L-MT-07-036, License Amendment Request, Remove Table of Contents from Technical Specifications

License Amendment Request, Remove Table of Contents from Technical Specifications

ML071920501
Person / Time
Site:	Monticello
Issue date:	07/09/2007
From:	O'Connor T Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-07-036
Download: ML071920501 (11)

Text

Monticello Nuclear Generatinq Plant Operated by Nuclear Management Company, LLC July 9, 2007 L-MT-07-036 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket 50-263 License No. DPR-22 License Amendment Request: Remove Table of Contents from Technical Specifications Pursuant to 10 CFR 50.90, the Nuclear Management Company, LLC (NMC) proposes to revise the Monticello Nuclear Generating Plant (MNGP) Technical Specifications (TS) to remove the table of contents (TOC) from the TS and place it under licensee control. provides a description of the proposed change and includes the technical analysis and associated no significant hazards and environmental considerations. provides the TS TOC pages for information.

The NMC requests approval by July 2008, with an implementation period of 60 days.

This letter makes no new commitments or changes to any existing commitments.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Minnesota Official.

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

resident, Monticello Nuclear Generating Plant ent Company, LLC Enclosures (2) cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce 2807 West County Road 75 Monticello, Minnesota 55362-9637 Telephone: 763.295.5151 Fax: 763.295.1454

ENCLOSURE 1 DESCRIPTION OF CHANGE LICENSE AMENDMENT REQUEST REMOVE TABLE OF CONTENTS FROM TECHNICAL SPECIFICATIONS I.0 DESCRIPTION Pursuant to 10 CFR 50.90, the Nuclear Management Company, LLC (NMC) proposes to revise the Monticello Nuclear Generating Plant (MNGP) Technical Specifications (TS) to remove the table of contents from the TS.

The table of contents (TOC) for the TS is not being eliminated, rather, following approval of this license amendment request (LAR), responsibility for maintenance and issuance of updates to the TS TOC will transfer from the U.S. Nuclear Regulatory Commission (NRC) to the NMC. The TOC will no longer be included in the NRC issued TS and as such will no longer be part of the TS (Appendix A to the Operating License). A TOC for the TS will be maintained under NMC (licensee) control. The TOC will be issued by NMC in conjunction with the implementation of future NRC approved TS amendments.

2.0 BACKGROUND

During implementation of the full-scope alternative source term (AST) license amendment (Amendment 148) (Reference I ) it was identified that the TS TOC should have been revised to reflect incorporation of an additional specification that was added with the AST amendment. Several licensees TS, as described within Section 4 of this LAR, do not include a TOC as part of the TS. From discussions with the Monticello NRC Project Manager it was concluded that an acceptable alternative to submitting a LAR to revise the TS TOC to include the new AST specification, was to submit a LAR to remove the TOC from the TS.

Once this LAR was approved by the NRC, the NMC would revise the TOC to add the AST related specification.

3.0 PROPOSED CHANGE

Remove the TOC from the TS and place it under NMC (licensee) control. Note that no TS pages will be issued as a result of this proposed change.

4.0 TECHNICAL ANALYSIS

The TOC does not meet the criteria specified in 10 CFR 50.36 requiring its inclusion within a plant's TS, 10 CFR 50.36(b) states:

Each license authorizing operation of a production or utilization facility of a type described in § 50.21 or § 50.22 will include technical specifications.

The technical specifications will be derived from the analyses and Page 1 of 6

ENCLOSURE 1 evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to § 50.34. . . .

10 CFR 50.36(c) states the "Technical specifications will include items in the following categories:". Review of 10 CFR 50.36 indicates that a TOC was not listed in the regulation as one of the categories.

10 CFR 50.36(a) indicates that a license application may provide other information associated with the TS, and gives an example of this information, i.e.,

the TS Bases, but 10 CFR 50.36(a) also clearly indicates that the Bases are not a part of the TS.

Each applicant for a license ... shall include in his application proposed technical specifications in accordance with the requirements of this section.

A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.

The TOC references where specific TS sections can be found throughout the TS, but does not contain technical information required by 10 CFR 50.36. Since the TOC does not include information required by 10 CFR 50.36 to be reviewed by the NRC staff, inclusion of a TOC within the TS is optional, and is not required by the regulation. Removal of the TOC from the TS therefore constitutes an administrative change, and is therefore acceptable.

Additionally, the "Writer's Guide for Plant-Specific Improved Technical Specifications (ITS) (Reference 2) was reviewed for guidance. The writer's guide refers to the TOC as "Technical Specification Front Matter," which also includes the Title Page and List of Effective Pages. The writer's guide describes the TS content as starting with Chapter 1, "Use and Application" and does not include the TOC as part of the content of the TS.

