ML23089A121
| ML23089A121 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/30/2023 |
| From: | Grady C Virginia Electric & Power Co (VEPCO) |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 23-102 | |
| Download: ML23089A121 (1) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 MAR 3 0 2023 10 CFR 50.59(d)(2)
U.S. Nuclear Regulatory Commission Serial No.23-102 Attn: Document Control Desk SPS/MMT RO Washington, DC 20555-0001 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 ANNUAL CHANGES, TESTS, AND EXPERIMENTS REPORT REGULATORY COMMITMENT EVALUATION REPORT Virginia Electric and Power Company hereby submits the annual report of Changes, Tests, and Experiments pursuant to 10 CFR 50.59(d)(2) implemented at Surry Power Station. The Attachment provides the descriptions and summaries of the Regulatory Evaluations and the Regulatory Commitment Change Evaluations completed in 2022.
Should you have any questions regarding this report, please contact Michael M. True, Jr. at (757) 365-2446.
Very truly yours, o~~
Cathy Grady Director Nuclear Safety & Licensing Surry Power Station Attachment Commitments made in this letter: None cc: United States Nuclear Regulatory Commission, Region II Marquis One Tower, Suite 1200 245 Peachtree Center Avenue, NE Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector Surry Power Station
Serial No.23-102 10 CFR 50.59 Annual Report Page 1 of 10 Attachment Surry Units 1 & 2 10 CFR 50.59 Changes, Tests, and Experiments, and Regulatory Commitment Evaluations for 2022 SU-22-00107/Rev. 0 Regulatory Evaluation 04/07/2022
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Description:==
The proposed activity removes the pressurizer cubicle roof blocks (Elevation 67'-6") from the Unit 1 containment structure. The roof blocks were designed and installed to act as a missile barrier to protect the containment liner from internally generated missiles originating from the pressurizer cubicle. The Design Engineering Change Package (DECP), SU 00107, evaluates the notable high-energy component missile sources within the pressurizer cubicle and has determined that all of the sources are not credible to generate missiles; therefore, the pressurizer cubicle roof blocks are not required to function as a missile barrier.
With the pressurizer cubicle roof blocks removed, the pressurizer cubicle temperature is expected to decrease due to natural circulation with the containment air. The maximum temperature change the pressurizer cubicle could theoretically experience would be the delta temperature between the pressurizer cubicle and the containment temperature (~30° F). However, based on the heat input of the pressurizer within the cubicle, a 30° F change is not likely. The DECP and ETE-SU-2021-0004 "Pressurizer Safety Valve Lift Pressure Due to Temperature Change" has evaluated the expected temperature change and any potential impact on the PORV's, PSVs, and all other components within the cubicle and has determined that there is negligible impact and the change will not result in any of the components operating outside of their design limits.
Summary:
The proposed activity removes the pressurizer cubicle roof blocks (Elevation 67'-6") in the Unit 1 Containment structure. The roof blocks were designed and installed to act as a missile barrier to protect the containment liner from internally generated missiles originating from the Pressurizer Cubicle. The DECP, SU-22-00107, evaluates all of the potential missile sources within the pressurizer cubicle and has determined that the industry experience concerning bolted fasteners demonstrates that failure modes do not result in missile generation. Additionally, the station safeguards ensure that degraded Reactor Coolant system pressure boundary components are identified quickly and properly addresses long before progressive failure could occur. As a result, components within the pressurizer cubicle are not creditable missile sources; therefore, the Pressurizer cubicle roof blocks are not required to function as a missile barrier.
The activity of permanently removing the pressurizer roof blocks provides benefits to the station.
Serial No.23-102 10 CFR 50.59 Annual Report Page 2 of 10 Heavy load lift reduction: The removal of the pressurizer missile shield roof reduces the risk of heavy load lifting and movement. The pressurizer missile shield roof blocks are removed each refueling outage and during forced outages. The risk associated with this lifting and movement will be eliminated.
