ML13168A373
| ML13168A373 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/28/2013 |
| From: | Beltz T A Plant Licensing Branch III |
| To: | Fili K D Northern States Power Co |
| Beltz T A | |
| References | |
| TAC ME9423 | |
| Download: ML13168A373 (25) | |
Text
UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 August 28, 2013 Mrs. Karen D. Fili Site Vice President Monticello Nuclear Generating Plant Northern States Power Company -Minnesota 2807 West County Road 75 Monticello, MN 55362-9637 MONTICEllO NUCLEAR GENERATING PLANT -ISSUANCE OF AMENDMENT NO. 175 TO REVISE THE RENEWED FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS (TAC NO. ME9423)
Dear Mrs. Fili:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 175 to Renewed Facility Operating License (FOl) No. DPR-22 for the Monticello Nuclear Generating Plant (MNGP). The amendment consists of changes, to the Renewed FOl and technical specifications (TSs) in response to your application dated August 21,2012, as supplemented by letters dC3;ted November 7,2012, and March 22,2013. The amendment corrects typographical errors, makes editorial changes, removes obsolete information, clarifies the fuel storage capacity to revise and align the existing FOl and TSs, and corrects a pagination error from a previously-issued license amendment.
A copy of our related safety evaluation is also enclosed.
The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. Sincerely, Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing . Office of Nuclear Reactor Regulation Docket No. 50-263
Enclosures:
Amendment No. 175 to DPR-22 Safety Evaluation cc w/encls: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 NORTHERN STATES POWER DOCKET NO. MONTICELLO NUCLEAR GENERATING AMENDMENT TO RENEWED FACILITY OPERATING Amendment No. 175 Renewed License No. DPR-22 The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment by Northern States Power Company (NSPM, the licensee), dated August 21,2012, as supplemented on November 7,2012, and March 22, 2013, 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 license 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: Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 175, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.
Enclosure 1
-2 The license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License DPR-22 and Technical Specifications Date of Issuance:
August 28, 2013 ATTACHMENT TO LICENSE AMENDMENT NO. RENEWED FACILITY OPERATING LICENSE NO. DOCKET NO. Replace the following pages of the Renewed Facility Operating License No. DPR-40 and the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. License Page REMOVE INSERT Front Page Front Page 3 -4 Technical Specifications (Appendix A) REMOVE INSERT Cover Page Cover Page 1.1-4 1.1-4 1.1-5 1.1-5 3.3.1.1-7 3.3.1.1-7 3.7.4-1 3.7.4-1 3.7.4-3 3.7.4-3 5.5-10 5.5-10 5.6-2 5.6-2 Additional Conditions (Appendix C) C-1 C-1 C-3 C-3 RENEWED FACILITY OPERATING LICENSE MONTICELLO NUCLEAR GENERATING UNIT MONTICELLO, NORTHERN STATES POWER DOCKET NO. November 8, 2006
-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); 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. 175, are hereby incorporated in the license. NSPM shalf . 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. -+-tJ::H:.H 175 Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p)(2).
The combined set of plans which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Monticello Nuclear Generating Plant Physical Security, Training and Qualification, and Safeguards Contingency Plan," with revisions submitted through May 12, 2006. NSPM shall fully implement and maintain in effect all provisions of the Commission-approved Northern State,s Power Company -Minnesota (NSPM) Cyber Security Plan (CSP), including changes made pursuant to , the authority of 10 CFR 50.90 and 10 CFR 50.54(p).
The'NSPM CSP was approved by License Amendment No. 166. 4. Fire Protection NSPM shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Safety Analysis Report for the facility and as approved in the SER dated August 29,1979, and supplements dated February 12,1981 and October 2, 1985, subject to the following provision:
NSPM may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. 5. Emergency Preparedness Plan NSPM shall follow and maintain in effect emergency plans which meet the standards 0(10 CFR 50.47(b) and the requirements in 10 CFR 50, Appendix E, including amendments and changes made pursuant to the authority of 10 CFR 50.54(q).
The licensee shall meet the requirements of 10 CFR 50.54(s), 50.54(t), and 50.54(u).
- 6. TMI Action Plan NSPM has satisfactorily met all TMI-2 Lessons Learned Category "A" requirements applicable to the facility.
