ML082730528

Issuance of Amendment 158 Regarding Control Rod Notch Surveillance Test Frequency and Clarification of a Frequency Example

ML082730528
Person / Time
Site:	Monticello
Issue date:	11/19/2008
From:	Tam P Plant Licensing Branch III
To:	O'Connor T Northern States Power Co
Tam P
Shared Package
ML082730542	List: ML082730528
References
TAC MD9489
Download: ML082730528 (14)

Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 November 19, 2008 Mr. Timothy J. O'Connor Site Vice President Monticello Nuclear Generating Plant Northern States Power Company 2807 West County Road 75 Monticello, MN 55362-9637 MONTICELLO NUCLEAR GENERATING PLANT -ISSUANCE OF AMENDMENT REGARDING CONTROL ROD NOTCH SURVEILLANCE TEST FREQUENCY AND CLARIFICATION OF A FREQUENCY EXAMPLE (TAC NO. MD9489)

Dear Mr. O'Connor:

The Commission has issued the enclosed Amendment No. 158 to Renewed Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant. The amendment consists of changes to the Technical Specifications in response to your application dated April 22, 2008.

The amendment revises (1) the control rod notch surveillance frequency in Section 3.1.3, "Control Rod Operability," and (2) one example in Section 1.4, "Frequency," to clarify the applicability of the 1.25 surveillance test interval extension.

These changes were done pursuant to the previously approved Technical Specification Task Force (TSTF) change traveler 475, "Control Rod Notch Testing Frequency and SRM [Source Range Monitor] Insert Control Rod Action," Revision 1. A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely,

/'/'7 -, /v, "-Pefer S. Tam, enior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263

Enclosures:

1. Amendment No. 158 to DPR-22 2. Safety Evaluation cc w/encls: Distribution via Listserv.

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 NORTHERN STATES POWER COMPANY* DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment 1\10. 158 License No. DPR-22 The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment by Nuclear Management Company, LLC* (the licensee), dated April 22, 2008, 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; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 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. 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 Renewed Facility Operating License No. DPR-22 is hereby amended to read as follows:

• After September 22,2008, Nuclear Management Company, LLC, transferred its operating authority to its parent, Northern States Power Company, a Minnesota corporation (NSPM). By letter dated September 3,2008 (Accession No. ML082470648), NSPM stated that it accepts responsibility for all actions before the NRC staff which were previously initiated or addressed by Nuclear Management Company.

-2Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 158, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications. This license amendment is effective as of its date of issuance and shall be implemented within 90 days after issuance. Lois M. James, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance:

November 19, 2008 ATTACHMENT TO LICENSE AMENDMENT NO. RENEWED FACILITY OPERATING LICENSE NO. DOCKET NO. Replace the following page of Renewed Facility Operating License DPR-22 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change. REMOVE INSERT 3 3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 1.4-4 1.4-4 3.1.3-2 3.1.3-2 3.1.3-3 3.1.3-3 3.1.3-4 3.1.3-4 3.1.4-3 3.1.4-3

-3 Pursuant to the Act and 10 CFR Part 70, NSPM to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operations, as described in the Final Safety Analysis Report, as supplemented and amended, and the licensee's filings dated August 16, 1974 (those portions dealing with handling of reactor fuel) and August 17, 1977 (those portions dealing with fuel assembly storage capacity); Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess, but not separate, such byproduct and special nuclear material as may be produced by operation of the facility. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission, now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1775 megawatts (thermal). Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 158, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications. Physical Protection NSPM shall implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Renewed License No. DPR-22 Amendment No. 158 Frequency 1.4 1.4 Frequency EXAMPLES (continued)

EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY NOTE--Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 225% RTP. Perform channel adjustment. 7 days The interval continues, whether or not the unit operation is

< 25% RTP between performances. As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is

< 25% RTP, this Note allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after power reaches 2 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance was not performed within the 7 day interval (plus the extension allowed by SR 3.0.2), but operation was

< 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (pius the extension allowed by SR 3.0.2) with power 225% RTP. Once the unit reaches 25% RTP, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would be allowed for completing the Surveillance. If the Surveillance was not performed within this 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval (pius the extension allowed by SR 3.0.2), there would then be a failure to perform a Surveillance within the specified Frequency. and the provisions of SR 3.0.3 would apply.

