ML14189A409

ISFSI 10 CFR 50.59, 10 CFR 72.48 Charge Report for 2013

ML14189A409
Person / Time
Site:	Millstone
Issue date:	06/30/2014
From:	Armstrong L Dominion, Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
14-311
Download: ML14189A409 (24)

Text

Dominion Nuclear Connecticut, Inc.

Rope Ferry Rd., Waterford, CT 06385 Mailing Address: P.O. Box 128 Waterford, CT 06385 dom.com U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

/Y Dominioni JUN,3 0 2014 Serial No.

MPS Lic/GJC Docket Nos.

License Nos.14-311 RO 50-245 50-336 50-423 72-47 DPR-21 DPR-65 NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNITS 1. 2.3. AND ISFSI 10 CFR 50.59, 10 CFR 72.48 CHANGE REPORT FOR 2013, AND COMMITMENT CHANGE REPORT FOR 2013 Pursuant to the provisions of 10 CFR 50.59(d)(2), the report for changes made to the facility for Millstone Power Station Unit 2 (MPS2) and Unit 3 (MPS3) are submitted via Attachments 1 and 2 respectively for the year 2013. There were no changes made to the facility for Millstone Power Station Unit 1 (MPS1) and the Independent Spent Fuel Storage Installation (ISFSI).

During 2013 there were no commitment changes for MPS1, MPS2, MPS3 or the ISFSI.

This constitutes the annual Commitment Change Report consistent with the Millstone Power Station's Regulatory Commitment Management Program.

If you have any questions or require additional information, please contact Mr. William D. Bartron at (860) 444-4301.

Sincerely, L. J. Armstrong Director, Nuclear Station Safety and Licensing Fy 1P

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 Page 2 of 2 Attachments:

2 Commitments made in this letter: None.

cc:

U. S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 S. J. Giebel NRC Project Manager Millstone Unit 1 U. S. Nuclear Regulatory Commission Two White Flint North, Mail Stop T-8 F5 11545 Rockville Pike Rockville, MD 20852-2738 L. A. Kauffman Health Physicist-DNMS U. S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 M. C. Thadani NRC Project Manager Millstone Units 2 and 3 U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 B-1 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 10 CFR 50.59 REPORT FOR 2013 Millstone Power Station Unit 2 Dominion Nuclear Connecticut, Inc. (DNC)

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 1 of 10 Millstone Power Station Unit 2 (MPS2)

S2-EV-10-0001, Revision 1 SIA File 0901238.301, Revision 0 SIA File 0901238.302, Revision 0 SIA File 0901238.303, Revision 0 SIA Report 0901238.401, Revision 0 ETE-MP-2010-0004, Revision 0 MPS2 Leak Before Break (LBB) Analysis for Alloy 82/182 Weld Overlays Revision 1 of this evaluation was issued to address two U. S. Nuclear Regulatory Commission (NRC) inspection comments documented within condition report CR505247. Specifically, during the NRC Permanent Plant Modification Inspection conducted in February of 2013, the NRC made the following observations on Evaluation

$2-EV-1 0-0001, Revision 0, which addressed leak before break (LBB) for weld overlays on dissimilar metal welds on the unisolable Reactor Coolant System (RCS) piping at MPS2:

" The LBB analysis performed by Structural Integrity Associates (SIA) contained potential non-conservatisms associated with the material properties (aging) of the reactor coolant piping and with the stratified thermal moments on the pressurizer surge line. These potential non-conservatisms were discussed in the original evaluation. The evaluation concluded that if the analysis was performed with the potentially more conservative values, the LBB limits would remain acceptable, based on consultation with SIA. While this conclusion is not challenged, there was insufficient documented basis provided by SIA to support the conclusion.

" The credited response time for LBB leakage detection was originally one gpm in one hour. The change made in 2010 documented two changes in leak detection capability associated with LBB (i.e., the rate was reduced and the trending time was increased). The sensitivity was changed to 0.25 gallons per minute (gpm) over a trending period of approximately one week. The rate of 0.25 gpm was adequately addressed within the original evaluation, but the trending time was not.

In response to the first item, the NRC staff reviewed and approved the pressurizer surge line LBB analysis in Letter A14085, dated May 4, 1999. (1) In the review process, the NRC had questions about the material properties utilized by MPS2 in the subject submittal. The NRC staff modified the MPS2 surge line analysis and performed their own analysis of the surge line. The NRC concluded that the difference in results were acceptable and approved the MPS2 surge line LBB. The material properties utilized by (1) Letter from Eaton, Sr., R. (NRC) to Necci, R. P. (NNECo), "STAFF REVIEW OF THE SUBMITTAL BY NORTHEAST NUCLEAR ENERGY COMPANY TO APPLY LEAK-BEFORE-BREAK STATUS TO THE PRESSURIZER SURGE LINE, MILLSTONE NUCLEAR POWER STATION, UNIT 2 (TAC NO. MA4126)"

Docket No. 50-336, May 4, 1999.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 2 of 10 SIA in the proposed 2010 surge line weld overlay LBB analysis were compared with the lower bound generic values utilized by the NRC in 1999. The material properties used in 2010 are the same as the material properties used in the original LBB analysis approved by the NRC in 1999 and are considered to be equivalent and acceptable.

