ML21026A142
| ML21026A142 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/23/2021 |
| From: | Richard Guzman NRC/NRR/DORL/LPL1 |
| To: | Stoddard D Dominion Energy Nuclear Connecticut |
| Guzman R | |
| References | |
| EPID L-2020-LLA-0039 | |
| Download: ML21026A142 (18) | |
Text
February 23, 2021 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO. 2 ISSUANCE OF AMENDMENT NO. 342 RE: REVISION TO TECHNICAL SPECIFICATION TABLE 3.3-11, ACCIDENT MONITORING INSTRUMENTATION (EPID L-2020-LLA-0039)
Dear Mr. Stoddard:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 342 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (Millstone 2), in response to your application dated March 3, 2020, as supplemented by letter dated September 17, 2020.
The amendment revises the Millstone 2 Technical Specification (TS) Table 3.3-11, Accident Monitoring Instrumentation, ACTION 3, to add an alternate method for determining if there is loss of coolant through a power-operated relief valve or pressurizer safety valve flow path when any of the three valve position indications (i.e., Instruments 4, 5, and 6) become inoperable.
The amendment also includes two corrections in TS Table 3.3-11 ACTIONS 4.a and 4.b for Instrument 7, Containment Pressure (Wide Range) and Instrument 9, Containment Water Level (Wide Range).
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336
Enclosures:
- 1. Amendment No. 342 to DPR-65
- 2. Safety Evaluation cc: Listserv
DOMINION ENERGY NUCLEAR CONNECTICUT, INC.
DOCKET NO. 50-336 MILLSTONE POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 342 Renewed License No. DPR-65
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Dominion Energy Nuclear Connecticut, Inc.
(the licensee) dated March 3, 2020, as supplemented by letter dated September 17, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-65 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 342 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of the date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: February 23, 2021 James G.
Danna Digitally signed by James G. Danna Date: 2021.02.23 11:14:12 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 342 MILLSTONE POWER STATION, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Insert 3
3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 3/4 3-33 3/4 3-33 3/4 3-34 3/4 3-34 Connecticut, in accordance with the procedures and limitations set forth in this renewed operating license; (2)
Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; (5)
Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at steady-state reactor core power levels not in excess of 2700 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 342 are hereby incorporated in the renewed license.
The licensee shall operate the facility in accordance with the Technical Specifications.
Renewed License No. DPR-65 Amendment No. 342
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 342 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION, UNIT NO. 2 DOCKET NO. 50-336
1.0 INTRODUCTION
By letter dated March 3, 2020, as supplemented by letter dated September 17, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20065K976 and ML20261H598, respectively), Dominion Energy Nuclear Connecticut, Inc. (DENC, the licensee), submitted a license amendment request (LAR) for the Millstone Power Station, Unit No. 2 (MPS2 or Millstone 2). The LAR proposed to revise Millstone 2 Technical Specification (TS) Table 3.3-11, Accident Monitoring Instrumentation, ACTION 3, to add an alternate method for determining if there is loss of coolant through a power-operated relief valve (PORV) or pressurizer safety valve (PSV) flow path when any of the three valve position indications (i.e., Instruments 4, 5, and 6) become inoperable. The LAR also proposes two grammatical corrections in TS Table 3.3-11 ACTIONS 4.a and 4.b for Instrument 7, Containment Pressure (Wide Range) and Instrument 9, Containment Water Level (Wide Range).
The supplemental letter dated September 17, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on April 21, 2020 (85 FR 22185).
2.0 REGULATORY EVALUATION
2.1
System Description
The proposed changes to the Millstone 2 TS are associated with the design and operation of the accident monitoring instrumentation, specifically for valve position indication for the PORVs, PORV block valves, and the PSVs.
The Millstone 2 reactor coolant system (RCS) is protected against overpressurization by control and protective circuits such as pressurizer pressure high reactor trip, two PORVs (RC-402 and RC-404), and two PSVs (RC-200 and RC-201) connected to the top of the pressurizer.
The two PORVs relieve sufficient pressure to avoid opening of the PSVs. Each flow path has a PORV and a PORV block (isolation) valve (RC-403 and RC-405). The PORV block valve in each train is normally open. The steam discharged by the relief valves is piped to the quench tank where it is condensed. Valve position indication for the PORVs is provided at the main control panel in the control room. In accordance with Section Ill of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, the RCS is protected from overpressure by two spring-loaded PSVs. The discharge from the PSVs is also piped to the quench tank.
