ML23353A171
| ML23353A171 | |
| Person / Time | |
|---|---|
| Site: | Callaway  (NPF-030) |
| Issue date: | 01/18/2024 |
| From: | Mahesh Chawla Plant Licensing Branch IV |
| To: | Diya F Ameren Missouri, Union Electric Co |
| Chawla M | |
| References | |
| EPID L-2022-LLA-0176 | |
| Download: ML23353A171 (1) | |
Text
January 18, 2024 Mr. Fadi Diya Senior Vice President and Chief Nuclear Officer Ameren Missouri Callaway Energy Center 8315 County Road 459 Steedman, MO 65077
SUBJECT:
CALLAWAY PLANT, UNIT NO. 1 - ISSUANCE OF AMENDMENT NO. 237 TO CLARIFY SUPPORT SYSTEM REQUIREMENTS FOR THE RESIDUAL HEAT REMOVAL SYSTEM AND CONTROL ROOM AIR CONDITIONING SYSTEM UNDER TECHNICAL SPECIFICATIONS 3.4.8, 3.7.11, AND 3.9.6 (EPID L-2022-LLA-0176)
Dear Mr. Diya:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 237 to Renewed Facility Operating License No. NPF-30 for the Callaway Plant, Unit No. 1 (Callaway). The amendment consists of changes to the technical specifications (TSs) in response to your application dated December 1, 2022, as supplemented by letters dated October 16, 2023, and November 20, 2023.
The amendment revises the Callaway Licensing Basis (i.e., the final safety analysis report and TSs), to allow use of one train of the normal, non-safety-related service water system to solely provide cooling water support for one of two redundant trains of TS-required equipment when both equipment trains are required to be operable during cold shutdown/refueling conditions.
The supported equipment/systems affected by the proposed change are the residual heat removal system and control room air conditioning system, as applicable, during Modes 5 and 6.
The applicable/affected TS limiting conditions for operation are TS 3.4.8, RCS [Reactor Coolant System] Loops - Mode 5, Loops Not Filled; TS 3.7.11, Control Room Air Conditioning System (CRACS); and TS 3.9.6, Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level.
F. Diya 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/
Mahesh L. Chawla, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483
Enclosures:
- 1. Amendment No. 237 to NPF-30
- 2. Safety Evaluation cc: Listserv
UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 237 License No. NPF-30
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Union Electric Company (UE, the licensee),
dated December 1, 2022, as supplemented by letters dated October 16, 2023, and November 20, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions 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. NPF-30 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan*
The Technical Specifications contained in Appendix A, as revised through Amendment No. 237 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This amendment is effective as of its date of issuance, and shall be implemented within 90 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License No. NPF-30 and Technical Specifications Date of Issuance: January 18, 2024 Jennivine K.
Rankin Digitally signed by Jennivine K. Rankin Date: 2024.01.18 10:02:03 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 237 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-30 CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483 Replace the following pages of Renewed Facility Operating License No. NPF-30 and 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.
Renewed Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 3.4-16 3.4-16 3.7-32 3.7-32 3.9-11 3.9-11
Renewed License No. NPF-30 Amendment No. 237 (3)
UE, 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)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
UE, pursuant to the Act and 10 CFR Parts 30, 40 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 license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to 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 UE is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan*
The Technical Specifications contained in Appendix A, as revised through Amendment No. 237 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Environmental Qualification (Section 3.11, SSER #3)**
Deleted per Amendment No. 169.
Amendments 133, 134, & 135 were effective as of April 30, 2000 however these amendments were implemented on April 1, 2000.
The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
CALLAWAY PLANT 3.4-16 RCS Loops - MODE 5, Loops Not Filled 3.4.8 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 RCS Loops - MODE 5, Loops Not Filled LCO 3.4.8 Two residual heat removal (RHR) loops shall be OPERABLE and one RHR loop shall be in operation.
NOTES --------------------------------------------
1.
All RHR pumps may be removed from operation for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided:
a.
The core outlet temperature is maintained at least 10°F below saturation temperature.
b.
No operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SDM of LCO 3.1.1; and c.
No draining operations to further reduce the RCS water volume are permitted.
2.
One RHR loop may be inoperable for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided that the other RHR loop is OPERABLE and in operation.
3.
The Service Water system may serve as the alternate source of cooling water to the Essential Service Water system for support of the second required RHR train, i.e., the RHR train not supported by the emergency diesel generator required per Technical Specification 3.8.2, provided the plant is not in a reduced-inventory, hot-core condition.
APPLICABILITY:
MODE 5 with RCS loops not filled.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One RHR loop inoperable.
A.1 Initiate action to restore RHR loop to OPERABLE status.
Immediately Amendment No. 237 (continued)
CALLAWAY PLANT 3.7-32 Amendment No. 237 CRACS 3.7.11 3.7 PLANT SYSTEMS 3.7.11 Control Room Air Conditioning System (CRACS)
LCO 3.7.11 APPLICABILITY:
MODES 1, 2, 3, 4, 5, and 6, During movement of irradiated fuel assemblies.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One CRACS train inoperable.
A.1 Restore CRACS train to OPERABLE status.
30 days B.
Required Action and associated Completion Time of Condition A not met in MODE 1, 2, 3, or 4.
B.1 Be in MODE 3.
AND 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B.2 Be in MODE 5.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
Two CRACS trains shall be OPERABLE.
