ML110350155

Issuance of Amendment Regarding the Control Rod Assemblies (Tac ME3991)

ML110350155
Person / Time
Site:	Watts Bar
Issue date:	02/25/2011
From:	John Lamb Watts Bar Special Projects Branch
To:	Krich R Tennessee Valley Authority
Lamb J, 415-1727
References
TAC ME3991
Download: ML110350155 (20)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 25, 2011 Mr. R. M. Krich Vice President, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga,TN 37402-2801 SUB~IECT:

WATTS BAR NUCLEAR PLANT, UNIT 1 -ISSUANCE OF AMENDMENT REGARDING THE CONTROL ROD ASSEMBLIES (TAC NO. ME3991)

Dear Mr. Krich:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 86 to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant, Unit 1. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated May 28,2010, as supplemented December 1, 2010.

The amendment revises TS 4.2.2, "Control Rod Assemblies," to include silver-indium-cadmium material in addition to the boron carbide control rod material.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

n G. Lamb, Senior Project Manager tts Bar Special Projects Branch vision of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosures:

1. Amendment No. 86 to NPF-90

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 86 License No. NPF-90

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the Tennessee Valley Authority (the licensee) dated May 28, 2010, as supplemented December1, 2010, 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 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.

- 2

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 Facility Operating License No. NPF-90 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 86, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance, and shall be implemented no later than 30 days from the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~~.~

Stephen J. Campbell, Chief Watts Bar SpeCial Projects Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-90 and the Technical Specifications Date of Issuance: February 25, 2011

ATTACHMENT TO LICENSE AMENDMENT NO. 86 FACILITY OPERATING LICENSE NO. NPF-90 DOCKET NO. 50-390 Replace page 3 of Operating License No. NPF-90 with the attached page 3.

Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

REMOVE INSERT 4.0-1 4.0-1

- 3 (4)

TVA, 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, instrument calibration, or other activity associated with radioactive apparatus or components; and (5)

TVA, 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 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 TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 86 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Safety Parameter Display System (SPDS) (Section 18.2 of SER Supplements 5 and 15)

Prior to startup following the first refueling outage, TVA shall accomplish the necessary activities, provide acceptable responses, and implement all proposed corrective actions related to having the Watts Bar Unit 1 SPDS operational.

(4)

Vehicle Bomb Control Program (Section 13.6.9 of SSER 20)

During the period of the exemption granted in paragraph 2.D.(3) of this license, in implementing the power ascension phase of the approved initial test program, TVA shall not exceed 50% power until the requirements of 10 CFR 73.55(c)(7) and (8) are fully implemented. TVA shall submit a letter under oath or affirmation when the requirements of 73.55(c)(7) and (8) have been fully implemented.

Amendment No. 86

4.0 DESIGN FEATURES 4.1 Site 4.1.1 Site and Exclusion Area Boundaries The site and exclusion area boundaries shall be as shown in Figure 4.1-1.

4.1.2 Low Population Zone (LPZ)

The LPZ shall be as shown in Figure 4.1-2 (within the 3-mile circle).

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy or Zirlo fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) 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. For Unit 1, Watts Bar is authorized to place a maximum of 704 Tritium Producing Burnable Absorber Rods into the reactor in an operating cycle.

4.2.2 Control Rod Assemblies The reactor core shall contain 57 control rod assemblies. The control material shall be either silver-indium-cadmium or boron carbide with silver indium cadmium tips as approved by the NRC.