NSPM shall make a timely submittal in response to the letter dated October 31, 1980 regarding TMI requirements from Darrell G, Eisenhut, Director, Division of Licensing, Office of Nuclear Reactor Regulation to All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits (NUREG-0737), 7. Repairs to the Recirculation System Piping The repairs to the recirculation system piping are approved and the unit is hereby authorized to return to power operation, subject to the following condition:
Prior to the startup of Cycle 11, NSPM shall submit by August 1, 1983 for the Commission's review and approval, a program for inspection and/or modification of the recirculation system piping. Renewed License No. DPR-22 Amendment No . .:tOO, 175 APPENDIX RENEWED FACILITY OPERATING LICENSE MONTICELLO NUCLEAR GENERATING UNIT MONTICELLO, NORTHERN STATES POWER DOCKET NO. October 29,2006 Definitions 1.1 Definitions OPERABLE -OPERABILITY PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) RATED THERMAL POWER (RTP) REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME SHUTDOWN MARGIN (SDM) STAGGERED TEST BASIS A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.5. RTP shall be a total reactor core heat transfer rate to reactor coolant of 1775 The RPS RESPONSE TIME shall be that time interval from initiation of any RPS channel trip to the de-energization of the scram pilot valve solenoids.
The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that: The reactor is xenon free; The moderator temperature is 68°F; and All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM. A STAGGERED TEST BASIS.shall consist of the testing of . one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
1.1-4 Amendment No. 146, 172, 175 1.1 Definitions 1.1 Definitions THERMAL POWER TURBINE BYPASS SYSTEM RESPONSE TIME THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. The TURBINE BYPASS SYSTEM RESPONSE TIME shall be that time interval from when the main turbine trip solenoid is activated until 80% of the turbine bypass capacity is established.
The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
Monticello 1.1-5 Amendment No. -Me, 175 RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 4) Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION 0.1 REQUIREMENTS VALUE 3(0)Neutron Flux -High F SR 3.3.1.1.1
- s 122% RTP SR 3.3.1.1.2 SR 3.3.1.1.4 SR 3.3.1.1.6 SR 3.3.1.1.11(f)(9)
SR 3.3.1.1.15 3(C)Inop. 1,2 G SR 3.3.1.1.4 NA SR 3.3.1.1.15 2-0ut-Of-4 Voter 1,2 2 G SR 3.3.1.1.1 NA SR 3.3.1.1.4 SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15 3(C)OPRM Upscale 2:20% RTP SR 3.3.1.1.1 As specified SR 3.3.1.1.4 in COLR SR 3.3.1.1.6 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.16 . Reactor Vessel Steam 1,2 2 G SR 3.3.1.1.4
- s 1075 psig Dome Pressure -SR 3.3.1.1.7 SR 3.3.1.1.9 SR 3.3.1.1.12 SR 3.3.1.1.14
- 4, Reactor Vessel Water 1,2 2 G SR 33.11.1 :e: 7 inches Level-SR 3.3.1.1.4 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.14 Each APRM / OPRM channel provides inputs to both trip systems. (Not Used.) If the as-found channel setpoint is not the Nominal Trip Setpoint but is conservative with respect to the Allowable Value, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. The instrument channel setpoint shall be reset to the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.
The NTSP and the methodology used to determine the NTSP are specified in the Technical Requirements Manual.
3.3.1.1-7 Amendment No. 146, 159, 175 CREF System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Filtration (CREF) System LCO 3.7.4 Two CREF subsystems shall be OPERABLE.
NOT E The control room envelope (CRE) boundary may be opened intermittently under administrative control. APPLICABILITY:
MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREF subsystem inoperable for reasons other than Condition B. A.1 CREF subsystem to OPERABLE status. 7 days B. One or more CREF subsystems inoperable due to inoperable CRE boundary in MODE 1,2, or 3. , B.1 AND B.2 AND B.3 Initiate action to implement mitigating actions. Verify mitigating actions ensure CRE occupant exposures to radiological, chemical and smoke hazards will not exceed limits. Restore CRE, boundary to OPERABLE status. Immediately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 90 days Monticello 3.7.4-1 Amendment No. 146, 160, 175 3.7.4 CREF System ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME F. Two CREF subsystems inoperable during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs. One or more CREF subsystems inoperable due to an inoperable CRE boundary during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs. --------------------N 0 T E LCO 3.0.3 is not applicable.
F.1 Suspend movement of recently irradiated fuel assemblies in the secondary containment.
