NLS2007048, License Amendment Request for a One-Time Exception to the Five-Year Test Frequency for a Single Safety Valve

License Amendment Request for a One-Time Exception to the Five-Year Test Frequency for a Single Safety Valve

ML072350122
Person / Time
Site:	Cooper
Issue date:	08/16/2007
From:	Minahan S Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2007048
Download: ML072350122 (17)

Text

Nebraska Public Power District Always the're when you need us 50.90 NLS2007048 August 16, 2007 U.S. Nuclear Regulatory Comm-ission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

License Amendment Request for a One-Time Exception to the Five-Year Test Frequency for a Single Safety Valve Cooper Nuclear Station, Docket No. 50-298, DPR-46

Dear Sir or Madam,

The purpose of this letter is for the Nebraska Public Power District (NPPD) to request an amendment to Facility Operating License DPR-46 in accordance with the provisions of 10 CFR 50.4 and 10 CFR 50.90 to revise the Cooper Nuclear Station (CNS) Technical Specifications (TS). This request revises TS Section 5.5.6, Inservice Testing Program to allow a one-time exception of the five-year frequency requirem-ent for setpoint testing of safety valve MS-RV-7OARV.

The five-year interval for this safety valve expires on March 10, 2008, during the current operating cycle (Cycle 24). This request is based on the need to performn testing of this valve when shutdown. The next refueling outage is scheduled to begin in April 2008. The requested one-time exception involves an extension of 90 days from the current due date of March 10, 2008, to June 8, 2008. The surveillance requirement (SR) being extended is SR 3.4.3.1 as it specifically pertains to safety valve MS-RV-7OARV.

NPPD requests Nuclear Regulatory Commission (NRC) approval of the proposed TS change and issuance of the requested license amendment by February 8, 2008. Approval by that date is needed to avoid the need to start planning to shut down CNS on or before March 10, 2008.

Failure to obtain the requested amendment prior to March 10, 2008 would require an unnecessary shutdown of CNS. The amendment will be implemented within 30 days of issuance of the amendment. The one-time extension proposed in this amendment request expires upon shut down in the next refueling outage. provides a description of the proposed TS change, the techniical analysis basis for the change, the no significant hazards consideration evaluation pursuant to 10 CFR 50.91(a)(1),

and the environmental impact evaluation pursuant to 10 CFR 51.22. Attachment 2 provides COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 A D www.nppd.com I(-IeYL

NLS2007048 Page 2 of 3 marked up pages with the specific changes to the Current CNS TS. Attachment _3

)provides the revised TS pages in final formnat. No TS Bases pages are affected by this amendment reqluest.

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 226 dated October _3

)1, 2006, have been incorporated into this request. This request is submitted under oath 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 to the NRC Region IV office and the CNS Resident Inspector are also being provided in accordance wvith 10 CFR 50.4(b)(1I).

Should you have any questions concerning this matter, please contact David Van Der Kamnp, Acting Licensing Manager, at (402) 825-2904.

I declare under penalty of perjury that the foregoing is true and correct.

Executed On:

op~<d Date Sincerely, etewart B. Minaa Vice President - Nuclear and Chief Nuclear Officer

/rr Attachments cc:

Regional Administrator w/ attachments USNRC - Region IV Cooper Project Manager w/ attachments USNRC - NRR Project Directorate IV-lI Senior Resident Inspector w/ attachments USNRC -CNS

N LS2007048 Page 3 of 3 Nebraska Health and Human Service w/! attachments Department of Regulation and Li censure NPG Distribution Nv/o attachments CNS Records w/! attachments

NLS2007048 Attachment I Page I of 9 ATTACHMENT I LICENSE AMENDMENT REQUEST FOR A ONE-TIME EXCEPTION TO THE FIVE-YEAR TEST FREQUENCY FOR A SINGLE SAFETY RELIEF VALVE COOPER NUCLEAR STATION DOCKET NO. 50-298, DPR-46 Revised Technical Specification Page 5.0-10 1.0 Description 2).0 Proposed Change 3.0 Backgr-ound 4.0 Technical Analysis 5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requiremnents/Criteria 6.0 Environmental Consideration 7.0 References

NLS2007048 Attachment I Page 2 of 9 1.0 Description This license amendment request (LAR) proposes a one-time extension of setpoint testing of a single safety valve at the Nebraska Public Power District (NPPD) Cooper Nuclear Station (CNS). This extension w~ould be allowed by addition of an exception to the provision that prohibits application of the 25 percent extension of surveillance intervals of Surveillance Requirement (SR) 3.0.2 in Technical Specifications (TS) Section 5.5.6.,

Inservice Testing Program. The requested extension will extend the selpoint testing out to the next scheduled refueling outage, but no later than June 8, 2008.