(continued)

Watts Bar Unit 1 4.0-1 Amendment 8,40,48, 67, 77, 86

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 86 TO FACILITY OPERATING LICENSE NO. NPF-90 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-390

1.0 INTRODUCTION

By application dated May 28, 2010 (Agencywide Document Management Systems (ADAMS)

Accession No. ML101540065), as supplemented December 1, 2010 (ADAMS Accession No. ML103400311), Tennessee Valley Authority (TVA) requested an amendment to Facility Operating License No. NPF-90 Watts Bar Nuclear Plant (WBN), Unit 1 and Appendix A, Technical Specifications (TSs), of the Facility Operating License. The proposed change would revise TS 4.2.2, "Control Rod Assemblies," by including silver-indium-cadmium (Ag-In-Cd) material in addition to the boron carbide (B4C) control rod material. The WBN Unit 1 existing Rod Cluster Control Assemblies (RCCAs) are configured with a hybrid design B4C absorber material with Ag-In-Cd tips. TVA proposes to replace these RCCAs with Enhanced Performance (EP) RCCAs containing only Ag-In-Cd as the absorber material. This RCCA material replacement necessitates a change to TS 4.2.2, "Control Rod Assemblies" to include Ag-In-Cd material in addition to B4C control rod material.

TVA's supplementary submittal dated December 1, 2010, provided clarifying information that did not change the scope of the proposed amendment as described in the original notice of proposed action published in the Federal Register on July 27,2010 (75 FR 44026) and did not change the initial proposed no significant hazards determination.

2.0 REGULATORY EVALUATION

2.1

System Description

The RCCAs each consist of a group of individual neutron absorber rods fastened at the top end to a common hub or spider assembly. These assemblies contain full length neutron absorber material to control the reactivity of the core under operating conditions.

The CRDMs are of the magnetic jack type. Control rods are positioned by electro-mechanical (solenoid) action utilizing gripper latches, which engage grooved drive rods which in turn are coupled to the RCCAs. The CRDMs are so designed that upon a loss of electrical power to the coils, the RCCA is released and falls by gravity to shutdown the reactor.

- 2 The RCCAs are divided into two categories: control and shutdown. The control groups compensate for reactivity changes due to variations in operating conditions of the reactor, power and temperature variations. Two criteria have been employed for selection of the control group.

First, the total reactivity worth must be adequate to meet the nuclear requirements of the reactor. Second, in view of the fact that these rods may be partially inserted at power operation, the total power peaking factor should be low enough to ensure that the power capability is met.

The control and shutdown group provides adequate shutdown margin, which is defined as the amount of negative reactivity by which the core would be subcritical at hot shutdown if all RCCAs are tripped, assuming that the highest worth assembly remains fully withdrawn and assuming no changes in xenon or boron concentration.

The current absorber materials used in the control rods are B4 C pellets plus Ag-In-Cd alloy slugs that are essentially "black" to thermal neutrons. These materials have sufficient additional resonance absorption to Significantly increase their worth. The B4 C pellets are stacked on top of the extruded Ag-In-Cd slugs and are sealed in stainless steel tubes to prevent them from coming in direct contact with the coolant. In construction, the B4C pellets and the Ag-In-Cd slugs are inserted into cold-worked stainless steel tubing that is then sealed at the bottom and the top by welded end plugs. Sufficient diametral and end clearance is provided to accommodate relative thermal expansions and material swelling.

2.2 Description of Proposed Change Currently, TS 4.2.2 states: The reactor core shall contain 57 control rod assemblies. The control material shall be boron carbide with silver indium cadmium tips as approved by the NRC."

The proposed change is to revise TS 4.2.2 to state: The reactor core shall contain 57 control rod assemblies. The control material shall be either silver-indium-cadmium or boron carbide with silver indium cadmium tips as approved by the NRC."

The switch from the Hybrid B4C design RCCAs to the EP Ag-In-Cd RCCAs returns WBN Unit 1 to the original design basis of Ag-In-Cd RCCAs.

2.3 Regulatory ReqUirements The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the license amendment request (LAR) to evaluate the TS amendment request and to confirm that the use of methodologies is within NRC approved ranges of applicability and verify that the results of the analyses are in compliance with the requirements of the following General Design Criteria (GDC) specified in Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50.

WBN Unit 1 was designed to meet the intent of the "Proposed General Design Criteria for Nuclear Power Plant Construction Permits" published in July 1967. The WBN Unit 1 construction permit was issued in January 1973. The WBN Unit 1 Updated Final Safety Analysis Report (UFSAR), however, addresses the NRC GOC published as Appendix A to 10 CFR Part 50 in July 1971, including Criterion 4 as amended October 27, 1987.