Reduction in ambient temperature: The removal of the pressurizer missile shield roof will reduce the ambient temperature in the pressurizer cubicle. This will reduce component thermal aging effects.
Facilitates inspection: The removal of the pressurizer missile shield roof will aid inspection of component located on top of the pressurizer. Bolted connections can be inspected at hot shutdown conditions with the RCS at operating pressure. Should a leak be present, this should help aid in its identification.
Outage dose savings: The removal of the pressurizer missile shield roof will reduce man-rem exposure.
A 10CFR50.59 Screen was performed which determined that, the permanent removal of the pressurizer cubicle roof blocks will adversely affect the design function of the containment liner. The containment liner is a "Defense-in-Depth" barrier which is protected extensively through the usage of engineered features such as the usage of missile barriers to prevent internally generated missiles from striking the containment liner. Although the DECP has determined that the high-energy components within the pressurizer cubicle are not credible at generating missiles, the removal of the missile barrier safety feature is adverse and requires a 10CFR50.59 Evaluation. Both the creditability for missile generation and the removal of the missile barrier are considered linked changes when evaluated against the effects on the containment liner's design functions. The design functions are described below:
UFSAR Section 15.5.1.11 describes pipe ruptures in pressure piping and equipment producing missiles that are shielded from striking the containment liner using reinforced concrete walls and floors.
UFSAR Section 15.5.1.8 describes the steel liner of the containment structure being designed to function as a gastight membrane and being protected from potential interior missiles by interior concrete shield walls. The steel liner is designed to withstand the effects of all temperature, earthquake, and pressure loads, including the effect of the sub-atmospheric operating pressure.
The creditability for missile generation and the removal of the pressurizer cubicle roof blocks missile barrier were evaluated against the effects on the containment liner's design function using the eight (8) 10 CFR 50.59 questions. The evaluation concluded that the activity does not result in more than a minimal increase in the frequency and consequences of an accident, and the likelihood of occurrence and consequences of a malfunction of an SSC important to safety, as previously evaluated in the SAR (i.e., Part II, Questions,1-4); does not create the possibility for an accident or malfunction to an SSC important to safety of a different type than any previously evaluated in the SAR (i.e., Part II, Questions 5-6); result
Serial No.23-102 10 CFR 50.59 Annual Report Page 3 of 10 in a design basis limit for a fission product barrier as described in the SAR being exceeded or altered (i.e., Part II, Question 7) nor result in departure from a method of evaluation described in the SAR used in establishing the design bases or in the safety analyses (i.e.,
Part 11, Question 8). The activity can be implemented without prior NRC approval.
Serial No.23-102 10 CFR 50.59 Annual Report Page 4 of 10 SU-18-00170, Rev. 0 Regulatory Evaluation 08/24/2022
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Description:==
Engineering Change (EC) SU-18-00170 will replace the existing SPS Unit 2 Westinghouse Mark I AEH TCS with a Westinghouse TCS. The replacement TCS is provided by Westinghouse and is based on the Ovation state-of- the-art Distributed Control System (DCS) platform manufactured by Emerson Process Management and customized for use as a TCS. The TCS provides a speed control mode, a flow control loop (referred to as open loop), a high pressure turbine 1st stage impulse pressure control loop, and a megawatt control loop. Frequency control, which provides over-grid-frequency correction by shedding load, is placed in service automatically after synchronization (i.e., when the turbine-generator is connected to the grid) and has an 8 rpm dead band. All of the automatic control loops can be taken out of, and placed into, service and are designed to fail to the open loop control mode. The Westinghouse TCS controls the turbine using a Distributed Control System (DCS), the existing servo-mechanism, the existing hydraulic valve actuators and linear variable differential transformers (LVDT). The existing Governor Valve Management, Full Arc mode of control, will be maintained with the same governor valve (flow vs. lift) curves. The four (4) valves operate in unison. Provisions for routine turbine valve testing will be maintained. Highly reliable components, fault detection techniques and trip avoidance measures are incorporated into the TCS design to minimize susceptibility to a turbine trip from a failed component and minimize exposure to the single point vulnerabilities (SPVs) remaining in the upgraded control system.