2.0 Proposed Change This LAR proposes to revise TS Section 5.5.6, Inservice Testing Program, by adding a subparagraph to allow a one-time exception of the five-year frequency requirement for setpoint testing of safety valve (SV) MS-RV-70ARV in order to coincide with the Refueling Outage (RFO) 24 schedule. The specific change is to add the following as subparagraph I under paragraph b:

One-time Exceptioni: Seipoint testing of safety valve A'IS-R V-ZARKV as required byv ASME OM Code Mandatory Appendix 1, paragraph I-1326. mnay-be dela ' ed until start of Cycle 24 refuelinlg outage, but no later than June 8, 2008 (90 days fr-oi expiration of the 5-year interv'al on March 10, 2008).

There are no TS Bases for TS Section 5.5.6.

3.0

Background

The Nuclear System Pressure Relief Systemn consists of eight relief valves (also referred to as safe~ty/relief valves; LSRV]), and three SVs. These valves are located on the main steamn lines within the drywell, between the reactor pressure vessel (RPV) and the first main steamn isolation valve. The SRVs discharge to the suppression pool through piping connected to the valve. The SRVs provide the following three functions:

I. Overpressure relief operation. By automatic opening, the SRVs limit the pressure rise in the RPV and prevent opening of the SVs.

2. Overpressure safety operation. By automatic opening, the SRVs augmrent the SVs by opening to prevent nuclear systemn overpressurization.

3. Depressurization operation. The SRVs are opened automatically or manually as part of the Emergency Core Cooling System (ECCS) function.

The SVs open automnatically on pressure to protect against overpressure of the nuclear systemn. The SVs discharge directly to the interior of the drywvell.

NLS2007048 Attachment I Page 3 of 9 The safety objective of the pressure relief system is to prevent over-pressurization of the nuclear system., thereby protecting the reactor coolant pressure boundary from failure, and helping to prevent uncontrolled release of fission products. The automatic depressurization feature works in conjunction with the ECCS to re-flood the core, thereby protecting the nuclear fuel fromn failure due to overheating.

The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (GM Code), Mandatory Appendix 1, paragraph 1-1320, requires that Class I pressure relief valves shall be tested at least once every five years. However, it was discovered that a SV currently installed in the Main Steam (MS)

System will exceed the five-year test frequency Outlined in GM Code, Mandatory Appendix I, paragraph 1-1320.

The required extension is related to setpoint testing required by SR 3.4.3. 1. This SR requires v'erifying the safety function lift setpoints of the SRVs and SVs. Thle frequency of SR 3.4.3.1 is specified as "in accordance with the 1ST program." The purpose of the SR is to ensure that thle subject valves will open at the pressures assumed in thle safety analysis contained in the SRV Setpoint Tolerance Analysis for CNS, October 1998.

Demonstrating the SRV and SV safety function lift settings requires removing the valves fr-om the plant and shipping them to an offsite test facility. As a result the plant must be in cold shutdown to performn this testing.

The required setpoint testing of this valve had been considered for performance during RFO-23 in fall of 2006. When the scope of work for RFO-23 was being determined, RFO-24 was unofficially scheduled to start in March 2008. Based on that planned outage start date this testing of the valve was postponed until RFO-24.

This condition of MS-RV-70ARV exceeding the five-year frequency for setpoint testing has been entered into the CNS Corrective Action Programn.

The above informiation explains hlow this condition occurred and why the proposed TS change is necessary.

4.0 Technical Analysis TS Section 3.4.3, "Safety/Relief Valves (SRVs) and Safety Valves (SVs)," requires the SRVs and 5\\Is to be op~erable in Reactor Modes 1, 2, and 3. Because setpoint testing of SRVs and SVs Is conducted by bench testing of the valve, the SRV or SV mnust be removed from the plant. As a result the setpoint testing can only be conducted when the plant is shutdown.

The interval for SRV/SV testing specified in GM Code Appendix I is five years.