In the NRC's review process for approval of the pressurizer surge line LBB analysis, the NRC staff questioned the thermal stratification forces and moments utilized by MPS2 in the subject submittal. In response to the NRC, additional evaluations to address the discrepancies in the loads were performed. The NRC staff modified the MPS2 surge line analysis and performed their own analysis of the surge line. The NRC concluded that the difference in results were acceptable and approved the MPS2 surge line LBB (Letter Al 4085, dated May 4, 1999). Relative to the moments utilized in the proposed 2010 analysis, these moments are from the 1998 SIA surge line LBB analysis which was approved by the NRC in 1999 and are considered to be equivalent and acceptable.

Based upon review of the above, these items would not constitute a departure from a method of evaluation described in the Updated Final Safety Analysis Report (UFSAR) used in establishing the design bases or in the safety analysis.

In response to the second item, Revision 1 modified the text to make plain that the 0.25 gpm LBB leakage detection capability sensitivity was for a trending period of one week. Although this leak detection trend time is longer than the previous leak detection time, the leak detection trend time of one week is deemed to be acceptable due to the relatively longer time required to progress from a leak to a break. A change in the RCS leakage detection capability for LBB is not considered by the NRC to be a change in methodology. Therefore, the implementation of the 0.25 gpm limit with an approximate one week trend time does not constitute a change in methodology.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 3 of 10 MPS2 S2-EV-1 1-0004, Revision 0 MP2-UCR-2011-013 MPS2 Main Turbine Electro-Hydraulic Control (EHC) System Digital Upgrade Note: This summary is resubmitted only to reflect change package MP2-UCR-2011-013 modified the MPS2 UFSAR during this reporting period.

The existing General Electric Mark I Electro-Hydraulic Control (EHC) system for turbine control was replaced with a modern, distributed, General Electric Mark Vie Digital Electro-Hydraulic Control (EHC) system. The new EHC system is a Triple Modular Redundant (TMR), fault tolerant design (including I/O and networking) which provides high reliability and supports online maintenance and testing. The existing Turbine Supervisory Instrumentation (TSI) system will also be replaced with a digital microprocessor based Bently Nevada 3500 system that will interface with the EHC system. The Bently Nevada system will be used for indications and alarms and will have no automatic turbine trip functions.

This evaluation addresses those portions of Design Change MP2-10-01016 where the 10 CFR 50.59 screening determined a design function was adversely affected because the change was judged to fundamentally alter the existing means of performing or controlling design functions: The following changes were considered:

" Single train analog to TMR digital control, since the digital controls contain different failure modes than the previously installed analog system,

" Conversion from hard controls to soft controls because it involved more than minimal differences in the Human Machine Interface (HMI),

• Change from diverse mechanical and electrical turbine trip mechanisms to redundant electrical turbine trip mechanisms.

The upgraded EHC system is more reliable than the original system. The software program undergoes a detailed validation and verification process, consistent with industry standards, and includes factory acceptance testing, on-site acceptance testing, and post modification testing to assure software integrity. The graphics displays and control features of the HMI workstations were developed in accordance with industry standards and provide several advantages over the previously installed controls. As a result, the new EHC system does not result in a more than minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR. Changing from diverse mechanical and electrical turbine trip mechanisms to redundant electrical turbine trip mechanisms impacted the probability that a turbine missile event may occur following a turbine overspeed event caused by the failure of the EHC system. The overspeed protection system reliability of the installed design has been evaluated by the manufacturer. The evaluation concluded the probability of an overspeed event is less for the new design than for the previously installed control system. As such, the change

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 4 of 10 did not result in a more than minimal increase in the likelihood of occurrence of a malfunction of a system, structure, or component (SSC) important to safety previously evaluated in the UFSAR. The modification does not increase the radiological dose consequences of any accident or malfunction of SSC important to safety previously evaluated in the UFSAR, does not introduce the possibility of an accident of a different type, does not result in a malfunction with a different result or an increased challenge to a fission product barrier than already analyzed in the UFSAR. The change does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Protective actions such as reactor scram required to respond to an EHC failure are provided by other systems external to the EHC and EHC control interfaces. The protective actions for the Reactor Protection System inputs were not modified by this activity. The protection systems are fully redundant and separate from the EHC system.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 5 of 10 MPS2