2.2 Description of Proposed Changes 2.2.1 Changes to TS Table 3.3-11, ACTION 3 The licensee proposes to revise Millstone 2 TS Table 3.3-11, Accident Monitoring Instrumentation, ACTION 3, as shown in the LAR and supplement. (Note: Proposed added text is italicized)
With any individual valve position indicator inoperable, obtain quench tank temperature, level and pressure information, and monitor discharge pipe temperature once per shift to determine valve position. With the number of OPERABLE accident monitoring instrumentation channels less than the required Minimum Channels OPERABLE in Table 3.3-11 and one or more of the above mentioned quench tank parameters or discharge pipe temperatures unavailable, either restore the inoperable accident monitoring instrumentation channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if repairs are feasible without shutting down or:
- 1) Initiate an alternate method for determining if there is a loss of coolant through an inadvertently open valve; and
- 2) Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the actions taken, the cause of the malfunction, the plans for restoring the accident monitoring instrumentation channel to OPERABLE status; and
- 3) Restore the accident monitoring instrumentation channel to OPERABLE status by the end of the next scheduled refueling outage.
This ACTION is not required if the PORV block valve is closed with power removed in accordance with Specification 3.4.3.b or 3.4.3.c.
2.2.2 Changes to TS Table 3.3-11, ACTION 4 The licensee also proposes a minor grammatical correction in two locations within TS Table 3.3-11 ACTIONS 4.a and 4.b. This correction changes the wording in affect to in effect for both locations.
2.3 Regulatory Requirements and Guidance 2.3.1 Regulatory Requirements The NRC staff identified the following requirements as applicable in its review of the LAR:
Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36, Technical specifications, specifies the Commissions regulatory requirements related to the content of TSs. Specifically, 10 CFR 50.36(a)(1) requires, in part, that [e]ach applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. The regulation at 10 CFR 50.36(c)(2)(ii) requires, in part, that TS limiting conditions for operation (LCOs) be established for items meeting one or more of the four specific criteria. LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When LCOs are not met, the licensee shall shut down the reactor or follow the remedial action permitted by the TSs.
Millstone 2 was designed and licensed in accordance with Atomic Energy Commission (AEC) proposed General Design Criteria (GDC) published in July 1967. However, the LAR states:
Although MPS2 was designed and licensed to the GDC, as issued on July 11, 1968, DENC has attempted to comply with the intent of the newer GDC to the extent possible, recognizing previous design commitments.
The NRC staff identified 10 CFR Part 50 Appendix A, GDC 13, Instrumentation and control, as being applicable to the proposed amendment. GDC 13 requires that instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
2.3.2 Regulatory Guidance Documents The NRC staff also considered the following guidance documents in its review of the LAR:
Regulatory Guide (RG) 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following An Accident, Revision 2, December 1980 (ADAMS Accession No. ML060750525), describes a method acceptable to the NRC for complying with regulations to provide instrumentation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plant.
NUREG-0578, TMl-2 [Three Mile Island Unit 2] Lessons Learned Task Force Status Report and Short-Term Recommendations, July 1979 (ADAMS Accession No. ML090060030), contains a set of short-term action recommended by the TMI-2 Lessons Learned Task Force following the event at TMI-2 involving a loss of feedwater transient and a partially mitigated loss-of-coolant accident (LOCA) with significant core damage.
NUREG-0737, Clarification of TMI Action Plan Requirements, November 1980 (ADAMS Accession No. ML051400209), incorporates in one document, all TMI-related items approved for implementation by the Commission.
3.0 TECHNICAL EVALUATION
3.1 Technical Specification Changes The NRC staffs review focused on determining whether the TSs, as amended, would provide reasonable assurance of adequate protection of public health and safety in conditions where accident monitoring instrumentation is relied upon. Specifically, the proposed change revises Millstone 2 TS Table 3.3-11, Accident Monitoring Instrumentation, ACTION 3 for valve position indication for the PORV, PORV block valve, and PSV. The licensee also proposes a minor grammatical correction in two locations of TS Table 3.3-11 ACTIONS 4.a and 4.b for Instrument 7, Containment Pressure (Wide Range) and Instrument 9, Containment Water Level (Wide Range).
3.1.1 Proposed Changes to TS Table 3.3-11, ACTION 3 The proposed change modifies ACTION 3 to include the proposed actions, as shown above in Section 2.2 of this safety evaluation. In the LAR, the licensee provides the following information:
Valve position for the pressurizer PORVs, PORV block valves, and PSVs is one of several indications that operators have available to determine whether the valves are open or closed. Other plant parameters are available that provide direct indication of a lifted or closed valve, or can be used indirectly to assist in diagnosing if a loss of RCS inventory is occurring and where it is occurring.