Precedent A license amendment was issued for the Waterford Steam Electric Station (Waterford 3) to remove the TS Index (which corresponds to the TOC) from the Waterford 3 TS on May 9, 2005 (Reference 3). Also, as discussed in the response (Reference 4) to an NRC request for additional information for this LAR (Reference 5), Arkansas Nuclear One (ANO) and Grand Gulf during their ITS conversions were issued TS which did not include a TOC. Both AN0 and the Grand Gulf plants created and maintain a TOC and periodically provide updated pages to controlled copy holders.

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ENCLOSURE I General Discussion on Updating TS The TS TOC will be maintained and revised in a similar manner to the TS Bases (which are controlled under TS Bases Control Program in the Administrative Controls section of the TS) and the Technical Requirement Manual (TRM) in accordance with NMC administrative procedures. Updating and issuance of the TOC to reflect changes due to NRC approved license amendments will be in accordance with these procedures, consistent with the current processes for controlling TS Bases and TRM changes, both of which include a TOC.

Holders of copies of the TS receive periodic updates of the TOC pages. The distribution process requires that each time the NMC receives an approved change to the TS from the NRC, or makes a change to the TS Bases or the TRM; a transmittal is made with the accompanying changes to all controlled copy holders, which includes offsite organizations, such as the NRC, that maintain controlled copies of the TS. This ensures that all stakeholders are informed of any changes to the TOC.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Determination In accordance with the requirements of 10 CFR 50.90, the Nuclear Management Company, LLC (NMC) requests an amendment to remove the table of contents (TOC) from the Technical Specifications (TS) and place it under licensee control.

NMC has evaluated the proposed amendment in accordance with 10 CFR 50.91 against the standards in 10 CFR 50.92 and has determined that the operation of the facility in accordance with the proposed amendment presents no significant hazards. NMCJsevaluation against each of the criteria in 10 CFR 50.92 follows.

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

No.

The proposed change is administrative and affects control of a document, the TOC, listing the specifications in the plant TS. Transferring control from the NRC to NMC (the licensee) does not affect the operation, physical configuration, or function of plant equipment or systems. It does not impact the initiators or assumptions of analyzed events; nor does it impact the mitigation of accidents or transient events. The change has no impact on, and hence cannot increase, the probability or consequences of an accident previously evaluated.

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ENCLOSURE I

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

No.

The proposed change is administrative and does not alter the plant configuration, require installation of new equipment, alter assumptions about previously analyzed accidents, or impact the operation or function of plant equipment or systems. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

No.

The proposed change is administrative. The TOC is not required by regulation to be in the TS. Removal does not impact any safety assumptions or have the potential to reduce a margin of safety as described in the TS Bases. The change involves a transfer of control of the TOC from the NRC to NMC. No change in the technical content of the TS specifications is involved. Consequently, transfer from the NRC to NMC has no impact on the margin of safety, and hence cannot involve a significant reduction in the margin of safety.

Based on the above, the NMC has determined that operation of the facility in accordance with the proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92(c), in that it does not: ( I ) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

5.2 Applicable Regulatory Requirements 10 CFR 50.36(a) states that "Each applicant for a license .. . shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications." Similar to the TS Bases, the TOC does not meet the criteria specified in 10 CFR 50.36 for inclusion within a plant's TS.

Page 4 of 6

ENCLOSURE I NMC has evaluated this change against the applicable regulatory requirements as described herein. Based on this, there is reasonable assurance that the health and safety of the public, following approval of this change is unaffected.

6.0 ENVIRONMENTAL EVALUATION NMC has determined that the proposed amendment does not change any requirements with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. The proposed amendment does not involve (i) a significant hazards consideration, or (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for a categorical exclusion set forth in 10 CFR 51.22(~)(9).Therefore, NMC concludes pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

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ENCLOSURE 1

7.0 REFERENCES

1. NRC letter to NMC, "Monticello Nuclear Generating Plant - lssuance of Amendment RE: Full-Scope Implementation of the Alternative Source Term Methodology (TAC No. MC8971),17dated December 7, 2006.

2. Technical Specifications Task Force, TSTF-GG-05-01, "Writer's Guide for Plant-Specific Improved Technical Specifications," dated June 2005.

3. U.S. NRC letter to Entergy Operations, Inc., "Waterford Steam Electric Station, Unit 3 - lssuance of Amendment Re: Modification of Technical Specification (TS) 5.3.1, Fuel Assemblies, TS 5.6.1, Criticality, TS 6.9.1 . I I.I, Core Operating Limits Reports, and Deletion of TS Index, (TAC No. MC3584),11dated May 9, 2005.

