ML042940080
| ML042940080 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/28/2005 |
| From: | Padovan L NRC/NRR/DLPM/LPD3 |
| To: | Thomas J. Palmisano Nuclear Management Co |
| Padovan L | |
| Shared Package | |
| ML050350094 | List: |
| References | |
| TAC MC1899 | |
| Download: ML042940080 (11) | |
Text
January 28, 2005 Mr. Thomas J. Palmisano Site Vice President Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT - ISSUANCE OF AMENDMENT RE: REVISION TO TECHNICAL SPECIFICATIONS TABLE 3.2.3 AND SECTION 3.7/4.7 (TAC NO. MC1899)
Dear Mr. Palmisano:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 141 to Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant. The amendment consists of changes to the Technical Specifications (TSs) in response to your application of January 30, 2004. The amendment changes the TS as follows:
Clarifies the permissive setpoint for the source range monitor detector-not-fully-inserted rod block bypass.
Corrects a typographical error in the surveillance requirement for suppression pool temperature monitoring.
Clarifies the setpoint for the pressure suppression chamber-reactor building vacuum breakers instrumentation.
Clarifies the operating force requirements for the pressure suppression chamber-drywell vacuum breakers surveillance test.
Makes corrections resulting from license Amendments 130 and 132.
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,
/RA/
L. Mark Padovan, Project Manager, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-263
Enclosures:
- 1. Amendment No. 141 to DPR-22
- 2. Safety Evaluation cc w/encls: See next page
ML042940080 Nrr-058
- Reviewed changes only OFFICE PDIII-1/PM PDIII-1/LA PDIII-1/LA**
EEIB/SC IROB/SC OGC*
PDIII-1/SC NAME LPadovan THarrris DClarke for THarris EMarinos TBoyce MHiggins MKotzalas DATE 12/30/04 12/30/04 1/24/05 8/6/04 1/14/05 1/19/05 1/27/05
Monticello Nuclear Generating Plant cc:
Jonathan Rogoff, Esquire Vice President, Counsel & Secretary Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 U.S. Nuclear Regulatory Commission Resident Inspector's Office 2807 W. County Road 75 Monticello, MN 55362 Manager, Regulatory Affairs Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637 Robert Nelson, President Minnesota Environmental Control Citizens Association (MECCA) 1051 South McKnight Road St. Paul, MN 55119 Commissioner Minnesota Pollution Control Agency 520 Lafayette Road St. Paul, MN 55155-4194 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Commissioner Minnesota Department of Health 717 Delaware Street, S. E.
Minneapolis, MN 55440 Douglas M. Gruber, Auditor/Treasurer Wright County Government Center 10 NW Second Street Buffalo, MN 55313 Commissioner Minnesota Department of Commerce 85 7th Place East, Suite 500 St. Paul, MN 55101-2198 Manager - Environmental Protection Division Minnesota Attorney Generals Office 445 Minnesota St., Suite 900 St. Paul, MN 55101-2127 John Paul Cowan Executive Vice President & Chief Nuclear Officer Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Nuclear Asset Manager Xcel Energy, Inc.
414 Nicollet Mall, R.S. 8 Minneapolis, MN 55401 November 2004
NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 141 License No. DPR-22 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Nuclear Management Company, LLC (the licensee), dated January 30, 2004, 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; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
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; D.
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 E.
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.
2.
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 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. 141, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Margie Kotzalas, Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: January 28, 2005
ATTACHMENT TO LICENSE AMENDMENT NO. 141 FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Replace the following pages of the 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 58a 58a 156 156 158 158 163 163 164 164 170 170 171a 171a
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 141 TO FACILITY OPERATING LICENSE NO. DPR-22 NUCLEAR MANAGEMENT COMPANY, LLC MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263
1.0 INTRODUCTION
The Nuclear Management Company, LLCs (NMCs) letter of January 30, 2004, requested changes to the Technical Specifications (TSs) for the Monticello Nuclear Generating Plant. The proposed amendment would change the TSs as follows:
Clarify the permissive setpoint for the source range monitor detector-not-fully-inserted rod block bypass.
Correct a typographical error in the surveillance requirement for suppression pool temperature monitoring.
Clarify the setpoint for the pressure suppression chamber-reactor building vacuum breakers instrumentation.
Clarify the operating force requirements for the pressure suppression chamber-drywell vacuum breakers surveillance test.
Make corrections resulting from license Amendments 130 and 132.