$2-EV-13-0001, Revision 0 ETE-NAF-2012-0152, Revision 0 MP2-UCR-2013-002 SP 2670, Revision 009 and SP 2670-001, Revision 010-04 MPS2 Main Steam Line Break (MSLB) and Loss of Coolant Accident (LOCA) Containment Analysis with Dominion GOTHIC Engineering Technical Evaluation (ETE) NAF-2012-0152 incorporates a reanalysis of the MPS2 LOCA and MSLB Containment Analysis using the NRC-approved Dominion GOTHIC methodology in topical report DOM-NAF-3-0.0-P-A and an 80'F Ultimate Heat Sink (UHS) temperature into the MPS2 licensing basis. The ETE implements new mass and energy release analyses for the LOCA and MSLB events to support closure of an open operability determination. This 10 CFR 50.59 evaluation also covers the corresponding changes to the MPS2 UFSAR and the associated surveillance.

The new LOCA and MSLB containment analysis using the Dominion GOTHIC methodology demonstrates that the containment design pressure of 54 pounds per square inch gage (psig) would not be exceeded. However, other results of the new LOCA analysis are not fully bounded by the existing LOCA engineering design basis limits:

" Post-LOCA Inside Containment and Engineered Safety Features (ESF) Room Equipment Environmental Qualification (EEQ) Profiles,

" Post-LOCA Reactor Building Closed Cooling Water (RBCCW) Inlet and Outlet Maximum Temperatures,

" Maximum Post-LOCA Sump Water Temperature Profile Emergency Core Cooling System (ECCS) and containment spray pump net positive suction head (NPSH) issue.

The inside containment EEQ profiles were modified using the Dominion GOTHIC methodology approved for this application. Using the revised RBCCW inlet temperature profile, a new EEQ temperature profile for the ESF Room was generated. These updated inside containment and ESF room temperature profiles were included in a revision to the EEQ environments specification and the associated equipment qualification records and post accident operating time calculations.

Using the Dominion GOTHIC methodology, revised maximum RBCCW and service water inlet and outlet temperatures throughout the systems were recalculated based on an 80'F UHS temperature. These increased RBCCW and service water temperatures were evaluated to ensure the structural integrity of the RBCCW and service water piping, equipment nozzles and pipe supports.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 6 of 10 The maximum sump water temperature versus time was recalculated using the Dominion GOTHIC methodology. Regarding ECCS and containment spray pump NPSH available calculations, the maximum sump water temperature increased following Sump Recirculation Actuation Signal (SRAS) from 207'F to 234°F. In order to continue to comply with Safety Guide I (Regulatory Guide 11), the minimum sump water level is determined assuming a breach of the containment integrity allowing some sump water to flash to steam when containment pressure and temperature is reduced to saturated conditions at 14.7 pound per square inch absolute (psia). This minimum sump water level is then used as an input to ECCS and containment spray pump flow delivery and net positive suction head (NPSH) available calculation following SRAS. The ECCS and containment spray pump flow delivery and NPSH available results were not adversely impacted by the increase in sump water temperature.

Reanalyzing the UFSAR Section 14.8.2 MSLB and LOCA containment analyses and incorporating the results of these analyses into the MPS2 licensing and engineering design basis does not increase the frequency of occurrence of a MSLB, LOCA or any other accident or malfunction of a system, structure, or component (SSC) previously evaluated in the UFSAR. There is no impact on the radiological consequences associated with a MSLB, LOCA or any other accident or malfunction previously evaluated in the UFSAR. Reanalyzing the UFSAR Section 14.8.2 MSLB and LOCA containment analyses and incorporating the results of these analyses into the MPS2 licensing and engineering design basis does not create the possibility for an accident of a different type or a malfunction of a SSC important to safety with a different result than any previously evaluated in the UFSAR. The UFSAR Section 14.8.2 MSLB and LOCA containment analysis continue to demonstrate that the containment design pressure of 54 psig is not exceeded. As such the design basis limit for the containment fission product barrier is neither exceeded nor altered. The fuel and Reactor Coolant System fission product barriers are not impacted by the reanalysis.

