ML13072B341

Issuance of Amendment Revise Decay Time Technical Specification

ML13072B341
Person / Time
Site:	Millstone
Issue date:	06/04/2013
From:	James Kim Plant Licensing Branch 1
To:	Heacock D Dominion Nuclear Connecticut
Kim J
References
TAC ME4367
Download: ML13072B341 (12)

Text

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Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION UNIT NO.2-ISSUANCE OF AMENDMENT RE: REVISE DECAY TIME TECHNICAL SPECIFICATION (TAC NO. ME4367)

Dear Mr. Heacock:

The Commission has issued the enclosed Amendment No. 315 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No.2, in response to your application dated July 21,2010, as supplemented by letter dated July 19, 2011.

The amendment revises Technical Specification (TS) 3/4.9.3.1, "Decay Time" by reducing the minimum decay time for irradiated fuel prior to movement in the reactor vessel from 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. A reduction in the minimum decay time requirement is requested to provide additional flexibility in outage planning such that irradiated fuel can be moved from the reactor vessel to the spent fuel pool (SFP) earlier in an outage.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, r~

James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336

Enclosures:

1. Amendment No. 315 to DPR-65

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DOMINION NUCLEAR CONNECTICUT, INC.

DOCKET NO. 50-336 MILLSTONE POWER STATION, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 315 Renewed License No. DPR-65

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amend ment by the applicant dated July 21, 2010, as supplemented by letter dated July 19, 2011, 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

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 Renewed Facility Operating License No. DPR-65 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 315, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance, and shall be implemented within 60 days of issuance.

~HEN~C~~COMMISSION RObert!: Acting Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: June 4, 2013

ATTACHMENT TO LICENSE AMENDMENT NO. 315 RENEWED FACILITY OPERATING LICENSE NO. DPR~65 DOCKET NO. 50~336 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert Page 3 Page 3 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 3/4 9.3 3/4 9.3

- 3 Connecticut, in accordance with the procedures and limitations set forth in this renewed operating license; (2) 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; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, 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; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, 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 and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at steady-state reactor core power levels not in excess of 2700 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 315, are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

Renewed License No. DPR-65 Amendment NO.315

REFUELING OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3.1 The reactor shall be subcritical for a minimum of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to movement of irradiated fuel in the reactor pressure vessel.

APPLICABILITY: MODE 6.

ACTION:

With the reactor subcritical for less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel.

SURVEILLANCE REQUIREMENTS 4.9.3.1 The reactor shall be determined to have been subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

MILLSTONE - UNIT 2 3/4 9-3 Amendment No. H4, ~,315

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 315 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION, UNIT NO.2 DOCKET NO. 50-336

1.0 INTRODUCTION

By letter dated July 21,2010, (Agencywide Documents Access and Management System (ADAMS) Accession Number ML102240064), as supplemented by letter dated July 19, 2011 (ML11208B450), Dominion Nuclear Connecticut, Inc. (the licensee) proposed changes to the Technical Specifications (TSs) for Millstone Power Station Unit 2 (MPS2).

The proposed change revises Technical Specification (TS) Limiting Condition for Operation (LCO) 3.9.3.1, "Decay Time," by reducing the minimum decay time for irradiated fuel prior to movement in the reactor vessel from 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. The licensee also proposed a corresponding change to Surveillance Requirement (SR) 4.9.3.1. The licensee requested the reduction in the minimum decay time requirement to provide additional flexibility in outage planning by allowing earlier movement of irradiated fuel.

The supplemental letter dated July 19, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on April 2, 2013 (78 FR 19749).

2.0 REGULATORY EVALUATION

The MPS2 Final Safety Analysis Report (FSAR) considers the decay time of irradiated fuel in two analyses. Section 14.7.4, "Radiological Consequences of a Fuel Handling Accident (FHA),"

lists decay time among many assumptions used to determine the radiological consequences of FHAs in the Spent Fuel Pool (SFP) and inside containment during reactor refueling. Section 9.5, "Spent Fuel Pool Cooling (SFPC) system," includes in-reactor decay time as an assumption used to determine the decay heat rate of the fuel most recently discharged to the SFP. Both of these analyses have been included in TS Bases Section 3/4.9.3.

The current design basis analyses for FHAs in containment and in the SFP used a 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay time as an initial condition to establish the inventory of radioactive fission products within the gap of the fuel pins of the limiting fuel assembly. The Design-Basis Analysis used assumptions derived from Regulatory Guide 1.183, "Alternative Radiological Source Terms for

-2 Evaluating Design Basis Accidents at Nuclear Power Reactors," and the Nuclear Regulatory Commission (NRC) staff approved the analyses for implementation at MPS2 as part of License Amendment 284, "Selective Implementation of Alternate Source Term," September 20,2004 (ADAMS Accession Number ML042650362). Subsequent full implementation of the alternative source term assumptions pursuant to Amendment 298, "Full Implementation of AST," May 31, 2007 (ADAMS Accession Number ML071450053) did not alter the assumptions and initial conditions used in the FHA analyses.