AND .F.2 Initiate action to suspend OPDRVs. Immediately Immediately SU RVEI LLANCE FREQUENCY SR 3.7.4.1 Operate each CREF subsystem for 10 continuous hours with the heaters operating.
' 31 days SR 3.7.4.2 SR 3.7.4.3 Perform required CREF filter testing in accordance with the Ventilation Filter Testing Program (VFTP). , , " Verify each CREF subsystem actuates on an actual or simulated initiation signal. , In accordance with VFTP 24 months SR 3.7.4.4 Perform required CRE unfiltered air in-leakage testing in accordance with the Control Room , Envelope Habitability Program. In accordance with the Control Room Envelope Habitability Program Monticello 3.7.4-3 Amendment No. 146, 160, 175 Programs and Manuals 5.5 Programs and Manuals 5.5.10 Safety Function Determination Program (SFDP) (continued) A required system redundant to the support system(s) for the supported systems described in Specifications 5.5.10.b.1 and 5.5.10.b.2 above is also inoperable. The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are'those of the support system. 5,5.11 Primary Containment Leakage Rate Testing Program A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as . modified by approved exemptions.
This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions: The Type A testing Frequency specified in NEI 94-01, Revision 0, Paragraph 9.2.3, as "at least once per 10 years based on acceptable performance history" is modified to be "at least once per 15 years based on acceptable performance history." This change applies only to the interval following the Type A test performed in March 1993; The main steam line pathway leakage contribution is excluded from the sum of the leakage rates from Type Band C tests specified in Section III.B of 10 CFR 50, Appendix J, Option B, Section 6.4.4 of ANSI/ANS 56.8-1994, and Section 10.2 of NEI 94-01, Rev. 0; and The main steam line pathway leakage contribution is excluded from the overall integrated leakage rate from Type A tests specified in Section liLA 0(10 CFR 50, Appendix J, Option S, Section 3.2 of ANSIIANS 56.8-1994, and Section 8.0 and 9.0 of NEI 94-01, Rev. O. The calculated peak containment internal pressure for the design basis loss of coolant accident, P a , is 42 psig. The containment design pressure is 56 psig. The maximum allowable containment leakage rate, La, at P a , shall be 1.2% of containment air weight per day.
5.5-10 Amendment No. 146148,175 5.6.3 Reporting Requirements 5.6 Reporting Requirements CORE OPERATING LIMITS REPORT (COLR) (continued) Control Rod Block Instrumentation Allowable Value for the Table 3.3.2.1-1 Rod Block Monitor Functions 1.a, 1.b, and 1.c and associated Applicability RTP levels; Reactor Protection System Instrumentation Delta W value for Table 3.3.1.1-1, Function 2.b, APRM Simulated Thermal Power -High, Note b; and Reactor Protection System Instrumentation Period Based Detection Algorithm trip setpoints associated with Table 3.3.1.1-1, Function 2.f, OPRM Upscale. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel"; . (Not Used.) (Not Used.) NEDO-31960, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology";
and NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," August 1996. The COLR will contain the complete identification for each of the Technical Specification referenced topical reports used to prepare the COLR (I.e., report number, title, revision, date, and any supplements). The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, . Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety. analysis are met. 'The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6-2 Amendment No. 146, 159, 175 APPENDIX ADDITIONAL RENEWED FACILITY OPERATING LICENSE NO. Northern States Power Company shall comply with the following conditions on the schedules noted below: Amendment Number Additional Condition 98 The emergency operating procedures (EOPs) shall be changed to require manual isolation of torus and drywell sprays prior to the point where primary containment pressure would not provide adeq.uate net positive suction head (NPSH) for the emergency core cooling system (ECCS) pumps, change the caution statement regarding NPSH in the Primary Containment Pressure EOP to include cores spray pumps, and add a caution statement regarding NPSH considerations for pressure control while venting to control primary containment pressure.
