ML21021A349
| ML21021A349 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/03/2021 |
| From: | Michael Wentzel Plant Licensing Branch II |
| To: | Jim Barstow Tennessee Valley Authority |
| Wentzel M | |
| References | |
| EPID L-2020-LLA-0130 | |
| Download: ML21021A349 (24) | |
Text
March 3, 2021 Mr. James Barstow Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 354 AND 347 REGARDING REVISION TO TECHNICAL SPECIFICATION 4.2.2, CONTROL ROD ASSEMBLIES (EPID L-2020-LLA-0130)
Dear Mr. Barstow:
The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 354 to Renewed Facility Operating License No. DPR-77, and Amendment No. 347 to Renewed Facility Operating License No. DPR-79, for the Sequoyah Nuclear Plant, Units 1 and 2, respectively. These amendments are in response to your application dated June 12, 2020.
The amendments revise Technical Specification 4.2.2, Control Rod Assemblies, to permit the Sequoyah Nuclear Plant, Units 1 and 2, Cycle 25 cores to contain 52 full length control rods with no full-length control rod assembly in core location H-08.
A copy of our related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Michael J. Wentzel, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328
Enclosures:
- 1. Amendment No. 354 to DPR-77
- 2. Amendment No. 347 to DPR-79
- 3. Safety Evaluation cc: Listserv TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 354 Renewed License No. DPR-77
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated June 12, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (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-77 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 354, 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 its date of issuance and shall be implemented prior to entering MODE 5 from MODE 6 during startup from the Cycle 24 (U1R24) refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION Undine S. Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 3, 2021 Undine S.
Shoop Digitally signed by Undine S. Shoop Date: 2021.03.03 13:56:23 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 354 SEQUOYAH NUCLEAR PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Replace page 3 of the Renewed Facility Operating License with the attached page 3.
Replace the following page of the Appendix A Technical Specifications with the attached page.
The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Page Insert Page 4.0-1 4.0-1 Amendment No. 354 Renewed License No. DPR-77 (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 calibration or associated with radioactive apparatus or components; and (5)
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 Sequoyah and Watts Bar Unit 1 Nuclear Plants.
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 The Tennessee Valley Authority is authorized to operate the facility at reactor core power levels not in excess of 3455 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 354 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Initial Test Program The Tennessee Valley Authority shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Tennessee Valley Authoritys Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:
a.
Elimination of any test identified in Section 14 of TVAs Final Safety Analysis Report as amended as being essential; b.
Modification of test objectives, methods, or acceptance criteria for any test identified in Section 14 of TVAs Final Safety Analysis Report as amended as being essential;
Design Features 4.0 SEQUOYAH - UNIT 1 4.0-1 Amendment 348, 354 4.0 DESIGN FEATURES 4.1 Site Location The Sequoyah Nuclear Plant is located on a site near the geographical center of Hamilton County, Tennessee, on a peninsula on the western shore of Chickamauga Lake at Tennessee River mile (TRM) 484.5. The Sequoyah site is approximately 7.5 miles northeast of the nearest city limit of Chattanooga, Tennessee, 14 miles west-northwest of Cleveland, Tennessee, and approximately 31 miles south-southwest of TVA's Watts Bar Nuclear Plant.
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or M5 clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. Sequoyah is authorized to place a limited number of lead test assemblies into the reactor as described in the Framatome-Cogema Fuels report BAW-2328, beginning with the Unit 1 Operating Cycle 12.
4.2.2 Control Rod Assemblies
NOTE--------------------------------------------------
Operation with 52 full length control rod assemblies (with no control rod assembly installed in core location H-08) is permitted during Cycles 24 and 25.
The reactor core shall contain 53 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium, and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.
4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent; TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 347 Renewed License No. DPR-79 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated June 12, 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-79 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 347, 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 its date of issuance and shall be implemented prior to entering MODE 5 from MODE 6 during startup from the Cycle 24 (U2R24) refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION Undine S. Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 3, 2021 Undine S.
