ML13291A203
| ML13291A203 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/14/2013 |
| From: | James Kim Plant Licensing Branch 1 |
| To: | Heacock D Dominion Nuclear Connecticut |
| Kim J | |
| References | |
| TAC MF1364 | |
| Download: ML13291A203 (13) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 November 14, 2013
SUBJECT:
MILLSTONE POWER STATION UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: REVISE TECHNICAL SPECIFICATION 3/4.9.16, SHIELDED CASK (TAC NO. MF1364)
Dear Mr. Heacock:
The Commission has issued the enclosed Amendment No. 316 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2, in response to your application dated April 3, 2013.
The amendment would revise Technical Specification 3.9.16 "Shielded Cask," due to changes to the minimum decay time for fuel assemblies adjacent to the spent fuel pool cask laydown area.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Docket No. 50-336
Enclosures:
- 1. Amendment No. 316 to DPR-65
- 2. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
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. 316 Renewed License No. DPR-65
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A The application for amendment by the applicant dated April 3, 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; 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 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. 316, 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 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
~~
Attachment:
Changes to the License and Technical Specifications Date of Issuance:
rJovember 14, 2013 Robert H. Beall, Acting Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
ATTACHMENT TO LICENSE AMENDMENT NO. 316 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 Page 3 Insert 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 3/4 9-19 Insert 3/4 9-19 Connecticut, in accordance with the procedures and limitations set forth in this renewed operating license; (2)
Pursuant to the Act and 1 0 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. 316, 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.316
REFUELING OPERATIONS SHIELDED CASK LIMITING CONDITION FOR OPERATION 3.9.16 All fuel within a distance L from the center ofthe spent fuel pool cask laydown area shall have decayed for at least 90 days. The distance L equals the major dimension of the shielded cask.
APPLICABILITY:
Whenever a shielded cask is on the refueling floor.
ACTION:
With the requirements of the above specification not satisfied, do not move a shielded cask to the refueling floor. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.16 The decay time of all fuel within a distance L from the center ofthe spent fuel pool cask lay down area shall be determined to be ~ 90 days within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to moving a shielded cask to the refueling floor and at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafter.
MILLSTONE -UNIT 2 3/4 9-19 Amendment No.~'~'~' 2:4§-, 316
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 316 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-65
1.0 INTRODUCTION
DOMINION NUCLEAR CONNECTICUT. INC.
MILLSTONE POWER STATION, UNIT NO.2 DOCKET NO. 50-336 By letter dated April 3, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13102A133), Dominion Nuclear Connecticut, Inc., (the licensee) requested changes to the Technical Specifications (TSs) for Millstone Power Station Unit 2 (MPS2).
The proposed amendment request revises TSs 3.9.16 and 4.9.16, "Shielded Cask," to reduce the minimum decay time for fuel assemblies adjacent to the spent fuel pool cask laydown area from one year to 90 days.
2.0 REGULATORY EVALUATION
In 10 CFR 50.36, "Technical specifications," the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TS are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.
According to the licensee, the proposed change would revise the decay time as stated in LCO 3.9.16, "Shielded Cask," from 1 year to 90 days, which is consistent with the licensee's revised radiological analysis. In addition, associated surveillance requirement 4.9.16 would be revised to cite the decay time in LCO 3.9.16.
With the implementation of specific changes to the LCO Conditions and Surveillance requirements, the intent of the 10 CFR 50.36 regulatory requirements will continue be met.
The Nuclear Regulatory Commission (NRC) staff evaluated the radiological consequences of the postulated design basis accidents (DBAs) against the dose criteria specified in Title 10 of the Code of Federal Regulations Section 50.67 (10 CFR 50.67), "Accident source term," using the guidance described in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." The acceptance criteria for a fuel handling accident (FHA) are specified in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]
Edition," (SRP), Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms" and RG 1.183, Section 4.4, "Acceptance Criteria." The dose acceptance criteria for the FHA are a Total Effective Dose Equivalent (TEDE) of 6.3 rem for the worst two hours at the exclusion area boundary (EAB) and the outer boundary of the low population zone (LPZ), and 5 rem in the control room (CR) for the duration of the accident. The RG 1.183 provides guidance to licensees on acceptable application of alternative source term (AST) methodology, including acceptable radiological analysis assumptions.
