ML18184A434

From kanterella
Jump to navigation Jump to search
1, 2, 3, and ISFSI 10 CFR 50.59, 10 Crf 72.48 Change Report for 2017, and Commitment Change Report for 2017
ML18184A434
Person / Time
Site: Millstone  Dominion icon.png
Issue date: 06/21/2018
From: O'Connor M
Dominion Energy Co, Dominion Energy Nuclear Connecticut
To:
Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
18-226
Download: ML18184A434 (21)
v • d • e


Text

Dominion Energy Nuclear Connecticut, Inc.

-,;. Dominion Rt 156, Rope Ferry Road, Waterford, CT 06385 Dominion Energy.com j a,, Energy U.S. Nuclear Regulatory Commission JUN 2* 1 2019 Serial No.18-226 Attention: Document Control Desk MPS Lic/GJC RO Washington, DC 20555 Docket Nos. 50-245 50-336 50-423 72-47 License Nos. DPR-21 DPR-65 NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNITS 1, 2, 3, AND ISFSI 10 CFR 50.59, 10 CFR 72.48 CHANGE REPORT FOR 2017, AND COMMITMENT CHANGE REPORT FOR 2017 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 2017. The report for changes made to the facility for both MPS2 and MPS3 is submitted via attachment 3. The report for changes made to the Independent Spent Fuel Storage Installation (ISFSI) for MPS is submitted via attachment 4. There were no changes made to the facility for Millstone Power Station Unit 1 (MPS1 ). submits the commitment changes for MPS. 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. Jeffry A. Langan at (860) 444-5544.

Sincerely, M. J. O'Connor Director, Nuclear Station Safety and Licensing

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Page 2 of 2 Attachments: 5 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 T. H. Carter NRG Project Manager Millstone Unit 1 U.S. Nuclear Regulatory Commission Two White Flint North, Mail Stop T-8 F8 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 R. V. Guzman NRG Senior Project Manager Millstone Units 2 and 3 U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 C-2 11555 Rockville Pike Rockville, MD 20852-2738 NRG Senior Resident Inspector Millstone Power Station I_

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 1 10 CFR 50.59 REPORT FOR 2017 MILLSTONE POWER STATION UNIT 2 Millstone Power Station Unit 2 Dominion Energy Nuclear Connecticut, Inc. (DENC)

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 1/ Page 1 of 3 MPS2 MPS2-EVAL-2017-0001 Title Introduction of the AREVA Standard CE14 HTP Fuel at Millstone Power Station Unit2 A 10 CFR 50.59 Evaluation was performed for the introduction of the AREVA Standard CE 14 HTP fuel, a revised boron dilution event, and a revised post-scram main steamline break analysis. The 10 CFR 50.59 Screen identified five (5) items that required evaluation under the criteria of 10 CFR 50.59(c)(2). The five items are:

(1) Methodology change for cladding stress and corrosion limits, (2) Updated assessment of the spent fuel pool, (3) Methodology change calculating spent fuel pool heat loads, (4) Methodology change for evaluation of the Boron Dilution event, and (5) Methodology change for evaluation of the Post-Scram MSLB event.

Item (1) incorporates the methodology of BAW-10240(P)(A) into the Millstone Power Station Unit 2 (MPS2) licensing bases for evaluating the Standard CE14 HTP fuel with M5 cladding.

This activity is evaluated against criterion 10 CFR 50.59(c)(2)(viii). The activity does not result in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analysis and may be implemented without NRC approval. The cladding stress and corrosion limits applicable to M5 clad were approved as part of BAW-10240(P)(A). With a change in methodology, the change in the Design Basis Limit for Fission Product Barrier (DBLFPB) is allowed under the provisions of 10 CFR 50.59 and NRC approval is not required.