Monticello 1.4-4 Amendment 158

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3.1.3 Control Rod OPERABILITY ACTIONS (continued)

CONDITION Two or more withdrawn control rods stuck. One or more control rods inoperable for reasons other than Condition A or B.

,..----

_. REQUIRED ACTION Perform SR 3.1.3.2 each OPERABLE control AND Perform SR 3.1.1.1. Be in MODE 3. RWM may be bypassed allowed by LCO 3.3.2.1, required, to allow of inoperable control and continued FUlly insert inoperable control rod.

AND Disarm the -COMPLETION TIME 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 12 hours 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 4 hours 3.1.3-2 Amendment No. 44G, 158

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ACTIONS (continued)

CONDITION Not applicable when THERMAL POWER > 10% RTP. Two or more inoperable control rods not in compliance with banked position withdrawal seq4ence(BPWS) and not separated by two or more OPERABLE control rods. Required Action and associated Completion Time of Condition A, C, or 0 not met.

OR Nine or more control rods inoperable.

Control Rod OPERABILITY 3.1.3 COMPLETION TIME 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 4 hours 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> REQUIRED ACTION Restore compliance with BPWS. OR Restore control rod to OPERABLE status. Be in MODE

3. SURVEILLANCE REQUIREMENTS FREQUENCY SR 3.1.3.1 Determine the position of each control rod.

SR ------------------------------NOTE Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.

Insert each withdrawn control rod at least one notch.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 31 days 3.1.3-3 Amendment No. 44e,' 158 Control Rod OPERABILITY 31.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.3.3 SR 3.1.3.4 Verify each control rod scram time from fully withdrawn to notch position 06 is 7 seconds. Verify each control rod does not go to the withdrawn overtravel position.

In accordance with SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 Each time the control rod is withdrawn to "full out" position AND Prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect coupling Monticello 3.1.3-4 Amendment No. 158 Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of Control Rod Scram OPERABLE control rods with scram times not within the limits of this Table are considered "slow." Enter applicable Conditions and Required Actions of LCO 3.1.3, "Control Rod OPERABILITY," for control rods with scram times> 7 seconds to notch position 06. These control rods are inoperable, in accordance with SR 3.1.3.3 and are not considered "slow." NOTCH POSITION 46 36 26 06 SCRAM TIMES(a)(b) (seconds)

WHEN REACTOR STEAM DOME PRESSURE;:> 800 psig 0.44 1.08 1.83 3.35 Maximum scram time from fully withdrawn position based on de-energization of scram pilot valve solenoids at time zero. Scram times as a function of reactor steam dome pressure when < 800 psig are within established limits.

3.1.4-3 Amendment No. 446;-158 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 158 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY* MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263

1.0 INTRODUCTION

By application dated April 22, 2008 (Accession No. ML081140388)

Nuclear Management Company, LLC (NMC*, the licensee), requested changes to the Technical Specifications (TS) for Monticello Nuclear Generating Plant (MNGP).

The proposed amendment would (1) revise the TS surveillance requirement (SR) frequency in Specification 3.1.3, "Control Rod Operability," and (2) revise Example 1.4-3 in Section 1.4, "Frequency," to clarify the applicability of the 1.25 surveillance test interval extension.

The licensee stated that the proposed amendment is consistent with Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler, TSTF-475, Revision 1. The NRC staff published a notice of this TS improvement in the Federal Register on November 11, 2007 (72 FR 63935) as part of the Consolidated Line Item Improvement Process. The notice contains a model Safety Evaluation; since the licensee adopted TSTF-475 with minor plant-specific terminology variations, the NRC staff has substantially reproduced the model Safety Evaluation (following below) with plant-specific changes to reflect the licensee's application.