NOTE ----------------------------------------------
During MODES 5 and 6, and during movement of irradiated fuel assemblies, the Service Water system may serve as the alternate source of cooling water to the Essential Service Water system for support of the second required CRACS train, i.e., the CRACS train not supported by the emergency diesel generator required per Technical Specification 3.8.2.
CALLAWAY PLANT 3.9-11 Amendment No. 237 RHR and Coolant Circulation - Low Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level LCO 3.9.6 APPLICABILITY:
MODE 6 with the water level < 23 ft above the top of reactor vessel flange.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Less than the required number of RHR loops OPERABLE.
A.1 Initiate action to restore required RHR loops to OPERABLE status.
OR Immediately A.2 Initiate action to establish 23 ft of water above the top of reactor vessel flange.
Immediately (continued)
Two RHR loops shall be OPERABLE, and one RHR loop shall be in operation.
NOTES --------------------------------------------
The Service Water system may serve as the alternate source of cooling water to the Essential Service Water system for support of the second required RHR train, i.e., the RHR train not supported by the emergency diesel generator required per Technical Specification 3.8.2, provided the plant is not in a reduced-inventory, hot-core condition.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 237 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483
1.0 INTRODUCTION
By application dated December 1, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22335A507), as supplemented by letters dated October 16, 2023 (ML23289A214) and November 20, 2023 (ML23324A370), Union Electric Company, doing business as Ameren Missouri (the licensee), submitted a license amendment request (LAR) for Callaway Plant, Unit No. 1 (Callaway).
The proposed amendment would revise the Callaway licensing basis (i.e., the final safety analysis report (FSAR) and technical specifications (TSs)), to allow use of one train of the normal, non-safety-related service water (SW) system to solely provide cooling water support for one of two redundant trains of TS-required equipment when both equipment trains are required to be operable during cold shutdown/refueling conditions. The supported equipment/systems affected by the proposed change are the residual heat removal (RHR) system and control room air conditioning system (CRACS), as applicable, during Modes 5 and 6. The applicable/affected TS limiting conditions for operation (LCOs) are TS 3.4.8, RCS
[Reactor Coolant System] Loops - Mode 5, Loops Not Filled; TS 3.7.11, Control Room Air Conditioning System (CRACS); and TS 3.9.6, Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level.
The U.S. Nuclear Regulatory Commission (NRC or the Commission) staff performed a regulatory audit to gain an understanding of the information needed to support the LAR review and to develop requests for additional information (RAIs), if needed (ML23317A001). During the audit, the staff reviewed documents, responses to audit questions, and held discussions with the licensee and its contractors. As a result of the audit, the licensee supplemented the LAR by letters dated October 16 and November 20, 2023. After reviewing the supplemental information, the staff did not identify the need for any RAIs.
On February 21, 2023, the NRC staff published a proposed no significant hazards consideration (NSHC) determination in the Federal Register (88 FR 10559) for the proposed amendment.
Subsequently, by letter dated October 16, 2023, the licensee provided additional information
that expanded the scope of the amendment request as originally noticed in the Federal Register. Accordingly, the NRC published a second proposed NSHC determination in the Federal Register on November 13, 2023 (88 FR 77616), which superseded the original notice in its entirety. An additional supplemental letter, dated November 20, 2023, provided more information that clarified the application, did not expand the scope of the application as renoticed, and did not change the NRC staffs proposed no significant hazards consideration determination as published in the Federal Register (FR) on November 13, 2023 (88 FR 77616).
1.1 Purpose of Proposed Changes As stated in its letter dated, December 1, 2022, the licensee determined the need for the proposed change to the licensing basis to address differences of interpretation of the plant licensing basis by the licensee and that reflected in NRC Inspection Report 05000483/2022010 (ML22277A822). Before the inspection, the licensee interpreted that the use of the SW system (in lieu of safety-related essential service water (ESW) system to support one (and only one) of the two trains of RHR during Modes 5 and 6) was consistent with the Improved Standard Technical Specifications (STSs) in NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Revision 1 (ML13196A405) and their Bases as adopted by Callaway in the 1999-2000 timeframe. Following the inspection, the licensee considered the continued application of the previous interpretation regarding cooling water support for RHR in Modes 5 and 6 to be a change to the plants licensing basis. The licensees Title 10 of the Code of Federal Regulations (10 CFR) Section 50.59, Changes, tests and experiments, review determined that NRC approval is required based on a conservative determination that the change involves more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the FSAR, per 10 CFR 50.59(c)(2)(ii).
2.0 REGULATORY EVALUATION
2.1
System Description
In Section 2.2 of the enclosure to the LAR, the licensee describes the following systems that are involved in the proposed amendment: RHR system, CRACS, component cooling water (CCW) system, ESW system, and SW system.
RHR:
In Mode 5, with the Reactor Coolant System (RCS) loops not filled, the primary function of the reactor coolant is the removal of decay heat generated in the fuel and the transfer of this heat to the CCW system via the RHR heat exchangers.
The secondary function of the reactor coolant is to act as a carrier for the soluble neutron poison, boric acid. During RHR system operation, heat is removed from the RCS by circulating reactor coolant through the RHR heat exchangers where the heat is transferred to the CCW system. The heat sink for the CCW system is in turn normally provided by the SW system or ESW system, as determined by TS LCO requirements and system availability. The flow provided by one RHR loop is adequate for decay heat removal.