98 Finalize the additional containment analysis and associated NPSH evaluation which extends the existing long-term cause evaluation to a time when the required containment overpressure returns to atmospheric Changes to the requested long-term containment overpressure, if any, shall be promptly reported to the NRC prior to starting up the unit from the current maintenance.
outage. 98 Submit the results of the additional containment analysis and associate NPSH evaluation discussed above. Prior to starting up from the current maintenance outage, or August 1, 1997, whichever is later. Prior to up from . outage or August 1, 1997, whichever is later.. Within 90 days from the date of the plant startup from the cu rrent maintenance outage or November 1, 1997, whichever is later. C-1 Amendment No. 98, 110, 175 APPENDIX C -continued Amendment Number Additional Condition 102 All affected process computer and SPDS data points shall be changed to reflect uprate operating conditions. Control room simulator changes shall be completed in accordance with ANSIIANS 3.5-1985 Section 5.4.1. Simulator Performance Testing. and Monticello simulator configuration control procedures. Classroom and simulator training on new knowledge and abilities associated with the power uprate shall be provided in accordance with Monticello Training Center procedures. NSPM shall monitor plant operational parameters for uprate impacts on the PRA models. Control room simulator changes shall be verified against actual plant startup data. The applicable training programs and the simulator.
shall be modified.
or appropriate compensatory actions shall be taken, in accordance with the Monticello Training Center procedures toi reflect issues and discrepancies identified during startup testing. 102 The MNGP USAR shall be updated to reflect the changes associated with power uprate This update shall not include credit for suppression pool scrubbing in the MSIV leakage pathway in the revised LOCA analysis. . Prior to implementation of Amendment No.1 02 (prior to exceeding 1670 MWt). Prior to implementation of Amendment No. 102 (prior to exceeding 1670 MWt). Prior to implementation of Amendment No. 102 (prior to exceeding 1670 MWt). During and after the power uprate ascension test program. Within 3 months of completion of the power uprate ascension test program. Within 6 months of completion of the power uprate ascension test program. Within 9 months of completion of the power uprate ascension test program. Amendment No. 102, 110 , 175 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 175 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 to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated August 21,2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 12237A321), as supplemented by letters dated November 7,2012, and March 22,2013 (ADAMS Accession Nos. ML 12314A315 and ML 13085A017, respectively), Northern States Power Company -Minnesota (NSPM, the licensee) submitted a license amendment request for the Monticello Nuclear*Generating Plant (MNGP). The proposed amendment revises and aligns the existing Renewed Facility Operating License and the technical specifications (TSs) by correction of typographical errors, incorporating minor editorial changes, removal of obsolete information, and clarification of the utilized fuel storage capacity.
The supplemental letters dated November 7,2012, and March 22, 2013, provided additional information that did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on December 26,2012 (77 FR 76081).
2.0 REGULATORY EVALUATION
The NRC's regulatory requirements related to the content of the TSs are contained in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36, "Technical specifications." Pursuant to 10 CFR 50.36(b), each operating license issued by the Commission includes technical specifications.
In SECY-96-238, dated November 19, 1996 (ADAMS Legacy Library Accession No. 9611250030), the NRC staff informed the Commission of the intent to issue guidance to staff members for determining what action is necessary to correct a typographical error associated with power reactor TSs. In a Staff Requirements Memorandum dated December 17, 1996 (ADAMS Accession No. ML003754054), the Commission provided comments on the guidance and stated that it did not object to the proposed guidance.
The actual guidance was issued in a memorandum dated January 16, 1997 (ADAMS Accession No. ML 1032'60096).
-In general, typographical errors discovered in the TSs for which the origin of the error is unknown must be corrected through the normal processing of a license amendment request to change the TSs. An exception to this general rule is the case in which the NRC staff or the licensee can demonstrate that the error was introduced inadvertently in a particular license amendment and that the erroneous change was not addressed in the notice to the public nor reviewed by the staff. Under limited circumstances, a change that introduced a typographical error was not a proper amendment to the license because it was neither addressed in the notice nor reviewed, and correction of the typographical error is not a "change" to the TS. In these specific cases, a typographical error may be corrected by a letter to the licensee from the NRC staff instead of an amendment to the license, but does notpreclude issuance of a license amendment to resolve the discrepancy.
As mentioned in Section 1.0 above, the scope of the license amendment request is to correct typographical errors, incorporate minor editorial changes, remove of obsolete information, and clarify the utilized fuel storage capacity, and the changes affect the renewed facility operating license including the technical specifications. TECHNICAL EVALUATION Revise the Front Page of the Renewed Facility Operating License to Reference Northern States Power Company -Minnesota or NSPM and Remove Reference to "Nuclear Management Company, LLC" The licensee proposed to revise the cover page of MNGP Renewed Facility Operating License to remove the reference to "NUCLEAR MANAGEMENT COMPANY, LLC." The NRC approved and issued License Amendment No. 156 on September 22, 2008 (ADAMS Accession No. ML082590580).