Item (2) analyzed the spent fuel pool heat load to account for the AREVA Standard CE14 HTP fuel, discharge of 81 fuel assemblies versus 80, increase in the minimum in-core hold time, and 180 locations that are blocked per the spent fuel pool criticality analysis. This activity is evaluated against criteria 10 CFR 50.59(c)(2)(i) through 10 CFR 50.59(c)(2)(vii). The analysis of the spent fuel pool heat load does not change the frequency of occurrence an accident, likelihood of occurrence of a malfunction of a system, structure, or component (SSC) important*

to safety, change the consequences of an accident, consequences of a malfunction of a SSC, create an accident of a different type, failure modes/responses due to a malfunction of a SSC, or result in a DBLFPB being exceeded or altered at MPS2. The calculated spent fuel pool heat load accounting for the changes remains bounded by the conservative heat loads given in MPS2 Final Safety Analysis Report (FSAR) Section 9.5. All DBLFPBs associated with the spent fuel pool heat load were shown to be met. Therefore, the change may be implemented without NRC approval.

Item (3) changed an element of the method of evaluation for calculating spent fuel pool heat loads. This activity is evaluated against criterion 10 CFR 50.59(c)(2)(viii). The spent fuel pool heat load was calculated using the ORIGEN-ARP code instead of ORIGEN2 as described in MPS2 FSAR Section 9.5. The decay heat loads calculated with ORIGEN-ARP are essentially the same as decay heat loads calculated using ORIGEN2. Therefore, the change to the method of evaluation is allowed by the provisions of 10 CFR 50.59 and does not require NRC approval.

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 1/ Page 2 of 3 Item (4) changed the method of evaluation for the boron dilution event. This activity is evaluated against criterion 10 CFR 50.59(c)(2)(viii). The boron dilution event was analyzed following the NRC approved methodology documented in AREVA Topical Report EMF-2310(P)(A), Rev. 1. Therefore, the change to the method of evaluation is allowed by the

  • provisions of 10 CFR 50.59 and does not require NRG approval.

Item (5) changed an element of the method of evaluation for the post-scram main steam line break event. This activity is evaluated against criterion 10 CFR 50.59( c)(2)(viii). The modeling of auxiliary feedwater in the post-scram main steamline break event was changed to model flow

.. to both steam generators instead versus only modeling flow to the affected steam generator. A sensitivity case was run to evaluate the impact of the modeling change. The sensitivity case demonstrated that the results are essentially the same. Therefore, the change to the method of evaluation is allowed by the provisions of 10 CFR 50.59 and does not require NRC approval.

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 1/ Page 3 of 3 MPS2 MPS2-EVAL-2017-0002 Title Correction to the Axial Shape Index (ASI) Uncertainty Used in the FSAR Chapter 14 Safety Analysis for MPS2 A correction to the Axial Shape Index (ASI) uncertainty used in the FSAR Chapter 14 Safety Analysis is being implemented including a change to the Millstone Power Station Unit 2 (MPS2)

Final Safety Analysis Report (FSAR).

The overall ASI uncertainty is used in the MPS2 FSAR Chapter 14 Departure from Nucleate Boiling (DNB) Limiting Condition for Operation (LCO) verification associated with the FSAR Section 14.3.1 four reactor coolant pump loss of flow and FSAR Section 14.4.3.1 control element assembly drop events. An incorrect ASI tolerance for comparing the calculated ASI value to the value displayed on main control room board C04F was used as an input in determining this overall Axial Shape Index (ASI) uncertainty. The current FSAR DNB LCO verification analysis used an ASI tolerance for comparing the calculated ASI value to the value displayed on main control room board C04F of+/- 0.004 ASI units. The correct ASI tolerance is +/-

0.04 ASI units. This ASI tolerance is one of the inputs AREVA uses to determine the overall ASI uncertainty input to the DNB LCO verification analysis.

The analysis correcting the ASI tolerance error resulted in an increase of the overall ASI uncertainty from +/-0.029 ASI units to +/-0.053 ASI units. As shown in FSAR Table 14.0.7-5, this overall ASI uncertainty is one of several other uncertainties statistically combined that is used in the DNB LCO verification analysis. The results of the DNB LCO verification analysis utilizing the correct overall ASI uncertainty are presented in revised FSAR Figure 14.0.7-5. The figure demonstrates that the results of the DNB LCO verification analysis are not significantly impacted by the increase in the ASI uncertainty.

A 10 CFR 50.59 evaluation was required because a reanalysis of the DNB LCO verification portion of the FSAR Chapter 14 loss of RCS flow and control element assembly drop events was required to demonstrate adequate margin remained to the DNB limit.

