ML100610233
| ML100610233 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 03/18/2010 |
| From: | Lyon C Plant Licensing Branch IV |
| To: | Minahan S Nebraska Public Power District (NPPD) |
| Lyon C Fred, NRR/DORL/LPL4, 301-415-2296 | |
| References | |
| TAC ME0848 | |
| Download: ML100610233 (16) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 18, 2010 Mr. Stewart B. Minahan Vice President-Nuclear and CNO Nebraska Public Power District 72676 648A Avenue Brownville, NE 68321 SUB..IECT:
COOPER NUCLEAR STATION - ISSUANCE OF AMENDMENT RE: REVISE BATTERY RESISTANCE IN SURVEILLANCE REQUIREMENTS (TAC NO.
ME0848)
Dear Mr. Minahan:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 236 to Facility Operating License No. DPR-46 for the Cooper Nuclear Station.
The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated March 11, 2009, as supplemented by letters dated August 12 and December 21,2009, and March 5,2010.
The amendment would revise Surveillance Requirements 3.8.4.2 and 3.8.4.5 in TS Section 3.8.4, "DC [Direct Current] Sources - Operating," by adding a parameter of total battery resistance to the values of battery connection resistance. The proposed changes correct nonconservative TSs.
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, (77~~/t-f'(
Carl F. Lyon, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298
Enclosures:
- 1. Amendment No. 236 to DPR-46
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555-0001 NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 236 License No. DPR-46
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Nebraska Public Power District (the licensee).
dated March 11, 2009, as supplemented by letters dated August 12 and December 21,2009, and March 5, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this 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 Facility Operating License No. DPR-46 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 236, are hereby incorporated in the license. The licensee 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 45 days from the date of issuance. Implementation of the amendment shall include updating the Updated Safety Analysis Report and the Technical Specification Bases for the batteries. These updates shall include that the safety-related batteries shall be considered at 15 years to have reached 85 percent of expected service life.
FOR THE NUCLEAR REGULATORY COMMISSION 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. DPR-46 and Technical Specifications Date of Issuance: March 18, 2010
ATTACHMENT TO LICENSE AMENDMENT NO. 236 FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 Replace the following pages of the Facility Operating License No. DPR-46 and Appendix A, Technical Specifications with the enclosed revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Facility Operating License REMOVE INSERT Page 3 of 5 Page 3 of 5 Technical Specifications REMOVE INSERT 3.8-17 3.8-17 3.8-18 3.8-18 3.8-19 3.8-19
(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 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:
(1)
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2419 megawatts (thermal).
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 236, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Physical Protection The licensee 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 contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Cooper Nuclear Station Safeguards Plan," submitted by letter dated May 17, 2006.
(4)
Fire Protection The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Cooper Nuclear Station (CNS) Updated Safety Analysis Report and as approved in the Safety Evaluations dated November 29, 1977; May 23, 1979; November 21, 1980; April 29, 1983; April 16, 1984; June 1, 1984; January 3, 1985; August 21, 1985; April 10, 1986; September 9, 1986; November 7, 1988; February 3, 1989; August 15, 1995; and July 31, 1998, subject to the following provision:
The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
Amendment No. 236 Revised by letter dated March 5, 2007 3 of 5
3.8.4 DC Sources -
Operating SURVEILLANCE REQUI REM ENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage on float charge is:
- a.
~ 125 V for the 125 V batteries; and
- b.
~ 250 V for the 250 V batteries.
7 days SR 3.8.4.2 Verify no visible corrosion at battery terminals and connectors.
OR Verify battery connection resistance meets the limits specified in Table 3.8.4~1.
92 days SR 3.8.4.3 Verify battery cells, cell plates, and racks show no visual indication of physical damage or abnormal deterioration that degrades battery performance.
18 months SR 3.8.4.4 Remove visible corrosion and verify battery cell to cell and terminal connections are coated with anti-corrosion material.
18 months SR 3.8.4.5 Verify battery connection resistance meets the limits specified in Table 3.8.4-1.
18 months SR 3.8.4.6 Verify:
- a.
Each required 125 V battery charger supplies ~
200 amps at ~ 125 V for ~ 4 hows; and
- b.