- 3 The following are the GDC and regulation that apply for this LAR.

GDC-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 (SAFDL) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs)." The reactor core with its related coolant, control, and protection systems is designed to function throughout its design lifetime without exceeding acceptable fuel damage limits. The reactor trip system is designed to actuate a reactor trip for any anticipated combination of plant conditions when necessary to ensure that fuel design limits are not exceeded. The core design, together with reliable process and decay heat removal systems, provides for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and anticipated transient situations, including the effects of loss of reactor coolant flow, trip of the turbine-generator, loss of normal feedwater and loss of both normal and preferred power sources.

GDC-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 SAFDL are not possible or can be reliably and readily detected and suppressed." Power oscillations of the fundamental mode are inherently eliminated by the negative Doppler and non positive moderator temperature coefficients (MTCs) 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 MTCs. 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.

GDC-13, Instrumentation and Control, "Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for AOOs, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges."

Instrumentation and controls are provided to monitor and control neutron flux, control rod position, temperatures, pressures, flows, and levels as necessary to assure that adequate plant safety can be maintained. Instrumentation is provided in the reactor coolant system (RCS),

steam and power conversion system, the containment, engineered safety features systems, radiological waste systems and other auxiliaries. Parameters that must be provided for operator use under normal operating and accident conditions are indicated in the control room in proximity with the controls for maintaining the indicated parameter in the proper range.

GDC-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 AOOs, and with appropriate margin for malfunctions such as stuck rods, SAFDL are

- 4 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." Two reactivity control systems are provided. These are RCCAs and chemical shim (boric acid). The RCCAs are inserted into the core by the force of gravity. During operation, the shutdown rod banks are fully withdrawn. The full length control rod system automatically maintains a programmed average reactor temperature compensating for reactivity effects associated with scheduled and transient load changes. The shutdown rod banks along with the full length control banks are designed to shutdown the reactor with adequate margin under conditions of normal operation and AOOs thereby ensuring that SAFDL are not exceeded. The most restrictive period in core life is assumed in all analyses and the most reactive rod cluster is assumed to be in the fully withdrawn position. The boron system will maintain the reactor in the cold shutdown state independent of the position of the control rods and can compensate for xenon burnout transients.

GDC-27, Combined reactivity control system 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." Sufficient shutdown capability is provided to maintain the core subcritical for any anticipated cooldown transient (e.g., accidental opening of a steam bypass or relief valve, or safety valve stuck open). This shutdown capability is achieved by a combination of RCCA insertion and automatic boron addition via the emergency core cooling system 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.

GDC-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." The maximum reactivity worth of control rods and the maximum rate of reactivity insertion employing control rods and boron removal are limited to values that prevent rupture of the RCS boundary or disruptions of the core or vessel internals to a degree that could impair the effectiveness of emergency core cooling. The appropriate reactivity insertion rate for the withdrawal of RCCA and the dilution of the boric acid in the RCS are specified in the TSs for WBN Unit 1.

GDC-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 AOOs." The protection and reactivity control systems are designed to assure an extremely high probability of fulfilling their intended functions. The design prinCiples of diversity and redundancy coupled with a rigorous Quality Assurance

- 5 Program and analyses support accomplishing this probability as does operating experience in plants using the same basic design.

10 CFR 50.36(c)(4), Design features, states that design features to be included 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 10 CFR 50.36(c) (1), (2), and (3).

2.4 Important Precedents RCCAs with Ag-In-Cd material are currently used in many plants. As such, a large amount of operating experience has been gained with Ag-In-Cd used as an absorber material. Ag-In-Cd has proven to be an effective absorber material and RCCA's with Ag-In-Cd have shown very good operating results.

The following plants have RCCAs with Ag-In-Cd material: Indian Point Unit 1 - August 30, 1965; Oconee Units 1, 2, and 3 - October 23, 1978; Davis Besse - October 3, 1988; ANO Unit 1 November 8,1988; Callaway Unit 1 - February 14,1989; Millstone Unit 3 - June 28, 1989; Byron Units 1 and 2 and Braidwood Units 1 and 2 - July 18, 1989; South Texas Units 1 and 2 July 31, 1989; Vogtle Units 1 and 2 - February 20, 1990; and Wolf Creek - August 22, 1991.