The amendment approved the license transfer from the Nuclear Management Company to NSMP. The NRC staff considers that the change removes an obsolete reference and is consistent with NRC-approved License Amendment No. 156. Based on the above, the NRC staff considers the change to be editorial and corrective in nature, in that it removes an obsolete reference which should have been removed when License Amendment No. 156 was issued. Therefore, the NRC staff considers the change to be acceptable. Revise the Front Page of the Facility Operating License by adding "RENEWED" to "FACILITY OPERATING LICENSE The licensee proposed to add the word "RENEWED" in front of "FACILITY OPERATING r LICENSE" on the front page to the MNGP Facility Operating License. In November 2006, the NRC approved and issued Renewed Facility Operating License No. 22 (ADAMS Accession No. ML062760125).
The word "RENEWED" should have been incorporated when the NRC issued the renewed facility operating license in November 2006. Based on the above, the NRC staff considers this change to be editorial or corrective in nature and, therefore, is acceptable.
Enclosure 2
-Revise License Condition 2.B.2 to Remove Outdated Reference to Spent Fuel Pool (SFP) Capacity Letter The original capacity of the MNGP SFP was 740 fuel assemblies.
In a letter dated April 14, 1978, the NRC issued License Amendment No. 34 (ADAMS Accession No. ML020880176), approving an increase in SFP capacity from 740 to 2237 fuel storage locations.
Paragraph
- 2. B of the Provisional Operating License was revised to read as follows: Pursuant to the Act and 10 CFR part 70, 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 operation, 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 reactor fuel) and August 17, 1977 (those portions dealing with fuel assembly storage capacity);
In a letter dated June 5,2006, the NRC issued License Amendment No. 146 (ADAMS Accession No. ML061 070577), approving the conversion of the MNGP TSs to the improvedTS (ITS) format. The intent of the ITS conversion is to provide clearer and more readily understandable TS requirements to ensure safer operation of the unit. The amendment added a subsection to TS 4.3, "Fuel Storage," entitled TS 4.3.3, "Capacity," which read as follows: The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2237 fuel assemblies.
The licensee proposes to remove the reference to the August 17. 1977. filing associated with fuel assembly storage capacity in License Condition 2.B.2. The MNGP fuel assembly storage capacity information is currently controlled under TS 4.3.3 with the adoption of ITS. Therefore, the information in License Condition 2.8.2 referencing the August 17. 1977, letter, is outdated and redundant.
Based on the above. the NRC staff finds that removing the reference to the August 17,1977, filing associated with fuel assembly storage capacity in License Condition 2.8.2, to be acceptable. License Condition 2.C.5 of the Renewed Facility Operating License -Editorial Correction In License Condition 2.C.5, "Emergency Preparedness Plan," of the Renewed Facility Operating License. the word "hall" should be changed to "shall." The NRC staff considers that this change is an editorial correction and. therefore, is acceptable. Revise the Cover Page to Appendix A of the Monticello Nuclear Generating Plant Technical Specifications The licensee proposed to add the word "RENEWED" in front of "FACILITY OPERATING LICENSE" on the cover page to Appendix A of the MNGP TSs. Based on the evaluation in Section 3.2. above. the NRC staff considers this change to be corrective in nature and to be acceptable.
Additionally, the licensee proposed to delete the reference to "NUCLEAR MANAGEMENT COMPANY, LLC" on the cover page to Appendix A of the MNGP TSs. Based on the evaluation in Section 3.1, above, the NRC staff considers this change to be editorial or corrective in nature and to be acceptable. TS 3.3.1.1, "RPS Instrumentation" -Remove Operating Power Range Monitoring (OPRM) System Note in Table 3.3.1.1-1 License Amendment 159 was approved and issued on January 30, 2009 (ADAMS Accession No. ML083440681).
The amendment revised TS 3.3.1.1, "Reaetor Protection System Instrumentation," functions to reflect adoption of the Power Range Neutron Monitoring System (PRNMS) at MNGP. A 90-day monitoring period was requested for the Operating Power Range Monitoring (OPRM) System during which time the OPRM system would disabled such that an actuation would not provide a trip. The licensee requested this 90-day monitoring period to verify proper OPRM system operation and prevent spurious trips. Upon issuance of License Amendment 159, note (e) was added to TS Table 3.3.1.1. -1 stating the following: During the OPRM Monitoring Period the OPRM Upscale function In its August 21, 2012, application, the licensee stated that the OPRM monitoring period of the PRNMS is complete and the system is in full operation.