Shoop Digitally signed by Undine S. Shoop Date: 2021.03.03 13:57:02 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 347 SEQUOYAH NUCLEAR PLANT, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Replace page 3 of the Renewed Facility Operating License with the attached page 3.
Replace the following page of the Appendix A Technical Specifications with the attached page.
The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Page Insert Page 4.0-1 4.0-1 Amendment No. 347 Renewed License No. DPR-79 (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 calibration or associated with radioactive apparatus or components; and (5)
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 Sequoyah and Watts Bar Unit 1 Nuclear Plants.
C.
This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's 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 The Tennessee Valley Authority is authorized to operate the facility at reactor core power levels not in excess of 3455 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 347 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Initial Test Program The Tennessee Valley Authority shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Tennessee Valley Authority's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:
a.
Elimination of any test identified in Section 14 of TVA's Final Safety Analysis Report as amended as being essential;
Design Features 4.0 SEQUOYAH - UNIT 2 4.0-1 Amendment 327, 342, 347 4.0 DESIGN FEATURES 4.1 Site Location The Sequoyah Nuclear Plant is located on a site near the geographical center of Hamilton County, Tennessee, on a peninsula on the western shore of Chickamauga Lake at Tennessee River mile (TRM) 484.5. The Sequoyah site is approximately 7.5 miles northeast of the nearest city limit of Chattanooga, Tennessee, 14 miles west-northwest of Cleveland, Tennessee, and approximately 31 miles south-southwest of TVA's Watts Bar Nuclear Plant.
4.2 Reactor Core 4.2.1 Fuel Assemblies 4.2.2 The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or M5 clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. Sequoyah is authorized to place a limited number of lead test assemblies into the reactor as described in the Framatome-Cogema Fuels report BAW-2328, beginning with the Unit 2 Operating Cycle 10 core.
Control Rod Assemblies
NOTE--------------------------------------------------
Operation with 52 full length control rod assemblies (with no control rod assembly installed in core location H-08) is permitted during Cycles 24 and 25.
The reactor core shall contain 53 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium, and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.
4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent; SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 354 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO. 347 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328
1.0 INTRODUCTION
By application dated June 12, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20164A270), the Tennessee Valley Authority (TVA, the licensee) submitted a license amendment request (LAR) for the Sequoyah Nuclear Plant (Sequoyah),
Units 1 and 2. The proposed amendments would revise the Sequoyah, Units 1 and 2, Technical Specification (TS) 4.2.2, Control Rod Assemblies, to permit the Sequoyah, Unit 1, Cycle 25 (U1C25) and Sequoyah, Unit 2, Cycle 25 (U2C25) cores to contain 52 full-length control rods with no full-length control rod assembly in core location H-08.
2.0 REGULATORY EVALUATION
2.1
System Description
Section 3.1 of the enclosure to the LAR describes the control rods at Sequoyah, Units 1 and 2 as follows:
[Sequoyah] Unit 1 and [Sequoyah] Unit 2 each normally contain 53 full length control rod assemblies divided into four control banks (Control Banks A, B, C, D) and four shutdown banks (Shutdown Banks A, B, C, D). Of the eight banks, Control Bank D is used for reactivity control during normal at-power operation.
The remaining control banks are normally used for reactor startup and shutdown.
The shutdown banks provide additional negative reactivity to meet shutdown margin (SDM) requirements. During MODES 1 and 2, the shutdown banks are fully withdrawn from the core in accordance with TS 3.1.5 and as specified in the Core Operating Limits Report (COLR).
The H-08 control rod is part of Control Bank D and is located in the center of the core as shown in Figure 1 [of the enclosure to the LAR]. With the removal of the H-08 control rod, U1C24 and U2C24 cores contain 52 full length control rod assemblies as shown in the table to Figure 1 [of the enclosure to the LAR]. The H-08 control rod would remain removed for U1C25 and U2C25.