The NRC approved selective implementation of the alternate source term (AST) methodology for FHA dose consequence analysis at MPS2 by License Amendment No. 284 to Facility Operating License DPR-65 (ADAMS Accession No. ML042360671), dated September 20, 2004. The cask tip accident was called the "cask drop accident" in the original AST submittal. The NRC approved revision of the FHA dose consequence analysis at MPS2 by License Amendment No. 298 to Facility Operating License DPR-65 (ADAMS Accession No. ML071450053), dated May 31, 2007. This safety evaluation (SE) addresses the impact of the proposed changes on previously analyzed DBA radiological consequences and the acceptability of the revised analysis results.
3.0 TECHNICAL EVALUATION
3.1 Cask Tip Accident Radiological Analysis The MPS2 current licensing basis (CLB) cask tip accident analysis postulates that a spent fuel cask tips over during refueling, damaging all of the fuel rods in the potential impact area. The potential impact area includes 872 fuel storage locations, of which 688 locations are allowed to contain consolidated canisters which can hold the equivalent of two fuel assemblies. The MPS2 CLB cask tip accident analysis that was approved in Amendment No. 298 assumes 1560 fuel assemblies are damaged, of which 184 of the fuel assemblies in the impact area have decayed for one year and the remaining 1376 fuel assemblies (in the 688 consolidated canisters) have decayed for five years.
The design inputs and assumptions used in the revised cask tip accident analysis have not been changed from those assumed in Amendment No. 298, except for the following:
- 1. Number of damaged fuel assemblies
- 2. Decay time of impacted fuel assemblies
- 3. CR unfiltered in leakage The revised cask tip accident analysis assumes that 1593 fuel assemblies are damaged, of which, 217 fuel assemblies are assumed to have decayed for 90 days, and the remaining 1376 fuel assemblies have decayed for five years. The MPS2 increased the number of fuel assemblies assumed to have decayed for less than five years in order to accommodate a full core offload of 217 fuel assemblies in the impact area, which conservatively assumes there are 33 more fuel assemblies in the impact area than there are storage locations. The MPS2 decreased the decay time from one year to 90 days in order to relax operational constraints on the scheduling of spent fuel cask loading. The revised cask tip accident analysis has not changed the core inventory radioactivity per fuel assembly assumed in Amendment 298.
The most limiting CLB cask tip accident analysis analyzed CR dose assuming the CR ventilation system isolates 20 seconds after the accident from a signal from the CR ventilation radiation monitor, and that the CR emergency ventilation (CREV) system is manually placed into the filtered recirculation mode one hour after isolation. The CLB cask tip accident assumes 800 cubic feet per minute (cfm) of unfiltered inleakage into the CR prior to CR isolation, 200 cfm of unfiltered in leakage after CR isolation until the CREV system is placed in filtered recirculation, and 2250 cfm of filtered flow and 200 cfm of unfiltered in leakage after the CREV system is placed in the filtered recirculation mode. The revised cask tip accident analysis was performed considering both isolated and unisolated CR ventilation scenarios. The unisolated CR ventilation scenario was the most limiting for the revised cask tip accident analyses. The most limiting revised cask tip accident analysis assumes 1000 cfm of unfiltered in leakage into the CR for the duration of the accident.
The licensee evaluated the radiological consequences resulting from the postulated cask tip accident and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose requirements provided in 10 CFR 50.67 and accident dose criteria specified in SRP 15.0.1 and RG 1.183. The NRC staff finds that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance, as identified in Section 2.0 of this SE.
These assumptions are presented in Table 1 and the licensee's calculated dose results are in Table 2 opposite the applicable acceptance criteria. Each calculated dose is below the applicable acceptance criteria. The NRC staff concludes that the EAB, LPZ, and CR doses estimated by the licensee for the cask tip accident meet the applicable accident dose criteria and are, therefore, acceptable.