The NRC safety evaluation associated with the Dominion GOTHIC methodology identifies that the applications of the methodology described above are technically appropriate for the intended application and are acceptable. Additionally, the NRC safety evaluation associated with the DOM-NAF-3-0.0-P-A methodology concludes that this methodology is acceptable for use at all Dominion nuclear facilities, including MPS2. The use of the Dominion GOTHIC methodology for LOCA post-reflood mass and energy releases satisfies the NRC safety evaluation condition identified in DOM-NAF-3-0.0-P-A. There are no other safety evaluation conditions associated with these Dominion GOTHIC applications. Based on this, the MSLB and LOCA containment re-analyses do not result in departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses and can be implemented without prior NRC approval.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 7 of 10 MPS2

$2-EV-13-0002, Revision 0 MP2-12-01148-023 Control Element Assembly Position Display System (CEAPDS) Replacement The existing CEAPDS is being replaced with a new CEAPDS that will perform all the functions of the existing system. The existing CEAPDS digital computer reliability is declining. Parts are obsolete, and it is proving difficult to maintain.

The existing CEAPDS hardware and software is being replaced in the Spring 2014 refueling outage with a new Invensys Foxboro I/A based digital system and hybrid mesh network system. This new digital system has redundancies built in to provide a higher degree of reliability.

The CEAPDS functions are not safety related. In addition to the functions of the existing CEAPDS design, the new CEAPDS design:

Monitors Up pulse and Down pulse from each of the 61 Coil Power Programmers (CPP) Automatic Control Element Drive Mechanism Timer Modules (ACTMs) and displays on Control Element Assembly (CEA) Backup Pulse Count Screen.

Monitors pulse count inputs from 61 CPP lift coils and displays on Lift Coil Demands Screen. The lift pulses are provided by the CEA lift coil output to each CEA. This screen has an odometer (accumulation of counts) and a resettable counter display for each CEA. This screen replicates the mechanical impulse odometers associated with the ACTM lift coils. This screen can be used for ACTM troubleshooting purposes, e.g., too many pulses for the number of step movements. The mechanical impulse odometers in the CPP bins will be retained.

" Provides ACTM trouble alarm output for each ACTM and displays on ACTM Trouble Alarm Screen.

While the CEA pulse counting position indication is available in the new CEAPDS, this CEAPDS indication will not be used to satisfy the pulse counting indicator operability requirements of Technical Specification 3.1.3.3, since the new CEAPDS pulse-counting position indication will not meet the UFSAR Section 1.A (General Design Criterion 13) requirement of being independent from the CEAPDS reed switch position indication.

In UFSAR Chapter 14, the sequential uncontrolled CEA withdrawal event is analyzed with acceptance criteria based on the scenario being a moderate frequency event. A single out-of-sequence CEA withdrawal event is analyzed with less restrictive acceptance criteria based on the scenario being an infrequent event. The CEAPDS generated CEA motion inhibit signal is credited in the UFSAR Chapter 14 safety analysis to justify this less frequent event categorization.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 /Page 8 of 10 Since the new CEAPDS hardware and software is an Invensys Foxboro I/A based digital system and hybrid mesh network system, different failure modes are possible.

Due to these new potential failure modes potentially affecting a design function, this 10 CFR 50.59 evaluation was developed.

The likelihood of a failure of the replacement CEAPDS indications, deviation alarm or CEA motion inhibit signal is reduced from that of the existing system. As such, the frequency of occurrence of an accident or likelihood of a malfunction previously evaluated in the UFSAR is not increased. Since no reanalysis is involved, the replacement of the CEAPDS does not result in an increase in the radiological consequences of any accident or malfunction previously evaluated in the UFSAR. The CEAPDS replacement does not create the possibility of an accident of a different type or a malfunction with a different result than previously evaluated in the UFSAR. The change does not result in the fuel cladding, reactor coolant system, or containment fission product barrier design basis limits being exceeded or altered. Additionally, the proposed change does not involve a method of evaluation described in the UFSAR that is used in establishing the design bases or the safety analysis. Based on this, it is concluded that the proposed change can be implemented without prior NRC approval.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 9 of 10 MPS2

$2-EV-13-0003, Revision 0 MP2-12-01223-002 LBDCR 13-MP2-011 MP2-UCR-2013-009 MPS2 Temperature Indication Upgrades To Support 80'F UHS The previously existing local bi-metal temperature indicators at the Service Water (SW) inlets to the RBCCW heat exchangers were replaced with high accuracy resistance temperature detector (RTD) transmitters. This evaluation was written to support using the output of the new high accuracy digital temperature sensors, either in the control room from the plant process computer (PPC) communication link or locally at the RBCCW heat exchangers' service water inlet, as the primary means to perform the UHS temperature surveillance to comply with Technical Specification (TS) 3.7.11 "Ultimate Heat Sink" temperature limit.

The RBCCW service water inlet location is already identified in the UFSAR and TS bases as one of the acceptable measurement locations when the margin to the UHS temperature limits is decreased (UHS temperatures greater than 70 0F) and an accurate measurement of service water temperature is desired. The use of this location when there is a wider margin (UHS temperatures less than 70°F) between UHS temperature and the TS limit is therefore acceptable.