Each required 250 V battery charger supplies ~
200 amps at ~ 250 V for ~ 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
18 months (continued)
Cooper 3.8-17 Amendment No. 236
3.8.4 DC Sources ~ Operating SURVEILLANCE RE.QUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.4.7
.----------~----N()TES---------------*------------
- 1.
The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4.7 once per 60.months.
- 2.
This Surveillance shall not be performed in M()DE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR.
Verify battery capacity is adequate to supply, and 18 months maintain in ()PERABLE status*, the required emergency loads for the design duty cycle when subjected to a battery service test.
N()TE-------------------------------
This Surveillance shall not be performed in M()DE 1,.
2, or 3. However, credit may-be taken for unplanned events that satisfy this SR.
Verify battery capacity is ~ 90% of the manufacturer's 60 months rating when subjected to a performance discharge test or a modified performance discharge test.
AND 18 months when battery shows degradation or has reached 85%
. of expected life with capacity
. < 100% of manufacturer's rating 24 months when battery has reached 85% of the expected life with capacity
~ 100% of manufacturer's
.rating Cooper 3.8-18 Amendment No. 236
DC Sources Operating 3.8.4 Table 3.8.4-1 (page 1 of 1)
Battery Connection Resistance Limits PARAMETER LIMIT (MICRO-OHMS)
SYSTEM DIVISION Inter-cell connections
~ 150 Both 125 volt and 250 volt Both 1 and 2 Inter-rack connections
~280 Both 125 volt and 250 volt Both 1 and 2 Inter-tier connections
~ 150 Both 125 volt and 250 volt Both 1 and 2 Terminal connections
~ 150 Both 125 volt and 250 volt Both 1 and 2 Total battery
~3300 125 volt Both 1 and 2 resistance
< 6500 250 volt Both 1 and 2 Cooper 3.8-19 Amendment No. 236
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 236 TO FACILITY OPERATING LICENSE NO. DPR-46 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298
1.0 INTRODUCTION
By application dated March 11, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090750599), as supplemented by letters dated August 12 and December 21,2009, and March 5, 2010 (ADAMS Accession Nos. ML092300636, ML093630758, and ML100680486, respectively), Nebraska Public Power District (NPPD, the licensee), requested changes to the Technical Specifications (TSs) for Cooper Nuclear Station (CNS). The proposed changes would revise Surveillance Requirements (SR) 3.8.4.2 and 3.8.4.5 in TS Section 3.8.4, "DC [Direct Current] Sources - Operating," by adding a parameter of total battery resistance to the values of battery connection resistance. Specifically, the proposed changes would modify SR 3.8.4.2 and SR 3.8.4.5 by adding additional acceptance criteria to verify that total battery connection resistance is within pre-established limits. These limits ensure that the CNS batteries can perform their safety functions and will remain operable during postulated design basis events. The proposed changes correct nonconservative TSs.
The supplemental letters dated August 12 and December 21, 2009, and March 5, 2010, 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) staffs original proposed no significant hazards consideration determination as published in the Federal Register on May 5,2009 (74 FR 20752).
2.0 REGULATORY EVALUATION
The NRC staff applied the following regulatory requirements in its review.
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
- 2 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, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operations; (3) SRs; (4) design features; and (5) administrative controls. The regulations in 10 CFR 50.36(c)(3) specify that SRs are "requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."
General Design Criterion (GDC) 17, "Electric power systems," 1 of Appendix A, "General Design Criteria for Nuclear Power Plants," to Part 50 of 10 CFR requires, in part, that an onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components that are important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents. The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.
In 10 CFR Part 50, Appendix A, GDC 18, "Inspection and testing of electric power systems," requires, in part, that electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components.
The NRC staff used the following documents for technical guidance during its review.
Regulatory Guide (RG) 1.128, Revision 2, "Installation Design and Installation of Vented Lead-Acid Storage Batteries for Nuclear Power Plants," February 2007 (ADAMS Accession No. IVIL070080013) describes a method that the NRC staff considers acceptable for use in complying with the agency's regulations with regard to satisfying criteria for the installation design and installation of vented lead-acid storage batteries in nuclear power plants.