The purpose of the RHR system in Mode 6 is to remove decay heat and the stored thermal energy of the RCS, as required by General Design Criterion (GDC) 34 (10 CFR 50, Appendix A) [General Design Criteria for Nuclear Power
Plants], to provide mixing of borated coolant, and to prevent boron stratification.
Heat is removed from the RCS by circulating reactor coolant through the RHR heat exchangers where the heat is transferred to the CCW system. (The heat sink for the CCW system is in turn normally provided by the SW system or ESW system, as determined by TS LCO requirements and system availability). The coolant is then returned to the RCS via the RCS cold leg(s). Operation of the RHR System for normal cooldown decay heat removal is manually accomplished from the control room. The heat removal rate is adjusted by controlling the flow of reactor coolant through the RHR heat exchanger(s) and the bypass lines. Mixing of the reactor coolant is maintained by this continuous circulation of reactor coolant through the RHR System.
CRACS:
The CRACS consists of two independent and redundant trains that provide cooling of recirculated control room air. The CRACS is a subsystem to the Control Room Emergency Ventilation System, providing air temperature control for the control room. The CRACS is an emergency system, which also operates during normal unit operations. A single train will provide the required temperature control to maintain the control room [less than or equal to] 84°F [degrees Fahrenheit]. The design basis of the CRACS is to maintain the control room temperature for 30 days of continuous occupancy.
CCW:
The CCW system provides a heat sink for the removal of process and operating heat from safety-related components during a Design Basis Accident (DBA) or transient. The CCW system serves as a barrier to the release of radioactive byproducts between potentially radioactive systems and the ESW system, and thus to the environment.
The CCW system is arranged as two independent, full capacity cooling loops, and has isolatable non-safety-related components. The principal safety-related function of the CCW system is the removal of decay heat from the reactor via the RHR system. This may be during a normal or post-accident cooldown and shutdown. The design basis of the CCW system is for one CCW train to remove the heat from components important to mitigating the consequences of a loss of coolant accident (LOCA) or a main steam line break (MSLB). The CCW system is designed to perform its function with a single failure of any active component, assuming a LOOP [loss of offsite power] (except that during shutdown conditions, TS requirements are relaxed on the basis that it is not necessary to assume a single failure concurrent with a LOOP).
ESW:
The ESW system provides a heat sink for the removal of process and operating heat from safety-related components during a DBA or transient. During normal operation, and a normal shutdown, the ESW system also provides this function for various safety-related and non-safety related components and receives coolant flow from the non-safety related SW system.
The ESW system consists of two separate, 100% capacity, safety-related, cooling water trains. The ESW system also provides emergency makeup to the spent fuel pool and CCW system and is the backup water supply to the Auxiliary Feedwater System.
The principal safety-related function of the ESW system is the removal of decay heat from the reactor via the CCW system and removal of containment heat loads via the containment coolers. The ESW system is required to mitigate DBAs and transients that occur either with or without offsite power available. The design basis of the ESW system is for one ESW train, in conjunction with the CCW system and a 100% capacity containment cooling system, to remove accident generated and core decay heat following a design basis LOCA.
SW:
The SW system is a non-safety-related system which provides a source of heat rejection for plant auxiliaries which require cooling during normal plant operation and normal plant shutdown. The system also supplies cooling water to the safety-related ESW system during normal operation. The SW system provides pumped circulation of approximately 38,000 gpm [gallons per minute] of cooling water (using any 2 of the 3 system pumps) from the cooling tower basin through various main plant auxiliary heat exchange equipment and returns it to the circulating water system via a return line inside the plant.
The SW system provides sufficient cooling water for the heat removal from nonessential auxiliary plant equipment and from the ESW system over the full range of the normal reactor operation and normal shutdown.
2.2 Licensing Basis Changes 2.2.1 FSAR Changes In the enclosure to the LAR, the licensee proposes to add the following bracketed text/statements in the applicable subsections of the Callaway FSAR (ML21193A197), Standard Plant (SP), section 9.4.1, Control Building HVAC [Heating, Ventilation, and Air Conditioning].
FSAR SP 9.4.1.2.2, Component Description Both the control room air-conditioning unit and the Class 1E electrical equipment air-conditioning unit consist of high efficiency prefilters, a self-contained refrigeration system utilizing essential service water as the heat sink, centrifugal fans, and electric motor drivers. [(During cold shutdown or refueling conditions (Mode 5 or Mode 6 per the plant's Technical Specifications), one (and only one) of the two control room air-conditioning units may have normal service water solely aligned to it as its heat sink (i.e., without essential service water available),
as permitted per the plant's Technical Specifications for such conditions.)]
FSAR SP 9.4.1.2.3, System Operation Cooling water for the nonessential units is supplied by the central chilled water system (Section 9.4.10), and cooling water for the safety-related units is supplied
by the essential service water system [(except that during cold shutdown or refueling conditions (Mode 5 or Mode 6 per the plants Technical Specifications),
one (and only one) of the two control room air-conditioning units may have normal service water solely aligned to it as its heat sink (i.e., without essential service water available), as permitted per the plants Technical Specifications for such conditions)] (Section 9.2.1). Hot water for the control building supply air unit is supplied by the plant heating system (Section 9.4.9).
2.2.2 TS Changes In the enclosure to its supplement dated October 16, 2023, the licensee proposes to add the following Notes below to the respective LCOs:
NOTE to LCO 3.4.8:
- 3.