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 36(c)(2)(ii)(A) requires that a TS Limiting Condition for Operation (LCO) must be established for instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. In addition, 10 CFR 50.36(c)(3), Surveillance Requirements, specifies that TSs are to include surveillance requirements for testing, calibrating, or inspecting to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.
3.0 TECHNICAL EVALUATION
NMCs proposed changes, and the U.S. Nuclear Regulatory Commission (NRC) staffs evaluation of the changes, are discussed below.
3.1 TS Table 3.2.3, Instrumentation That Initiates Rod Block 3.1.1 Proposed TS Changes TS Table 3.2.3 is being revised to clarify the permissive setpoint for the source range monitor (SRM) detector-not-fully-inserted rod block bypass. Currently, TS Table 3.2.3 Allowable Bypass Conditions item a. states SRM Detector-not-fully-inserted rod block may be bypassed when the SRM channel rate is 100 cps [counts-per-second] or when all IRM [intermediate range monitor] range switches are above Position 2. The proposed change will modify the condition to read SRM Detector-not-fully-inserted rod block may be bypassed when the SRM channel rate is $ 100 cps or when all IRM range switches are above Position 2.
3.1.2 NRC Staff Evaluation The SRM detector-not-fully-inserted rod block is designed to assure that no control rod is withdrawn unless all SRM detectors are properly inserted when they must be relied on to provide the operator with neutron flux level information. A rod block is generated when any SRM detector is not fully inserted into the reactor core, the SRM count level is below 100 cps, and any intermediate-range monitor (IRM) range switch is on either of the two lowest ranges.
An automatic bypass of the rod block is enabled when the count rate on the SRM instrumentation exceeds the preset low-count level. The bypass allows the detector to be partially or completely withdrawn as reactor startup continues.
NMCs proposed change merely clarifies that automatic bypass of the rod block is enabled when the count rate on the SRM instrumentation exceeds 100 cps, rather than being equal to 100cps. Thus, the proposed change is administrative in nature, and is consistent with the design of the rod block and the bypass of the rod block. Therefore, the change is acceptable.
3.2 TS Section 4.7.A.1.b, Suppression Pool Volume and Temperature 3.2.1 Proposed TS Change TS Section 4.7.A.1.b, Suppression Pool Volume and Temperature, is being revised to correct a typographical error in the text. Currently, TS 4.7.A.1.b states Whenever there is indication of relief valve operation which adds heat to the suppression pool, the pool temperature shall be continually monitored and also observed and logged ever 5 minutes until the heat addition is terminated. NMC proposes to change the word ever to every.
3.2.2 NRC Staff Evaluation Since the proposed change only changes ever to every to make the sentence read correctly, the NRC staff considers this as an administrative change and therefore finds it acceptable.
3.3 TS Section 3.7.A.3, Pressure Suppression Chamber - Reactor Building Vacuum Breakers 3.3.1 Proposed TS Change NMC proposes changing TS Section 3.7.A.3 to clarify LCO requirements for the pressure suppression chamber - reactor building vacuum breakers. Currently, TS Section 3.7.A.3 states:
The set point of the differential pressure instrumentation which actuates the pressure suppression chamber - reactor building vacuum breakers shall be 0.5 psi [pounds per square inch]. NMC proposes to revise this sentence to read... shall be # 0.5 psi to indicate that the instrumentation must actuate before differential pressure exceeds 0.5 psi.
3.3.2 NRC Staff Evaluation These vacuum breakers work with the pressure suppression chamber - drywell vacuum breakers to limit the negative pressure in either the suppression chamber or the drywell to less than the design pressure of -2 psid (pounds per square inch differential). Operation of these systems will maintain the pressure differential less than 1 psid. NMCs analysis indicates that actuation of the instrumentation associated with the pressure suppression chamber - reactor building vacuum breakers at 0.5 psid will ensure that the pressure suppression chamber and drywell will not exceed their design pressure. Based on this, the NRC staff considers the proposed change acceptable.
3.4 TS Section 4.7.A.4.a.(4), Pressure Suppression Chamber - Drywell Vacuum Breakers 3.4.1 Proposed TS Change NMC proposes to revise TS Section 4.7.A.4.a.(4) to clarify vacuum breaker opening force.
Currently, the TS section 4.7.A.4.a.(4) says Once each operating cycle, the vacuum breakers shall be tested to determine that the force required to open each valve from fully closed to fully open does not exceed that equivalent to 0.5 psi acting on the suppression chamber face of the valve disc. NMC proposes to revise the sentence to read... equivalent to 0.5 psid....