Correcting the ASI tolerance and revising the FSAR Chapter 14 Safety Analyses that use that parameter as an input does not involve a physical change to any SSC. As such, no new failure modes are introduced that can cause a transient or an accident or a malfunction of a system, structure, or component (SSC) important to safety. As such, the likelihood of an accident or a malfunction of an SSC previously evaluated has not increased and the change does not introduce the possibility of an accident of a different type, or a malfunction with a different result.

The incorrect ASI tolerance was not used in any FSAR Chapter 14 event whose radiological consequences were previously evaluated. As such, there is no increase in the consequences of an accident previously evaluated. The change does not impact any equipment credited to mitig.ate the radiological dose consequences of any accident previously evaluated in the FSAR and there is no impact on the consequences of a malfunction of a SSC important to safety previously evaluated in the SAR. The reanalysis of the FSAR Chapter 14 DNB LCO verification using the same methodology in the previous analysis demonstrated that the fuel DNB limit is not violated and therefore the fuel clad fission product barrier design basis limit is not exceeded or altered. There is no impact on the RCS or containment fission product barriers.

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 2 10 CFR 50.59 REPORT FOR 2017 MILLSTONE POWER STATION UNIT 3 Millstone Power Station Unit 3 Dominion Energy Nuclear Connecticut, Inc. (DENC)

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 2/ Page 1 of 2 MPS3 MPS3-EVAL-2016-0007 Title Retirement of the Unit 3 Recirculation Tempering Line (3-CWS-060-046-4, 3-CWS-060-047-4, 3-CWS-060-027-4, and 3-CWS-060-26-4)

This activity involves the retirement of the Millstone Power Station Unit 3 (MPS3) recirculation tempering line (3-CWS-060-046-4, 3-CWS-060-047-4, 3-CWS-060-027-4, and 3-CWS-060 4). The line will be retired from valve 3CWS-V957 to the sluice gates. This equipment is retired in-place. The de-icing function is eliminated as it has been determined to not be needed for MPS3. The following existing design features assure that icing or ice blockage of the intake screens and pumps will not adversely affect safety-related facilities and water supplies. For ice to obstruct flow to the Intake and extend below the reinforcement curtain wall, the ice would need to be 7 feet thick, which is not a credible event for a salt water plant. If ice would obstruct flow to the Intake Structure, the Circulating Water pumps would trip on the differential water level across the screens. The Intake Structure entrance fluid velocity is below the minimum velocity needed to keep frazil ice suspended and submerged. Moreover, historical ice formation and studies support the continued low probability of ice formation as discussed in Final Safety Analysis Report (FSAR) sections 2.4.7 and 2.4.11.6. The license renewal function of the tempering line is to provide de-icing capability at the intake, if required. The Evaluation has concluded that the retirement of the tempering line will not more than minimally increase the likelihood of the formation of ice which would block flow to the safety related Service Water system. Therefore, the retirement of the tempering line removes the piping from the License Renewal program. This is a change from what was previously stated in NUREG-1838.

The 10 CFR 50.59 evaluation concluded the following:

  • The change did not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the FSAR.
  • The change did not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the FSAR.
  • The change did not result in more than a minimal increase in the consequences of an accident previously evaluated in the FSAR.
  • The change did not result in more than a minimal increase in the consequences of a malfunction of a SSC important to safety previously evaluated in the FSAR.
  • The change did not create a possibility for an accident of a different type than any previously evaluated in the FSAR.
  • The change did not create a possibility for a malfunction of a system, structure, or component (SSC)- important to safety with a different result than any previously evaluated in the FSAR.
  • The change did not result in a design basis limit for a fission product barrier as described in the FSAR being exceeded or altered.
  • The change did not result in departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses.