Regarding the digital nature of the new instrumentation, all the provided diagnostic and alerts available to Operations, coupled with the monitoring software to be added to the PPC, will make it highly unlikely that a critical failure associated with the new digital temperature monitoring system will go undetected for any appreciable time. Once detected, the operators will use either local RBCCW indication or some other suitable indication. The monitoring instrumentation has no automatic control function over any plant equipment and no other control functions are combined as part of this change.

The equipment, local indication or control room indication from the PPC does not interact with any automatic plant functions. Therefore, this activity does not create the possibility for an accident of a different type than any previously evaluated in the UFSAR. The temperature of the UHS is an initial condition used in the safety analysis and is enforced by limits described in TS 3.7.11. The actual UHS temperature limits are not being changed by this activity. The change in instrumentation used to monitor the temperature of the UHS cannot affect the frequency of the occurrence of an accident previously analyzed in the UFSAR, and does not result in an increase in the consequences of a malfunction of a SSC important to safety as previously evaluated in the UFSAR.

All SSCs that use service water for cooling are analyzed to adequately perform their safety functions at UHS temperatures allowed by TS 3.7.11. There is no change in the allowed temperature of the UHS, therefore the change in the selection of the equipment

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 10 of 10 monitoring the temperature of the UHS does not increase the likelihood of a malfunction of an SSC important to safety, nor does it increase the consequences of an accident previously evaluated in the UFSAR.

This activity has only to do with the measurement of UHS temperature. As such, it does not alter the design basis limit for a fission product barrier. There is no change in the allowed temperature of the UHS. The consequences of currently analyzed events are unchanged. Therefore, there is no change in the approach to the design basis limits for a fission product barrier from that already described in the UFSAR. This activity does not change any method of evaluation used in establishing the design bases or in the safety analysis. As such, the proposed change can be implemented without prior NRC approval.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 10 CFR 50.59 REPORT FOR 2013 Millstone Power Station Unit 3 Dominion Nuclear Connecticut, Inc. (DNC)

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 1 of 10 Millstone Power Station Unit 3 (MPS3)

$3-EV-13-0001, Revision 0 EVAL-ENG-RSE-M3C16, Revision 0 LBDCR No. 13-MP3-005 MP3-UCR-2013-005 MPS3 Implementation of Revised Fuel Rod Cladding Stress and Strain Limits The subject activity relates to the evaluation of the MPS3 Cycle 16 (M3C16) reload core design. The mechanical design of the Region 18 feed fuel for M3C1 6 is the same as that for the Region 17 assemblies, except that Optimized ZIRLO fuel rod cladding is used, replacing the ZIRLO cladding used in previous fuel regions. This resulted in implementation of revised fuel rod cladding stress and strain limits, applicable to Optimized ZIRLO. This is considered an adverse change to a method of evaluation and to a Design Basis Limit for a Fission Product Barrier (DBLFPB).

The use of the Optimized ZIRLO cladding in the fresh Region 18 fuel assemblies was approved for use at MPS3 by the U. S. Nuclear Regulatory Commission (NRC).(1) (2)

Westinghouse has assessed the impact of M3C16 operation on nuclear steam supply system (NSSS) accidents using NRC approved analysis methods. As part of the M3C16 Optimized ZIRLO cladding implementation, MPS3 is implementing the revised transient clad stress criterion for Westinghouse fuel as described in WCAP-1 0125-P-A, Addendum 1-A. This augments the existing method of evaluation. With the use of WCAP-10125-P-A, Addendum 1-A, the DBLFPB will now evaluate the cladding stress using American Society of Mechanical Engineers (ASME) pressure vessel guidelines.

The use of the revised transient clad stress criterion is a change to the method of evaluation as described in UFSAR.

The NRC-approved revised cladding stress criteria maintains the DBLFPB of one percent transient cladding strain criterion as described in the UFSAR. The current DBLFPB for cladding strain of one percent is not exceeded or altered. The revised cladding stress limit using the ASME pressure guidelines satisfies the design requirement not to exceed yield strength.

The former method of evaluation described in the UFSAR evaluated transient clad stress against the yield stress. The revised method of analysis documented in WCAP-10125-P-A, Addendum 1-A evaluates the cladding stress using ASME pressure vessel guidelines. The method of WCAP-10125-P-A, Addendum 1-A has been approved by the NRC for this intended application. Therefore, this activity does not result in a (1) Letter from Kim, J. (NRC) to Heacock, D. A. (Dominion), "MILLSTONE POWER STATION UNIT NO. 3 ISSUANCE OF AMENDMENT RE: THE USE OF OPTIMIZED ZIRLOTM FUEL ROD CLADDING (TAC NO. ME7663)," Docket No. 50-423, Dominion Serial No.12-626, September 24, 2012.