The Service Water system may serve as the alternate source of cooling water to the Essential Service Water system for support of the second required RHR train, i.e., the RHR train not supported by the emergency diesel generator required per Technical Specification 3.8.2, provided the plant is not in a reduced-inventory, hot-core condition.
NOTE to LCO 3.7.11:
During MODES 5 and 6, and during movement of irradiated fuel assemblies, the Service Water system may serve as the alternate source of cooling water to the Essential Service Water system for support of the second required CRACS train, i.e., the CRACS train not supported by the emergency diesel generator required per Technical Specification 3.8.2.
NOTE to LCO 3.9.6:
The Service Water system may serve as the alternate source of cooling water to the Essential Service Water system for support of the second required RHR train, i.e., the RHR train not supported by the emergency diesel generator required per Technical Specification 3.8.2, provided the plant is not in a reduced-inventory, hot-core condition.
2.3 Applicable Regulatory Requirements and Guidance The regulatory requirements associated with this LAR include the following:
The requirements for TSs are set forth in 10 CFR 50.36, Technical specifications.
Applications for licenses are required under 10 CFR 50.36(a)(1) to provide proposed TS as well as [a] summary statement of the bases or reasons for such specifications.
The regulation at 10 CFR 50.36(b) states that [t]he technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to § 50.34.
The regulations at 10 CFR 50.36(c), require that TSs includes, among other items, LCOs. The regulations in 10 CFR 50.36(c)(2)(i), contain requirements for LCOs and states, in part, that such TSs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The regulation further states that if an LCO is not met, the facility must be shut down, or other acceptable remedial action must be taken.
The regulation at 10 CFR 50.36(c)(2)(ii) states that LCOs must be established for each item that meets one or more of the following four criteria:
(A)
Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
(B)
Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
(C)
Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
(D)
Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
The regulation in 10 CFR 50.71, Maintenance of records, making of reports, requires the licensee to update the UFSAR periodically with a submittal that shall include evaluations performed by the licensee in support of approved license amendments.
GDC 4, Environmental and dynamic effects design bases states, in part, that:
Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.
GDC 19, Control room, states in part, that, A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.
GDC 34, Residual heat removal, states, in part, that, A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.
GDC 44, Cooling water, states, in part, that, A system be provided to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.
The NRC staffs guidance for the review of the TSs is in NUREG 0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor]
Edition, Chapter 16, Technical Specifications, Revision 3, dated March 2010 (ML100351425).
As described therein, as part of the regulatory standardization effort, the staff has prepared improved Standard Technical Specifications (STSs) for each of the LWR nuclear steam supply systems and associated balance of plant equipment systems. Special attention is given to TS provisions that depart from the improved STSs to determine whether proposed differences are justified by uniqueness in plant design or other considerations so that 10 CFR 50.36 is met.
NUREG-1431 contains the STSs applicable for Callaway.
3.0 TECHNICAL EVALUATION
3.1 Residual Heat Removal System (RHR Functions in Modes 5 and 6)
Section 2.2 of the enclosure to the LAR indicates that for applicable Mode 5 and Mode 6, the RHR system is designed to remove decay heat and to provide flow for the soluble neutron poison, boric acid, to mix with the RCS inventory.
During RHR system operation, heat is removed from the RCS by circulating reactor coolant through the RHR heat exchangers where the heat is transferred to the CCW system. The heat sink for the CCW system is provided by the ESW system. Both CCW and ESW systems are safety-related systems, which are designed to serve as a heat sink for the removal of process and operating heat from safety-related components during a DBA. The SW system is a non-safety-related system, which is designed to supply cooling water to plant loads, including the safety-related CCW system during normal plant operation and normal plant shutdown.
TS Requirements for the RHR in Modes 5 and 6 Current Callaway TSs 3.4.8 and 3.9.6 require that two RHR loops be OPERABLE and one RHR loop be in operation for both Mode 5 with RCS loops not filled, and Mode 6 with RCS loop at low water level. TS Bases 3.4.8 and 3.9.6 for the RHR LCOs in Modes 5 and 6 clarify that the reason for the TS inclusion is based on Criterion 4 of 10 CFR 50.36(c)(2)(ii), which is the
criterion based on risk. In addition, the Bases provide clarification of the requirements for an operable RHR loop, which states, as follows:
For TS 3.4.8: An OPERABLE RHR loop is comprised of an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required.
For TS 3.9.6: An OPERABLE RHR loop consists of an RHR pump, a heat exchanger, valves, piping, instruments and controls to ensure an OPERABLE flow path and to determine the RCS temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs. An OPERABLE RHR loop must be capable of being realigned to provide an OPERABLE flow path.
None of the paragraphs of TS Bases 3.4.8 and 3.9.6 quoted above for cooling water support of RHR Operability in Modes 5 and 6, refer to any cooling water supply (CCW, ESW, or SW systems), as each only defines Operability of an RHR loop within the scope of the RHR system.
During Modes 1, 2, 3, and 4, cooling water support of RHR Operability is clarified in the Bases for TSs 3.7.7 and 3.7.8, which state that the safety-related CCW and ESW systems must be considered to complete the link from RHR heat exchanger to the ultimate cooling water supply.