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 2/ Page 2 of 2 MPS3 MPS3-EVAL-2017-0001 Title Loading and Operation of Eight Lead Test Assemblies (LTAs) With AXIOM Cladding in Millstone Unit 3 The loading and operation of eight Lead Test Assemblies (LTAs) with AXIOM cladding In Millstone Unit 3 (MPS3) is being evaluated under 10 CFR 50.59. This evaluation was deemed necessary in the 10 CFR 50.59 Screen because the AXIOM LTAs are a test/experiment not specifically described In the MPS3 Final Safety Analysis Report (FSAR) and outside the reference bounds described in the FSAR. An exemption request from the requirements of 10 CFR 50.46 and Appendix K of 10 CFR Part 50 was submitted to the NRC in Reference 3 to allow the use of AXIOM cladding at MPS3. The NRC approved this exemption request in Reference 4. In addition, cycle-specific and generic engineering evaluations concluded that the AXIOM LTAs do not increase the frequency of occurrence of an accident or an system, structure, or component (SSC) important to safety malfunction, and have been shown to meet all design limits so that the consequences of an accident or SSC malfunction are not Increased. No new accident or SSC malfunctions are created by the AXIOM LTAs, due to the similarity with the resident fuel product. Engineering evaluations determined that no Design Basis Limit for Fission Product Barriers (DBLFPBs) were exceeded or altered. There was no departure from a method of evaluation described in the FSAR, as appropriate restrictions have been imposed (only eight LTAs, LTAs only in non-limiting core locations in accordance with MPS3 Technical Specification 5.3. 1).

Therefore, It was concluded that the loading and operation of the AXIOM LTAs may be implemented without further NRC review and approval In accordance with 10 CFR 50.59.

References:

1. Westinghouse Report, WCAP-9272-P-A, 'Westinghouse Reload Safety Evaluation Methodology," July 1985.
2. Westinghouse Letter, NF-NEU-17-49, 'Westinghouse AXIOM' Lead Test Assemblies - Revised 10 CFR 50.59 Input," dated August 23, 2017.
3. Dominion Correspondence, Dominion Serial No.16-242, "Dominion Nuclear Connecticut, Inc., MPS3, Proposed Exemption Request for the Use of AXIOM Cladding Material in Lead Test Assemblies," June 30, 2016 [Adams Accession No. ML16189A104).
4. NRC Correspondence, Dominion Serial No.17-212, "Millstone Power Station, Unit No. 3 - Exemption from the Requirements of 10 CFR 50.46 and Appendix K of 10 CFR Part 50, to allow the use of AXIOM cladding material In lead test assemblies CAC No. MF8210,"dated May 10, 2017.

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 3 10 CFR 50.59 REPORT FOR 2017 MILLSTONE POWER STATION UNITS 2 & 3 Millstone Power Station Units 2 and 3 Dominion Energy Nuclear Connecticut, Inc. (DENC)

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 3/ Page 1 of 6 MPS2&3 SG-EV-15-0002 Revision 1

Title:

Replacement of 345 kV Obsolete and Aging Electro-Mechanical and Solid State Protective Relaying This 10 CFR 50.59 evaluation reviewed the replacement of 345 kV obsolete and aging electro-mechanical and solid state protective relaying with newer state of the art microprocessor based digital relays. The replacement digital protective relays which were installed approximately 7 -

10 years ago are owned and maintained by the grid owner. The evaluation evaluated the replacement of electro-mechanical and solid state relays with microprocessor based digital protective relays.

The replacement protective relays are digital in nature. However, to ensure diversity in the protection scheme, the primary and backup digital protective relays are of a different manufacturer and design. Similar to the previous design, either the primary or backup protective digital relay sensing a fault condition will result in the protective action of isolating the faulted 345 kY line. The digital relays were installed because the existing electro-mechanical and solid state relays were obsolete and aging and the reliability of the existing relays was reduced.

The 10 CFR 50.59 evaluation concluded that the following:

  • The change did not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the Safety Analysis Report (SAR).
  • The change did not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a system, structure, or component (SSC) important to safety previously evaluated in the SAR.
  • The change did not result in more than a minimal increase in the consequences of an.

accident previously evaluated in the SAR.

  • The change did not result in more than a minimal increase in the consequences of a malfunction of a SSC important to safety previously evaluated in the SAR.
  • The change did not create a possibility for an accident of a different type than any previously evaluated in the SAR.
  • The change did not create a possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the SAR.
  • The change did not result in a design basis limit for a fission product barrier as described in the SAR being exceeded or altered.
  • The change did not result in departure from a method of evaluation described in the SAR used in establishing the design bases or in the safety analyses.