(2) Letter from Kim, J. (NRC) to Heacock, D. A. (Dominion), "MILLSTONE POWER STATION, UNIT 3 EXEMPTION FROM THE REQUIREMENTS OF TITLE 10 OF CODE OF FEDERAL REGULATIONS, PART 50, SECTION 50.46 AND APPENDIX K (TAC NO. ME7724)," Docket No. 50-423, Dominion Serial No.12-573, August 23, 2012.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 2 of 10 departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analysis and can be implemented for MPS3 without prior NRC approval. The revised cladding stress limit using the ASME pressure guidelines satisfies the original cladding stress limit DBLFBP. The revised cladding stress criteria have been approved by the NRC in WCAP-10125-P-A, Addendum 1-A. With a change in methodology, the change in the DBLFPB is allowed under the provisions of 10 CFR 50.59 and NRC approval is not required.

All other changes and analysis performed for MPS3 Cycle 16 operation were either previously approved by the NRC, or screened out in the 10 CFR 50.59 screening process. Therefore, the M3C16 reload and all analyses as described in EVAL-ENG-RSE-M3C1 6 can be implemented under the provisions of 10 CFR 50.59.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 3 of 10 MPS3

$3-EV-1 3-0002, Revision 0 ETE-NAF-2013-0008, Revision 0 MPS3 Implementation of Revised Analysis for Uncontrolled Rod Cluster Control Assembly (RCCA) Withdrawal from Subcritical Conditions (Updated Final Safety Analysis Report (UFSAR) 15.4.1)

Engineering Technical Evaluation (ETE) NAF-2013-0008 implements the MPS3 Rod Withdrawal from Subcritical (RWSC) analysis performed in Westinghouse Calculation CN-TA-06-99, Revision 1. Implementation of this analysis provides enhanced Departure from Nucleate Boiling (DNB) performance below the first mixing vane grid (MVG) through the use of the NRC-approved ABB-NV critical heat flux (CHF) correlation and its associated 95/95 departure from nucleate boiling ratio (DNBR) design limit.

Implementation of the CN-TA-06-99, Revision 1 MPS3 RWSC analysis identified two adverse effects which required evaluation under the requirements of 10 CFR 50.59:

" The revised RWSC analysis is evaluated at a higher core heat flux, 102% of the value assumed in the previous analysis. The higher power level reflects the licensed MPS3 NSSS thermal power of 3666 megawatts thermal (MWt), (3650 MWt core power), including the 2% power uncertainty contained within the RWSC event calculations. It is important to note that while the analysis is performed at a higher power level, there is no physical change to the operating conditions of MPS3 proposed as part of this activity.

The revised RWSC analysis employs the NRC-approved alternate W-3 CHF correlation, ABB-NV, to perform DNBR calculations below the first MVG. In using the ABB-NV CHF correlation, the NRC approved WCAP-14565-P-A, Addendum 2-P-A Revision 0 methodology is employed, which constitutes a methodology change and requires evaluation.

Evaluation of the increased power level determined the combined effect of all analysis modifications continues to satisfy the event acceptance criteria described in the UFSAR.

The change is not an initiator of any transient or malfunction, introduces no new failure modes or system, structure, or component important to safety alterations, and does not change or exceed a DBLFPB.

Evaluation of the use of NRC-approved ABB-NV methodology determined the change is permitted as its use within the MPS3 RWSC analysis is based on sound engineering practice, appropriate for the intended application, and within the limitations of the Safety Evaluation Report (SER). Use of the NRC-approved ABB-NV 95/95 DNBR design limit is allowed under 10 CFR 50.59 criterion (viii) since the ABB-NV methodology change is approved by the NRC for its intended use.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 4 of 10 The two changes described above do not impact the associated design function, exceed or alter a DBLFPB, or result in a departure in a method of evaluation as described in the UFSAR. Therefore, the CN-TA-06-99, Revision 1 RWSC analysis for MPS3 can be implemented without prior NRC approval.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 5 of 10 MPS3

$3-EV-13-0003, Revision 0 ETE-NAF-2013-0002, Revision 0 MPS3 Containment Reanalysis Due to Westinghouse Mass and Energy Errors The activity being performed is to implement a new set of design basis analyses that evaluates the MPS3 containment response to a Large Break Loss of Coolant Accident (LBLOCA) event. The revised analyses were necessitated by a series of errors in Westinghouse provided LBLOCA Mass and Energy (M&E) source term information to the Dominion GOTHIC analysis.