Therefore,.
the note is no longer applicable and may be . Based on the above, the NRC staff concludes that note (e) of TS Table 3.3.1.1-1 may now be removed because the 90-day monitoring period was completed in 2009. The NRC staff considers note (e) of TS Table 3.3.1.1-1 to be obsolete and moot. Therefore, removal is acceptable. TS 3.7.4, "Control Room Emergency Filtration (CREF) System" -Editorial Corrections In TS Section 3.7.4, "Control Room Emergency Filtration (CREF) System," two "AND" logical connectors associated with the Required Actions of Condition B, and an "OR" logical connector associated with Condition F, should be underlined.
These conditions were inadvertently introduced during issuance of License Amendment No. 160, dated March 17, 2009 (ADAMS Accession No. ML083640529).
' As discussed in TS Section 1.2, "Logical Connectors," logical connectors are used to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies.
The only logical connectors that appear in TSs are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings.
Several levels of logic may be used to state Required Actions, and these levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action.
-In its April 3, 2008, license amendment application (ADAMS Accession No. ML080950329), the licensee provided marked-up TS pages identifying proposed changes. The TS pages 3.7.4-1 and 3.7.4-3 were provided with mark-ups, and the logical connectors for the Required Actions of Condition B, 'and for Condition F, included the required underlines.
In the NRC-approved License Amendment No. 160, dated March 17, 2009, the underlines were inadvertently omitted from the logical connectors on TS pages 3.7.4-1 and 3.7.4-3. Based on the above, the NRC staff concludes that the two "AND" logical connectors associated with the Required Actions of Condition B, and an "OR" logical connector associated with Condition F, should be underlined.
'The NRC staff considers that this corrects an inadvertent omission and is considered editorial and, therefore, is acceptable. TS 5.5.11, "Primary Containment Leakage Rate Testing Program" -Editorial Correction In TS Section 5.5.11, "Primary Containment Leakage Rate Testing Program," as "s" is currently omitted from the end of the word "exception." The NRC staff considers this change to be an editorial correction and, therefore, is acceptable. TS 5.6.3, "CORE OPERATING LIMITS REPORT (COLR) -Correction of Incorrect Phrase License Amendment 159 was approved and issued on January 30,2009 (ADAMS Accession No. ML083440681).
The amendment revised TS 3.3.1.1, "Reactor Protection System Instrumentation," functions to reflect adoption of the Power Range Neutron Monitoring System (PRNMS) at MNGP. As part of the approved change, Note (b) to TS Table 3.3.1.1-1, "Reactor Protection System Instrumentation," was revised to reflect a new equation for the APRM [Average Power Range Monitor] Simulated Thermal Power -High function for single-loop operation.
Note (b) currently states the following: (b) 0.66 (W -Delta W) + 61.6% RTP, when reset for single loop operation . per LCO 3.4.1, "Recirculation Loops Operating." The cycle-specific value Delta W is specified in the Item a,5 of TS 5.6,3, "CORE OPERATING LIMITS REPORT (COLR)," states the following:
Reactor Protection System Instrumentation Delta W Allowable Value for Table 3.3.1.1-1, Function 2. b, APRM Simulated Thermal Power -High, Note b; and Note (b) to Table 3.3.1.1-1 correctly indicates that the cycle-specific value for Delta W is specified in the COLR. In the COLR Specification 5.6.3, however, Delta W is incorrectly identified as an Allowable Value. The licensee proposes to remove "Allowable" from the statement and change "Value" to "value," thus providing consistency between the terminology in Table 3.3.1.1-1 and TS 5.6.3. Item a.5 in TS 5.6.3, "CORE OPERATING LIMITS REPORT (COLR)," would be changed as follows:
-6 Reactor Protection System Instrumentation Delta W value for Table Function 2.b, APRM Simulated Thermal Power -High, Note b; Based on the above, the cycle-specific value for pelta W is not an Allowable Value as defined in the TSs and, therefore, the NRC staff finds that the proposed change to be acceptable. TS 5.6.3, "CORE OPERATING LIMITS REPORT (COLR) -Remove Reference to No Longer Utilized Analytical Methods The licensee proposed to remove the following reports for determining the core operating limits: NSPNAD-8608-A, "Reload Safety Evaluation Methods for Application to the Monticello Nuclear Generating Plant" NSPNAD-8609-A, "Qualification of Reactor Physics Methods for Application The licensee stated in its August 21,2012, application that the above analytical methods are no longer utilized for determination of the licensed operating limits at MNGP. Furthermore, the licensee provided additional technical justification in its November 7, 2012, supplemental letter: The NRC-approved Global Nuclear Fuels (GNF) / General Electric/Hitachi (GEH) licensing methods are currently used to determine operating limits for Monticello.