Summary:
The existing Surry Power Station Unit 2 Westinghouse Mark I Analog Electrohydraulic Control (AEHC) Main Turbine Control System (TCS) is obsolete and is replaced by a Westinghouse Digital TCS. The replacement TCS maintains the fundamental design function of the turbine control system which is to provide the capability to control the turbine from turning gear to rated speed and rated load, with the capability to monitor, detect and control undesirable operating conditions.
In accordance with NEI 96-07 Appendix D, of factors involving (1) the design attributes of the modified SSC, (2) the quality of the design processes and (3) the operating experience of the software and hardware used, the likelihood of a digital-related failure in the proposed TCS upgrade is judged to be sufficiently low. No prior NRG review and approval is necessary.
Serial No.23-102 10 CFR 50. 59 Annual Report Page 5 of 10 ETE-NAF-2022-0029/Rev. 0 Regulatory Evaluation 10/06/2022
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Description:==
The activity being reviewed is the implementation of the Westinghouse PADS code-based method into the Surry Unit 1 licensing basis, to account for thermal conductivity degradation (TCD) at Surry Power Station Unit 1 as documented in ETE-NAF-2022-0029.
Summary:
The primary purpose of the changes being implemented as part of this 10 CFR 50.59 Evaluation is to capture the impact of using PADS to generate fuel-related inputs on affected analyses for Surry Power Station (SPS) Unit 1. Implementation of PADS at Surry Power Station Unit 2 is planned to occur in support of their next refueling outage scheduled for the spring of 2023. This activity directly addresses the phenomena of thermal conductivity degradation (TCD). Previously, the NRC released Information Notice 2009-23, which requested licensees to review the information notice and determine if TCD had been adequately addressed at each operating facility. To address this concern, Westinghouse developed an improved fuel thermal model and fuel performance evaluation code and methodology documented in WCAP-17642-P-A, Revision 1, which has been approved by the NRC. The methodology and associated code, known as PADS, incorporates several improved fuel thermal performance models, including consideration of TCD with increasing burnup, compared to the current PAD 4.0 code and method described in the Surry Power Station Updated Final Safety Analysis Report (UFSAR).
This 10 CFR 50.59 activity implements the PADS code and method into Chapter 3 of the SPS UFSAR and implements an updated Chapter 14 safety analysis to account for the use of PADS inputs for SPS Unit 1. An Evaluation is required because methods of analysis are changed from those currently described in the SPS UFSAR and a Chapter 14 safety analysis was re-run to address the use of PADS inputs and to ensure the respective acceptance criteria continue to be met such that their individual design functions remain unaffected.
Because of this methodology change and potentially affected SAR-described design functions, which are considered adverse changes under 10 CFR 50.59, an Evaluation is required to assess each of the eight Evaluation criteria. The method change is solely evaluated under Criterion 8 consistent with NEI 96-07, Revision 1, Section 4.3.8, while the reanalyzed safety analysis is evaluated under Criteria 1 through 7.
The reanalyzed safety analysis continues to meet the applicable acceptance criteria ensuring the design functions remain unaffected. Because the 50.59 treatment of the reanalysis under Criteria 1 - 7 only involved input changes and all acceptance criteria continue to be met, there was no impact on frequency of occurrence of accidents or SSC malfunctions, no increase in radiological consequences of the accidents or via SSC malfunctions, and no potential for different accidents directly or via SSC malfunctions. The fuel melt limit, which is considered a DBLFPB, is altered as an acceptance criterion for safety analyses for which fuel melt is evaluated; however, the fuel melt limit is included in the approved WCAP-17642-P-A, Revision 1 topical, which is implemented en toto under
Serial No.23-102 10 CFR 50.59 Annual Report Page 6 of 10 Question 8 of the Evaluation. Finally, as the method of analysis for this event is unchanged, Criterion 8 of the Evaluation is not applicable.
