ML110100291

Cooper - License Amendment Request for Reducing the Number of Technical Specification 3.4.3 Required Safety Relief Valves

ML110100291
Person / Time
Site:	Cooper
Issue date:	01/05/2011
From:	O'Grady B J Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2010046
Download: ML110100291 (22)

Text

N Nebraska Public Power District Always there when you need us 50.90 NLS2010046 January 5, 2011 U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, D.C. 20555-0001

Subject:

License Amendment Request for Reducing the Number of Technical Specification

3.4.3 Required

Safety Relief Valves Cooper Nuclear Station; Docket No. 50-298, License No. DPR-46

Dear Sir or Madam:

The purpose of this letter is for Nebraska Public Power District (NPPD) to request an amendment to Facility Operating License DPR-46 under the provisions of 10 CFR 50.4 and 10 CFR 50.90 to revise the Cooper Nuclear Station (CNS) Technical Specification (TS) 3.4.3, Safety/Relief Valves (SRVs) and Safety Valves (SVs). The proposed amendment reduces the number of SRVs required to be OPERABLE for over-pressure protection from eight to five.NPPD has determined from the No Significant Hazards Consideration determination that this: change does not involve a significant hazard.NPPD requests approval of the proposed amendment by January 8, 2012, allowing for an approximate one year review by the Nuclear Regulatory Commission (NRC). Once approved, the amendment will be implemented within 60 days.Attachment 1 provides a description of the TS changes, the basis for the amendment, the No Significant Hazards Consideration evaluation pursuant to 10 CFR 50.9 1(a)(1), and the Environmental Impact evaluation pursuant to 10 CFR 51.22. Attachment 2 provides the proposed changes to the current CNS TS in marked up format. Attachment 3 provides the final typed TS pages to be issued with the amendment.

Attachment 4 provides conforming changes to the TS Bases for information.

Enclosure 1 provides a proprietary report from General Electric Hitachi (GEH) that supports relaxation of the number of required SRVs with an affidavit requesting non-disclosure to the public. Enclosure 2 provides a non-proprietary version of the GEH report for the public docket. No regulatory commitments are being made by this request.NPPD requests that Enclosure 1 be withheld from public disclosure in accordance with 10 CFR 2.390.COOPER NUCLEAR STATION PO. Box 98 / Brownville, NE 68321-0098 Telephone:

(402) 825-3871 / Fax: (402) 825-5211 www.nppd.com NLS2010046 Page 2 of 2 This proposed TS change has been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). Amendments to the CNS Facility Operating License through Amendment 236 issued March 18, 2010, have been incorporated into this request. This request is submitted under affirmation pursuant to 10 CFR 50.30(b).By copy of this letter and its attachments, the appropriate State of Nebraska official is notified in accordance with 10 CFR 50.91(b)(1).

Copies are also being provided to the NRC Region IV office and the CNS Senior Resident Inspector in accordance with 10 CFR 50.4(b)(1).

Should you have any questions concerning this matter, please contact David Van Der Kamp, Licensing Manager, at (402) 825-2904.I declare under penalty of perjury that the foregoing is true and correct.Executed on: (Date)Sincerely, Brian J. O'Grady Vice President

-Nuclear and Chief Nuclear Officer/em Attachments Enclosures cc: Regional Administrator w/Attachments and Enclosure 2 USNRC -Region IV Cooper Project Manager w/Attachments and Enclosures USNRC -NRR Project Directorate IV-1 Senior Resident Inspector w/Attachments and Enclosure 2 USNRC -CNS Nebraska Health and Human Services w/ Attachments and Enclosure 2 Department of Regulation and Licensure NPG Distribution w/o Attachments or Enclosures CNS Records w/Attachments and Enclosures ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© 4 ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© 4 Correspondence Number: NLS2010046 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document.

Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments.

Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITMENT COMMITTED DATE COMMITMENT NUMBER OR OUTAGE None 4 -~4 -I 4 -I 4 -I.I PROCEDURE 0.42 REVISION 27 PAGE 18 OF 25 NLS2010046 Attachment 1 Page 1 of 11 Attachment 1 License Amendment Request for Reducing the Number of Technical Specification

3.4.3 Required

Safety Relief Valves Cooper Nuclear Station; Docket No. 50-298, DPR-46 Revised Technical Specification Page 3.4-6 1.0 Summary Description

2.0 Detailed

Description

2.1 Proposed

Change 2.2 Need for Change 2.3 Technical Specification Bases Changes 3.0 Technical Evaluation

3.2 Updated

Safety Analysis Report (USAR) Safety Design Basis 3.3 Current TS Bases Applicable Safety Analysis 3.4 Analytical Methods, Standards, Data & Results 3.5 Technical Justification of Proposed Changes 3.6 USAR Accident Analysis Impact 3.7 Conclusion

4.0 Regulatory

Safety Analysis 4.1 Applicable Regulatory Requirements/Criteria

4.2 Precedent

4.3 No Significant Hazards Consideration

4.4 Conclusion

5.0 Environmental Consideration

6.0 References

NLS2010046 Attachment 1 Page 2 of 11 1.0

SUMMARY

DESCRIPTION This letter is a request to amend Facility Operating License DPR-46 for Cooper Nuclear Station (CNS). The proposed change would revise Technical Specifications (TS) to reduce the number of SafetyRelief Valves (SRVs) requiredby TS 3.4.3 to be OPERABLE for over-pressure protection (OPP) from eight to five. This will minimize unnecessary entries into Limiting Condition for Operation (LCO) 3.4.3 because of a specification that is more restrictive than needed to meet the SRV safety function of reactor coolant pressure boundary (RCPB) OPP or the anticipated transient without scram (ATWS) support function.Nebraska Public Power District (NPPD) requests approval of the proposed amendment by January 8, 2012, allowing an approximate one year review by the Nuclear Regulatory Commission (NRC). Once approved, CNS will implement the amendment within 60 days.2.0 DETAILED DESCRIPTION The following revisions are proposed to TS Section 3.4.3 2.1 Proposed Change LCO 3.4.3 currently states, "The safety function of 8 SRVs and 3 SVs shall be OPERABLE." This proposed change would replace the number 8 with 5 in LCO 3.4.3 for the number of SRVs required to be OPERABLE for OPP. The condition statement is also changed to "One or more required SRVs or SVs inoperable." 2.2 Need for Change This change is being proposed, because LCO 3.4.3 is more restrictive than needed to meet the SRV safety function of RCPB OPP or the ATWS support function.To minimize unnecessary entries into LCO 3.4.3, CNS commissioned General Electric Hitachi (GEH) to evaluate reactor vessel OPP in the event that one or more SRVs fail to open and remain completely closed. The GEH analysis is documented in Enclosures 1 and 2. The analysis focuses on American Society of Mechanical Engineers (ASME) OPP events and on ATWS events (Reference 6.3).2.3 Technical Specification Bases Changes Revised TS Bases are provided in Attachment 4 for NRC information.

These Bases revisions will be made as an implementing action pursuant toTS 5.5.10, TS Bases Control Program, following issuance of the amendment.

The TS Bases for pages B 3.4-15, B 3.4-16 and B 3.4-18 are revised to conform to the changes proposed for TS 3.4.3.

NLS2010046 Attachment 1 Page 3 of 11 3.0 TECHNICAL EVALUATION

3.1 System Description

CNS is a boiling water reactor (BWR) of General Electric BWR/4 design, with a Mark I containment.

The pressure relief system at CNS includes three Dresser safety valves (SVs) and eight Target Rock SRVs, all of which are located on the main steam lines, within the drywell, between the reactor vessel and the first main steam isolation valve (MSIV). The SVs discharge directly to the interior space of the drywell and the SRVs discharge to the suppression pool.The safety function (i.e., spring-lift function) of each Target Rock SRV is actuated by a pilot valve assembly.

In the event that Reactor Pressure Vessel (RPV) pressure rises to the nominal setpoint (NSP) of the SRV (1080 to 1100 psig), the pilot valve will open, admitting steam and allowing the reactor pressure itself to open the SRV and relieve pressure.