Extension of the fiv'e-year interval for SRV testing to coincide with a scheduled refueling outage is allowed by NUREG-1482, Guidelines for Inservice Testing at Nuclear Powver Plants. NUREG-1482, Section 3.1.3, Scheduling of Inservice Tests, discusses scheduling

NLS2007048 Attachment I Page 4 of 9 of tests. Table 3.2 in this section specifies the required frequency for 1ST activities of various tern-is, up to a maximumn tern-i of two years. This section also discusses the maximum extension of 25 percent of the test interval allowed by TS. However, CNS TS 5.5.6.b allows use of the 25 percent extension only for surveillances with intervals of two years or less. This section of NURFG-1482 also states: "H-owever, licensees should not extend the test intervals for safety and relief valves defined in Appendix I to the GM Code, other than to coincide wvith a refueling outage."

Setpoint testing of MS-RV-70ARV was last conducted on March 10, 2003. Thus, the five-year test frequency for this valve expires on March 10, 2008. The next refueling outage is currently scheduled to start in April 2008. If this outage starts as scheduled the requested exception would involve an extension of approximately one month for the setpoint testing. An extension of one month is an extension of less than two percent of the allowed interval of 60 months (five years). To accommodate delay in the start of the next refueling outage, NPPD is proposing an extension of 90 days. An extension of 90 days (three months) is a five percent extension of the allowed 60-month interval.

TI-i results of setpoint testing over the last ten-years were reviewed. This covers the last three performnances of the setpoint testing. The following table summarizes the results of ti-i as found setpoint testing performied on MS-RV-7OARV., as well as the as left values of the setpoint, for the most recent three tests. This shows that the tests were within the currently allowed as found tolerance (range) of plus-or-minus three percent around the nameplate test pressure of 1240 pslig. Following completion of the as found tests, the valve was refurbished and the setpoint left within plus-or-minus one percent.

Values of pressure in the table are in units of psig. The value il-the Deviation column is ti-i deviation fromn 1240 psig. Note that the as found acceptance criteria was revised in 1998 from plus-or-minus one percent to plus-or-minus three percent.

Summar-y of MS-RV-70ARV Setpoint Testing ASIFOUND AS LEFT Date Acceptable Results Deviation Date Setpoint Range April 9, 1227 to 1253 I " Act.- 1217

-1.85%

April 11, 1240 1997

(+ 1 %)

211 Act.- 1219 1997 October 9, 1202.S to I " Act.-]1252

+0.97%

October 15, 1242 1998 1277.2 (+/- 3%)

2' Act.- 1218 1998____

March 8, 1202.8 to Is' Act.- 1226

-1.13%

March 10, 1245 2003 1277.2 (+/- 3%)

2 n'Act.-

1227 2003____

Although the results of the as found test performied in April 1997 were outside the acceptable range at that time (plus-or-minus one percent) the results are within the acceptable range In use today (plus-or-minus three percent).

NLS2007048 Attachment I Page 5 of 9 Based on the results presented in the above table, the setpoint drift experienced by MS-RV-70ARV in the three most recent surveillances is acceptable. Based on that, there is reasonable expectation that the as found actuation setpoint of MS-RV-7OARV, following the requested extended peniod of five years and 90 days, would be within the current acceptable range of plus-or-m-inus three percent.

5.0 Regulatory Analysis 5.1 No Significant Hazards Consideration Nebraska Public Power District (NPPD) is requesting a revision to the Facility Operating License No. DPR-46 for Cooper Nuclear Station (CNS). The requested change proposes to add a provision to Technical Specification (TS) Section 5.5.6, Inservice Testing Program, to allow a one-timne extension of the interval for lperforning setpoint testing of one safety valve.

The Nuclear System Pressure Relief Systemn is comprised of eight safety-relief valves (SRVs) and three safety valves (SVs). Setpoint testing of these SRVs and SVs is perfonrmed by mneans of bench testing. This requires that the installed SRV or SV be removed from the Main Steam System. The plant must be shutdown to do this.

The interval for performning setpoint testing on SRVs and SVs is five years. The next refueling outage is currently scheduled to begin in April 2008. The five-year interval for one SV installed in CNS expires on March 10, 2008, approximately one month before the refueling outage is scheduled to start. To accommodate unanticipated delays in the start of the refueling outage, the requested extension is for a maximum of 90 days beyond the five-year Interval.

NPPD has evaluated whether or not a significant hazards consideration is involved wvith the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below.