Each control rod is moved by a full length CRDM [control rod drive mechanism]
consisting of a stationary gripper, movable gripper, and a lift pole. Three coils are installed external to the CRDMs to electromechanically manipulate the CRDM components to produce rod motion. The CRDMs are magnetic jacking type mechanisms that move the control rods within the reactor core by sequencing power to the three coils of each mechanism to produce a stepping rod motion. Rod position is achieved through a timed sequence of stationary, movable, and lift coil current. At each point in time during rod positioning, the control rod is being held by either the stationary gripper or movable grippers.
Should both sets of grippers be de-energized simultaneously, the corresponding control rod would drop into the core. The primary function of the CRDMs is to insert, withdraw, or hold control rods within the core to control average core temperature and to shut down the reactor. Mechanically, each control rod location includes a guide tube, which is an assembly that houses and guides the control rod through the upper internals.
The full length Rod Control System receives rod speed and direction signals from the Tavg control system (contained within the Distributed Control System). The automatic rod speed demand signal varies over the corresponding range of 5 to 45 inches per minute (8 to 72 steps/minute) depending on the magnitude of the error signal. The rod direction demand signal is determined by the positive or negative value of the error signal. Manual control is provided to move a control bank in or out at a prescribed fixed speed.
Note that the licensee uses the terms control rod and rod cluster control assemblies (RCCAs) synonymously.
2.2 Proposed TS Changes
The licensee proposed to modify the note in TS 4.2.2, as indicated below (added text indicated by underline):
NOTE--------------------------------------------------
Operation with 52 full length control rod assemblies (with no control rod assembly installed in core location H-08) is permitted during Cycles 24 and 25.
The U.S. Nuclear Regulatory Commission (NRC, the Commission) staff issued amendments under exigent circumstances approving the use of the above note for Cycle 24 for Units 1 and 2, by letters dated November 21, 2019 (ADAMS Accession No. ML19319C831), and April 23, 2020 (ADAMS Accession No. ML20108F049), respectively. The licensee is requesting an extension for Cycle 25 pending permanent removal via a separate fuel transition LAR.
2.3 Regulatory Requirements Section 50.36, Technical specifications, of Title 10 of the Code of Federal Regulations (10 CFR) establishes the regulatory requirements related to the content of TSs.
Paragraph 50.36(a)(1) of 10 CFR requires an application for an operating license to include proposed TSs. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the TSs.
Pursuant to 10 CFR 50.36(c), TSs for operating reactors are required, in part, to include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.
Paragraph 50.36(c)(4) of 10 CFR states that design features are those features of the facility, such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3) of the section.
Appendix A, General Design Criteria (GDC) for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The following GDC are applicable for this review:
Criterion 2 - Design bases for protection against natural phenomena Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.
Criterion 4 - Environmental and dynamic effects design bases 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. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
Criterion 10 - Reactor design The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
Criterion 11 - Reactor inherent protection The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.
Criterion 12 - Suppression of reactor power oscillations The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.
Criterion 23 - Protection system failure modes The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.
Criterion 25 - Protection system requirements for reactivity control malfunctions The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.
Criterion 26 - Reactivity control system redundancy and capability Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.
Criterion 27 - Combined reactivity control systems capability The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.
Criterion 28 - Reactivity limits The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold-water addition.
Criterion 29 - Protection against anticipated operational occurrences The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the licensees application to determine if the proposed changes are consistent with the guidance, regulations, and plant-specific design and licensing basis information discussed in Section 2.3 of this safety evaluation.
3.1 Evaluation of Proposed Changes The licensee stated that Framatome performs the reload licensing analysis for Sequoyah, Units 1 and 2, and applies NRC-approved codes and analytical methods to design the reload core. The NRC-approved codes and analytical methods used to generate the reload safety evaluation are included in TS 5.6.3, Core Operating Limits Report, and are also listed in the cycle-specific COLR. The licensee stated that the reload safety analysis methods are not invalidated by the removal of control rod H-08 from the core design because these methods are not dependent on a particular RCCA configuration. The licensee stated that Cycle-specific reload evaluations of TS limits, safety analysis limits, and operating limits without the H-08 control rod for Cycle 25 have not yet been completed and are typically not finalized until approximately 30 days prior to the start of the refueling outage. The licensee also stated that the reload analyses for Cycle 24 support the removal of H-08 for Cycle 25 given the following:
The U1C24 and U2C24 cores are similar in design to the expected design of the Cycle 25 cores (i.e., similar energy requirements, feed batch size and enrichment).