3.0 TECHNICAL EVALUATION

3.1 Introduction WBN Unit 1 existing RCCAs are configured with a hybrid design B4 C absorber material with Ag-In-Cd tips. TVA intends to replace the RCCAs with EP RCCAs containing only Ag-In-Cd as absorber material. This necessitates a change to TS 4.2.2, "Control Rod Assemblies."

Currently, TS 4.2.2 states:

The reactor core shall contain 57 control rod assemblies. The control material shall be boron carbide with silVer indium cadmium tips as approved by the NRC.

TVA proposes to revise the TS 4.2.2 to state:

The reactor core shall contain 57 control rod assemblies. The control material shall be either silver-indium-cadmium or boron carbide with silver indium cadmium tips as approved by the NRC.

The WBN Unit 1 reactor core is comprised of an array of 193 fuel assemblies that are similar in mechanical design, but different in fuel enrichment. Each fuel assembly consists of 264 fuel rods that are mechanically joined in a square array. The fuel rods are supported in intervals along their length by grid assemblies that maintain the lateral spacing between the rods throughout the design life of the assembly. The center position in the assembly is reserved for the incore instrumentation, while the remaining 24 positions in the array are equipped with guide thimbles joined to the grids and the top and bottom nozzles. Depending upon the position of the

- 6 assembly in the core, the guide thimbles are used as channels for insertion of RCCAs, neutron source assemblies, and burnable absorber rods. OthelWise, the guide thimbles are fitted with plugging devices to limit bypass flow.

The RCCAs each consist of a group of individual neutron absorber rods fastened at the top end to a common hub or spider assembly. These assemblies contain full length neutron absorber material to control the reactivity of the core under operating conditions. The absorber materials used in the control rods are B4C with Ag-In-Cd alloy slugs that are essentially "black" to thermal neutrons. Some of the existing RCCAs have reached the end of life and are required to be replaced. TVA is planning to replace these RCCAs with EP RCCAs that contains only Ag-In-Cd as the absorber materials.

In addition to the absorber material change, the replacement EP Ag-In-Cd RCCAs will be coupled with Control Rod Drive Mechanism (CRDM) drive rod shafts that are lighter than the current CRDM drive shaft coupled to the B4C RCCAs and will have a different reactivity or rod worth. The NRC staff evaluated changes in weight and reactivity of the CRDM drive shaft and the RCCAs, based on TVA's evaluation of the impacted areas demonstrating the safe operation of the plant with the replacement RCCAs with EP Ag-In-Cd absorber material.

3.2 Mechanical Evaluation 3.2.1 Ag-In-Cd Rods Ag-In-Cd alloy has been used as a predominant absorber material in the RCCA rodlets in Westinghouse designed pressurized-water reactors (PWRs), both domestically and internationally. Currently WBN Unit 1 is the only domestic Westinghouse PWR plant that does not use Ag-In-Cd as an absorber in the RCCA rodlets; instead, WBN Unit 1 uses B4C absorber.

TVA proposes to use the EP Ag-In-Cd absorber material in the WBN Unit1 RCCAs starting in spring 2011. From a mechanical perspective, the EP Ag-In-Cd RCCAs are advantageous for a number of reasons including a relatively high thermal conductivity and reduced swelling concerns since irradiation produces isotopes rather than gases. The proposed RCCAs utilize three new features that improve the performance of RCCAs relative to the existing hybrid B4C RCCA design at the WBN Unit 1. These features are (1) high purity rodlet cladding material, (2) an increase in the gap between the absorber tip and the cladding, and (3) industrial hard chrome plating on the surface of the control rodlets that improve margins to applicable wear criteria.