4. Entergy Operations, Inc., letter to U.S. NRC, "Supplement to Amendment Request NPF-38-258 to Modify Technical Specification (TS) 5.3.1, Fuel Assemblies and TS 6.9.1 .I 1.I,Core Operating Limits Report, Waterford Steam Electric Station, Unit 3," dated March 8, 2005.

5. Entergy Operations, Inc., letter to U.S. NRC, "License Amendment Request NPF-38-258 to Modify Technical Specification (TS) 5.3.1, Fuel Assemblies and TS 6.9.1 . I I. I , Core Operating Limits Report, Waterford Steam Electric Station, Unit 3," dated June 17, 2004.

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ENCLOSURE 2 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST REMOVE TABLE OF CONTENTS FROM TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATION TABLE OF CONTENTS PAGES (No changes provided for Information)

(3 pages follows)

TABLE OF CONTENTS Page Number 1.0 USE AND APPLICATION I.I Definitions ...........................................................................................................

I.1-1 I.2 Logical Connectors .............................................................................................. I.2-1 1.3 Completion Times ............................................................................................... 1.3-1 1.4 I.4-1 Frequency ...........................................................................................................

2.0 SAFETY LIMITS (SLs) .............................................................................................. 2.0-1

2. I Safety Limits 2.2 SL Violations 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY ..........................3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ......................................... 3.0.4 REACTIVITY CONTROL SYSTEMS NO CHANGES FOR INFORMATION .

SHUTDOWN MARGIN (SDM) .................................................................... 3. I .1.1 Reactivity Anomalies .................................................................................. 3.1.2-1 Control Rod OPERABILITY ........................................................................ 3.1.3.1 Control Rod Scram Times ........................................................................... 3.1.4.1 Control Rod Scram Accumulators ............................................................... 3.1.5.1 Rod Pattern Control .................................................................................... 3 .I .6-1 Standby Liquid Control (SLC) System......................................................... 3.1.7.1 Scram Discharge Volume (SDV) Vent and Drain Valves .............................3.1.8-1 3.2 POWER DISTRIBUTION LIMITS 3.2. I AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)............................................................................................ .3.2.1-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) ..........................................3.2.2-1 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) (Optional) ............................3.2.3-1 INSTRUMENTATION Reactor Protection System (RPS) Instrumentation ..................................3.3.1.1-1 Source Range Monitor (SRM) Instrumentation ........................................3.3.1.2-1 Control Rod Block Instrumentation ......................................................... .3.3.2.1-1 Feedwater Pump and Main Turbine High Water Level Trip Instrumentation ................................................................................. .3.3.2.2-1 Post Accident Monitoring (PAM) Instrumentation ..................................... 3.3.3.1-1 Alternate Shutdown System ..................................................................... 3.3.3.2-1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation ............................................................3.3.4.1-1 Emergency Core Cooling System (ECCS) Instrumentation......................3.3.5.1-1 Reactor Core Isolation Cooling (RCIC) System Instrumentation ..............3.3.5.2-1 Primary Containment Isolation Instrumentation........................................ 3.3.6.1-1 Secondary Containment Isolation Instrumentation ................................... 3.3.6.2-1 Low-Low Set (LLS) Instrumentation ......................................................... 3.3.6.3-1 Control Room Emergency Filtration (CREF) System Instrumentation .................................................................................. 3.3.7.1-1 Loss of Power (LOP) Instrumentation ...................................................... 3.3.8.1-1 Reactor Protection System (RPS) Electric Power Monitoring ...................3.3.8.2-1 Monticello i Amendment No. 146

TABLE OF CONTENTS Page Number REACTOR COOLANT SYSTEM (RCS)

Recirculation Loops Operating....................................................................3.4.1-1 Jet Pumps ..................................................................................................

3.4.2-1 SafetyIRelief Valves (SIRVs) ...................................................................... 3.4.3-1 RCS Operational LEAKAGE ....................................................................... 3.4.4-1 RCS Leakage Detection Instrumentation ....................................................3.4.5-1 RCS Specific Activity .................................................................................. 3.4.6-1 Residual Heat Removal (RHR) Shutdown Cooling System .Hot Shutdown..............................................................................................

3.4.7.1 Residual Heat Removal (RHR) Shutdown Cooling System .Cold Shutdown..............................................................................................

3.4.8.1 RCS Pressure and Temperature (PIT) Limits.............................................. 3.4.9.1 Reactor Steam Dome Pressure ............................................................... .3.4.10.1 EMERGENCY CORE COOLING SYSTEM (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ECCS .Operating...................................................................................... .3.5.1-1 ECCS .Shutdown....................................................................................... 3.5.2.1 RCIC System ..............................................................................................