An evaluation was needed because the corrected LBLOCA M&E releases resulted in increased containment peak pressure and increases to the long term containment electrical equipment environmental qualification (EEQ) temperature envelope. The change in the long-term temperature envelope is accommodated within the current post accident operating times.

The total reanalysis effort involved in addressing the Westinghouse LBLOCA M&E errors included revised M&E calculations by Westinghouse using the current UFSAR methods, containment response calculations performed by Dominion Nuclear Safety Analysis using the current UFSAR methods, piping stress calculations, and equipment EEQ assessments. The work is summarized in Engineering Technical Evaluation ETE-NAF-2013-0002, which is the implementing document for the analysis packages.

The only changes to the containment response analyses are the revised Loss of Coolant Accident (LOCA) M&E data from Westinghouse and a reduction in the Refueling Water Storage Tank (RWST) temperature from 100°F in the UFSAR analyses to at least 750F in the new calculations. The RWST temperature assumption is greater than the current maximum limit of 50°F in Technical Specification (TS) 3/4.5.4.

Three primary LBLOCA containment response calculations were performed:

" A calculation that determines peak containment pressures and temperatures for comparison against containment design, for input to EEQ assessments, and for input to containment leakage/radiological consequences evaluations.

• A calculation determining maximum containment sump temperature performance for Emergency Core Cooling System (ECCS) net positive suction head (NPSH) evaluations and input to long-term chemical effects calculations.

" A calculation of the temperature response of Quench Spray System (QSS) piping, Recirculation Spray System (RSS) piping and the containment liner. This calculation provides input to evaluations of spray piping stress and hanger performance.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 6 of 10 ETE-NAF-2013-0002 summarizes the results of these three calculations and that of downstream analyses with revised inputs from the GOTHIC containment analyses. The conclusion is that all containment design limits and equipment limits continue to be met.

A license amendment request was needed to change the Peak Accident Pressure (Pa) value contained in TS 6.8.4.f from 41.4 to 41.9 psig, thus, the LOCA short-term peak pressure calculation could be implemented only after NRC review and approval of the license amendment. All other containment analysis consequences remain within limits described in the UFSAR and may be implemented without prior NRC review and approval.

The other facets of the analyses that have corrected the Westinghouse containment M&E errors can be implemented in MPS3 design basis as a result of this 10 CFR 50.59 evaluation as the frequency of occurrence of the initiating LBLOCA event is independent of the analysis of the plant's response to its postulated occurrence. The evaluation of plant equipment needed to respond to and recover from a LBLOCA has been shown to meet the current post accident operating times. Despite the increase in LBLOCA Pa, the containment leakage rates assumed in the radiological consequences calculations have been verified by the successful performance of a 2011 containment Integrated Leak Rate Test which bounded the increased value of Pa. The analyses demonstrated that the response of the plant with the assumed malfunctions of system, structure, or components (SSC) important to safety continued to meet the containment leakage assumptions in the radiological consequences calculations. The accident initiator, the LBLOCA, is independent of containment's response. Response of the containment to the LBLOCA is essentially unchanged. SSCs required to immediately respond to the LBLOCA and those required to be operable for extended periods of time during plant recovery have been demonstrated to be operable in the manner currently credited. The analysis of the core response to a LBLOCA was not affected by the Westinghouse M&E error. The analyses have demonstrated acceptable containment barrier response. The containment response was analyzed with methods currently described in the MPS3 UFSAR and therefore the reanalysis can be accepted without prior NRC approval.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 /Page 7 of 10 MPS3 S3-EV-1 3-0004, Revision 0 MP3-13-01005-001 LBDCR 13-MP3-010 MPS3 Temperature Indication Upgrades To Support 80°F Ultimate Heat Sink (UHS)

The previously existing local indicators at the Service Water (SW) inlets to the Reactor Plant Component Cooling Water (RPCCW) heat exchangers were replaced with high accuracy resistance temperature detector (RTD) transmitters with digital displays for local indication and digital communication with the Plant Process Computer (PPC). This evaluation was written to support the use of these new temperature sensors either in the control room via the PPC or locally at the RPCCW heat exchangers to comply with TS 3.7.5 "Ultimate Heat Sink" temperature limit.

The RPCCW SW inlet, as one of the SW branch line locations, is already considered in the current TS bases as one of the acceptable measurement locations. Regarding the digital nature of the new instrumentation, all the provided diagnostic and alerts available to Operations coupled with the monitoring software to be added to the PPC will make it highly unlikely that a critical failure associated with the new digital temperature monitoring system will go undetected for any appreciable time. Once detected, the operators will use either local RPCCW indication or some other suitable indication. The monitoring instrumentation has no automatic control function over any plant equipment and no other control functions are combined as part of this change.