The improvement in the purity of the materials in EP RCCAs enhances rodlet corrosion resistance relative to the current RCCAs in the WBN Unit 1. A small increase in the gap between Ag-In-Cd portion of the absorber and the inner surface of the cladding in the EP Ag-In-Cd RCCAs at the lower end of the rodlet accommodates the possible swelling and minimizes the absorber-to-rodlet interaction thereby reduces absorber induced strain of the rod let.

The increase in weight of the EP RCCAs over the existing hybrid RCCAs is from 94 pounds (Ib) to 149 lb. TVA has demonstrated that the residual kinetic energy resulting from rapid insertion

- 7 (scram) of the heavier RCCAs is sufficiently absorbed to prevent impact damage to the fuel assemblies.

In support of the above design changes for the EP Ag-In-Cd RCCAs, TVA performed analyses and evaluations that demonstrate all of the mechanical design criteria are satisfied, including wear characteristics, sufficient rod to thimble tube diameter gaps, no surface boiling, maintaining rod centerline temperatures below acceptable limits, as well as ensuring that Conditions I, II, III, and IV loads meet all applicable stress and fatigue criteria.

3.2.2 Control Rod Drive Mechanism In addition to the absorber material change, the proposed EP RCCAs are coupled with CRDM drive rod shafts that are lighter than the existing CRDM drive rod shafts. As an individual component, the heavy drive weighs 170 Ibs and the standard drive weighs 136 Ibs. The main advantage of a lighter CRDM drive is that the unlatching and latching of the heavy drive rods take substantially longer than the standard drive rods, causing additional critical path time and an associated increase in dose to personnel. The heavy drive rods were intended to be mated with the lighter B4C RCCAs, and the standard drive rod was the original design for the Ag-In-Cd RCCAs. The subsequent B4 C RCCAs required heavier drive rods to keep the overall driveline weight approximately the same as the standard drive rod/Ag-ln-Cd RCCA combination.

The lighter drive rod meets the design criteria for the drive rod. The drive rod weight is specifically used in the missile analysis presented in Section 3.5.1 of the WBN Unit 1 Updated Final Safety Analysis Report (UFSAR). The conclusions of the missile analysis presented in the UFSAR remain applicable for the change to the standard lighter drive rod.

The overall driveline weight increases from 264 Ib for heavy CRDM with the B4C RCCA combination to 285 Ib for the standard CRDM with the EP Ag-In-Cd RCCA combination. The driveline weight is a consideration in the vibration and seismic analysis for the American Society of Mechanical engineers Code CRDM pressure boundary design report. The WBN Unit 1 CRDM design report is an enveloping design report for driveline weights up to 320 Ib; therefore, the current WBN Unit 1 CRDM design report envelopes the driveline weight change.

3.2.3 Rod Drop Time The combined weight of the CRDM drive rod/RCCA has an impact on the rod drop insertion time. The rod drop time is an input in to safety analyses and is also specified in the TS. The TS limit is 2.7 seconds. TVA evaluated the rod drop time for WBN Unit 1 for the past four cycles and found it to be 1.496 seconds. With the increased weight of the Ag-In-Cd RCCAs, and the drive shaft at WBN Unit 1, the rod drop time is expected to reduce by 0.20 second and will continue to meet the TS limit of 2.7 seconds.

3.2.4 Mechanical Evaluation Conclusion The principal difference that has some impact on GDC-10 is the difference in weight with regard to the EP Ag-In-Cd RCCAs. The EP Ag-In-Cd RCCAIstandard drive line weight continues to

- 8 meet the rod drop time of 2.7 seconds limit listed in TS 3.1.5, "Rod Group Alignment Limits."

The NRC staff finds that GOC-10 is satisfied.

The NRC staff finds the proposed change to revise TS 4.2.2 to include Ag-In-Cd material in addition to the B4C control rod material is acceptable regarding the mechanical design.

3.3 Nuclear Evaluation The proposed replacement of current RCCA absorber material from B4C with Ag-In-Cd results in different reactivity insertion characteristics at WBN Unit 1. TVA has performed a review of all nuclear design input parameters used in the safety analyses and identified those parameters that exhibit minimal margin to their limiting condition and were potentially impacted by the replacement EP Ag-In-Cd RCCA. These parameters are discussed in the sections below.