3.2 Changes toTS 3.9.16 and TS 4.9.16 The TS 3.9.16 limits placing a shielded cask on the refueling floor whenever fuel assemblies without sufficient decay time are within striking distance of the spent fuel pool cask laydown area. The TS 4.9.16 requires verification that all fuel assemblies within striking distance of a shielded cask have sufficient decay time prior to moving a shielded cask onto the refueling floor and every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafter. The proposed amendment to TS 3.9.16 and TS 4.9.16 would revise the decay time for fuel assemblies within striking distance of a shielded cask in the spent fuel pool cask laydown area from one year to 90 days. The licensee also proposes to renumber the TS 3. 9.16.1 and TS 4. 9.16.1 to remove the subsection number (i.e., ".1 ") because TS 3.9.16.2 and TS 4.9.16.2 were deleted by Amendment 274 dated April1, 2003 (ML030910485).
As discussed in Section 3.1 above, the licensee evaluated the radiological consequences resulting from the postulated cask tip accident using a full core offload with a decay time of 90 days and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose requirements provided in 10 CFR 50.67 and accident dose criteria specified in SRP 15.0.1 and RG 1.183. The NRC staff reviewed the licensee's assumptions and inputs and found that the doses estimated by the licensee for the cask tip accident meet the applicable accident dose criteria. Therefore, the staff concludes these TS 3.9.16 and TS 4.9.16 changes are acceptable with respect to the radiological consequences of DBAs. In addition, the proposed changes to renumber TS 3.9.16.1 and TS 4.9.16.1 to remove the subsection.1 are minor corrective changes to reflect that each TS contains only one item consistent with a previous amendment that deleted TS 3.9.16.2 and TS 4.9.16.2.
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 35062).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). In addition, the amendment makes editorial, corrective or other minor revisions. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(1 0). 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 Contributors: M. Kichline D. Duvigneaud Date: November 14, 2013 Table 1 Data and Assumptions for the Cask Tip Accident Number of Fuel Assemblies Damaged:
Decay Time for Damaged Assemblies:
Unisolated Control Room Ventilation Flow:
Isolated Control Room Ventilation Flow:
Gap Fractions:
Time = 0 seconds Time = 20 seconds Time = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 20 seconds 1-131 Kr-85 Other noble gases Other halogens Pool Decontamination Factor (DF):
Noble gases lodines (effective DF)
Release Point:
Radial Peaking Factor:
Duration of Release:
1593 assemblies 217 assemblies decayed for 90 days 1376 assemblies decayed for 5 years 1000 cfm unfiltered flow for duration of accident Normal unfiltered intake flow = 1000 cfm Control room ventilation isolates Intake flow= 0 cfm Assumed unfiltered inleakage = 200 cfm Control room emergency ventilation (CREV) starts CREV filtered recirculation flow = 2250 cfm Assumed unfiltered inleakage = 200 cfm 12 percent 30 percent 10 percent 10 percent 1
200 Enclosure Building I Containment Ground 1.0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Table 2 Radiological Consequences of a Cask Tip Accident TEDE (rem)
Calculated Acceptance Location Dose (rem)
Criteria (rem)
Exclusion Area Boundary {EAB)
.5 6.3 Low Population Zone (LPZ)
.1 6.3 Control Room.(CRJ
.8 5.0
Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 November 14, 2013
SUBJECT:
MILLSTONE POWER STATION UNIT NO.2 -ISSUANCE OF AMENDMENT RE: REVISE TECHNICAL SPECIFICATION 3/4.9.16, SHIELDED CASK (TAC NO. MF1364)
Dear Mr. Heacock:
The Commission has issued the enclosed Amendment No. 316 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2, in response to your application dated April 3, 2013.
The amendment would revise Technical Specification 3.9.16 "Shielded Cask," due to changes to the minimum decay time for fuel assemblies adjacent to the spent fuel pool cask laydown area.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Docket No. 50-336
Enclosures:
- 1. Amendment No. 316 to DPR-65
- 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
PUBLIC RidsNrrLAKGoldstein RidsNrrPMMillstone RidsAcrsAcnw_MaiiCenter Resource Accession No.: ML13291A203 OFFICE LPL 1-1/PM LPL 1-1/LA NAME JKim KGoldstein DATE 10/22/13 10/22/13 Sincerely, Ira/
James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsRgn1 Mail Center RidsNrrDoriDpr Resource RidsNrrDraAadb Resource RidsOgcMaiiCenter Resource RidsNrrDorllpl1-1 Resource R. McKinley, Rl LPLI-1 R/F
- See memo dated October 3 2013 AADB/BC STSB/BC OGC LPL 1-1/BC(A)
JDozier*
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