The Dresser SVs are held closed by a large mechanical spring.Once the Dresser SV NSP is reached (1240 psig), vessel pressure will act against the spring (directly on the valve disc) to open the SV. Together, the SVs and the spring-lift function of the SRVs provide overpressure protection for the reactor vessel as required by the ASME Code.The pilot-actuated, SRV spring-lift function of the Target Rock SRVs is completely independent of the pressure relief function of these valves. In the Automatic Depressurization System (ADS) and Low Low Set (LLS) modes, the SRVs are electro-pneumatically actuated by direct current powered ADS solenoids, which will admit nitrogen from the ADS accumulators to open the SRVs for pressure control.The ADS serves as backup to the High Pressure Coolant Injectionsystem and is actuated by completely different logic/trip signals. The first lift of any SRV coincident with a high pressure scram signal will arm the LLS function.

On second and subsequent lifts, the LLS valves will take over, controlling reactor pressure and minimizing SRV cycling. The ADS and LLS functions are governed by LCO 3.5.1, ECCS -Operating, and LCO 3.6.1.6, LLS Valves, which are not impacted by this License Amendment Request (LAR). This LAR only addresses the safety (spring-lift) mode of SV & SRV operation under LCO 3.4.3.3.2 Updated Safety Analysis Report (USAR) Safety Design Basis (Reference 6.1)The safety objective of the pressure relief system is to prevent over-pressurization of the nuclear system; this protects the RCPB from failure which could result in the uncontrolled release of fission products.

In addition, the automatic depressurization feature of the pressure relief system acts in conjunction with the Emergency Core Coolant System for reflooding the core. This protects the reactor fuel cladding from failure due to overheating.

NLS2010046 Attachment 1 Page 4 of 11 The safety design basis of the pressure relief system is: 1. The pressure relief system shall prevent overpressurization of the nuclear system in order to prevent failure of the RCPB due to pressure.2. The pressure relief system shall provide automatic depressurization so that the Low Pressure Coolant Injection and the Core Spray systems can operate to protect the fuel barrier.3. The SRV discharge piping is designed to accommodate forces resulting from relief action and shall be supported for reaction loads due to flow at maximum relief discharge capacity so that system integrity is maintained.

4. The pressure relief system is designed to withstand adverse combinations of loadings and forces resulting from operation during abnormal, accident, or special event conditions.

5. The pressure relief system is designed for testing prior to nuclear system operation and for verification of the operability of the pressure relief system.6. Positive position indication of the SRVs shall be provided in the control room.3.3 Current TS Bases Applicable Safety Analysis (Reference 6.2)The OPP system must accommodate the most severe pressurization transient.

Evaluations have determined that the most severe transient is the closure of all MSIVs, followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position).

A separate analysis has been performed assuming a setpoint tolerance of +/- 3% for the SRVs and SVs. For the purpose of the setpoint tolerance analysis, eight SRVs and three SVs are assumed to operate in the safety mode. The analysis results demonstrate that the design SRV and SV capacity, with a setpoint tolerance of +/- 3%, is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig = 1375 psig). LCO 3.4.3 helps to ensure that the acceptance limit of 1375 psig is met during the most severe design basis pressure transient.

From an over-pressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above.SRVs and SVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

NLS2010046 Attachment 1 Page 5 of 11 3.4 Analytical Methods, Standards, Data & Results The GEH analysis (Enclosures 1 and 2) assume: (a) The number of SRVs Out-of-Service (OOS) is variable, and (b) The three SVs on the RPV function as designed.The SVs and SRVs are assumed to lift at their nominal, spring-lift setpoint plus 3%, the upper end of the allowable code tolerance.

For the SVs, this is 1240 psig x 1.03 1277.2 psig for all cases analyzed by GEH, as specified in Surveillance Requirement (SR) 3.4.3.1. For the SRVs, the spring-lift setpoint is different for each valve group but also includes the 3% tolerance.