The requirements of 10 CFR 50.36 (c)(2)(ii) specify that LCOs be established for equipment or conditions satisfying anyone of four criteria:

Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

As an initial condition of a Design-Basis Accident Analysis (the fuel handling accident analysis) that assumes the failure of a fission product barrier (the fuel cladding), the decay time satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

The NRC Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants provided guidance for NRC staff review. The NRC SRP listed specific acceptance criteria derived from applicable general design criteria and other NRC regulations and a method acceptable to the staff to demonstrate compliance with those acceptance criteria for various SSCs at commercial Light-Water Reactors (LWRs). The review criteria in SRP Section 9.1.3, Revision 2, specified the SFP cooling system considerations for SFP coolant temperature control, including the following statement:

The largest heat load placed on the SFPCCS [spent fuel pool cooling and cleanup system] heat exchangers is imposed by refueling offloads, which are deliberate, planned evolutions. As a result, if necessary for adequate cooling of the fuel, factors that increase heat load (e.g., power increases, decay time reductions, or storage capacity increases) may be offset by operational factors that reduce heat load (e.g., longer decay times or transfer of fewer fuel assemblies to the SFP) or that increase heat removal capability (e.g., scheduling

-3 offloads for periods of reduced ultimate heat sink temperature or optimizing cooling system performance).

Thus, credit for lower cooling water temperature has been an accepted method of controlling SFP temperature at higher heat loads resulting from shorter decay times.

3.0 TECHNICAL EVALUATION

The MPS2 reactor discharges fuel to a single SFP in support of refueling. Preparation for refueling of the reactor involves the following activities: reactor shutdown and cooldown, removal of the reactor vessel head and upper internals, removal of the flange covering the fuel transfer path through the containment wall, and flooding of the refueling cavity. Once preparations are complete and the minimum in-reactor decay time has been satisfied, operators may begin transfer of fuel through the flooded refueling cavity and fuel transfer path to the SFP. The amendment request proposed a reduction in the minimum decay time prior to beginning fuel movement in TS LCD 3.9.3.1 and SR 4.9.3.1 from 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. The reduction in minimum decay time increases the quantity of radioactive fission products remaining in the fuel and the rate of decay heat generation within the fuel at the time fuel handling begins.

To satisfy the requirements of 10 CFR 50.36, the decay time of irradiated fuel used as an initial condition in the FHA dose consequence analysis must be consistent with or bound the value specified in the TS limiting condition for operation (I.e., TS LCD 3.9.3.1). The NRC approved the analysis of record for implementation at MPS2 as part of License Amendment 284, "Selective Implementation of Alternate Source Term," September 20,2004. This analYSis used a decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> as the initial condition of the analysis. The proposed revision to TS LCD 3.9.3.1 specifies that the reactor be subcritical for a minimum of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to movement of irradiated fuel in the reactor pressure vessel. The proposed changes to TS LCD 3.9.3.1 and TS SR 4.9.3.1 are consistent with the FHA consequence analysis in Amendment 284. Therefore, the NRC staff finds that the proposed changes are acceptable.

The NRC staff requested documents associated with any licensing basis changes to the FSAR Section 9.5.2 or revisions to the TS 3/4.9.3.1 bases related to revised spent fuel pool heating load analyses. As an attachment to the letter dated July 19, 2011, the licensee provided the screening evaluation of the change to Section 9.5 of the FSAR, which included evaluation of fuel discharges at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay and a change to the bases of TS 3/4.9.3.1. Based on the screening, the licensee determined that these changes could be made without a license amendment under the provisions of 10 CFR 50.59.

The licensee indicated that following issuance of the license amendment associated with this request, plant procedures will allow a variable subcriticality time as a function of Reactor Building Closed Cooling Water (RBCCW) inlet temperature. The variable subcriticality decay time will not be less than the proposed 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay time required to support the FHA dose consequence analysis assumptions and included in the proposed revision to TS 3/4.9.3.1.

The NRC staff evaluated the effects of the reduced decay time on decay heat generation rates.