The licensee's proposed changes are based on the NRC-approved TSTF change traveler TSTF-475, Revision 1, that revised the reference Standard Technical Specifications (STS) by:

(1) revising the frequency of SR 3.1.3.2, notch testing of each fully withdrawn control rod, from 7 days after the control rod is withdrawn and THERMAL POWER is greater than the Low Power Setpoint (LPSP) of the Rod Worth Minimizer (RWM) to "31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM" (NUREG-1433 and NUREG-1434) and (2) revising Example 1.4-3 in Section 1.4 "Frequency" to clarify that the 1.25 surveillance test interval extension in SR 3.0.2 is applicable to time periods discussed in NOTES in the "SURVEILLANCE" column in addition to the time periods in the "FREQUENCY" column (NUREG-1430 through NUREG-1434).

• After September 22, 2008, Nuclear Management Company, LLC, transferred its operating authority to its parent, Northern States Power Company, a Minnesota corporation (NSPM). By letter dated September 3, 2008 (Accession No. ML082470648). NSPM stated that it accepts responsibility for all actions before the NRC staff which were previously initiated or addressed by Nuclear Management Company.

-2 The purpose of the subject surveillances is to confirm control rod insertion capability, which is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. Control rods and the control rod drive (CRO) mechanism (CROM) by which the control rods are moved are components of the CRO system (CROS), which is the primary reactivity control system for the reactor. By design, the CROM is highly reliable with a tapered design of the index tube which is conducive to control rod insertion. A stuck control rod is an extremely rare event and industry review of plant operating experience did not identify any incidents of stuck control rods while performing a rod notch surveillance test. The purpose of these proposed TS revisions is to reduce the number of control rod manipulations and, thereby, reduce the opportunity for reactivity control events.

The purpose of the change to Example 1.4-3 in Section 1.4 "Frequency" is to clarify the applicability of the 25 percent allowance of SR 3.0.2 to time periods discussed in NOTES in the "SURVEILLANCE" column as well as to time periods in the FREQUENCY" column. REGULATORY EVALUATION Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix A, General Oesign Criterion (GOG) 29, Protection against anticipated occurrence, requires that the protection and reactivity control systems be designed to assure an extremely high probability of accomplishing their safety functions in an event of anticipated operational occurrences. The design relies on the CROS to function in conjunction with the protection systems under anticipated operational occurrences, including loss of power to all recirculation pumps, tripping of the turbine generator, isolation of the main condenser, and loss of all offsite power. The CROS provides an adequate means of inserting sufficient negative reactivity to shut down the reactor and prevent exceeding acceptable fuel design limits during anticipated operational occurrences. Meeting the requirements of GOC 29 for the CROS prevents occurrence of mechanisms that could result in fuel cladding damage such as severe overheating, excessive cladding strain, or exceeding the thermal margin limits during anticipated operational occurrences.

Preventing excessive cladding damage in the event of anticipated transients ensures maintenance of the integrity of the cladding as a fission product barrier. TECHNICAL EVALUATION The NRC staff previously reviewed the following information provided by the TSTF to support the staff's review and approval of TSTF-475, Revision 1.

Specifically, the following documents were reviewed during the NRC staff's evaluation: TSTF letter TSTF-04-07 (Reference 1) -Provided a description of the proposed changes in TSTF-475 that changes the weekly rod notch frequency to monthly and clarify the applicability of the 25 percent allowance in Example 1.4-3. TSTF letter TSTF-06-13 (Reference 4) -Provided responses to the NRC staff's request for additional information (RAI) on (1) industry experience with identifying stuck rods, (2) tests that would identify stuck rods, (3) continue compliance with Service Information Letter 139, (4) industry experience on collet failures, and (4) applying the 25 percent grace period to the 31 day control rod notch SR test frequency.