As previously indicated, WCAP-17642-P-A, Revision 1 (Westinghouse's PAD5 model and code) is being implemented as the new fuel performance evaluation methodology at SPS Unit 1 which includes incorporation into the safety analyses. PAD5 is NRG-approved and offers improvements over the current PAD 4.0 fuel performance code/method, namely the consideration of the effects of TCD with increasing burn up. The change in methodology is not considered a departure from a method of evaluation because PAD5 is NRG-approved, is being used for its intended application, is used within the limitations outlined in the WCAP-17642-P-A, Revision 1 SER, and it offers technical improvements over PAD 4.0. Only Criterion 8 applies to this change because it only involves a change in the method of evaluation.
Implementation of PAD5, including the reanalyzed safety analysis and use of the new fuel melt limit, may be implemented without NRG approval under the provisions of 10 CFR 50.59.
Serial No.23-102 10 CFR 50.59 Annual Report Page 7 of 10 SU-22-00151, Rev. 1 Regulatory Evaluation 10/28/2022
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Description:==
Design Change SU-22-00151 is installing an interposing relay in series with an "a" contact and a "b" contact associated with the upstream feeder breakers (15D1 and 15F1) for the Unit 1 Emergency Bus supply breakers (15H8 and 15J8) breaker closing circuits. An "a" contact from the interposing relay will then be wired in place of the relocated "a" contact from the upstream feeder breaker (15D1 and 15F1) for breakers 15H8 and 15J8. A "b" contact from the interposing relay will be wired in place of the relocated "b" contact from the upstream feeder breaker (15D1 and 15F1) for breakers 15H8 and 15J8.
Summary:
SU-22-00151 is installing an interposing relay in the 15J8 and 15H8 Emergency Bus Normal Supply Breakers to address appendix R concerns with wiring going to the 15D1 and 15F1 Transfer Bus Supply Breakers. The interposing relay will energize and deenergize in conjunction with the position of 15D1 and 15F1 to provide the interlock between 15J8/15D1 and 15H8/15F1 breakers.
An evaluation is required due to the potential adverse effect of having an additional electrical component within the breaker operating scheme for 15J8 and 15H8 Emergency Bus Normal Supply Breakers.
Does the activity result in more than a minimal increase in the frequency of occurrence of an accident previous evaluated in the SAR? - No. The modification installs a relay in the Emergency Bus Normal Supply Breakers. There is no impact to any accident frequency.
Does the activity 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 SAR? - No. The interposing relay energizes and deenergizes in conjunction with the position of 15D1 /15F1 to provide the interlock between 15D1/15J8 and 15F1/15H8. In the event of a failure, the Emergency Bus Normal Supply Breaker would open, resulting in the EOG supplying the Emergency Bus. This condition is bounded in the accident analysis.
Does the activity result in more than a minimal increase in the consequences of an accident previously evaluated in the SAR? - No. A failure of the fuses or relay could result in the Emergency Bus Normal Supply Breaker opening and the EOG supplying the Emergency Bus. This condition is bounded by the assumption in the accident analysis of a complete loss of offsite power.
Does the activity result in more than a minimal increase in the consequences of a malfunction of a SSC important to safety previously evaluated in the SAR. No. In the event of a relay failure, the Emergency Bus Supply Breaker would open and the EOG would supply the Emergency Bus. As this condition is bounded by the accident analysis, there is no increase in consequences of a malfunction.
Serial No.23-102 10 CFR 50.59 Annual Report Page 8 of 10 Does the activity create a possibility for an accident of a different type than any previously evaluated in the SAR? No. The installation of a relay in the Emergency Bus Normal Supply Breaker does not introduce any new accident type.