Based on the above information of the RHR requirements in the current Callaway TSs for Modes 1 through 6, the licensee states that the LCO sections of the Bases for the ESW system only address its Operability requirements for Modes 1, 2, 3, and 4 and do not explicitly address Operability requirements for Modes 5 and 6. However, the NRC staff noted in the second paragraphs of the Applicability sections of the Callaway TS Bases for TSs 3.7.7 and 3.7.8 that the Bases state In MODES 5 and 6, requirements for the CCW system are determined by the systems it supports, and In MODES 5 and 6, requirements for the ESW system are determined by the systems it supports, respectively.
Proposed TS Changes to Supporting System Requirements for RHR in MODES 5 and 6 To meet the requirements in TSs 3.4.8 and 3.9.6 for Modes 5 and 6, the licensee proposed that one train (referred to as the first train) must be supported by a safety-related diesel generator (DG) with the associated safety-related ESW train aligned to provide the ultimate cooling water supply. The other train (referred to as the second train), which is not required to have an Operable DG, may be aligned to the associated SW train to provide its ultimate cooing water supply. The proposed Note would allow crediting the use of a non-safety-related normal offsite power source and non-safety-related SW system as the supporting systems for the second required RHR train. The clarifications of the proposed RHR requirements are added to the LCO section of TS Bases 3.4.8 and 3.9.6 for applicable Modes 5 and 6. Specifically, they state the Service Water System may serve as the heat sink for one of the two required RHR loops, provided that one train of the Essential Service Water System serves as the heat sink of the other RHR loop.
The NRC staff reviewed the proposed first Operable RHR train requirements and finds them acceptable because the required emergency DG system and the associated ESW system are safety-related systems, and therefore meet the intent of the TS requirements of an Operable train.
The NRC staff reviewed the acceptability of the proposed second RHR train in MODES 5 and 6, which allows use of a non-safety-related normal offsite power source and SW system and discusses its evaluation in sections 3.3 and 3.4 of this report for shutdown risk insights and proposed TS changes, respectively.
3.2 CRACS TS LCO 3.7.11 requires two CRACS trains be OPERABLE during MODES 1, 2, 3, 4, 5, and 6, and during movement of irradiated fuel assemblies. TS Bases 3.7.11 for CRACS LCO clarifies that [t]wo independent and redundant trains of the CRACS are required to be OPERABLE to ensure that at least one is available, assuming a single failure disabling the other train. Total system failure could result in the equipment operating temperature exceeding limits in the event of an accident. TS Bases 3.7.11 for CRACS LCO also states that, [t]he CRACS is considered to be OPERABLE when the individual components necessary to maintain the control room temperature are OPERABLE in both trains. These components include the refrigeration compressors, heat exchangers, cooling coils, fans, and associated temperature control instrumentation.
FSAR SP section 9.4.1.2.3, states that [c]ooling water for the nonessential units is supplied by the central chilled water system (Section 9.4.10), and cooling water for the safety-related units is supplied by the essential service water system (Section 9.2.1), which the licensee proposes to change to allow using one (and only one) of the SW system for one CRACS train as heat sink.
TS LCO 3.7.8 requires two ESW trains be OPERABLE during MODES 1, 2, 3, and 4; however, TS Bases 3.7.8 applicability section states that in MODES 5 and 6, requirements for the ESW system are determined by the systems it supports. As stated above, TS LCO 3.7.11 requires two CRACS trains be OPERABLE during MODES 1, 2, 3, 4, 5, and 6, and during movement of irradiated fuel assemblies. With these TS requirements, the ESW is required to support CRACS during Modes 5 and 6.
If one CRACS train is inoperable, Condition A of TS 3.7.11 requires restoring the train to OPERABLE status within 30 days. Two independent and redundant trains are required to be OPERABLE to ensure that at least one is available, assuming a single failure disabling the other train. Total system failure could result in the equipment operating temperature exceeding limits in the event of an accident. During Modes 5 and 6, TS 3.8.2 requires at least one operable offsite power circuit and one operable DG. The offsite circuit can provide power to one or both trains of safety-related shutdown equipment (e.g., ESW/CRACS) and the non-safety-related SW. The DG can provide support to only one train of safety-related shutdown equipment to which it is connected. With the licensees proposed changes, one CRACS train will continue to be supported by a safety-related ESW system, which can be powered by its preferred offsite power circuit source, or an emergency DG while the other train is supported by a non-safety-related SW system powered by the offsite power circuit.
Requirements for LOOP are discussed in FSAR SP section 3.1.2, Additional Single Failure Assumptions (ML21193A186), which contains lists of assumptions, including the following, made in addition to postulating the initiating event in designing for and analyzing for DBAs:
- a.
The events are assumed not to result from a tornado, hurricane, flood, fire, loss of offsite power, or earthquake.
- e.
All offsite power is simultaneously lost and is restored within 7 days (except that for events postulated to occur during MODE 5, MODE 6, and/or during movement of irradiated fuel assemblies when the plant is in MODE 5 or MODE 6 or with the core fully offloaded, such as a fuel handling accident, a loss of all offsite power is not required to be assumed in addition to a single failure).
As stated in Item e., during a DBA that is postulated to occur in MODES 5 or 6, or during movement of irradiated fuel assemblies, a LOOP is not required to be assumed in addition to a single failure. The SW system will continue to be powered by offsite power. If a single failure disables either of the supporting ESW or SW systems, the other cooling system will be available to support one of the CRACS trains safety function. Therefore, with the licensee proposed changes to the FSAR regarding operability of CRACS in MODES 5 and 6, and a corresponding change to the TS with a NOTE for LCO 3.7.11, as stated in section 2.2 above, the CRACS will continue to perform its safety function.