3.3.1 Rod Worth and Rod Insertion Limit RCCA rod worth is dependent on the fuel loading pattern and the core conditions of the measurement. Using a representative fuel loading pattern and comparing the total RCCA bank worth under the same core conditions, TVA has determined that EP Ag-In-Cd are worth approximately 7-percent less than the B4C hybrid RCCAs. This difference is nearly the same at the beginning-of-life and end-of-life. Because both RCCAs utilize Ag-In-Cd absorber material at their tips, and the replacement RCCAs are thicker than the current RCCAs, the lead bank (Control Bank 0) has higher worth with the replacement RCCAs when it is partially inserted to the power dependent insertion limits (POlL) at 100-percent power. The replacement Ag-In-Cd RCCAs are observed to be worth 10-percent more than the B4C hybrid RCCAs for this portion of lead bank insertion. The lead bank rod worths are determined based on actual calculated values using modeled Ag-In-Cd RCCAs, increased to account for appropriate uncertainties and cycle-to-cycle variation. The updated withdrawing control rod worth values bound both the current RCCAs and the replacement RCCAs.

The impact of the new rod worth values on the POlL depends on the fuel loading pattern and varies from cycle to cycle. The POlL are inputs to multiple reload safety checks and these limits are validated for plant operation. It is generally expected that the POlL will not change as a result of the RCCA replacement.

3.3.2 Shutdown Margin Shutdown margin is defined as the amount of negative reactivity by which the core would be subcritical at hot shutdown if all RCCAs are tripped, assuming that the highest worth assembly remains fully withdrawn and assuming no changes in xenon or boron concentration.

Shutdown margin is confirmed each cycle by using the cycle-specific calculations per the Westinghouse Reload Safety Evaluation Methodology that has been approved by the NRC staff.

TVA has performed a nuclear design evaluation for representative fuel loading pattern using the EP Ag-In-Cd RCCAs and showed that there is sufficient margin to meet the SOM requirement.

- 9 3.3.3 Axial Power Shape Peaking Factors The behavior of axial power shapes are used in the confirmation of several safety parameters.

The axial power shapes are impacted in an adverse manner due to changes in the reactivity insertion characteristics of the replacement Ag-In-Cd RCCAs. Axial power shapes at WBN Unit 1 are evaluated using the Relaxed Axial Offset Control (RAOC) Methodology. Power peaking factors are generated using a variety of RCCA insertions for each cycle of operation.

The RAOC analyses show that both RCCA types produce similar margin to the design limits. A plot of difference in power peaking factors versus core height generated by TVA shows that the maximum difference is 0.09 in a nonlimiting axial region of the core. For most of the core, the power peaking factors are shown to be not sensitive to the RCCA material (Le., the difference is less than 0.02). TVA evaluates and confirms power peaking factors each cycle of operation. If sufficient margin cannot be shown, TVA will restrict plant operations in order to improve the margin in the evaluation.

3.3.4 Nuclear Evaluation Conclusion The change in absorber material from B4C to Ag-In-Cd affects GDC-10. The reactivity difference was addressed for the impact on core neutronics and safety analyses. It was determined that the reactivity change can be accommodated within the bounds of the current safety analysis limits using approved NRC methodologies. Future core designs will use NRC-approved methodologies as the means to demonstrate the continued safe operation of the plant with the EP Ag-In-Cd RCCAs. The NRC staff finds that GDC-10 is satisfied.

Axial power oscillations are controlled using the RCCAs. The EP Ag-In-Cd RCCAs can be effectively used to control reactivity oscillations. The ability to reliably detect and suppress power oscillations is unaffected by the proposed changes. The NRC staff finds that GDC-12 is satisfied.

The existing systems and components used for monitoring and control of RCCA positions are unaffected by the proposed changes and will be equally effective and relied upon for the control of the EP Ag-In-Cd RCCAs. The NRC staff finds that GDC-13 remains satisfied.