1 The 1967 Proposed GDC as described in the CNS updated safety analysis report (USAR), Appendix F, are the licensing basis for CNS; however, the NRC staff concluded in its 1973 Safety Evaluation Report for CNS that the intent of the 1971 Final Rule for 10 CFR Part 50, Appendix A, had also been met.
- 3 RG 1.129, Revision 2, "Maintenance, Testing, and Replacement of Vented Lead Acid Storage Batteries for Nuclear Power Plants," dated February 2007 (ADAMS Accession No. ML063490110), describes a method that the NRC staff considers acceptable for use in complying with the agency's regulations with regard to the maintenance, testing, and replacement of vented lead-acid storage batteries in nuclear power plants.
RG 1.212, "Sizing of Large Lead-Acid Storage Batteries," dated November 2008 (ADAMS Accession No. ML082740047), describes a method that the NRC staff considers acceptable for use in complying with requirements and regulations on the criteria for the sizing of large lead-acid storage batteries for use in nuclear power plants.
Construction of CNS predates the 1971 issuance of 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants." Appendix F, "Conformance to AEC proposed General Design Criteria," of CNS FSAR discusses that CNS is designed to conform to the proposed general design criteria published in the July 11, 1967, Federal Register, except where commitments were made to specific GDC.
3.0 TECHNICAL EVALUATION
3.1 Background and Evaluation The electrical power system at CNS consists of various Alternating Current and Direct Current (DC) systems, as described in Chapter 8 of the CNS USAR. Two of the DC power systems are the 125 volt (V) system and the 250 V system. These two systems provide both motive and control power to selected safety-related and non-safety related equipment. The systems are designed to have sufficient independence, redundancy, and testability to perform their safety functions, assuming a single failure.
Each of these systems provides two independent onsite sources of DC power for startup operation, shutdown, and the loads required for station safety. The loss of anyone source will not prevent safe shutdown of the station. The safety objective of these two systems is to provide an uninterruptible source of power to supply all normal and emergency 125 V DC and 250 V DC control and power loads under all conditions.
The 125 V DC and 250 V DC systems each have two subsystems. The Division 1 and Division 2 125 V DC subsystems each consists of a 125 V battery, battery charger, and DC distribution system. The Division 1 and Division 2 250 V DC subsystems each consists of a 250 V battery, battery charger, and distribution system. Each 125 V battery has 58 individual cells. Each 250 V battery has 120 individual cells.
The batteries and battery chargers are designed and installed to Class 1E requirements. The batteries and battery chargers are seismically qualified and are located in the Control Building, which is a Class I seismic structure. The licensee noted in its application dated March 11, 2009, that these batteries are designed, tested, and maintained in accordance with the Institute of Electrical and Electronics Engineers (IEEE) Standard (Std.) 450-1995, "IEEE Recommended
-4 Practice for Maintenance, Testing, and Replacement of Large Stationary Type Power Plant and Substation Lead Storage Batteries," IEEE Std. 535-1979, "IEEE Standard for Qualification of Class 'I E Lead Storage Batteries for Nuclear Power Generating Stations," and IEEE Std. 485 1983, "I EEE Standard for the Sizing of Large Stationary Type Power Plant and Substation Lead Storage Batteries."
The existing TS SRs 3.8.4.2 and 3.8.4.5 for CNS currently require the licensee to verify that battery connection resistance is less than specified limits for the individual parts of the safety related batteries (Le., inter-cell, inter-tier, inter-rack, and terminal connections). The licensee proposed modifying SRs 3.8.4.2 and 3.8.4.5 by adding additional acceptance criteria to verify total battery connector resistance is within pre-established limits that ensure the CNS batteries can perform their safety functions and will remain operable during postulated design basis events. The licensee proposed this change to restore conservatism to the CNS TSs and to ensure assumptions in the CNS USAR accident analyses remain valid. This change is necessary since the parameters currently specified in the existing SRs do not, by themselves, ensure that the batteries will be maintained in a condition such that they are able to perform their safety function, since the total resistance of the battery, as determined by summing the values of resistance for these individual parts, could exceed the value of total battery resistance reflected in the load and voltage study calculations.