The digital relays provide the same protective action as the relays which they replaced and do not introduce any new operator intervention or burden with the change. The digital relays are periodically checked in accordance with Table 1-1 of Standard PRC-005-2 and the Nuclear Electric Insurance Limited (NEIL) standard. Further, these relays provide remote alarm notification via the Supervisory Control And Data Acquisition (SCADA) system which is monitored by the grid operator. The installation of the newer state of the art digital relays was performed to maint replacement of 345 kV obsolete and aging electro-mechanical and solid state protective relaying ain the reliability of the transmission line protection system by replacing

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 3/ Page 2 of 6 aging and obsolete electromechanical and solid state protective relays with newer state of the art microprocessor based digital relays. The newer digital relays are of a proven design and are used throughout by the grid owner for similar applications. These specific protective relays which provide the 345 kV transmission line protection have been in operation for approximately 7 to 10 years of service and have been shown to be reliable.

Both MPS2 and MPS3 Final Safety Analysis Reports (FSARs) require updating to reflect the replacement of the existing electro-mechanical and solid state protection relays with digital protection relays. The FSARs identify these protection relays as being electromechanical and solid state relays when in fact they are now digital based protection relays.

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 3/ Page 3 of 6 MPS2 &3 MPSO-EVAL-2017-0001

Title:

Reconfiguration of a 345kv Transmission Line Connected to The Millstone Power Station (MPS) Switchyard by the Transmission Operator - Eversource Energy The activity being evaluated is reconfiguration of a 345kV transmission line connected to the Millstone station switchyard by the transmission operator - Eversource Energy. The current 3-terminal Millstone - Haddam - Beseck 348 line is being reconfigured into two 2-terminal lines, 348 line - Millstone to Haddam, and 3252 line - Haddam to Beseck, in conjunction with other 345kV transmission system improvements in eastern Connecticut. The scope also includes changing the line protection scheme to reflect two terminal lines. The transfer trip devices between Millstone and Beseck substations will no longer be required and will be removed, and the corresponding line protection relay settings will be modified to reflect the new line configurations.

The proposed design is consistent with the regional transmission system standards authority, Northeast Power Coordinating Council, requirements that are credited in the licensing bases for MPS Unit 2 (MPS2) and MPS Unit 3 (MPS3) and described in the Final Safety Analysis .Reports (FSARs).

The activity introduces a new transmission system alignment that has potential for impacting functionality of the 348 transmission line when the new 3252 line is out of service. An outage of the new 3252 line will remove the 345kV source to the existing 348 line. As a result, a new technical requirement is being added to each units Technical Requirements Manual (TRM) to address this condition by applying limitations on station output for 3252 line outages that match the existing limitation associated with 348 line outages.

  • The change did not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the SAR, since the design of the reconfigured 348 line complies with the Northeast Power Coordinating Council, Inc. (NPCC) standards that form the bases for the MPS2 and MPS3 design and licensing bases, and since a new technical requirement is added to the 3252 line to match the TRM of the existing 348 line.
  • The change did not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a system, structure, or component (SSC) important to safety previously evaluated in the SAR, since the reconfigured 348 line complies with the Design Basis/Licensing Basis (DB/LB) of the existing FSARs, and since outages of the new 3252 line are managed the same as the existing 348 line.
  • The change did not result in more than a minimal increase in the consequences of an accident previously evaluated in the Safety Analysis Report (SAR), since design basis events assume either offsite power is available or lost and therefore configuration of the 348 line does not impact the assumptions of the accidents previously evaluated and will not result in an increase in the radiological dose consequences of any accident previously evaluated.
  • The change did not result in more than a minimal increase in the consequence of a malfunction of a SSC important to safety previously evaluated in the SAR, since the offsite power system will continue to function consistent with the design bases as described in the SAR safety analyses, and radiological consequences will not increase.

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 3/ Page 4 of 6

  • The change did not create a possibility for an accident of a different type than previously evaluated in the SAR. The proposed design affects SSCs associated with the offsite power availability, which is already considered extensively in the existing analysis.
  • The change did not create a possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the SAR. Result of malfunctions of the proposed design are either unchanged or are previously evaluated in the SAR.
  • The change did not result in a design bases limit for a fission product barrier as described in the SAR being exceeded or altered.
  • The change did not result in a departure from a method of evaluation described in the SAR used in establishing the design bases or used in the safety analysis. The proposed activity does not affect or involve any methods of evaluation.
  • Both MPS2 and MPS3 FSARs require update to reflect the reconfiguration of the 348 transmission line.