The Cycle 25 analyses results will not vary appreciably from the U1C24 and U2C24 results.
The U1C24 and U2C24 cores demonstrate that a core designed assuming 53 control rods (i.e., assuming the H-08 control rod present) can still meet all required safety analysis acceptance criteria when the H-08 control rod is removed.
Designing the Cycle 25 cores with the H-08 control rod removed assures the core designs will meet all acceptance criteria.
The U1C24 and U2C24 margin assessments form a strong technical justification for the H-08 RCCA removal extension.
In the two previous LARs (referenced above in Section 2.2) where the licensee proposed operation with only 52 control rods for Cycle 24, the licensee provided details on important safety parameters (shutdown margin, boron concentration and boron worth, moderator temperature coefficient, etc.) and the impact on the Updated Final Safety Analysis Report (UFSAR) Chapter 15 Accident Analyses. The NRC staffs review concluded that the licensees proposed use of 52 control rod assemblies in Sequoyah, Units 1 and 2, for Cycle 24 was acceptable because all cycle-specific safety analysis limits were met, the design change is consistent with the current design basis, and does not challenge the safety analyses detailed in Chapter 15 of the UFSAR. Given that the reload core design methodology requires the performance of cycle-specific thermal-hydraulic and limiting transient analysis checks to ensure the core does not reduce the existing safety analysis margin, the NRC staff finds that the operating and safety analysis limits with the use of only 52 control rods for the Cycle 25 cores will be assured.
3.1.1 Impact of the Flow Restrictor 3.1.1.1 Thermal-Hydraulic Impacts The licensee stated that when control rods H-08 and their associated driveshafts were removed from service, flow restrictors were installed in the H-08 control rod guide tubes in the reactor vessel upper internals. Installation of the flow restrictor ensures the flow area and hydraulic resistance normally provided by the driveshaft in the guide tube is maintained. In the previous LARs, the licensee performed bypass flow analyses to determine the impact of removing the control rod in core location H-08. The licensee stated that this analysis shows that the core bypass flow increases slightly but remains below the analyzed bounding value, and therefore all DNB analyses remain bounding. The NRC staff previously concluded that the licensees installation of a flow restrictor is acceptable because the flow restrictor maintains the thermal-hydraulic configuration of the reactor vessel upper internals. The NRC staff concludes that this finding does not change for the current LAR proposal to operate for an additional cycle.
3.1.1.2 Structural Evaluation The licensee stated that the evaluation of the potential design impacts of removing control rod H-08 and associated drive shaft in support of the previous LARs for Cycle 24 remain valid for Cycle 25 because the flow restrictors installed for Cycle 24 will remain installed for Cycle 25.
The NRC staff agrees that the licensees evaluations performed for Cycle 24 remain valid for Cycle 25 because no other changes to the core are being made that would impact those evaluations. The NRC staffs evaluation of the relevant issues is below.
Dynamic Analysis The licensee stated that the changes in reactor coolant system water volume and metal mass for each Unit are negligible due to the removal of the control rod and the installation of the flow restrictor and that the effect on the dynamic analysis is negligible. The NRC staff agrees that the change in system mass due to the removal of control rod H-08 is negligible. Therefore, the NRC staff finds that the impact on the dynamic analyses that predicts the stresses in the CRDM, reactor vessel, vessel supports, and reactor internals when subjected to seismic or loss-of-coolant accident (LOCA) excitations is negligible.
Further, the licensee stated that there is no impact on the functionality or structural integrity of the reactor vessel upper internals with the removal of the control rod drive shaft and RCCA at core location H-08 when a flow restrictor is installed in its place. Therefore, the licensee concluded that there is no impact on the current reactor vessel internals analyses. The NRC staff agrees that there should not be significant impact on the current reactor vessel internals analysis because the hydraulic loads on reactor internals will be similar, as the flow restrictor provides similar flow and pressure loss at core location H-08 for Units 1 and 2.