Based on the above, the NRC staff concludes that there is reasonable assurance that at least one CRACS train will be available, assuming a single failure disabling the other train, and thus the control room equipment operating temperature will be maintained within allowable limits in the event of an accident because at least one train of CRACS will be required to be supported by only safety related equipment. Therefore, the staff finds that the licensees proposed FSAR and TS changes related to CRACS continues to satisfy GDC 4, 19 and 44, and are therefore acceptable.
3.3 Shutdown Risk Insights In section 2.2 of the LAR, the licensee states that in Mode 5, with the RCS loops not filled, the primary function of the RHR system is to remove decay heat generated in the fuel and transfer it to the CCW system via the RHR heat exchangers, as required by GDC 34. The secondary function of the reactor coolant is to act as a carrier for the soluble neutron poison, boric acid. In Mode 6, the function of the RHR system is to remove decay heat and stored thermal energy of the RCS to the CCW system via the RHR heat exchangers, as required by GDC 34. The heat sink for the CCW system is normally provided by the SW or ESW system, as determined by TS LCO requirements and system availability. According to GDC 34, the RHR system must transfer the decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. Extended loss of the decay heat removal function would result in RCS boiling, an inadvertent change to Mode 4, and challenge to the specified acceptable fuel design limits. To prevent such a scenario, per the licensees TS, two trains of the RHR system are required to be Operable in Modes 5 and 6, with one train in operation.
The licensee stated that while its evaluation against 10 CFR 50.59(c)(2)(ii) did identify a more than minimal increase in the likelihood of malfunction of the RHR system, the reliability of an RHR loop supported by an ESW train with non-Class 1E power is not viewed to be significantly different than the reliability of an RHR loop supported by a SW train (in lieu of an ESW train). In its supplement dated October 16, 2023, the licensee stated that the SW system is not specifically analyzed for protection from hazards and events such as tornadoes, floods, missiles, pipe breaks, fires, and seismic events compared to the ESW system.
In its supplement dated October 16, 2023, the licensee provided a revised LCO that stipulated, via a Note in the LCO, that [t]he Service Water system may serve as the alternate source of
cooling water to the Essential Service Water system for support of the second RHR train, i.e.,
the RHR train not supported by the emergency diesel generator required per Technical Specification 3.8.2, provided the plant is not in a reduced inventory, hot-core condition. The NRC staff notes that hot core, mid-loop conditions, are a subset of reduced inventory, hot core conditions.
In its supplement dated November 20, 2023, the licensee stated that Callaway uses the same definition for reduced inventory as that within the Nuclear Management and Resources Council (NUMARC) 91-06 Guidelines for Industry Actions to Assess Shutdown Management (ML14365A203), and Generic Letter (GL) 88-17 Loss of Decay Heat Removal, dated October 17, 1988 (ML031200496), which is the pressurized water reactor (PWR) condition with fuel in the reactor vessel and level lower than 3 feet below the reactor vessel flange, which corresponds to the RCS level of 64 inches (i.e., 64 above the bottom of the inner diameter of the RCS hot leg piping) and hot core is defined as the present fuel cycle core within the reactor before refueling. Thus, the revised LCO precludes the SW system from being used to support RHR when the reactor vessel level is less than 3 feet below the reactor vessel flange before refueling when the time to core boiling and core damage are the shortest. The licensee also provided Notes to the affected TS Bases to clarify that the Service Water System may serve as the heat sink for one of the two required RHR loops, provided that one train of the Essential Service Water System serves as the heat sink for the other RHR loop and that this allowance is dependent upon the plant not being in a reduced-inventory, hot-core condition. In addition, the licensee stated in the TS bases, Since a reduced inventory, hot-core condition is an elevated risk condition for the plant during Mode 5, the restriction in the Note for this condition ensures the decay heat removal function performed by the RHR system is supported by the safety-related Essential Service Water System (in lieu the non-safety Service Water System) and also defined hot core and reduced inventory in its site procedures.
GL 88-17 and NUMARC 91-06 contain key shutdown voluntary initiatives credited in SECY-97-168, Issuance for Public Comment of Proposed Rulemaking Package for Shutdown and Fuel Storage Pool Operation, that are necessary to reduce shutdown risk beyond what is provided by adherence to TSs. In Staff Requirements Memorandum (SRM) to SECY 97-168, the Commission expects the NRC staff to continue to monitor licensee performance, through inspections and other means, in the area of shutdown operations to ensure that the current level of safety is maintained. Following this expectation, the NRC staff reviewed the licensees implementation of GL 88-17 during the audit on PWR reduced inventory operation after refueling when the licensee performs vacuum refill of the RCS. The staff also reviewed the licensees implementation of NUMARC 91-06 during the audit regarding containment closure.
In its supplement dated October 16, 2023, the licensee stated that as part of the site's response to GL 88-17, two additional injection systems (i.e., safety injection (SI) and centrifugal charging pumps (CCPs)), are available to respond following a potential loss of decay heat removal and that the site ensures the SI and CCPs providing the additional injection sources are on the same division as the operable emergency diesel generator (EDG) and ESW train.