Operators are trained to diagnose these parameters as part of determining if an RCS leak is occurring and these diagnostic skills can also be used to determine if the loss of inventory is from a PORV or PSV.
Examples of direct parameters for diagnosing an inadvertently opened or stuck open PORV or PSV include PORV discharge pipe (tailpipe) temperature, PSV discharge pipe (tailpipe) temperature, quench tank level, quench tank pressure, and quench tank temperature. If a PORV or PSV is opened, the following alarms would be expected in the control room within seconds:
High PORV discharge pipe temperature (setpoint at 165 °F [degrees Fahrenheit])
High PSV discharge pipe temperature (setpoint at 155 °F)
High quench tank level (setpoint at 55%)
High quench tank pressure (setpoint at 10 psig [pounds per square inch])
High quench tank temperature (setpoint at 120 °F)
Since multiple, independent instruments are available to the operator to provide indication of an inadvertently opened PORV or PSV, loss of one or more of these instruments would not preclude the operator from correctly diagnosing the condition.
An example of using indirect parameters to diagnose that the RCS is leaking through an inadvertently opened PORV or PSV might occur if the operators observed decreasing RCS pressure combined with a loss of RCS subcooling, which would indicate that saturated conditions are being approached. For this example, operators would check RCS pressure and pressurizer level, along with charging and letdown mismatch, to determine if the status of these parameters are indicative of a loss of inventory. If inventory is being lost, the operators would then check containment parameters and radiation monitors to determine if RCS inventory is being discharged to containment. Then the operators would check for other possible RCS inventory loss paths, such as auxiliary building sumps, radiation monitors, and tank levels for signs of inter-system leakage outside containment. If no signs of inter-system leakage exist, the leakage path may be a lifting PORV or PSV, which can be confirmed by available quench tank parameters or discharge pipe temperatures. For example, a loss of RCS inventory combined with low RCS pressure, decreasing reactor vessel level and high pressurizer level is indicative that the loss is through a PORV or PSV (bubble has shifted from pressurizer to core). Conversely, closure of a PORV or PSV would be indicated by recovery of level in both the reactor vessel and pressurizer and ultimately the recovery of RCS pressure to shut-off head of the high pressure safety injection pumps, in addition to downward trends of discharge pipe temperatures.
If a PORV is inadvertently opened, the operator would take action to isolate the PORV and PORV block valve from the control room. If successful closure of a PORV or PORV block valve cannot be confirmed because the position indication was already inoperable, isolation can be recognized by observation of an indicated reduction in available discharge pipe temperature, quench tank parameters, or other expected changes in plant parameters. If a PSV is inadvertently opened or a PORV flow path is inadvertently opened and cannot be isolated, EOP [Emergency Operating Procedure] 2532, Loss of Coolant Accident, would be entered.
Staff Evaluation of Changes to TS Table 3.3-11, ACTION 3 The NRC staff reviewed the LAR and the supplemental letter dated September 17, 2020 (hereafter referred as the supplement), based on the Millstone 2 Updated Final Safety Analysis Report (UFSAR). If the PORV, PORV block valve or SRV position indication is inoperable, TS Table 3.3-11, ACTION 3 requires the operators to obtain quench tank temperature, level and pressure information, and monitor discharge pipe temperature once per shift to determine valve position. Based on the current wording of ACTION 3, should any of these four single-channel instruments fail, the licensee would be required to enter TS 3.0.3 and initiate actions to place the unit in hot shutdown. The licensee states that this is overly restrictive and that operators can use the available direct and indirect parameters to identify if there is loss of coolant through a PORV or PSV, so that restorative or mitigative actions can be taken. In the licensees supplement, the licensee states:
[r]eceipt of alarms from any available direct parameter channels and the corresponding procedural actions can reliably diagnose gross leakage through an inadvertently open PORV or PSV flow path.
Through observation of other normally monitored reactor coolant system (RCS) parameters in the control room, the position of a PORV or PSV can also be reasonably ascertained without any direct parameters.
MPS2 operators are trained on these monitoring strategies as part of the Initial License Training (ILT) and Licensed Operator Requalification (LOR).
The normal pressurizer operating pressure and temperature are 2,250 pounds per square inch absolute (psia) and 653 °F (per Table 4.3-6 of the UFSAR). Both the PORVs and PSVs discharge to the quench tank which is in containment at much lower pressure and temperature.
The quench tank normal operating pressure and temperature are 3 psig and 120 °F, respectively (per Table 4.3-7 of the UFSAR). If a PORV or SRV opens, the temperature and pressure in both the discharge piping and quench tank would increase.