Note that GEH uses the term Analysis of Record to refer to internal GEH documents used prior to CNS changing its Licensed Thermal Power to 2419 MWth.ASME OPP Events (Section 3.0 of Enclosures 1 and 2)Initial conditions and inputs are assumed as follows. The limiting single failure pressurization event for ASME OPP is the MS1V closure with high flux scram. The MSIVs are assumed to close in three seconds, the minimum time allowed by the TS.Licensing Limits for this event are 1375 psig peak vessel pressure (110% of RPV design pressure) and 1337 psig dome pressure safety limit, respectively, as defined in ASME Boiler and Pressure Vessel Code. For conservatism, 15 psi margin below each limit is used as the analysis success criterion.

For the ASME OPP analysis, reactor power is assumed to be 2428.6 MWth, which is 2% over the Original Licensed Thermal Power (OLTP) of 2381 MWth. Maximum and minimum core flows of 105%and 76.8% for Cycle 26 End of Cycle (EOC) exposure were both evaluated.

The ASME OPP analyses were performed with three SRVs OOS. Each case of SRVs OOS was evaluated for variations in SRV setpoints to account for possible drift. The maximum drift (MAX) allowed in the analysis was to 1246.3 psig, which is 3% above a setpoint of 1210 psig.For three SRVs bOS, the ASME OPP analysis results showed success with substantial margin to the dome pressure safety limit, the peak vessel pressure limit;and the more conservative analysis goals. This result proved true for all cases up to the MAX drift.ATWS Events (Section 4.0 of Enclosures 1 and 2)The ATWS analysis considered Pressure Regulator Failing Open and MSIV Closure events, which have been shown to be limiting.

For ATWS, the peak allowable vessel pressure is 1500 psig, the ASME Code Service Level C Limit. Based on this, the analysis success criterion was set at 1490 psig to allow margin for future fuel and hardware changes. The ATWS analysis was performed at 2419 MWth,(1.6%

above OLTP), core flows at 76.8% and 105% of rated, and Beginning of Cycle and EOC exposure.

Cladding temperature and cladding oxidation were restricted to 2200'F and 17% local oxidation, respectively, consistent with 10 CFR 50.46. With respect to NLS2010046 Attachment 1 Page 6 of 11 success criteria, the ATWS long-term containment pressure and suppression pool temperature were compared to current CNS post-Loss of Coolant Accident limits of 56 psig and 208'F respectively.

For three SRVs OOS, all parameters met the ATWS success criteria with a setpoint drift of +70 psi or less. Seventy (70) psi drift is equivalent to an SRV lift pressure from 1184.5 psig to 1205.1 psig over a range of SRV NSPs of 1080 to 1100 psig, respectively.

3.5 Technical

Justification of Proposed Changes The GEH analysis shows success for both the ASME OPP and the ATWS events for three SRVs OOS for OPP safety function if SRV drift is less than 70 psi above ASME code tolerance.

LCO 3.4.3 allows a maximum SRV setpoint tolerance of +/-33 psig including drift. This supports submittal of this request to allow up to three SRVs OOS for OPP safety function as determined by 18-month surveillance testing.The GEH analysis is valid for the currently licensed reactor thermal power of 2419 MWth at CNS.3.6 USAR Accident Analysis Impact The effect of this change to TS 3.4.3 was evaluated for the USAR specified safety design criteria as follows.1. The GEH analysis (Enclosures 1 and 2) showed that only five valves were needed to accomplish the over-pressure protection required to prevent RCPB failure.2. Current CNS design basis analysis supports the ADS pneumatic pressure relief function requirement in TS 3.5.1. It did not require revision, because it was originally performed considering only five of the six TS required SRVs OPERABLE for ADS function.

Therefore, this is bounded by current analysis.3. Current design basis analysis did not require revision to address the reaction loading of discharge piping, because the maximum relief discharge capacities, and discharge piping restraints are not changed. Compared to valve operation atSRV NSPs, the OPP analysis increase in SRV critical steam flow is small and only for a short duration.

The flow difference only lasts until the RPV pressure is relieved to the NSP. The mass flow increase is relatively small and would not impact SRV discharge piping loads. This transient is bounded by current analysis.