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The function of SRVs and SVs is to p~rev'ent overpressurization of the reactor coolant systemn (RCS) during transients and abnormnal operation that could cause increases in RCS pressure. They are also used to depressurize the RCS when needed to allow injection of water from the high-volumne, lowv-pressure Emergency Core Cooling System (ECCS) Low Pressure Coolant Injection mode of the Residual Heat Removal Systemn into the reactor pressure vessel (RPV) as part of mitigation of an accident. Actuation or failure to actuate of a SRV or SV is not an initiator of any accident previously evaluated. Thus, this

NLS2007048 Attachm-ent 1 Page 6 of 9 proposed amendment would not result in a significant increase In thle probability of an accident previously evaluated.

A range or tolerance of plus-or-minus three percent of the setpoint pressure is acceptable for the results of setpoint testing. A 90-day extension of the interval for setpoint testing of one SV is not expected to result in actuation of the SV outside of its acceptable setpoint range. However, even if the single SV whose test interval is being extended did actuate outside of its acceptable range, it is not expected that this would result in a significant degradation in the ability of the Nuclear Systemn Pressure Relief Systemn to perform-its safety function, since the remaining eight SRVs and two other SVs would be unaffected by the proposed extension of the testing interval for the single SV.

The proposed change does not modify the design of or alter thle operation of systemns or components used in mitigating design basis accidents. Thus, this pr-oposed amendment wvould not result in a significant increase in the consequences of any accident previously evaluated.

Based onl the above, it is concluded that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

A new or different kind of accident fromn any previously evaluated might result from a modification of the plant design by either addition of a new systemn or removal of an existing systemn, or a change in how any of the plant systemns function during the operation of the plant. The proposed change does not modify the plant design, nor does it alter the operation of the plant or equipment involved in either routine plant operation or in the mitigation of thle design basis accidents.

Based on the above, it is concluded that the proposed change does not create the possibility of a newv or different kind of accident fromn any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The margin of safety applicable to this issue would be the margin between thle pressure at which the SRVs and SVs would actuate and the allowvable ASME Code overpressure limit of 1,375 psig (1 10 percent of vessel design pressure,

NLS2007048 Attachment I Page 7 of 9 1250 psig). This margin would be impacted if the setpoint at which the applicable SV actuated experienced drift greater than the allowable plus-or-minus three percent of the setpoint pressure. This is not expected to occur based on the results demonstrated by the setpoint testing conducted over the last ten years. Those results were two actuations of the SV at a pressure below the nameplate rating with less than two percent deviation, and one actuation at a pressure above the namneplate rating wvith less than one percent deviation. H-owever, even if this one SV did experience setpoint drift greater than the allowable plus-or-minus three percent, there wvould not be a sgificant reduction in the margin since it is expected that the remaining eight SRVs and the two other SVs wvould actuate within the allowable setpoint tolerance and begin to reduce RCS pressure as needed. Furthermnore., the proposed extension will not result in a change to the steamn discharge capacity and characteristics of the applicable SV.

Based on the above, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, NPPD concludes that the proposed amendment presents no sgnificant hazards consideration under the standards set forth in 10 CFR 50.92, paragraph (c), and, accordingly, a finding of no significant hazards consideration is justified.

5.2 Applicable Regulatory Requirements/Criteria A. ASME Boiler and Pressure Vessel (B&PV) Code Section III requires that the RPV be protected fr-om-overpressure during upset conditions by self-actuated SVs.

CNS complies with this ASME code requirement through the eight SRVs and three SVs in the Nuclear System Pressure Relief System. The requested extension does not significantly challenge in any manner the continued compliance with this code requirement.

B. ASME OM Code, Mandatory Appendix 1, Paragraph 1-1320, "Test Frequencies,, Class I Pressure Relief Valves," subparagraph (a) "5-Year Test Interval," requires that Class I pressure relief valves be tested at least once every five years. It states that the test interval for any individual valve shall not exceed five years.

The requested one-time extension of testing of the single SV *Involves an exception to this code requirement. Only the single SV is affected. The other two SVs and the eight SRVs in the CNS Nuclear Systemn Pressure Relief System, continue to comply with this code requirement.

NLS2007048 Attachment I Page 8 of 9 Conclusion The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be mnet. NPPD has determined that the proposed change does not require any exemptions or relief from regulatory requiremnents, other than the TS.

Applicable regulatory requirements will continue to be mnet, adequate defense-in-depth will be maintained, and sufficient safety margins will be maintained.

In conclusion, based on the considerations discussed above., (I ) there is reasonable assurance that the health and safety of the public will not be endangered by operation InI the proposed manner, (2) such activities will be conducted in compliance with the Commi-ission~s regulations, and (3) the Issuance of the amendment will not be Minimical to the common defense and security or to the health and safety of the public.