In its supplement dated October 16, 2023, the licensee stated that outage activities and risk are managed using the process described in NUMARC 91-06, as implemented via site procedure APA-ZZ-00315, appendix E, Configuration Risk Management Program - Shutdown, and that this process recognizes use of either the SW or ESW systems to provide cooling water flow to the CCW system and its supported systems (including RHR and CRACS). The licensee explained that additional risk management actions are specified in the procedure to protect the SW pumps and their support systems including barriers, signs, and ribbon for periods where
only one train of ESW is available and SW is used as the sole support system for the opposite train.
Regarding the licensees containment closure capability consistent with NUMARC 91-06, the licensee stated in its supplement dated October 16, 2023 that the closure of the containment equipment hatch and containment equipment hatch missile shield is controlled by site procedure APA-ZZ-00750, Hazard Barrier Program, attachment 4. The licensee stated that the containment equipment hatch and containment equipment hatch missile shield closure is estimated to take approximately 30 minutes.
The licensee further explained that the procedure sets a weather monitoring distance for each missile shield and equipment hatch to be closed based upon an assumed storm translation speed of 70 miles per hour (mph), which ties to a weather monitoring distance of 40 miles.
Additionally, in its supplement dated November 20, 2023, the licensee stated that site procedure APA-ZZ-00150, appendix M, Containment Closure, contains guidance for when containment closure needs to be established in Modes 5, 6, or with loops not filled.
Section 4.1.1 of this procedure includes the following:
When greater than 64 inches RCS level (not in reduced inventory),
containment hatches and penetrations must be capable of being closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or the time calculated for T-Boil, whichever is less. If an opening cannot be closed within these time constraints, it will remain closed.
During reduced inventory, containment hatches and penetrations must be capable of being closed within 30 minutes or the time calculated for T-Boil, whichever is less. If an opening cannot be closed within these time constraints, it will remain closed.
Additional administrative controls must be instituted if the demonstrated time to close the equipment hatch is greater than one-half the time calculated for T-boil. For example:
o Prestage the equipment hatch closure team at the equipment hatch, o Limit what can be in the equipment hatch, such that any obstructions, e.g., cables and hoses, SEALAND boxes, cart or Hatch Transfer System rails that could prevent closure, can be quickly removed.
The NRC staffs review finds that the licensees proposed request to allow the use of one train of the normal, non-safety-related SW system to solely provide cooling water support for one of two required trains of the RHR system and CRACS during Modes 5 and 6 is acceptable for this application because: (1) The more than a minimal increase in the likelihood of occurrence of a malfunction of the RHR system is mitigated by the revised LCO that states that [t]he Service Water system may serve as the alternate source of cooling water to the Essential Service Water system for support of the second required RHR train, i.e., the RHR train not supported by the emergency diesel generator required per Technical Specification 3.8.2, provided the plant is not in a reduced inventory, hot-core condition, and (2) key shutdown risk voluntary initiatives from GL 88-17 on PWR reduced inventory operation and NUMARC 91-06 on containment closure have been met through the licensees procedures as directed in the SRM to SECY 97-168.
3.4 Review of Proposed TS Changes The NRC staffs review of the proposed TS changes evaluated whether: (1) the TSs would continue to be derived from the licensees analyses and evaluations per 10 CFR 50.36(b);
(2) the LCO would continue to list the lowest functional capability or performance levels of equipment required for safe operation of the facility per 10 CFR 50.36(c)(2), and (3) TS provisions that depart from the improved STSs are justified by uniqueness in plant design or other considerations so that 10 CFR 50.36 would continue to be met.
3.4.1 TS Derived from Licensees Evaluations The NRC staff reviewed the licensees evaluations of the proposed changes in sections 3.1 through 3.3 of this safety evaluation. The staff determined that the licensees evaluations provided appropriate justification for addition of the respective Notes because the Notes would restrict the use of non-safety-related SW to support only one train of ESW and to situations when the plant is not in a reduced inventory, hot-core condition, thereby limiting the risk if a malfunction of the systems occurred. Therefore, the staff determined that the proposed changes to the TSs are derived from the licensees evaluations. Given the 10 CFR 50.71 requirement for the licensee to update the FSAR with evaluations in support of license amendments, the staff determined the TSs will continue to be derived from the FSAR. Therefore, the proposed TS changes will continue to meet the requirements of 10 CFR 50.36(b).
3.4.2 LCO Lists Lowest Functional Capability The NRC staff reviewed the proposed Notes for the respective TSs. The Notes modify the LCO requirements for each respective LCO.
The licensee justified the proposed Note for TS 3.7.11 in its LAR and supplements. On page 4 of attachment 2 to the October 16, 2023, supplement, the licensee stated:
The one FSAR described DBA that may be postulated to occur during Mode 5 or Mode 6 is a fuel handling accident (FHA). The proposed changes do not affect the systems/functions required to mitigate the dose consequences of an FHA.
(Control room dose is mitigated by the Control Room Emergency Ventilation System and not by CRACS.) Therefore, the proposed changes do not involve any significant increase in the consequences of the FHA as previously described in the FSAR.
The NRC staff reviewed the justifications and determined the proposed Note below the LCO statement for TS 3.7.11 creates an allowance to use the non-safety-related SW system to support one train of CRACS, with the restriction that SW cannot be used to support the CRACS train supported by the EDG required per TS 3.8.2. The staff determined the LCO, as modified by the allowance and its associated restrictions, will still list the lowest functional capability or performance levels of equipment required for safe operation of the facility.