The NRC staff reviewed the proposed changes for technical adequacy. In its application dated March 11, 2009, the licensee stated that it has evaluated the event-specific load profiles for the safety-related batteries at CNS to address the loss-of-offsite power/loss-of-coolant accident (LOOP/LOCA) event and the station blackout (SBO) event. The licensee further noted that the calculations determined the terminal voltages at the devices and verified that adequate voltage and current exists for these devices to perform their safety function for the LOOP/LOCA and SBO events. This evaluation includes verifying values of resistance for all parts of the battery, as well as the total battery resistance. The licensee uses the resistance values to calculate the acceptable voltages at the devices.
To verify the licensee's statements, the NRC staff reviewed the licensee's calculations for the total battery resistance limits that were referenced in the TS Bases for TS SRs 3.8.4.2 and 3.8.4.5. The licensee provided the following calculations, without appendices, in its letter dated August 12, 2009: NEDC 87-131C, Revision 9, "125 VDC Division I Load and Voltage Study,"
NEDC 87-131D, Revision 8, "125 VDC Division II Load and Voltage Study," NEDC 87-131A, Revision 9, "250 VDC Division I Load and Voltage Study," and NEDC 87-131 B, Revision 8, "250 VDC Division II Load and Voltage Study." The staff reviewed these calculations and determined that additional clarifying information was needed. Specifically, in its request for additional information (RAI) letter dated November 25, 2009 (ADAMS Accession No. ML093220807), the staff requested the licensee to provide the technical basis for the resistance values in the calculations (Le., resistance values for inter-cell, inter-tie, inter-rack, terminal, and the total battery). In its response to the staff dated December 21, 2009, the licensee provided a detailed discussion on how the various connection resistances were derived including how the battery manufacturer's acceptance limits were applied. Based on its review of the licensee's response, the staff determines that the licensee has adequately incorporated the battery manufacturer's acceptance limits into the CNS load and voltage studies identified above. The staff also requested the licensee to clarify other inputs, assumptions, and formulas that were in the
- 5 calculations. Based on its review of the licensee's response dated December 21,2009, the NRC staff concludes that the proposed resistance values are acceptable.
In its RAI letter dated July 10, 2009 (ADAMS Accession No. ML091880757), the NRC staff also requested the licensee to (1) explain the use of the words 'not applicable' in proposed TS Table 3.8.4-1 and to (2) describe how battery limits will be monitored during the performance of revised SRs 3.8.4.2 and 3.8.4.5. In its August 12, 2009, response to the staff's request, the licensee recognized that the proposed language in TS Table 3.8.4-1 was confusing and proposed replacing the words 'not applicable' with 'Both 125 volt and 250 volt' under the System column heading and 'Both 1 and 2' under the Division column heading in proposed TS Table 3.8.4-1. The NRC staff concludes that this change clarifies the applicability of the associated TS, and, therefore, is acceptable. In response to the second part of the staff's request, the licensee stated that battery resistance limits of the station batteries are currently monitored through performance of CNS Surveillance Procedure 6.EE.609, "125V/250V Station Battery Inter-cell Connection Testing." The licensee stated that, "Upon issuance of this license amendment, the method of monitoring battery resistance, including total resistance, will not change; however, Surveillance Procedure 6.EE.609 will require revision to make the limits consistent with revised TS SRs 3.8.4.2 and 3.8.4.5". The NRC staff concludes that this is acceptable, since the procedure uses the sum of the inter-cell connectors, the inter-tier cables and connectors, the inter-rack cables and connectors, and the terminal connections to determine total resistance.
In the revision summary section of these calculations, the licensee noted that the aging coefficient is raised from 0.80 (80 percent) to 0.90 (90 percent) consistent with TS SR 3.8.4.8.