3.4.2 NRC Staff Evaluation NMCs proposed change clarifies that the vacuum breakers must actuate with a differential pressure of 0.5 psi acting on the pressure suppression chamber face of the valve disc. The current TS surveillance requirement could be considered unclear about the operating force requirements for vacuum breaker testing. Specifying the requirements for operating force on one side of the valve disc does not ensure that a differential pressure exists across the valve.
The proposed TS change clarifies the surveillance requirement to indicate that the vacuum breaker must fully open with an operating force not to exceed 0.5 psid across the pressure suppression chamber face of the valve. This is intended to demonstrate that the vacuum breaker will function as designed. On this basis, the NRC staff finds the proposed change acceptable.
3.5 Correct License Amendments (LAs) 130 and 132 3.5.1 Proposed TS Change NMC proposes to revise TS Section 3.7.A.2.a(1), Primary Containment Integrity, to correct an omission from LA 130. Currently, this section says Primary Containment Integrity as defined in Section 1, shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 212 EF and fuel is in the reactor vessel, except as specified in 3.7.A.2.a (2) or 3.7.A.2.a (3). The proposed change modifies the LCO to read... in 3.7.A.2.a(2),
3.7.A.2.a(3) or 3.7.D.
3.5.2 NRC Staff Evaluation Revising TS Section 3.7.A.2.a(1) to add an exception to allow using TS 3.7.D corrects the previously inadvertently-omitted exception, and allows the operating personnel to enter an available LCO condition which does not impact the safe operation of the plant. This change is needed to eliminate confusion and clarify the conditions for using TS 3.7.D, and therefore is acceptable to the NRC staff.
3.5.3 Proposed TS Change NMC proposes to revise TS Section 3.7.D.1, Primary Containment Isolation Valves (PCIVs),
to correct an omission from LA 130. Currently, this section says During reactor power operating conditions, all Primary Containment automatic isolation valves and all primary system instrument line flow check valves shall be operable except as specified in 3.7.D.2. NMCs proposed change modifies the LCO to read... in 3.7.D.2. and 3.7.D.3.
3.5.4 NRC Staff Evaluation This change adds an exception to TS 3.7.D.1 to allow operators to use the requirements of TS 3.7.D.3 for purge and vent valve operation and isolation. The TS 3.7.D.3.a exception allows plant operators to inert and deinert the primary containment using the 18-inch purge and vent valves. It also allows using the 2-inch purge and bypass vent valve bypass line for purging and venting through the standby gas treatment system when primary containment integrity is required. The TS 3.7.D.3.b exception permits continued reactor operation when purge and vent valve leakage limits are exceeded, provided that the following occurs within the subsequent 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:
Valve leakage is restored to within leakage limits, or The penetration causing the limit to be exceeded is isolated using the means specified in TS 3.7.D.3.b.
This change is acceptable to the NRC staff as it corrects a previously inadvertently-omitted exception to allow the use of an approved TS change. This change does not impact the safe operation of the plant, and enhances the ability of the operators to enter the appropriate LCO condition.
3.5.5 Proposed TS Change TS Section 4.7.D.4 is being revised to remove an obsolete reference attributed to LA 132.
Currently, this section contains the following statement: The seat seals of the drywell and suppression chamber 18-inch purge and vent valves shall be replaced at least once every six operating cycles. If periodic Type C leakage testing of the valves performed per surveillance requirement 4.7.A.2.b identifies a common mode test failure attributable to seat seal degradation, then the seat seals of all drywell and suppression chamber 18-inch purge and vent valves shall be replaced. The proposed change removes the words performed per surveillance requirement 4.7.A.2.b, and corrects a typographical error after the identification of paragraph 4 from, to. so that it reads 4.
3.5.6 NRC Staff Evaluation TS 4.7.A.2.b no longer exists. Revising TS 4.7.D.4 to remove this reference will eliminate confusion and clarify the conditions for using TS 4.7.D.4. The other change corrects the format of the TS. The NRC staff finds the proposed changes acceptable as they correct inadvertent oversights and a typographical error that occurred during relocation of the TS under LA 132.
On the basis of its review of the licensees application for amendment, the NRC staff has concluded that NMC has conformed to 10 CFR 50.36 as the proposed changes either clarify or correct previous LAs and do not affect the intent of the TSs. Therefore, the proposed TS changes are acceptable to the staff.
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 CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The 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 (69 FR 19573). 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 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.
Principal Contributor: H. Garg Date: January 28, 2005