Flow Restrictor Design The licensee stated that the installed flow restrictor is a standard component used to hydraulically simulate the CRDM drive shaft clearance with the guide tube housing opening.
This will establish hydraulically equivalent flow conditions in the upper internals when the drive shaft is removed. The licensee performed a generic structural analysis of the restrictor plate/orifice assembly using a bounding pressure differential load for the faulted service condition (i.e., LOCA). This analysis conservatively assumed no orifice holes in the assembly to maximize the differential pressure load. The analysis demonstrated that all membrane and bending, bearing, and shear stress intensities satisfy the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code),Section III. The licensee also demonstrated that bolting preload was adequate to resist assembly separation for maximum LOCA pressure loads. The NRC staff agrees that the generic analysis bounds the Sequoyah plant-specific service conditions.
Further, the NRC staff notes that the materials used for the flow restrictor assembly conform to the ASME BPV Code,Section II, Part A. The licensee stated in the LAR that the material is Type 304 stainless steel for restrictor assembly, as well as the guide tube, and is compatible with fluid conditions in the reactor vessel upper internals. The NRC staff agrees that there will be no differential thermal expansion effects because the restrictor assembly and the guide tube are the same material.
The licensee also discussed the design to prevent the possibility of loose parts in the reactor coolant system (RCS). Installation of the restrictor is controlled to ensure that the required hex bolt preload is obtained, securely locking the flow restrictor in place at the top of the guide tube.
A locking cup, which is tack welded to the flow restrictor, is crimped onto the hex bolt to prevent hex bolt rotation. The NRC staff agrees that the capture features of the flow restrictor (i.e.,
locking fingers, hex bolt cup, and hex bolt preload) provide reasonable assurance that the flow restrictor is securely installed and will not result in the generation of loose parts.
The licensee stated that the reactor internals are designed and analyzed to the requirements of UFSAR Section 3.9.3, NSSS Components Not Covered by the ASME Code. The basis for the design stress and deflection criteria is summarized in Section 4.2.2.5, Design Criteria Basis, of the UFSAR. While the restrictor assembly does not perform a core support or safety function, it is classified as American National Standards Institute Safety Class 3. All of the calculated stresses are within the ASME BPV Code allowable stress limits. The NRC staff has determined that the flow restrictor assembly materials, fabrication, and design analysis meet the intent of ASME BPV Code, Subsection NG, which is consistent with the Sequoyah, Units 1 and 2, design bases described in the UFSAR.
Based on the review of the information provided by the licensee, the NRC staff has determined that there is reasonable assurance that the flow restrictor will maintain its structural integrity without generating loose parts, consistent with the ASME BPV Code, and that the design allowable stress limits are utilized in accordance with the Code.
3.2 Compliance with applicable GDC The NRC staffs review of the applicable GDC listed in Section 2.3 above are summarized below.
GDC 2 - Design Bases for Protection Against Natural Phenomena The NRC staff finds that the current CRDM dynamic stress evaluations due to seismic and LOCA excitations remain valid because the impact of the mass change or reduction is not significant.
GDC 4 - Environmental and Missile Design Bases The licensee evaluated the removal of the control rod assembly and the addition of the flow restrictor. The NRC staff reviewed this evaluation and determined that existing analyses for most reactor coolant system subcomponents remain bounding, and the stresses associated with the flow restrictor are within the ASME BPV Code limits. Further, there will be no differential thermal expansion effects, and the capture features of the flow restrictor (i.e., locking fingers, hex bolt cup, and hex bolt preload) provide reasonable assurance that the flow restrictor is securely installed, and will not result in the generation of loose parts because acceptable component design limits will not be exceeded during normal operation, including the effects of anticipated operational occurrences.
GDC 10 - Reactor Design The licensee previously performed a redesign reload analysis for Cycle 24 in accordance with the methods described in TS 5.6.3 and confirmed that the fuel design limits are not exceeded during any condition of normal operation; including the effects of anticipated operational occurrences with control rod H-08 removed.