-3 Boiling Water Reactor Owners Group (BWROG) letter BWROG-06036 (Reference 5) -Provided the General Electric (GE)

Nuclear Energy Report, "CRD Notching Surveillance Testing for Limerick Generating Station," in which CRD notching frequency and CRD performance were evaluated. TSTF letter TSTF-07-19 (Reference 6) -Provided response to the NRC staff's RAI on CRD performance in Control Cell Core designed plants, including TSTF-475, Revision 1. The CRD System at MNGP is the primary reactivity control system for the reactor. The CRD System, in conjunction with the Reactor Protection System, provides the means for the reliable control of reactivity changes to ensure under all conditions of normal operation, including anticipated operational occurrences that specified acceptable fuel design limits are not exceeded. Control rods are components of the CRD System that have the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System. The CRD System at MNGP consists of a CRDM, by which the control rods are moved, and a hydraulic control unit (HCU) for each control rod. The CRDM is a mechanical hydraulic latching cylinder that positions the control blades. The CRDM is a highly reliable mechanism for inserting a control rod to the full-in position. The collet piston mechanism design feature ensures that the control rod will not be inadvertently withdrawn. This is accomplished by engaging the collet fingers, mounted on the collet piston, in notches located on the index tube. Due to the tapered design of the index tube notches, the collet piston mechanism will not impede rod insertion under normal insertion or scram conditions. The collet retainer tube (CRT) is a short tube welded to the upper end of the CRD which houses the collet mechanism which consist of the locking collet, collet piston, collet return spring and an unlocking cam. The collet mechanism provides the locking/unlocking mechanism that allows the insert/withdraw movement of the control rod. The CRT has three primary functions: (a) to carry the hydraulic unlocking pressure to the collet piston, (b) to provide an outer cylinder, with a suitable wear surface for the metal collet piston rings, and (c) to provide mechanical support for the guide cap, a component which incorporates the cam surface for holding the collet fingers open and also provides the upper rod guide or bushing. The NRC staff approved TSTF-475 which revised the TS SR 3.1.3.2, "Control Rod OPERABILITY" in the Standard Technical Specifications (NUREG-1433 and NUREG-1434) from 7 days to monthly based on the following: (1) slow crack growth rate of the CRT; (2) the improved CRT design; (3) a higher reliable method (scram time testing) to monitor CRD scram system functionality; (4) GE chemistry recommendations; and (5) no known CRD failures have been detected during the notch testing exercise, therefore, the NRC staff concluded that the changes would reduce the number of control rod manipulations thereby reducing the opportunity for potential reactivity events while having a very minimal impact on the extremely high reliability of the CRD system. The following paragraphs describe the bases for the staff's approval of TSTF-475:

According to the BWROG, at the time of the first CRT crack discovery in 1975, each partially or fully withdrawn operable control rod was required to be exercised one notch at least once each week. It was recognized that notch testing provided a method to demonstrate the integrity of the CRT. Control rod insertion capability was demonstrated by inserting each partially or fully

-4withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. It was determined that during scrams, the CRT temperature distribution changes substantially at reactor operating conditions. Relatively cold water moves upward through the inside of the CRT and exits via the flow holes into the annulus on the outside. At the same time, hot water from the reactor vessel flows downward on the outside surface of the CRT. There is very little mixing of the cold water flowing from the three flow holes into the annulus and the hot water flowing downward. Thus, there are substantial through-wall and circumferential temperature gradients during scrams which contribute to the observed CRT cracking.

Subsequently, many boiling-water reactors (BWRs) have reduced the frequency of notch testing for partially withdrawn control rods from weekly to monthly. The notch test frequency for fully withdrawn control rods are still performed weekly. The change for partially withdrawn control rods was made because of the potential power reduction required to allow control rod movement for partially withdrawn control rods, the desire to coordinate scheduling with other plant activities, and the fact that a large sample of control rods are still notch tested on a weekly basis. The operating experience related to the changes in CRD performance also provided additional justification to reduce the notch test frequency for the partially withdrawn control rods. In response to NRC's RAI and to support their position to reduce the CRD notch testing frequency, the BWROG provided plant data and a GE Nuclear Energy report entitled, "CRD Notching Surveillance Testing for Limerick Generating Station" (CRDNST). The GE report provided a description of the cracks noted on the original design CRT surfaces.

These cracks, which were later determined to be intergranular, were generally circumferential, and appeared with greatest frequency below and between the cooling water ports, in the area of the change in wall thickness.