A number of factors affect the expected life of a battery (e.g., maintenance and discharge cycling); however, the standard industry practice is to use the typical battery life curve (Le., age versus capacity curve) for flooded type cells. Based on its review of standard battery age versus capacity curve, which indicates average cell performance, the NRC staff noted that it would expect a drop in the expected service life after aging the batteries to 90 percent of rated capacity. In its RAI letter dated November 25, 2009, the NRC staff requested the licensee to describe the impact of this change on the expected life of CNS batteries (e.g., conclusions drawn from the battery age versus capacity curve for the batteries). The NRC staff understands that the CNS batteries were originally sized to produce required capacity at 80 percent of nameplate rating, corresponding to a 20-year qualified life (expected service life). In its response dated December 21, 2009, the licensee noted that, as loads have increased over time, the required battery capacity is now closer to 90 percent of the nameplate rating. The licensee further stated that its battery vendor does not estimate battery life for 90 percent capacity. Therefore, the licensee exercised engineering judgment to conservatively establish 15 years as the point at which the batteries have reached 85 percent of expected life and defined battery degradation to be when capacity drops by more than 5 percent (normally 10 percent as defined by IEEE Std. 450-1995) relative to the capacity on the previous performance test or when the battery capacity reaches less than or equal to 95 percent of manufacturers rating (normally 90 percent or when it is 10 percent below the manufacturer's rating as defined by IEEE Std. 450-1995).
Based on the above, the NRC staff concludes that the expected service life reduction that was proposed by the licensee is consistent with the staff's assessment that was derived from
- 6 standard battery life versus capacity curves. Therefore, the NRC staff concludes that 15 years is acceptable to be considered the point at which the CNS safety-related batteries have reached 85 percent of expected service life as a result of the proposed change.
This battery age of 15 years (i.e., 85 percent of expected life) advances by 2 years (currently 17 years, which is 85 percent of a 20 year expected service life) the point at which the frequency for performance of battery capacity verification per SR 3.8.4.8 changes from 60 months to 18 months. Two years is considered a sufficient reduction in expected life to account for the increased loading on the batteries. This new, more restrictive, definition of battery degradation is a graded approach to ensure that decisions to replace batteries are made before the batteries reach the 90 percent capacity limit of SR 3.8.4.8. The NRCstaff concludes that the licensee's new definition for battery degradation is conservative and provides additional assurance to account for uncertainty with its engineering judgment to establish 15 years as the point at which the batteries have reached 85 percent of expected service life. Based on this information and the inclusion in the licensing basis of 15 years as being considered the point at which the CNS safety-related batteries have reached 85 percent of expected service life, the NRC staff concludes that the proposed changes are acceptable.
3.2 Conclusion The NRC staff evaluated the licensee's request to modify SR 3.8.4.2 and SR 3.8.4.5 by adding additional acceptance criteria to verify that total battery connection resistance is within pre established limits.
Based on the above, the staff concludes the licensee's proposed changes to the CNS TSs provide reasonable assurance of the continued availability of the required electrical power to shut down the reactor and to maintain the reactor in a safe condition after an anticipated operational occurrence or a postulated design-basis accident. Furthermore, the staff concludes that the proposed TS changes are in accordance with 10 CFR 50.36 and meet the CNS design basis. Therefore, the NRC staff concludes that the proposed changes are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regUlations, the Nebraska State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes an inspection or a surveillance requirement. 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 May 5,2009 (74 FR 20752). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to
- 7
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: M. McConnell Date: March 18, 2010
March 18, 2010 Mr. Stewart B. Minahan Vice President-Nuclear and CNO Nebraska Public Power District 72676 648A Avenue Brownville, NE 68321
SUBJECT:
COOPER NUCLEAR STATION - ISSUANCE OF AMENDMENT RE: REVISE BATTERY RESISTANCE IN SURVEILLANCE REQUIREMENTS (TAC NO.
ME0848)
Dear Mr. Minahan:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 236 to Facility Operating License No. DPR-46 for the Cooper Nuclear Station.
The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated March 11, 2009, as supplemented by letters dated August 12 and December 21, 2009, and March 5, 2010.
The amendment would revise Surveillance Requirements 3.8.4.2 and 3.8.4.5 in TS Section 3.8.4, "DC [Direct Current] Sources - Operating," by adding a parameter of total battery resistance to the values of battery connection resistance. The proposed changes correct nonconservative TSs.
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 by Mohan C. Thadani forI Carl F. Lyon, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298
Enclosures:
- 1. Amendment No. 236 to DPR-46
- 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
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