Redundancy and capability for the RCCAs to control reactivity is not impacted and remains bounded by maintaining the operational restrictions. The NRC staff finds that GDC-26 remains satisfied.

The ability of the EP Ag-In-Cd RCCAs to control reactivity is not impacted and remains bounded by maintaining the operational restrictions assumed in the reload analysis and shown to be acceptable by the reload process. The NRC staff finds that GDC-28 remains satisfied since the absorber materials used in the control rod design are either (1) all Ag-In-Cd alloy rods, or (2) boron carbide pellets and Ag-In-Cd alloy slugs and the absorber materials are essentially "black" to thermal neutrons.

The NRC staff finds the proposed change to revise TS 4.2.2 to include Ag-In-Cd material in addition to the B4C control rod material meets 10 CFR 50.36(c)(4).

- 10 3.4 Safety Analyses 3.4.1 Main Streamline Break with Coincident Rod Withdrawal at Power The hot full power streamline break coincident with rod withdrawal at power accident was reanalyzed to demonstrate that the departure-from-nucleate boiling (DNB) design basis is met for RCCA absorber material change. This transient was analyzed using the Revised Thermal Design Procedure (RTDP) and the WRB-2M correlation. The DNB ratio (DNBR) results show substantial margin above the RDTP design limit DNBR of 1.23 (WBN-1 FSAR Table 4.4-1).

3.4.2 Non-Loss of Coolant Accidents (Non-LOCA)

The following non-LOCA events that are related to the RCCAs have been dispositioned by TVA based on one or more of the applicable criteria listed below:

1. The analysis used bounding control rod worths for the withdrawing banks that remain bounding with the replacement RCCAs.

2. The rod drop time currently assumed in the analysis is not impacted by the replacement RCCAs.

3. The rod insertion characteristics (reactivity insertion versus time) during a scram (plant trip) for the replacement RCCAs are bounded by the insertion characteristics for the current RCCAs, confirmed on a reload basis consistent with Westinghouse report WCAP-9273-P-A.

List of Non-LOCA Events Dispositioned and the Applicable Criteria Rod withdrawal at subcritical (1, 2, and 3)

Rod withdrawal at power (1, 2, and 3)

Partial loss of flow (2 and 3)

• Loss of load/turbine trip (2 and 3)

• Loss of normal feedwater and loss of offsite power (2 and 3)

Feedwater malfunction (2 and 3)

Excessive load increase (2 and 3)

RCS depressurization (2 and 3)

• Complete loss of flow (2 and 3)

Feedline break (2 and 3)

Locked rotor (2 and 3)

Rod ejection (2 and 3) 3.4.3 Loss of Coolant Accidents The current licensing basis small break LOCA (SBLOCA) analysis for WBN Unit 1 uses 10 CFR Part 50 Appendix K evaluation model. The Appendix K SBLOCA analysis is impacted by changes in RCCA rod drop time and/or RCCA geometry. However, the replacement of B4C RCCAs with heavy CRDMs with EP Ag-In-Cd RCCAs does not change the specific RCCA geometry modeled in the current SBLOCA analysis. In addition, the new RCCA rod drop time is

- 11 within the TS limit. Therefore, the CRDM and RCCA change is bounded by the SBLOCA 10 CFR Part 50 Appendix K evaluation model.

The current best estimate LOCA (BELOCA) analysis does not credit or model control rod insertion. Therefore, the replacement RCCAs will not impact the BELOCA licensing basis.

3.4.4 Safety Analyses Conclusion The current design of the reactor control systems includes an adequate capability for reactivity control using only the EP Ag-In-Cd RCCAs. The design of the replacement EP Ag-In-Cd RCCAs meet all the same performance requirements of the current design and will not introduce any new effect that could adversely impact the performance of the RCCAs.

Therefore, the reactivity control systems remain capable 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. The NRC staff finds that GDC-27 is satisfied.

It was determined that the reactivity change can be accommodated within the bounds of the current safety analysis limits using approved NRC methodologies. Future core designs will use NRC approved methodologies as the means to demonstrate the continued safe operation of the plant with the EP Ag-In-Cd RCCAs.