Subsequently, cracks associated with residual stresses were also observed in the vicinity of the attachment weld. Continued circumferential cracking could lead to 360 degree severance of the CRT that would render the CRD inoperable which would prevent insertion, withdrawal or scram. Such failure would be detectable in any fully or partially withdrawn control rod during the surveillance notch testing required by the TSs. To a lesser degree, cracks have also been noted at the welded joint of the interim design CRT, but no cracks have been observed in the final improved CRT design.

Neither the BWROG nor the NRC staff was able to find evidence of a collet housing failure since 1975. To date, operating experience data shows no reports of a severed CRT at any BWR. No collet housing failures have been noted since 1975. On a numerical basis for instance, based on BWROG assumption that there are 137 control rods for a typical BWR/4 and 193 control rods for a typical BWR/6, the yearly performance would be 6590 rod notch tests for a BWR/4 plant and 9284 for a BWR/6 plant. For example, if all BWRs operating in the US are taken into consideration, the yearly performances of rod notch data would translate into approximately 240,000 rod notch tests without detecting a failure. In addition, the intergranular stress-corrosion cracking crack growth rates were evaluated, at Limerick Generating Station, using GE's PLEDGE model with the assumption that the water chemistry condition is based on GE recommendations. The model is based on fundamental principles of stress corrosion cracking which can evaluate crack growth rates as a function of water oxygen level, conductivity, material sensitization and applied loads. It was determined that the additional time of 24 days represented an additional 10 mils of growth in total crack length. The small difference in growth rate would have little effect on the behavior between one notch

-5test and the next subsequent test. Therefore, from the materials perspective based on low crack growth rates, a decrease in the notch test frequency would not affect the reliability of detecting a CRDM failure due to crack growth. Also, the BWR scram system has extremely high reliability. In addition to notch testing, scram time testing can identify failure of individual CRD operation resulting from IGSCC-initiated cracks and mechanical binding. Unlike the CRD notch tests, these single rod scram tests cover the other mechanical components such as scram pilot solenoid operated valves, the scram inlet and outlet air operated valves, and the scram accumulator, as well as operation of the control rods. Thus, the primary assurance of scram system reliability is provided by the scram time testing since it monitors the system scram operation and the complete travel of the control rod. Also, the HCUs, CRD drives, and control rods are tested during refueling outages, approximately every 18-24 months. Based on the data collected during the preceding cycle of operation, selected CRDs are inspected and, as required, their internal components are replaced.

Therefore, increasing the CRD notch testing frequency to monthly would have very minimal impact on the reliability of the scram system. The licensee stated in its application that it has reviewed the basis for the NRC staff's acceptance of TSTF-475, Revision 1, and concluded that the basis is applicable to MNGP, and supports its adoption of the TSTF-475 changes into the MNGP TS. The NRC staff also reviewed the TSTF-475, Revision 1 basis, and similarly concluded that the basis for the TSTF is applicable to MNGP, and therefore, the TSTF is appropriate for adoption by the licensee. In addition, the NRC staff reviewed the licensee's proposed changes against the corresponding changes made to the STSs TSTF-475, Revision 1, which the NRC staff has found to satisfy applicable regulatory requirements, as described above. The proposed changes would: (1) revise the TS control rod notch surveillance frequency in TS 3.1.3, "Control Rod Operability," and (2) revise one Example in Section 1.4 "Frequency" to clarify the applicability of the 1.25 surveillance test interval extension. The NRC staff found that the proposed changes are consistent with the changes approved by the NRC staff in TSTF-475, Revision 1, but with minor plant-specific terminology variations. The NRC staff, therefore, finds these changes acceptable.

The NRC staff has reviewed the licensee's proposal to amend existing MNGP TS sections SR 3.1.3.2, "Control Rod Operability," and Example 1.4-3 of Section 1.4, "Frequency," applicable to SR 3.0.2. The NRC staff has concluded that the TS revisions will have a minimal effect on the high reliability of the CRD system while reducing the opportunity for potential reactivity events; thus, the proposed amendment meets the requirement of 10 CFR Part 50, Appendix A, GDC 29, and will clarify the applicability of the 1.25 provision in SR 3.0.2.