While the Cycle 25 reload analysis will not be completed until shortly before the next refueling outage, the existing core reload design process using NRC-approved methods assures that the fuel design limits will be met; therefore, the NRC staff finds that the requirements of GDC 10 continue to be met.
GDC 11 - Reactor Inherent Protection The licensee stated that the fuel temperature coefficient is negative, and the moderator temperature coefficient of reactivity is non-positive for power operating conditions, thereby providing negative reactivity feedback characteristics.
The NRC staff finds that this criterion remains satisfied because removal of control rod H-08 does not impact the ability to detect or control core power distribution, and the at-power nuclear reactivity feedback coefficients remain unchanged.
GDC 12 - Suppression of Reactor Power Oscillations The licensee states that power oscillations of the fundamental mode are inherently eliminated by the negative Doppler and non-positive moderator temperature coefficient of reactivity.
Oscillations, due to xenon spatial effects, in the radial, diametral and azimuthal overtone modes are heavily damped due to the inherent design and due to the negative Doppler and non-positive moderator temperature coefficients of reactivity. Oscillations due to xenon spatial effects in the axial first overtone mode may occur. Assurance that fuel design limits are not exceeded by xenon axial oscillations is provided as a result of reactor trip functions using the measured axial power imbalance as an input. Oscillations due to xenon spatial effects in axial modes higher than the first overtone are heavily damped due to the inherent design and due to the negative Doppler coefficient of reactivity.
The NRC staff finds that this criterion remains satisfied as the previous safety analysis for Cycle 24 with control rod H-08 removed demonstrated that it will not result in power oscillations, which would result in conditions exceeding specified acceptable fuel design limits. While the Cycle 25 reload analysis will not be completed until shortly before the next refueling outage, the existing core reload design process using NRC approved methods assures that the fuel design limits will be met; therefore, the NRC staff finds that the requirements of GDC 12 continue to be met.
GDC 23 - Protection System Failure Modes The licensee stated that the Protection System is designed with due consideration of the most probable failure modes of the components under various perturbations of the environment and energy sources. Each reactor trip channel is designed on the de-energize-to-trip principle; thus loss of power, disconnection, open channel faults, and the majority of internal channel short-circuit faults cause the channel to go into its tripped mode. The removal of control rod H-08 from the reactor vessel does not impact the fail-safe function of the remaining 52 control rods, which will still reliably maintain an adequate reactor shutdown capability.
The NRC staff finds that this criterion remains satisfied by maintaining the control rod insertion capability with the remaining 52 control rods.
GDC 25 - Protection System Requirements for Reactivity Control Malfunctions The licensee stated that the Protection System is designed to limit reactivity transients so that fuel design limits are not exceeded. The previous Cycle 24 redesign reload analyses, performed according to methods referenced in TS 5.6.3, confirmed that the fuel design limits are not exceeded. The same analysis will be performed for the Cycle 25 cores.
The NRC staff finds that this criterion remains satisfied as the reactor trip function remains fully capable of performing its function with 52 control rods, and fuel design limits were not exceeded for previously analyzed malfunctions of the reactivity control systems with the removal of control rod H-08 for Cycle 24. While the Cycle 25 reload analysis will not be completed until shortly before the next refueling outage, the existing core reload design process using NRC-approved methods assures that the fuel design limits will be met; therefore, the NRC staff finds that the requirements of GDC 25 continue to be met.
GDC 26 - Reactivity Control System Redundancy and Capability The licensee stated that two Reactivity Control Systems are provided, including the rod cluster control assemblies and chemical shim (boration). The RCCA are inserted into the core by the force of gravity. The boron chemical shim is unaffected and will maintain the reactor in the cold shutdown state independent of the position of the control rods and can compensate for all xenon burnout transients.
The NRC staff finds that this criterion remains satisfied as the licensees previous Cycle 24 analysis has demonstrated that removal of control rod H-08 does not impact the ability of the reactivity control system to perform its function. While the Cycle 25 reload analysis will not be completed until shortly before the next refueling outage, the existing core reload design process using NRC-approved methods assures that the fuel design limits will be met; therefore, the NRC staff finds that the requirements of GDC 26 continue to be met.