These GNF/GEH methods are referred to in the TSs (and COLR) and have been used for Cycles 22 through 26 (approximately 10 years). The use of the NAD methods was discontinued due to these methods being superseded by the GNF/GEH methods. Based on the above, the NRC staff finds that removal of these two analytical methodologies from TS 5.6.3 is acceptable as they are no longer being utilized for determination of core operating limits. TS 1.1. "Definitions" -Correction of Error of Omission Associated with Issuance of License Amendment No. 172 On February 27, 2013, the NRC issued License Amendment No. 172 (ADAMS Accession No. ML 13025A 155), in response to NSPM's application dated January 20, 2012 (ADAMS Accession No. ML 12033A175), as supplemented by letter dated December 7,2012 (ADAMS Accession No. ML 12349A21 0). Subsequent to issuance of the approved amendment request, the licensee informed the NRC that two definitions had been omitted from TS Section 1.1. Specifically, the definitions of "STAGGERED TEST BASIS" and "THERMAL POWER" were missing. In the licensee's January 20,2012, application, the definition of "PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)" would be added to TS Section 1.1, page 1.1-4. As a result of this new definition on page 1.1-4, the aforementioned two definitions were forced off the page. The licensee did not provide a revised page 1.1-5 with the two definitions, and this omission was not identified by the NRC staff prior to issuing License Amendment No. 172.
-7 Based on the above, the NRC staff finds that omission of definitions "STAGGERED TEST BASIS" AND "THERMAL POWER" was a result of an error that occurred during issuance of License Amendment No. 172. Pages 1.1-4 and 1.1-5 of TS Section 1.1 will be re-issued to include the missing definitions.
4.0 STATE CONSUL rATION 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 CONSIDERATION The amendment changes a requirement with respect to installation or use of a facility component 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 types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (77 FR 76081, dated December 26,2012).
Accordingly.
the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The NRC staff 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) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributor:
T. Beltz Date of issuance:
AUgllS t 28, 2013 August 28, 2013 Mrs. Karen D. Fili Site Vice President Monticello Nuclear Generating Plant Northern States Power Company -Minnesota 2807 West County Road 75 Monticello, MN 55362-9637 MONTICELLO NUCLEAR GENERATING PLANT -ISSUANCE OF AMENDMENT NO. 175 TO REVISE THE RENEWED FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS (TAC NO. ME9423)
Dear Mrs. Fili:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 175 to Renewed Facility Operating License (FOL) No. DPR-22 for the Monticello Nuclear Generating (MNGP). The amendment consists of changes to the Renewed FOL and technical specifications (TSs) in response to your application dated August 21, 2012, as supplemented by letters dated November 7,2012, and March 22, 2013. The amendment corrects typographical errors, makes editorial changes, removes obsolete information, clarifies the fuel storage capacity to revise and align the existing FOL and TSs, and corrects a pagination error from a previously-issued license amendment.
A copy of our related safety evaluation is also enclosed.
The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. Sincerely, IRA! Terry A. Beltz, 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. 175 to DPR-22 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
PUBLIC RidsNrrDorlDpr Resource RidsNrrLASRohrer Resource LPL3-1 r/f RidsNrrDorlLpl3-1 Resource RidsRgn3MailCenter Resource RidsAcrsAcnw_MailCTR Resource RidsNrrPMMonticello Resource RidsNrrDssStsb Resource ADAMS Accession No MI13168A373
-OFFICE LPL3-1/PM LPL3-1/LA STSB/BC SRXB/BC OGC (NLO) LPL3-1/BC LPL3-1/PM NAME TBellz SRohrer RElliott KWood for CJackson DRoth RCarlson TBellz DATE 07/30/13 08/06/13 08/19/13 08/09/13 08/22/13 08/27/13 08/28/13 OFFICIAL RECORD COpy