The NRC staff confirmed that the Millstone 2 plant configuration does not require all four of the listed direct indications to determine valve position. If the direct PORV or SRV position indicator is inoperable, any one of the following instruments individually would allow the operators to determine that a PORV or an SRV is open: quench tank pressure, quench tank temperature, quench tank level, and discharge pipe temperatures. Given that there are multiple independent instruments available, the NRC staff finds that it is not necessary for all the above instruments to be available for the operators to determine PORV or SRV position.
The proposed changes provide additional operational flexibility for the operators to monitor valve position using a combination of methods, by using the available direct parameters and indirect parameters as described in the LAR and the supplement, when the primary valve indication becomes inoperable. The four direct parameters each have existing alarms to provide direct and expeditious indication in the control room. The indirect parameters are normally monitored parameters in the control room and can reasonably provide indication of valve position or flow in the PSV or PORV discharge lines. Also, the existing operator actions and training already address the use of these alternative methods. Based on the NRC staffs review, the staff finds that the proposed changes are consistent with the current design basis and the direct and indirect parameters provide a reliable alternative method to monitor valve position. Therefore, the NRC staff finds these changes to TS Table 3.3-11, ACTION 3 to be acceptable.
Additionally, the proposed change to TS Table 3.3-11, ACTION 3 would allow 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to restore the inoperable PORV, PORV block valve, or PSV position indication when one or more of the quench tank instruments (level, pressure, or temperature) or discharge pipe temperature instruments are unavailable. If the instrumentation cannot be restored within the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the licensee would have to submit a special report (pursuant to TS 6.9.2) which must be submitted within 10 days after the 48-hour completion time has expired. The purpose of the special report is to notify the NRC of the inoperability and describe the alternate methods in effect and outline the cause of the malfunction and plans for restoring the affected position indicators to operable status. The NRC staff finds that the 48-hour completion time is acceptable in that it is consistent with completion times that apply to other accident monitoring instrumentation when the number of operable channels is less than the minimum channels operable requirements. This includes containment pressure (wide range), containment water level (wide range), core exit thermocouples, and reactor vessel coolant level. The NRC staff finds that submittal of a special report is appropriate if the position indication instrumentation is not made operable within the 48-hour completion time and is consistent with similar requirements for other inoperable accident monitoring instrumentation such as containment pressure (wide range), containment water level (wide range), and reactor vessel coolant level.
3.1.2 Proposed Changes to TS Table 3.3-11, ACTION 4 The licensee also proposes minor grammatical corrections for two locations within TS Table 3.3-11 ACTIONS 4.a and 4.b for Containment Pressure (Wide Range) and Containment Water Level (Wide Range). Specifically, the licensee proposes the correction change to the wording in affect to in effect.
Staff Evaluation of Changes to TS Table 3.3-11, ACTION 4 The NRC staff confirms that the proposed changes to ACTION 4 are grammatical changes and are technically correct. Therefore, the NRC staff finds these changes to TS Table 3.3-11, ACTIONS 4.a and 4.b to be acceptable.
3.2 Technical Specifications, 10 CFR 50.36 The NRC staff reviewed the LAR and the supplement against the requirements of 10 CFR 50.36(c)(2)(ii). Based on its review, the NRC staff finds that the proposed changes provide TSs with the necessary LCOs required for safe operation of the facility under the criteria specified by 10 CFR 50.36(c)(2)(ii). Additionally, as described above, the NRC staff reviewed the LAR and the supplement based on the UFSAR, and determined the proposed changes are consistent with the current design basis. Therefore, the NRC staff concludes that the proposed changes meet the requirements of 10 CFR 50.36.
3.3 Appendix A of 10 CFR Part 50, GDC 13 Instrumentation and control The NRC staff reviewed the LAR and the supplement against GDC 13. As mentioned above, Millstone 2 was designed and licensed in accordance with the AEC-proposed GDCs published in July 1967. However, DENC has attempted to comply with the intent of the newer GDCs, published in February 1971, to the extent possible. Based on NRC staffs review, the staff finds that the proposed changes provide the instrumentation and controls to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Additionally, the proposed changes provide appropriate controls to maintain these variables and systems within prescribed operating ranges. The NRC staff also reviewed the LAR and the supplement based on the UFSAR, and determined the proposed changes are consistent with the current design basis. Therefore, the NRC staff concludes that the proposed changes meet the requirements of 10 CFR Part 50 Appendix A, GDC 13.