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 3/ Page 5 of 6 MPS2 &3 MPSO-EVAL-2016-0002

Title:

Replacement of 345 kV Obsolete Electro-Mechanical Breaker Failure Relaying With State of the Art Microprocessor Based Digital Relays The activity being evaluated was the replacement of 345 kV obsolete electro-mechanical breaker failure relaying with state of the art microprocessor based digital relays. The replacement digital relays are owned and maintained by the grid design authority. The evaluation was prepared to address the potential introduction of a new common cause failure (due to the replacement of electro-mechanical with microprocessor based digital protective relays) and the incorporation of multiple protective relaying functions into a single digital device.

The replacement protective relays are digital in nature. The proposed design is consistent with the requirements of the governing standard, including incorporation of diversity in the design of the protection scheme. Specifically, the primary and backup digital breaker failure relays were produced by different manufacturers and employ different designs. Similar to the previous design, the primary breaker failure digital relay sensing a failure of a 345kV breaker to open on demand, will result in a trip signal to the next in-line circuit breaker to clear a faulted 345 kV line.

New to the design is the provision of backup breaker failure relays that will provide a protective function redundant to the primary breaker failure relay. The digital relays provide the same protective action as the relays which they replaced and do not introduce any new operator intervention or burden with the change. The newly installed breaker failure relays provide remote alarm notification monitored by the grid operator.

The 10 CFR 50.59 evaluation concluded that the following:

  • The change did not result in more than a rninimal increase in the frequency of occurrence of an accident previously evaluated in the Safety Analysis Report (SAR). The proposed design is proven and in accordance with applicable design standards.
  • The change did not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a system, structure, or component (SSC) important to safety previously evaluated in the SAR. The proposed design is proven and in accordance with applicable design standards.
  • The change did not result in more than a minimal increase in the consequences of an accident previously evaluated in the SAR The proposed activity does not affect SSCs associated with accident consequences.
  • The change did not result in more than a minimal increase in the consequences of a malfunction of a SSC important to safety previously evaluated in the SAR The proposed activity does not affect SSCs associated with accident consequences.
  • Ttie change did not create a possibility for an accident of a different type than any previously evaluated in the SAR The proposed design affects SSCs associated with offsite power availability, which is already considered extensively in existing accident analysis.
  • The change did not create a possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the SAR. Results of malfunctions of the proposed design are either unchanged or are previously evaluated in the SAR

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 3/ Page 6 of 6

  • The change did not result in a design basis limit for a fission product barrier as described in the SAR being exceeded or altered. The SSCs in the proposed design is not associated with Design Basis Limit for Fission Product Barriers (DBLFPBs).
  • The change did not result in departure from a method of evaluation described in the SAR used in establishing the design bases or in the safety analyses. The proposed activity does not affect or involve any methods of evaluation. Both MPs2 and MPs3 FSARs require updating to reflect the replacement of the existing electro-mechanical relays with digital protection relays.

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and qommitment Change Report for 2017 Attachment 4 10 CFR 50.59 REPORT FOR 2017 MILLSTONE POWER STATION ISFSI Millstone Power Station ISFSI Dominion Energy Nuclear Connecticut, Inc. (DENC)

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 4/ Page 1 of 1 MPS ISFSI MPSO-EVAL-2017-0001 A Title The activity being evaluated is the addition of shims under some of the corners of HSMs 33-52 to address imperfections in the ISFSI pad and allow for minimizing gaps between HS Ms. The design function of the HSM is to provide missile protection, biological shielding and provide for passive heat removal of the dry storage canister. A 10 CFR 72.48 evaluation was required due to an adverse change in the Goefficient of friction between the HSM and the ISFSI pad; additionally, since a calculation was required to ensure the design function of the HSM still met the requirements of the UFSAR a 10 CFR 72.48 evaulation was required.

The design function that is affected by the activity is missile protection and natural phenomena accidents. Missile protection was not affected since the smallest tornado generated missile is inches in diameter and the maximum gap analyzed is 2 inches. The stability of the HSM under seismic, tornado and flood accidents were the only accidents possibly affected by the activity.