The licensee used the ORIGEN2 decay heat model and conservative assumptions to calculate the overall decay heat generation rate within the SFP. For the normal refueling batch offload of 80 assemblies at a discharge rate of 6 assemblies per hour, the licensee determined that the peak heat load would increase from 14.52 million BTUs per hour (MBTU/hr) to 16.18 MBTU/hr

-4 as the initial decay time decreases from 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. For the normal refueling full core offload of 217 assemblies at a discharge rate of 6 assemblies per hour, the licensee determined that the peak heat load would increase from 30.90 MBTU/hr to 34.59 MBTU/hr as the initial decay time decreases from 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. The NRC staff determined that the decay heat rate for the full core discharge analysis would experience a greater increase in decay heat as a result of the reduced decay time because the recently discharged full core contributes a much larger share of the overall decay heat generation (approximately 84 percent at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />) than for the refueling batch offload (approximately 66 percent at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />).

Using simplified approximations of decay heat from Branch Technical Position [Auxiliary Systems Branch] ASB 9-2, "Residual Decay Energy for Light-Water Reactors for Long-Term Cooling," July 1981 (ADAMS Accession Number ML052350549), the NRC staff confirmed that the heat load increase calculated by the licensee that would result from the reduced decay time (approximately 14 percent for the most recently discharged fuel) was appropriate.

The SFPC system removes decay heat from the SFP. The spent fuel cooling system consists of one seismically qualified cooling train, which includes two full-capacity pumps and two heat exchangers in parallel. Heat is removed from the spent fuel cooling system heat exchangers by the safety-related Component Cooling Water (CCW). Operators can align the shutdown cooling system to the SFP to provide additional cooling capacity under full core offload conditions.

The SFPC system design basis described in Section 9.5 of Revision 28.1 to the MPS2 FSAR specifies that the pool temperature will be maintained at no more than 150°F. The design basis analyses included normal refueling partial core discharges with cooling provided by the SFPC system only and normal refueling full core discharges with cooling provided by the SFPC system assisted by the shutdown cooling system. The analyses of SFP temperature for discharges with 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> decay indicated adequate cooling would be provided at CCW temperatures no higher than 85°F at a fuel transfer rate of 6 assemblies per hour. The analyses of SFP temperature for discharges with 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay indicated that adequate cooling would be provided at CCW temperatures no higher than 75°F at a fuel transfer rate of 6 assemblies per hour.

The NRC staff compared the maximum increase in calculated decay heat generation rate resulting from the reduced decay time to the increased cooling capability provided by the specified reduction in CCW temperature. With the SFP at its design maximum temperature of 150°F, the decrease in CCW temperature from 85°F to 75°F would increase the heat removal capability by 15 percent with all other parameters unchanged. At lower SFP temperatures, the increase in heat removal capability resulting from the same CCW temperature decrease would be greater. The additional heat removal capability provided by the CCW temperature reduction would bound the maximum increase in decay heat generation rate resulting from the proposed decrease in decay time. Therefore, the NRC staff finds that the higher decay heat from the proposed decrease in decay time would be adequately compensated by increased heat removal capacity provided by the specified lower CCW temperatures, consistent with the guidance in SRP Section 9.1.3.

As described above, the NRC staff finds that the proposed decrease in decay time from 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> in TS LCO 3.9.3.1 and associated TS SR 4.9.3.1 are acceptable based on evaluation of the FHA consequence analysis in Amendment 284. The licensee indicated that following issuance of the license amendment associated with this request, plant procedures will allow a variable subcriticality time as a function of RBCCW inlet temperature. This approach

- 5 has been accepted by the NRC staff as an appropriate means to control SFP temperature, as specified among the review criteria in SRP Section 9.1.3.

Therefore, the staff finds that the proposed decrease in decay time at lower CCW temperatures would be acceptable with respect to changes in spent fuel decay heat rate. Based on staff's analysis, the staff concludes that reducing the minimum decay time for irradiated fuel prior to movement in the reactor vessel from 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Connecticut 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 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 (78 FR 19749).

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) 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: S. Jones Date: June 4, 2013

June 4,2013 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION UNIT NO.2-ISSUANCE OF AMENDMENT RE: REVISE DECAY TIME TECHNICAL SPECIFICATION (TAC NO. ME4367)

Dear Mr. Heacock:

The Commission has issued the enclosed Amendment No. 315 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No.2, in response to your application dated July 21, 2010, as supplemented by letter dated July 19, 2011.

The amendment revises Technical Specification (TS) 3/4.9.3.1, "Decay Time" by reducing the minimum decay time for irradiated fuel prior to movement in the reactor vessel from 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. A reduction in the minimum decay time requirement is requested to provide additional flexibility in outage planning such that irradiated fuel can be moved from the reactor vessel to the spent fuel pool (SFP) earlier in an outage.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/raJ James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336

Enclosures:

1. Amendment No. 315 to DPR-65

2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

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