The licensee justified the proposed Notes for TSs 3.4.8 and 3.9.6 in its LAR and supplements.
On page 23 of attachment 1 to the October 16, 2023, supplemental letter the licensee stated:
It is not the site's intent to utilize the SW system to support one of the RHR trains in lieu of having an operable ESW train for mid-loop operations during the hot core period. The intent is to allow the capability to utilize the SW system in
support of one of the two TS 3.4.8-required RHR trains when not in mid-loop conditions or following the refueling period of the outage, when the core is considered a cold core. The TS changes proposed per this LAR supplement reflect this intent.
On page 3 of attachment to the November 20, 2023, supplement the licensee stated:
A time limit was not considered to be necessary or included in the proposed Notes since the Notes already stipulate that Service Water is not allowed during the most elevated risk condition for the plant during a typical refueling outage, i.e., a hot core at reduced inventory. In addition, the overall time that the plant is in these LCOs (i.e., when the Note may be applied, exclusive of the hot-core, reduced-inventory condition) is typically a fraction of the overall time that a plant is in a refueling outage.
The NRC staff reviewed the justifications and determined the proposed Notes below the LCO statement for TS 3.4.8 and TS 3.9.6 create an allowance to use the non-safety-related SW system to support one train of RHR, with the restrictions that SW cannot support the train aligned to EDG power support and the allowance cannot be used when the plant is in a reduced-inventory, hot core condition. The staff determined the LCO, as modified by the allowance and its associated restrictions, will still list the lowest functional capability or performance levels of equipment required for safe operation of the facility.
Therefore, the NRC staff determined that the LCOs, as modified by the allowances in the proposed Notes, will continue to meet the requirements of 10 CFR 50.36(c)(2).
3.4.3 TS Departures from STSs The NRC staff compared the Callaway TSs, as amended by the proposed Notes, to the STSs in NUREG-1431 (the STSs applicable for Callaway). The staff determined the proposed Notes are a departure from the STSs. The staff determined that the departure from STSs is justified by the licensees statements in the October 16, 2023, and November 20, 2023, supplemental letters regarding the restrictions for the application of the allowance provided in the Notes because the staff determined the Notes restrict when non-safety related equipment can be used to support RHR and CRACS to the lowest risk periods.
3.4.4 TS Acceptability Conclusion The NRC staff concluded that the proposed changes to the TSs are acceptable because: (1) the TS would continue to be derived from the licensees analyses and evaluations; (2) The affected LCOs would continue to list the lowest functional capability or performance levels of equipment required for safe operation of the facility; and (3) while the proposed TS Notes do depart from the improved STSs, they are justified by the limited circumstances under which they can be applied so that 10 CFR 50.36 would continue to be met.
3.5 Technical Conclusion The NRC staff reviewed the proposed LAR, as supplemented, for NRC approval of a change to Callaways licensing basis (i.e., the FSAR and TSs), to allow use of one train of the normal, non-safety-related SW system to provide cooling water support for one of two redundant trains of TS-required equipment when both equipment trains are required to be Operable during cold
shutdown/refueling conditions. The supported equipment/systems affected by the proposed change are the RHR system and CRACS, as applicable during Modes 5 and 6.
The NRC staff performed an audit to review the supporting documents, which resulted in the licensee supplementing the LAR twice. The staff reviewed impacts of the proposed FSAR and TS changes on the RHR system and CRACS. The staff reviewed the shutdown risk insights on RHR functions in Modes 5 and 6. The staff also reviewed the proposed revisions to the Callaway FSAR and TS.
As discussed above, the NRC staff review finds that the proposed changes will not impact continued compliance with GDCs 4, 19, 34, and 44. The staff finds the corresponding changes to the Callaway FSAR and TS are acceptable because they are consistent with the proposed RHR system, and CRACS licensing basis changes, and the TS will continue to meet the requirements of 10 CFR 50.36.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Missouri State official was notified of the proposed issuance of the amendment on December 7, 2023. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in 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 that the amendment involves NSHC, as published in the Federal Register on February 21, 2023 (88 FR 10559). Subsequently, by letter dated October 16, 2023, the licensee provided additional information that expanded the scope of the amendment request as originally noticed in the Federal Register. Accordingly, the Commission issued a second proposed finding that the amendment involves NSHC in the Federal Register published on November 13, 2023 (88 FR 77616), which superseded the original notice in its entirety. There had been no public comment on such finding. Accordingly, the amendments meet 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: H. Wagage, NRR G. Curran, NRR S. Sun, NRR M. Hamm, NRR M. Pohida, NRR K. Tetter, NRR Date: January 18, 2024
- concurrence by email OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA*
NRR/DEX/EEEB/BC*
NRR/DRA/APLC/BC*
NAME MChawla PBlechman WMorton (VKGoel for)
SVasavada DATE 12/14/2023 12/21/2023 12/18/2023 12/7/2023 OFFICE NRR/DSS/SCPB/BC*
NRR/DSS/SNSB/BC*
NRR/DSS/STS/BC(A)*
OGC NAME BWittick PSahd SMehta (ARussell for)
MWright DATE 12/18/2023 12/16/2023 12/18/2023 1/10/2024 OFFICE NRR/DORL/LPL4/BC*
NRR/DORL/LPL4/PM*
NAME JRankin MChawla DATE 1/18/2024 1/18/2024