NLS2010046 Attachment 1 Page 7 of 11 4. Current design basis analysis did not require revision to address the ability to withstand adverse combinations of loadings and forces from abnormal, accident or special event conditions, including asymmetrical torus loading. For the case of three SRVs OOS for OPP, the effect would be no greater than current analyses, because total steam discharged from the SRVs would be less. Thus, the new condition is bounded by current analysis.5. Since no physical changes are made to the SRVs, the LLS function and the ability to test or manually operate them is unchanged.

6. Since no physical changes are made to the SRVs, position indication is still available to the control room.Revising TS 3.4.3 by reducing the number of SRVs required to be OPERABLE for OPP from eight to five based on the GEH analysis in Enclosures 1 and 2 continues to preserve the safety design bases of the pressure relief system.3.7 Conclusion After the ATWS analysis demonstrated that three SRVs OOS for OPP safety function was acceptable, the other analyses were evaluated with that boundary condition.

The MSIV closure-flux scram event met all required pressure limits over the specified range of SRV setpoints.

The ATWS results showed that criteria of peak cladding temperature, containment pressure, and suppression pool temperature were met for all SRV setpoints.

Only the peak vessel pressure was noticeably affected by changing SRV setpoints.

The highest SRV setpoint drift that meets the vessel pressure analysis limits is +70 psi. The proposed change to TS 3.4.3 does not affect the SRV relief mode NSPs, or +/- 3% code tolerance for drift, and no changes are being made to valve mechanisms.

Therefore, based on the GEH analysis, the existing CNS containment and steam line integrity analyses are not adversely affected by the proposed TS change to enable CNS to operate with fewer SRVs in the safety mode function than currently specified TS 3.4.3.In summary, the proposed change is technically sound, conservative, and continues to maintain an acceptable level of safety.4.0 REGULATORY SAFETY ANALYSIS 4.1 Applicable Regulatory Requirements/Criteria Construction of CNS predated the 1971 issuance of 10 CFR 50, Appendix A,"General Design Criteria for Nuclear Power Plants." CNS USAR Appendix F,"Conformance to AEC Proposed General Design Criteria," discusses that CNS is designed to conform to the proposed general design criteria (GDC) published in the July 11, 1967, Federal Register, except where commitments were made to specific NLS2010046 Attachment 1 Page 8 of 11 1971 GDC. It notes that the Atomic Energy Commission accepted CNS conformance with these proposed GDC.The following is a discussion of the applicable regulations and the Draft GDC from USAR Appendix F, along with a discussion of continued conformance.

4.1.1 10 CFR 50.36, Technical Specifications 10 CFR 50.36(b) requires that each license authorizing operation of a utilization facility to include TS. 10 CFR 50.36(c) specifies the categories that are to be included in TS. 10 CFR 50.36(c)(2) identifies LCOs as one of the categories to be included in TS. 10 CFR 50.36(c)(2)(ii) states: "A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:" ..."(C) Criterion

3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The reduction in the number of TS required OPERABLE SRVs does not impact the pressure relief system's ability to perform its safety related functions.

Thus, LCO 3.4.3 will continue to be met. Therefore, CNS continues to meet this regulation with the proposed changes to TS LCO 3.4.3.4.1.2 USAR Appendix F Criterion 9, Reactor Coolant Pressure Boundary"The reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime." Since the physical configuration, operation, and setpoints of the SRVs are not changed, the reduction of the number of SRVs required by TS 3.4.3 to be OPERABLE has no impact on this criterion, and it continues to be satisfied.

4.1.3 USAR Appendix F Criterion 14, Core Protection Systems"Core protection systems, together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits." Since the physical configuration, operation, and setpoints of the SRVs are not changed, the reduction of the number of SRVs required by TS 3.4.3 to be NLS2010046 Attachment 1 Page 9 of 11 OPERABLE has no impact on this criterion, and it continues to be satisfied.