6.0 Environmental Consideration 10 CFR 51.22(c) provides categories of actions which are categorical exclusions from performing an environmental assessment. An action which is a categorical exclusion does not require an environmnental assessment or an environmental impact statemnent. 1 0 CFR 51.22(c)(9) allows as a categorical exclusion issuance of an amendment to a license for a reactor pursuant to 10 CFR Part 50 which changes a SR provided that (I ) the amendment involves no significant hazards consideration, (2) there is no significant change HIn the types or significant icesinthe amounts of any effluents that may be released off-site, and (3) there is no significant increase in individual or cumulative occupational radiation exposure.

NPPD has reviewed the proposed license amendment and concludes that it mneets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). The basis for this determnination is as follows:

I. The proposed license amendment does not involve significant hazards as described previously in the No Significant Hlazards Consideration Evaluation.

2. The proposed license amendment does not Introduce any new equipment, nor does it require any existing equipment or systems to perform a different type of fuinction than they are presently designed to perform. NPPD has concluded that this proposed change does not result in a signifiatcnginheypsosgifat increase in the amounts of any effluents that may be released off-site.

3. These changes do not adversely affect plant systems or operation and therefore, do not significantly increase individual or cumulative occupational radiation exposure beyond that already associated with normal operation.

NLS2007048 Attachment I Page 9 of 9 Therefore, pursuant to 10 CFR 5 1.22(c), no environmental impact statement or environmental assessment needs to be prepared in connection with issuance of the proposed license changes.

7.0 References 7.1 ASME Boiler and Pressure Vessel Code Section III; Article 9, "Protection Against Overpressure" 7.2 NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants"

NLS2007048 Page I of 2 ATTACHMENT 2 LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION FOR A ONE-TIME EXTENSION OF FIVE-YEAR TEST FREQUENCY FOR MS-RV-70ARV COOPER NUCLEAR STATION DOCKET NO. 50-298, DPR-46 Technical Specification Page - Markup Formnat 5.0-10

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves:

a.

Testing Frequencies applicable to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME GM Code) and applicable Addenda are as follows:

ASME GM Code and applicable Addenda I. One-time Exception:

terminology for Required Frequencies Setpoint testing of safety inservice testing for performing inservice valv____________ctivtie testing activities GM Code Mandatory Weekly At least once per 7 days Appedix,

m agay edlaye Monthly At least once per 31 days uINtIl start of Cycle 24 Qatryo vr refueling outage, but no 3 months At least once per 92 days later than June 8, 2008, Semiannually or (90 days from expiration every 6 months At least once per 184 days of the 5-year interval on Every 9 months At least once per 276 days March 10, 2008).

Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b.

The provisions of SIR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the nsricTesting Program for performing inservice testing activities;,

c.

The provisions of SR 3.0.3 are applicable to inservice testing activities;, and

d.

Nothing in the ASME GM Code shall be construed to supersede the requirements of any TS.

5.5-7

'Ventiia lion i.Filter Testing Program,(V FTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. Tests described in Specifications 5.5.7.a, 5-5.7.b, and 5-5.7.c shall be performed once per 18 months for standby service or after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of Amend ment,*5.-1 5.0-10

NLS2007048 Page 1 of'2 ATTACHMENT 3 LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION FOR A ONE-TIME EXTENSION OF FIVE-YEAR TEST FREQUENCY FOR MS-RV-70ARV COOPER NUCLEAR STATION DOCKET NO. 50-298, DPR-46 Technical Specification Page - Final Typed Format 5.0-10

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves:

a.

Testing Frequencies applicable to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda are as follows:

ASME OM Code and applicable Addenda terminology for inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days

b.

The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities;

1. One-time Exception: Setpoint testing of safety valve MS-RV-7OARV, as required by ASME OM Code Mandatory Appendix 1, paragraph 1-1 320, may be delayed until start of Cycle 24 refueling outage, but no later than June 8, 2008 (90 days from expiration of the 5-year interval on March 10, 2008).

c.

The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d.

Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

5.5.7 Ventilation Filter Testina Proaram (VFTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. Tests described in Specifications 5.5.7.a, 5.5.7.b, and 5.5.7.c shall Cooper 5.0-10 Coopr 5.-10Amendment

ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS@

0 ATTACHMVENT 3 LIST OF REGULATORY COMMITMVENTS© Correspondence Number: NLS2007048 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.

1-

4.

4.

4.

4.

.4.

.4.

IPROCEDURE 0.42 REVISION 22 PAGE 18OF 25