Therefore, the NRC staff concludes that the proposed amendment is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATIONS This amendment changes a requirement with respect to use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the

-6types, of any effluent 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 (73 FR 52419). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment. 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. REFERENCES Letter TSTF-04-07 from the Technical Specifications Task Force to the NRC, TSTF-475 Revision 0, "Control Rod Notch Testing Frequency and SRM Insert Control Rod Action," August 30,2004, ADAMS Accession No. ML042520035. NUREG-1433, "Standard Technical Specifications General Electric Plants, BWR/4, Revision 3," August 31,2003. Letter TSTF-07-19, Response from the Technical Specifications Task Force to the NRC, "Request for Additional Information (RAI) Regarding TSTF-475 Revision 0, "Control Rod Notch Testing Frequency and SRM Insert Control Rod Action," dated February 28,2007, (TSTF-475 Revision 1 is an enclosure), ADAMS Accession No.

ML071420428. Letter TSTF-06-13 from the Technical Specifications Task Force to the NRC, "Response to NRC Request for Additional Information Regarding TSTF-475, Revision 0," dated July 3,2006, ADAMS Accession No. ML0618403421. Letter BWROG-06036 from the BWR Owners Group to the NRC, "Response to NRC Request for Additional Information Regarding TSTF-475, Revision 0," dated November 16, 2006, with Enclosure of the GE Nuclear Energy Report, "CRD Notching Surveillance Testing for Limerick Generating Station," dated November 2006, ADAMS Accession No. ML063250258. Letter TSTF-07-19 from the Technical Specifications Task Force to the NRC, "Response to NRC Request for Additional Information Regarding TSTF-475, Revision 0," dated May 22,2007, ADAMS Accession No. ML071420428. Principal Contributor: R. Grover November 19, 2008 November 19, 2008 Mr. Timothy J. O'Connor Site Vice President Monticello Nuclear Generating Plant Northern States Power Company 2807 West County Road 75 Monticello, MN 55362-9637 MONTICELLO NUCLEAR GENERATING PLANT -ISSUANCE OF AMENDMENT REGARDING CONTROL ROD NOTCH SURVEILLANCE TEST FREQUENCY AND CLARIFICATION OF A FREQUENCY EXAMPLE (TAC NO. MD 9489)

Dear Mr. O'Connor:

The Commission has issued the enclosed Amendment No. 158 to Renewed Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant. The amendment consists of changes to the Technical Specifications in response to your application dated April 22, 2008.

The amendment revises (1) the control rod notch surveillance frequency in Section 3.1.3, "Control Rod OPERABILITY," and (2) one example in Section 1.4, "Frequency," to clarify the applicability of the 1.25 surveillance test interval extension.

These changes were done pursuant to the previously approved Technical Specification Task Force (TSTF) change traveler 475, "Control Rod Notch Testing Frequency and SRM [Source Range Monitor] Insert Control Rod Action," Revision 1. A copy of our related safety evaluation is also enclosed.

The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, lRAI Peter S. Tam, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263

Enclosures:

1. Amendment No. 158 to 2. Safety cc wtencls: ,Ateditis: ,*.!Jbstribution via DISTRIBUTION PUBLIC LPL3-1 rtf RidsNrrDorlLpl3-'1 Resource RidsNrrPMPTamResource RidsNrrLATHarrisResource RidsOGCRpResource RidsAcrsAcnw_MailCTRResource RidsNrrDirsSbpb G. Hill, OIS ,I( RidsNrrDorlDpr RidsRgn3MailCenterResource '> R. Grover, NRR C))w Q, r-c Package Accession No.: ML082730542 f Amendment Accession No.: ML082730528 o Tech. Sec. a e Accession No.: MLO NRRIITSB/BC.A OFFICIAL RECORD COPY 101)108 1 10 08 THarri$Ji RElliott)' . LPL3-1/LA 101i1!08 LPL3-1/PM PTa OFFICE NAME DATE /I , *," I