The replacement EP Ag-In-Cd RCCAs have been evaluated by TVA against design criteria applicable to the existing RCCAs, and it was determined that they will function as required during AOOs (Le., uncontrolled RCCA withdrawal from a subcritical condition, uncontrolled RCCA withdrawal at power, RCCA misalignment, loss of external electrical load and/or turbine trip, loss of offsite power to the station auxiliaries, excessive heat removal due to feedwater system malfunctions, excessive load increase incident, and accidental depressurization of the RCS). The Ag-In-Cd RCCAIstandard drive line weight continues to meet the rod drop time of 2.7 seconds limit listed in TS 3.1.5. The NRC staff finds that GDC-29 is satisfied.

The NRC staff finds the proposed change to revise TS 4.2.2 to include Ag-In-Cd material in addition to the B4C control rod material is acceptable regarding the safety analyses.

3.5 Summary and Conclusions The NRC staff has evaluated the LAR for replacement RCCAs based on mechanical, nuclear, and safety analyses provided by TVA. The NRC staff has determined that TVA has demonstrated compliance with all of the mechanical design criteria for the replacement RCCAs.

The NRC staff finds that WBN Unit 1 is in compliance with all the mechanical design criteria for the replacement RCCAs. The NRC staff has reviewed TVA's evaluation of nuclear design and safety analyses and has determined that the nuclear design models and safety analysis methodologies used in these evaluations for the replacement EP Ag-IN-Cd RCCAs are consistent with NRC-approved methodologies. Therefore, the NRC staff finds the proposed change to revise TS 4.2.2 to include Ag-In-Cd material in addition to the B4C control rod material is acceptable.

- 12

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to 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 no significant hazards consideration, and there has been no public comment on such finding (75 FR 44026). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0

'CONCLUSION The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's 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 Contributor: Mathew Panicker Date: February 25, 2011

February 25, 2011 Mr. R. M. Krich Vice President, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 SUB.JECT:

WATTS BAR NUCLEAR PLANT, UNIT 1 -ISSUANCE OF AMENDMENT REGARDING THE CONTROL ROD ASSEMBLIES (TAC NO. ME3991)

Dear Mr. Krich:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 86 to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant, Unit 1. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated May 28, 2010, as supplemented December 1, 2010.

The amendment revises TS 4.2.2, "Control Rod Assemblies,>> to include silver-indium-cadmium material in addition to the boron carbide control rod material.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRA!

John G. Lamb, Senior Project Manager Watts Bar Special Projects Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosures:

1, Amendment No. 86 to NPF-90

2. Safety Evaluation cc w/encls: Distribution via Listserv NRC Distribution: See next page ADAMS Accession No. ML110350155

• via memorandum wwNo IlegaI 0 b'lJecfIon sub'lJect to comments OFFICE NRRlLPWB/PM NRRlLPWB/LA DSS/SNPB/BC DIRS/ITSB/BC DSS/SRXB/BC OGC NRRlLPWB/BC NAME JLamb BClayton AMendiola*

RElliott AUlyses MSpencer-SCampbell DATE 02/23/11 02/23/11 02107111 02111/11 02/09/11 2122111 2/25/11 OFFICIAL AGENCY RECORD

Letterto R. M. Krich from John G. Lamb dated February 25, 2011

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNIT 1 -ISSUANCE OF AMENDMENT REGARDING THE CONTROL ROD ASSEMBLIES (TAC NO. ME3991)

NRC DISTRIBUTION:

PUBLIC LPWB rtf RidsNrrDirsltsb Resource RidsNrrDorlDpr Resource RidsNrrDorlLp_WB Resource RidsNrrPMWattsBar1 Resource RidsNrrLABClayton Resource RidsNrrDssSrxb Resource RidsNrrDssSnpb Resource RidsOgcRp Resource RidsRgn2MailCenter Resource RidsAcrsAcnw_MailCTR Resource MPanicker, NRRlDSS/SNPB GWaig, NRRIDIRSIITSB