4.1.4 USAR Appendix F Criterion 37, Engineered Safety Features Basis for Design"Engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems. As a minimum, such engineered safety features shall be designed to cope with any size reactor coolant pressure boundary break up to and including the circumferential rupture of any pipe in that boundary assuming unobstructed discharge from both ends." Since the physical configuration, operation, and setpoints of the SRVs are not changed, the reduction of the number of SRVs required by TS 3.4.3 to be OPERABLE has no impact on this criterion, and it continues to be satisfied.

4.2 Precedent

Sufficiently similar precedent was not found. NPPD requests this amendment request be evaluated on its own merit.4.3 No Significant Hazards Consideration 10 CFR 50.91(a)(1) requires that licensee requests for operating license amendments be accompanied by an evaluation of no significant hazard posed by issuance of the amendment.

Nebraska Public Power District (NPPD),has evaluated this proposed amendment with respect to the criteria given in 10 CFR 50.92(c), The following is the evaluation required by 10 CFR 50.91 (a)(1).NPPD is requesting an amendment of the Operating License for the Cooper Nuclear Station (CNS) to revise Technical Specification (TS) 3.4.3, Safety/Relief Valves (SRVs) and Safety Valves (SVs). The proposed amendment reduces the number of SRVs required to be OPERABLE from eight to five, based on an analysis of over-pressure protection (OPP) events and anticipated transient without scram (ATWS)events that showed acceptable results with up to three SRVs out of service.1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No.This change to the TS does not change any physical configuration or setpoint of SRV lift. The probability of any accident would also be unchanged.

Therefore, the proposed change does not result in a significant increase in the probability of an accident previously evaluated.

NLS2010046 Attachment 1 Page 10 of I1 The change does reduce the number of SRVs originally assumed to be OPERABLE in design basis accident mitigation calculations.

However, analysis has shown that reducing the number of SRVs required to be OPERABLE from eight to five based on the General Electric Hitachi (GEH)analysis continues to preserve the safety design bases of the nuclear pressure relief system. Therefore, the change does not involve a significant increase in the consequences of an accident previously evaluated.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No.The proposed amendment does not change the design function or operation of the SRVs. The change does not create the possibility of a new or different kind of accident due to a credible new failure mechanism, malfunction, or accident initiator not considered in the design and licensing bases, because it does not change physical configuration or operation.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?Response:

No.The safety margins affected by this proposed change are the OPP pressure relief margin to Reactor Coolant System Pressure Boundary design pressure and the SRV steam flow to required ATWS mitigation steam flow. The GEH analysis performed to support this change demonstrates the safety margin after the change, though less than before the change, is still sufficient to ensure maximum pressure and required steam flows are within analysis success criteria.

The analysis success criteria are, in turn, below the accident and transient limits. Thus, the margin reduction is not significant.

The change does not exceed or alter a design basis or safety limit, and it does not significantly reduce the margin of safety.Therefore, the proposed change does not involve a significant reduction in a margin of safety.Based on the responses to the above questions, NPPD concludes that the proposed NLS2010046 Attachment 1 Page 11 of 11 amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusion

In conclusion, based on the considerations discussed above, (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.5.0 ENVIRONMENTAL CONSIDERATION 10 CFR 51.22 provides criteria for, and identification of, licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment or environmental impact statement.

10 CFR 51.22(c)(9) identifies an amendment to an operating license for a reactor which changes an inspection or a surveillance requirement as a categorical exclusion provided that operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amount of any effluents that may be released offsite, or (3) result in a significant increase in individual or cumulative occupational radiation exposure.NPPD review has determined that the proposed amendment, which would change a Technical Specification, does not involve (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluent that might be released offsite, or (3) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1 CNS Updated Safety Analysis Report Section IV-4 6.2 CNS Technical Specification Bases 3.4.3, Safety Analyses 6.3 ASME Boiler and, Pressure Vessel Code, 1968 Section III NLS2010046 Attachment 2 Page 1 of 2 Attachment 2 Proposed Technical Specification Revisions (Markup)Cooper Nuclear Station, Docket No. 50-298, DPR-46 Revised Technical Specification Page 3.4-6 SRVs and SVs 3.4.3 kj) 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.3 Safety/Relief Valves (SRVs) and Safety Valves (SVs)LCO 3.4.3 The safety function of ?/SRVs and 3 SVs shall be OPERABLE.APPLICABILITY:

MODES 1,'2, and 3.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SRVs or A.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SVs inoperable.