The method of evaluation used to validate the installation of shims under the HSM(s) in NUH004-0279, Rev. 2 was identical to that used in the Updated Final Safety Analysis Report (UFSAR). Although the coefficient of friction was changed to model the shims installation all acceptance criteria (ANSI 57.9 - 1984) were met and the factors of safety exceeded 1.1, as required by code.

A review of 10 CFR 72.48(c)(2)(i-viii) shows that all eight criteria were answered in the negative.

The activity does not result in a more than minimal increase in the frequency or likelyhood of an occurrence of an accident or malfunction of a system, structure, or component (SSC) important to safety.

The activity does not result in a more than minimal increase in the consequences of an accident or malfunction of a SSC important to safety.

  • The activity does not create the possiblity of a different accident or a malfunction of an SSC with a different result from that previously evaluated.

The activity does not affect any design basis limit for a fission product barrier.

The activity does not depart from the method of evaluation described in the UFSAR.

It is also noted here that the MPS 10 CFR 72.212 Report (ETE-NAF-2016-0030) will require revision to include this evaluation (and Screen MPSO-SCRN-2017-0260) and inclusion of Calculation NUH004-0279, Rev. 2. Corrective Action item PA3142710 has been initiated to track revision of ETE-NAF-2016-0030.

This evaluation was performed using guidance from NEI 96-07, Appendix B.

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 5 COMMITMENT CHANGE REPORT FOR 2017 MILLSTONE POWER STATION Millstone Power Station Dominion Energy Nuclear Connecticut, Inc. (DENC)

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 5/ Page 1 of 2 Cleaning and Inspection Frequency of the MP2 RBCCW Heat Exchangers - RCR-43029 Two commitments contained in Millstone Power Station Unit 2 (MPS2) Ultimate Heat Sink License Amendment Request (Dominion Letter Serial 13-227, dated May 3, 2013 were changed. They are:

  • Administrative controls will be established for cleaning of the Reactor Building Closed Cooling Water (RBCCW) heat exchangers at a 3-month interval.
  • For Generic Letter (GL) 89-13, cleaning and inspection schedules for the RBCCW heat exchangers (X-18A/B/C) and the emergency diesel generator heat exchangers (X-83A/B, X-53A/B, X-45A/B) will be altered such that those heat exchange~s are cleaned annually prior to each summer. Additionally, if necessary, the RBCCW cleaning frequency may increase based on observed RBCCW outlet temperatures and DP surveillance results.

The first commitment was changed to allow a 6-month interval for the inspection and cleaning of the RBCCW heat exchangers. The second was changed to require that the RBCCW and EOG heat exchangers be cleaned within 4 months of the UHS temperature rising above 75 degrees F.

Engineering developed an Engineer Technical Evaluation (ETE-MP-2016-1136, RBCCW Heat Exchanger Cleaning and Inspection Frequency and Allowable Tube Plugging) that provides the justification for these commitment changes. The ETE is based on review of past performance testing and review of historical maintenance for these heat exchangers. The conclusion from these reviews allowed extending the cleaning and inspection frequency of the RBCCW heat exchangers from three (3) months to six (6) months, allowed the changes to the two commitments described above and allows for a 10% tube plugging limit for the RBCCW heat exchangers.

Serial No.18-226 10 CFR 50.59 Change Report for 2017 and Commitment Change Report for 2017 Attachment 5/ Page 2 of 2 Review, Approval and Control of Leak Injection Type Repairs - RCR-01894 MPS, in RCR-01894, committed to treat all leak injection type repairs as Temporary Configuration Changes (TCC) and require Facility Safety Review Committee (FSRC) (formerly Plant Operations Review Committee (PORC)) review and approval.

The change to the commitment changed the requirement to treat all leak injection type repairs as TCCs and require FSRC review, to using the Fleet Leak repair process for all leak injection type repairs and only require FSRC review for leak injection type repairs on safety related components.

Leak injection type repairs will be performed using a Fleet Process (ER-AA-106). TCCs are no longer required. The Fleet process incorporated the critical reviews that previously were performed by MPSs TCC process. The Fleet process (ER-AA-106) was revised to ensure all leak injection type repair packages are reviewed in accordance with 10 CFR 50.59/72.48 and reviewed and approved by FSRC.

Retrieved from "https://kanterella.com/w/index.php?title=ML18184A434&oldid=2401649"