The equipment, local indication or control room indication from the PPC does not interact with any automatic plant functions. The temperature of the UHS is an initial condition used in the safety analysis and is enforced by limits described in TS 3.7.5.

The actual UHS temperature limits are not being changed by this activity. The change in instrumentation used to monitor the temperature of the UHS cannot affect the frequency of the occurrence of an accident previously analyzed in the UFSAR, and does not result in an increase in the consequences of a malfunction of a SSC important to safety as previously evaluated in the UFSAR.

All SSCs that use service water for cooling are analyzed to adequately perform their safety functions at UHS temperatures allowed by TS 3.7.5. There is no change in the allowed temperature of the UHS, therefore the change in the selection of the equipment monitoring the temperature of the UHS does not increase the likelihood of a malfunction of an SSC important to safety, nor does it increase the consequences of an accident previously evaluated in the UFSAR.

This activity has only to do with the measurement of UHS temperature. As such it does not alter the design basis limit for a fission product barrier. There is no change in the allowed temperature of the UHS. The consequences of currently analyzed events are unchanged. Therefore, there is no change in the approach to the design basis limits for

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 8 of 10 a fission product barrier from that already described in the UFSAR. This activity does not change any method of evaluation used in establishing the design bases or in the safety analysis. Therefore, the proposed change can be implemented without prior NRC approval.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 9 of 10 MPS3 S3-EV-13-0005, Revision 0 MP3-13-01183-000 MP3-UCR-2013-018 MPS3 Residual Heat Removal System (RHS) Cross-Train Suction Motor Operated Valve Breaker Normal Alignment Change Fire Safe Shutdown circuit analysis of the RHS system pumps "cross-train" suction isolation valves, 3RHS*MV8701 B and 3RHS*MV8702A, with respect to Information Notice 92-18 and current circuit analysis techniques identifies the potential inability to establish cold shutdown in the required time frame.

The cable routes, compared to the fire areas credited train and circuit analyzed per the current accepted method, shows three fire areas where alignment of shutdown cooling could be in jeopardy. Two cables for 3RHS*MV8701 B route through fire areas ESF-3 and ESF-1 0. A postulated fire in one of these areas could cause a spurious close that also bypasses the torque limit switch resulting in valve damage and the valve being incapable of re-opening from the control switch or the manual hand-wheel. Two cables for 3RHS*MV8702A route through fire area ESF-7. A postulated fire in this area could cause a spurious close that also bypasses the torque limit switch resulting in valve damage and it being incapable of re-opening from the control switch or the manual hand-wheel. Either of the cases, for the affected area, could result in no credible method to cool the plant to cold shutdown.

The current activity provides mitigation of the postulated fire induced valve damage by maintaining 3RHS*MV8701 B and 3RHS*MV8702A circuit breakers, MCC32-3U(FI FB) and MCC32-4T(F7KT) "OFF" while in MODES 1, 2, or 3 or during periods in which the RHS is not aligned to perform its residual heat function.

With proper isolation of the low pressure RHS from the high pressure reactor coolant system (RCS), the RHS is not an initiator for any accidents currently presented in the UFSAR. The proposed change increases the isolation with two of the three valves closed and unpowered. [Prior to the change one of the three in series RHR suction valves was closed and unpowered in operating modes 1, 2, and 3]

The introduction of a new operator action is evaluated in assessing whether the change increases the likelihood of occurrence of a malfunction of a SSC important to safety.

The change meets the NEI 96-07 guidance of being simple, included in plant operating procedures, is verified as being able to be performed and has no interactions with other plant systems.

The RHS does perform a safety function of providing low head safety injection (LHSI) to the RCS in the event of a LBLOCA. However, in its alignment to provide injection, the RHS suction valves are closed and flow is direct from the RWST to the RHS suction.

Serial No.14-311 10 CFR 50.59 and Commitment Change Report for 2013 / Page 10 of 10 Therefore, the change neither results in an increase in the consequences of a malfunction of a SSC important to safety or an increase in the consequences of an accident previously evaluated in the UFSAR.

With the RHS isolated from the RCS in higher operating MODES, the system is not an initiator for any accident. The activity increases the isolation in that two of the three RHS suction valves will be closed and unpowered. The change therefore does not create a possibility for an accident of a different type than any previously evaluated in the UFSAR. The approach of the plant to exceeding (or to the point to which they are exceeded) the design basis limits for fission product barriers depend in part upon the safety function of the RHS to provide LHSI if needed in response to an accident. The ability of the RHS to provide that flow is unchanged by this activity. Therefore there is no change in the approach to any fission product barriers. Additionally, the change has no relationship to establishing the basic limits for the fission product barriers. The activity is not related to any method of evaluation discussed in the UFSAR.

Consequently, this change can be implemented without prior NRC approval.