3.4 Regulatory Guide 1.97, Revision 2 The NRC staff reviewed the LAR and the supplement using guidance from RG 1.97, Revision 2.
The current Millstone 2 licensing basis demonstrates conformance with RG 1.97 Revision 2.
Based on the NRC staffs review, the staff finds that the proposed changes provide indications of plant variables that are required by the control room operating personnel during accident situations to provide information to the operators that will enable them to determine the potential for occurrence of a gross breach of the barriers to radioactivity release (i.e., fuel cladding, RCS pressure boundary, and containment) and to determine if a gross breach of a barrier has occurred. The NRC staff also reviewed the LAR and the supplement based on the UFSAR, and determined the proposed changes are consistent with the current design basis. Therefore, the NRC staff concludes that the proposed changes conform with the applicable portions of RG 1.97.
3.5 NUREG-0737, Clarification of TMI Action Plan Requirements As stated by the licensee in the LAR, the accident monitoring instrumentation contained in TS Table 3.3-11 was added to the Millstone 2 TSs to meet the requirements of NUREG-0578, TMl-2 Lessons Learned Task Force Status Report and Short-Term Recommendations, and NUREG-0737, Clarification of TMI Action Plan Requirements. As noted in NUREG-0737, Position II.D.3 (Direct Indication of Relief and Safety Valve Position), the purpose of RCS relief and safety valve position indication is to provide the control room with unambiguous indication of valve position (i.e., open or closed) so that appropriate operator actions can be taken.
The NRC staff reviewed the LAR using guidance from NUREG-0737. Specifically, the NRC staff reviewed the power supply information provided in supplement for the valve position indication, annunciators and direct parameters. The supplement states that, the loss of any single power circuit would not cause the loss of valve position indication, annunciators and direct parameters and the plant configuration allows for the use of more than one alternating current bus to power the direct parameters. Additionally, an uninterruptible power supply was installed in 2010 for the primary panel, VR11, which powers the direct parameters and improved overall reliability. Therefore, the proposed changes conform with Position II.D.3, Direct Indication of Relief And Safety Valve Position, to provide RCS relief and safety valves with a position indication in the control room derived from a reliable valve-position detection device or a reliable indication of flow in the discharge pipe. The NRC staff also reviewed the LAR and the supplement based on the UFSAR, and determined that the proposed changes are consistent with the current design basis. As discussed in Section 3.1.1 of this safety evaluation, operators are already being trained regarding monitoring strategies to use available direct and indirect parameters as part of the Initial License Training and Licensed Operator Requalification programs in the event that primary valve position indication is not available. Therefore, the NRC staff concludes that the proposed changes comply with the applicable portions of NUREG-0737.
NRC Staff Conclusion The licensee proposed to modify TS Table 3.3-11, Accident Monitoring Instrumentation, ACTION 3 to address unnecessary restrictions for monitoring valve position when either the PORV, PORV block valve or SRV position monitoring indications become inoperable. The NRC staff reviewed the proposed change and found that it is not necessary for all currently specified instrumentation (PORV discharge pipe temperature, SRV discharge pipe temperature, quench tank level, quench tank pressure, and quench tank temperature) to be available for operators to determine if either a PORV or SRV is open as any one of these instruments would be sufficient.
Based on the NRC staffs review, the staff determined that the proposed changes are consistent with the current design basis and the direct and indirect parameters provide a reliable alternative method to monitor valve position. Therefore, the NRC staff finds these changes to TS Table 3.3-11, ACTION 3 to be acceptable. The NRC staff also finds that the two editorial changes to ACTIONS 4.a and 4.b are grammatical corrections in nature and are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Connecticut State official was notified of the proposed issuance of the amendment. The State official provided comments via e-mail on January 21, 2021, which were considered in the NRC staffs review. The NRC staffs response to the State officials comments, and the State officials reply, can be viewed in ADAMS at Accession No. ML21032A109.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 or changes SRs.
The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding, which was published in the Federal Register on April 21, 2020 (85 FR 22185), that the amendment involves no significant hazards consideration, and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: C. Cheung D. Rahn R. Beaton Date: February 23, 2021
ML21026A142 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DEX/EICB/BC NAME RGuzman JBurkhardt MWaters DATE 01/26/2021 01/29/2021 11/12/2020 OFFICE NRR/DSS/SNSB/BC NRR/DSS/STSB/BC OGC - NLO w/revisions NAME SKrepel VCusumano STurk DATE 05/19/2020 02/01/2021 02/18/2021 OFFICE NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME JDanna RGuzman DATE 02/22/2021 02/23/2021