclear Project No.2 (TAC No. M94226)," dated March 4, 1997 (ADAMS Accession No. ML022120577).
-8 with 10 CFR 50.36(c)(3) requirements. Furthermore, the NRC staff reviewed precedents relating to the exclusion of channel check SRs on instrumentation and determined the proposed change is consistent with those NRC-approved precedents in its approach to deviate from the STS. Therefore, the NRC staff concludes that the proposed change meets the requirements of 10 CFR 50.36,10 CFR 50, Appendix A, GDC 56, and NUREG-1433, Revision 3, and, therefore, is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Washington 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 the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20, and makes editorial, corrective, or other minor revisions. 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 June 1, 2010 (75 FR 30444). 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 Contributors: B. Lee P. Chung Date: March 18, 2011
March 18,2011 Mr. Mark E. Reddemann Chief Executive Officer Energy Northwest P.O. Box 968 (Mail Drop 1023)
Richland, WA 99352-0968 SUB~IECT: COLUMBIA GENERATING STATION -ISSUANCE OF AMENDMENT RE:
CHANGE TO TECHNICAL SPECIFICATIONS RELATING TO TRAVERSING IN-CORE PROBE CONTAINMENT ISOLATION INSTRUMENTATION (TAC NO.
ME3713)
Dear Mr. Reddemann:
The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 220 to Facility Operating License No. NPF-21 for the Columbia Generating Station. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated March 29, 2010, as supplemented by letter dated January 14, 2011.
The amendment revises TS 3.3.6.1, "Primary Containment Isolation Instrumentation," by deleting channel check Surveillance Requirement 3.3.6.1.1 from TS Table 3.3.6.1-1, "Primary Containment Isolation Instrumentation," for the traversing in-core probe isolation instrumentation.
A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely,
/ra/{LWilkins for)
Balwant K. Singal, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosures:
- 1. Amendment No. 220 to NPF-21
- 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
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