Does the activity create a possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the SAR? No. A failure of the relay or contacts to operate has two possible results. One is the Emergency Bus Supply Breaker opening and the EOG powering the bus. This is bounded by the assumptions of a loss of offsite power in the accident analysis. The second possibility is that the breaker closing interlock between 15D1/15J8 and 15F1 and 15H8 would cease to function. The closing of 15J8 and 15H8 is governed by operating procedures that require synchronizing across and closing the Emergency Bus Supply Breaker prior to manually closing.
Does the activity result in a design basis limit for a fission product barrier as described in the SAR being exceeded or altered? No. In the event of a relay failure, the Emergency Bus Supply Breaker would open and the EOG would supply the Emergency Bus. As this condition is bounded by the accident analysis, a fission product barrier limit is neither altered nor exceeded.
Does the activity result in departure from a method of evaluation described in the SAR used in establishing the design bases or in the safety analysis? No. There is no change to the method of evaluation.
Serial No.23-102 10 CFR 50.59 Annual Report Page 9 of 10 Commitment Change Evaluation 10/19/2022 Original Commitment Summary:
Commitment Change Evaluation Associated with SN 89-390A, dated July 14, 1989 Corrective Actions That Will Be Taken To Avoid Further Violations To ensure that design inputs used in calculations, etc. undergo a proper verification, the governing NDCM procedure for calculations will be revised to include an Attachment that will be filled out and included in every new or revised calculation. This attachment will ask the preparer a number of questions to ensure he has given proper consideration to the source and the validity of his design inputs. The preparer will be required to issue the calculation as "Preliminary - Requires Confirmation" until the questions can be answered satisfactorily.
Appropriate engineering personnel and A/Es supporting North Anna and Surry will receive training on the problems identified by the SSFI and on the procedural changes which were made to help prevent recurrence of the problems.
Revised Commitment Summary:
The calculation procedure, CM-AA-CLC-301, requires the preparer to document the definition of design inputs and their sources and the calculation review standard, DNES-AA-GN-1001, has the reviewer ensure design inputs were correctly incorporated into the design.
A specific Attachment documenting these steps have occurred is no longer required.
Therefore, the new commitment is: To ensure that design inputs used in calculations undergo a proper verification, the procedure governing the creation of engineering calculations shall require documentation of the definition and source of design inputs and design verification of calculations shall require ensuring correct incorporation of design inputs into the design.
Justification:
Dominion Energy has made substantial improvements to the design control and calculation processes since 1989. The procedural requirements identified immediately above render the committed Attachment redundant to the steps required for generating a calculation.
Removal of the commitment for a specific Attachment reduces unnecessary administrative burden. The intent of the commitment to ensure the validity of design inputs is maintained by these procedural controls without a specific Attachment.
Serial No.23-102 10 CFR 50.59 Annual Report Page 10 of 10 Commitment Change Evaluation 11/28/2022 Original Commitment Summary:
SN 20-220: NRC Safety Evaluation (ML20076A576)
Revising commitment due dates from the NRC 10 CFR 50.54(f) Flood Hazard Assessment Letter and Safety Evaluation (SE)
Revised Commitment Summary:
Change the due dates of the U1/U2 flooding commitments made in Surry's response to the NRC 10 CFR 50.54(f) letter [Fukushima Accident]. Commitments involved numerous flood-related modifications, procedure changes, and training. The original completion dates for the commitments were 12/13/2022 (U1) and 6/6/2023 (U2). The new completion date for both units is 11/1/2023 NRC Response to Extension Request (ADAMS Accession No. ML22342A424)
SPS U1 U2; Flood Mitigation Modification Change in Completion Date; ML23005A240 NRC staff has reviewed the submittal and the conditions contributing to the delay and revised commitment for the associated completion date and finds the extension reasonable and acceptable. This letter acknowledges the revised proposed completion date of November 1, 2023.