3.5.3.1 CONTAINMENT SYSTEMS NO CHANGES .FOR INFORMATION Primary Containment .............................................................................. .3.6. I. 1.1 Primary Containment Air Lock ................................................................. 3.6.1.2.1 Primary Containment Isolation Valves (PCIVs) ........................................ 3.6.1.3-1 Drywell Air Temperature ..........................................................................3.6.I.4-1 Low-Low Set (LLS) Valves ...................................................................... .3.6. I.5-1 Reactor Building-to-Suppression Chamber Vacuum Breakers .................3.6.1.6-1 Suppression Chamber-to-Drywell Vacuum Breakers ............................... 3.6.1.7-1 Residual Heat Removal (RHR) Drywell Spray .........................................3.6.1.8-1 Suppression Pool Average Temperature ................................................ .3.6.2.1-1 Suppression Pool Water Level................................................................. 3.6.2.2-1 Residual Heat Removal (RHR) Suppression Pool Cooling.......................3.6.2.3-1 Primary Containment Oxygen Concentration ........................................... 3.6.3.1-1 Secondary Containment ..........................................................................3.6.4.1-1 Secondary Containment Isolation Valves (SCIVs) ...................................3.6.4.2-1 Standby Gas Treatment (SGT) System ...................................................3.6.4.3-1 PLANT SYSTEMS Residual Heat Removal Service Water (RHRSW) System..........................3.7.1-1 Emergency Service Water (ESW) System and Ultimate Heat Sink (UHS) ............................................................................................

3.7.2-1 Emergency Diesel Generator Emergency Service Water (EDG-ESW) System ............................................................................. 3.7.3-1 Control Room Emergency Filtration (CREF) System .................................. 3.7.4-1 Control Room Ventilation System ...............................................................3.7.5-1 Main Condenser Offgas .............................................................................. 3.7.6-1 Main Turbine Bypass System ..................................................................... 3.7.7-1 Spent Fuel Storage Pool Water Level ......................................................... 3.7.8-1 Monticello ii Amendment No. 146

TABLE OF CONTENTS Page Number ELECTRICAL POWER SYSTEMS AC Sources .Operating .............................................................................. 3.8.1-1 AC Sources .Shutdown.............................................................................. 3.8.2.1 Diesel Fuel Oil, Lube Oil, and Starting Air ................................................... 3.8.3-1 DC Sources .Operating.............................................................................. 3.8.4-1 DC Sources .Shutdown ............................................................................. 3.8.5.1 Battery Parameters ..................................................................................... 3.8.6.1 Distribution Systems .Operating ................................................................ 3.8.7.1 Distribution Systems .Shutdown ................................................................ 3.8.8.1 REFUELING OPERATIONS Refueling Equipment Interlocks .................................................................. 3.9.1.1 Refuel Position One-Rod-Out Interlock ....................................................... 3.9.2.1 Control Rod Position ................................................................................... 3.9.3-1 Control Rod Position Indication................................................................... 3.9.4-1 Control Rod OPERABILITY - Refueling ...................................................... 3.9.5-1 Reactor Pressure Vessel (RPV) Water Level .............................................. 3.9.6-1 Residual Heat Removal (RHR) - High Water Level ..................................... 3.9.7-1 Residual Heat Removal (RHR) - Low Water Level ...................................... 3.9.8-1 3.10 SPECIAL OPERATIONS NO CHANGES .FOR INFORMATION 3.10.1 Inservice Leak and Hydrostatic Testing Operation .................................... 3.10.1.1 3.10.2 Reactor Mode Switch Interlock Testing ..................................................... 3.10.2-1 3.10.3 Single Control Rod Withdrawal - Hot Shutdown ........................................ 3.10.3-1 3.10.4 Single Control Rod Withdrawal - Cold Shutdown ...................................... 3.10.4-1 3.10.5 Single Control Rod Drive (CRD) Removal - Refueling ..............................3.10.5-1 3.10.6 Multiple Control Rod Withdrawal - Refueling ............................................. 3.10.6-1 3.10.7 Control Rod Testing - Operating ............................................................... 3.10.7-1 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling ........................................ 3.10.8-1 4.0 DESIGN FEATURES ................................................................................................ 4.0.1 4.1 Site Location 4.2 Reactor Core 4.3 Fuel Storage 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility ......................................................................................................

5.1-1 5.2 Organization ........................................................................................................

5.2-1 5.3 Unit Staff Qualifications .......................................................................................

5.3-1 5.4 Procedures .........................................................................................................

.5.4-1 5.5 Programs and Manuals ...................................................................................... .5.5-1 5.6 Reporting Requirements ...................................................................................... 5.6-1 5.7 High Radiation Area ............................................................................................

5.7-1 Monticello iii Amendment No. 146