GDC 27 - Combined Reactivity Control Systems Capability The licensee stated that reactivity control is achieved by a combination of RCCA and automatic boron addition via the ECCS with the most reactive control rod assumed to be fully withdrawn.
Manually controlled boric acid addition is used to supplement the RCCA in maintaining the shutdown margin for the long-term conditions of xenon decay and plant cooldown.
The NRC staff finds that this criterion remains satisfied with the removal of control rod H-08 as the previous licensee analysis for Cycle 24 has demonstrated that the ability of the reactivity control systems to reliably control reactivity changes and that adequate SDM is maintained when considering the highest stuck rod worth. Licensee evaluations of the removal of control rod H-08 during Cycle 24 demonstrate that SDM and safety analysis limits are met throughout the fuel cycle. While the Cycle 25 reload analysis will not be completed until shortly before the next refueling outage, the existing core reload design process using NRC-approved methods assures that the fuel design limits will be met; therefore, the NRC staff finds that the requirements of GDC 27 continue to be met.
GDC 28 - Reactivity Limits The licensee stated that the appropriate reactivity insertion rate for the withdrawal of RCCA and the dilution of the boric acid are controlled by the TSs. The specification includes or references appropriate graphs that show the permissible mutual withdrawal limits and overlap of functions of the several RCCA banks as a function of power.
The NRC staff finds that this criterion remains satisfied as the licensees previous Cycle 24 analysis with removal of control rod H-08 demonstrates trip reactivity insertion rate, SDM, and the safety analysis limits remain met for the UFSAR Chapter 15 analysis. While the Cycle 25 reload analysis will not be completed until shortly before the next refueling outage, the existing core reload design process using NRC approved methods assures that the fuel design limits will be met; therefore, the NRC staff finds that the requirements of GDC 28 continue to be met.
GDC 29 - Protection Against Anticipated Operational Occurrences The licensee stated that the Protection and Reactivity Control Systems are designed to ensure an extremely high probability of fulfilling their intended functions. The design principles of diversity and redundancy coupled with a rigorous Quality Assurance Program and analyses support this probability, as does operating experience in plants using the same basic design.
The staff finds that this criterion remains satisfied as the removal of control rod H-08 does not impact the ability of the reactivity control systems to perform their safety functions. The mechanical removal of the control rod drive shaft and RCCA do not have any mechanical impact on the function of the remaining 52 control rods. The remaining 52 control rods are also not impacted by the related electrical changes when control rod H-08 is removed. Therefore, the staff finds that there remains a high probability control rod insertion continues to exist under anticipated operational occurrences, even with the removal of the H-08 control rod through Cycle 25.
3.3 Technical Conclusion The licensee proposed to modify the Note associated with TS 4.2.2 to allow operation with 52 full-length control rod assemblies (with no control rod assembly installed in core location H-08) for one additional cycle. The NRC staff concludes that the licensees proposed use of 52 control rod assemblies in Sequoyah, Units 1 and 2 for Cycle 25, including the use of a flow restrictor, is acceptable because the design change is consistent with the current design basis and is not expected to challenge the safety analyses detailed in Chapter 15 of the UFSAR.
Because the licensees technical specifications are derived from its UFSAR analyses, and the proposed change does not adversely affect the licensees UFSAR analyses or change any applicable TS Limiting Condition for Operation and Surveillance Requirements, the NRC staff finds that the TS, as amended by the proposed change, will continue to meet the requirements in 10 CFR 50.36. The NRC staff concludes that the licensee used methods consistent with regulatory requirements and guidance identified in Section 2.3 above. The NRC staff also finds the proposed use of 52 control rod assemblies continues to meet the requirements of GDC 10, 11, 12, 14, 23, 25, 26, 27, 28, and 29.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Tennessee State official was notified of the proposed issuance of the amendments on January 21, 2021. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20, Standards for Protection Against Radiation. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on August 11, 2020 (85 FR 48572).
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 amendments.
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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: R. Beaton Y. Wong Dated: March 3, 2021
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