AND A.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> J Cooper 3.4-6 Amendment No. I-&

NLS2010046 Attachment 3 Page 1 of 2 Attachment 3 Proposed Technical Specification Revisions (Re-Typed)

Cooper Nuclear Station, Docket No. 50-298, DPR.-46 Revised Technical Specification Pages 3.4-6 SRVs and SVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.3 Safety/Relief Valves (SRVs) and Safety Valves (SVs)LCO 3.4.3 APPLICABILITY:

The safety function of 5 SRVs and 3 SVs shall be OPERABLE.MODES 1,2, and 3.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SRVs or SVs inoperable.

AND A.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Cooper 3.4-6 Amendment No.

NLS2010046 Attachment 4 Page 1 of 4 Attachment 4 Proposed Technical Specification Bases Revisions (Information Only)Cooper Nuclear Station, Docket No. 50-298, DPR--46 Revised Technical Specification Bases Pages B 3.4-15 B 3.4-16 B 3.4-18 SRVs and SVs B 3.4.3 BASES Ref. 7rcuar-rent-APPLICABLE analysis wa performe assuming a setpoint tol rance of + 3% for the SAFETY ANALYSES SRVs and Ss (Ref. 3. For the purpose of the setpoint tolerance (continued) analysis, SRVs and 3 SVs are assumed to operate in the safety mode. The analysis results demonstrate that the design SRV and SV capacity, with a setpoint tolerance of + 3%, is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig = 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the most severe design basis pressure transient.

From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above. Reference 4 discusses additional events that are expected to actuate the SRVs and SVs.SRVs and SVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 5)./ and Ref. 7 LCO The safety function of 4 SR s an SVs are required tore OPERABLE to satisfy the assumptions of the safety analysis (Ref. 3 .The requirements of this LCO, as they apply to the SRVs, are applicable only to the capability of the SRVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function).

The SRV and SV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied.

The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions.

The transient evaluations in Reference 3 are based on these setpoints, but also include the additional uncertainties of +/- 3% of the nominal setpoint to provide an added degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.Cooper B 3.4-15 SRVs and SVs B 3.4.3 BASES APPLICABILITY In MODES 1, 2, and 3, a4l SRVs an SVs must be OPERABLE, since considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The SRVs and SVs may be required to provide pressure relief to limit peak reactor pressure.In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents.

In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure.

The SRV and SV function is not needed during these conditions.

ACTIONS A.1 and A.2 of/-- the _required With the safety function of one or more RVs or SVs inoperable, a transient may result in the violation of the ASME Code limit on reactor pressure.

If the safety function of one or more SRVs or SVs is inoperable, the plant must be brought to a MO E in which the LCO does not apply. To achieve this status, the plant mu be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating expe ence, to reach required plant conditions from full power conditions in an o derly manner and without challenging plant systems. -of the required SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance requires that the SRVs and SVs will open at the pressures assumed in the safety analysis of Reference

3. The demonstration of the SRV and SV safety function lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

The SRV setpoint is +/- 3% for OPERABILITY; however, the valves are reset to +/- 1% during the Surveillance to allow for drift.Cooper B 3.4-16 SRVs and SVs B 3.4.3 BASES REFERENCES 1.2.3.4.5.6.ASME Boiler and Pressure Vessel Code,Section III.USAR, Section IV-4.9.NEDC-31628P, SRV Setpoint Tolerance Analysis for Cooper Nuclear Station, October 1988.USAR,Section XIV.10 CFR 50.36(c)(2)(ii).

ASME Code for Operation and Maintenance of Nuclear Power Plants.\_7. NEDC 10-032 Revision 1, Acceptance of GE SRV Out-of-Service Reports, October 29, 2010.Cooper B 3.4-18 Q4Q8A 0