ML15068A386

Issuance of an Amendment on Rod Position Indication Technical Specification Surveillance Requirements

ML15068A386
Person / Time
Site:	Robinson
Issue date:	05/27/2015
From:	Martha Barillas Plant Licensing Branch II
To:	Glover R Duke Energy Progress
Barillas M DORL/LPL2-2 301-415-2760
References
TAC MF1966
Download: ML15068A386 (17)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Richard Michael Glover Site Vice President H. B. Robinson Steam Electric Plant Duke Energy 3581 West Entrance Road Hartsville, SC 29550 May 27, 2015

SUBJECT:

H. B. ROBINSON STEAM ELECTRIC PLANT UNIT NO. 2 - ISSUANCE OF AMENDMENT ON ROD POSITION INDICATION TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS (TAC NO. MF1966)

Dear Mr. Glover:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 241 to Renewed Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant Unit No. 2 (HBR). This amendment changes the HBR Technical Specifications (TSs) in response to Duke Energy Progress, lnc.'s application dated June 7, 2013, as supplemented by letter dated July 24, 2014.

This amendment deletes Surveillance Requirements (SRs) 3.1.7.1, 3.1.7.2, and 3.1.7.3 of HBR TS 3.1.7, "Rod Position Indication," and renumbers SR 3.1.7.4 as SR 3.1.7.1.

The NRC staff's related safety evaluation is enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Docket No. 50-261

Enclosures:

1. Amendment No. 241 to DPR-23

2. Safety Evaluation cc w/enclosures: Distribution via Listserv Sincerely, IRA!

Martha Barillas, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY PROGRESS, INC.

DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 241 Renewed License No. DPR-23

1.

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

A.

The application for amendment by Duke Energy Progress, Inc. (the licensee),

dated June 7, 2013, as supplemented by letter dated July 24, 2014, 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 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.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 3.8. of Renewed Facility Operating License No. DPR-23 is hereby amended to read as follows:

8. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 241 are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days.

Attachment:

Changes to Renewed Operating License No. DPR-23 and the Technical Specifications Date of Issuance: May 27, 2015 FOR THE NUCLEAR REGULA TORY COMMISSION Shana R. Helton, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO. 241 RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace page 3 of Renewed Facility Operating License No. DPR-23 with the attached page 3.

Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages 3.1-18 3.1-19 Insert Pages 3.1-18 3.1-19 D.

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; E.

Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

3.

This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; 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:

A.

Maximum Power Level The licensee is authorized to operate the facility at a steady state reactor core power level not in excess of 2339 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 241 are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

(1)

For Surveillance Requirements (SRs) that are new in Amendment 176 to Final Operating License DPR-23, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 176. For SRs that existed prior to Amendment 176, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 176.

Renewed Facility Operating License No. DPR-23 Amendment No. 241

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.1. 7.1 Perform CHANNEL CALIBRATION of the ARPI System.

HBRSEP Unit No. 2 3.1-18 Rod Position Indication 3.1.7 FREQUENCY 18 Months Amendment No. 241

PAGE IS INTENTIONALLY BLANK HBRSEP Unit No. 2 3.1-19 Rod Position Indication 3.1.7 Amendment No. 241

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 241 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DUKE ENERGY PROGRESS, INC.

H.B. ROBINSON STEAM ELECTRIC PLANT UNIT NO. 2 DOCKET NO. 50-261

1.0 INTRODUCTION

By letter dated June 7, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13182A019), as supplemented by letter dated July 24, 2014 (ADAMS Accession No. ML14219A010), Duke Energy Progress, Inc. (the licensee), submitted a request for changes to the H. B. Robinson Steam Electric Plant Unit No. 2 (HBR), Technical Specifications (TSs). The licensee requested to delete Surveillance Requirements (SRs) 3.1.7.1, 3.1.7.2, and 3.1.7.3 of HBR 3.1.7, "Rod Position Indication," and renumber SR 3.1.7.4 as SR 3.1.7.1. The changes would delete redundant requirements and eliminate a minimum of eight reactivity manipulations per year.

By letter dated July 8, 2014 (ADAMS Accession No. ML14183B598), the U.S. Nuclear Regulatory Commission (NRC or the Commission) staff requested additional information. By letter dated July 24, 2014, the licensee responded to the request. The supplemental letter dated July 24, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on August 20, 2013 (78 FR 51222).

2.0 REGULATORY EVALUATION

2.1 Description of the HBR Design and Technical Specifications Chapter 1 of the HBR Updated Final Safety Analysis Report (UFSAR) states that the control rods consist of clusters of stainless steel clad absorber rods and guide tubes located within the fuel assembly. The reactor is controlled by a coordinated combination of chemical shim and mechanical control rods. The automatic rod control system is designed to restore programmed average temperature following a scheduled or transient step change in load of 10 percent or a ramp change of 5 percent per minute within the range of 15 to 100 percent power.

Chapter 7 of the UFSAR states that there are 45 full length rod control cluster assemblies (RCCAs), which are divided into a shutdown group comprising two shutdown banks of 8 rod clusters each, and a control group comprising 4 control banks containing 8, 8, 8, and 5 rod clusters. Each control bank is divided into subgroups. Position indication for each RCCA type is the same. The automatic rod control system maintains the average coolant temperature by adjusting the RCCA positions.

Chapter 3 of the UFSAR states that the control rod drive mechanisms (CRDMs) are used for withdrawal and insertion of the RCCAs into the reactor core and to provide sufficient holding power for stationary support. Each assembly is an independent unit. Each drive is welded onto an adaptor on top of the reactor pressure vessel and is connected to the control rod (directly below) by means of a grooved drive shaft. Reactor coolant fills the pressure containing parts of the drive mechanism. All working components and the shaft are immersed in the reactor coolant. Three magnetic coils, which form a removable electrical unit and surround the rod drive pressure housing, induce magnetic flux through the housing wall to operate the working components. They move two sets of latches that lift or lower the grooved drive shaft. The three magnets are turned on and off in a fixed sequence by solid-state switches for the full length rod assemblies. The sequencing of the magnets produces step motion over the 144 inches of normal control rod travel in 5/8 inch increments, or approximately 230 steps from fully inserted to fully withdrawn.

Chapter 7 of the UFSAR states that HBR has separate digital and analog rod position indication (RPI). The digital system counts pulses generated in the rod drive control system. One counter is associated with each group (or subgroups) of RCCAs. Readout of the digital system is in the form of electronic add-subtract counters reading the number of steps of rod withdrawal with one display for each group or subgroup. These readouts are mounted on the control panel.

Additionally, an analog signal is produced for each RCCA by a linear position transmitter. An electrical coil stack is placed above the stepping mechanisms of the control rod magnetic jacks external to the pressure housing. When the associated control rod is at the bottom of the core, the magnetic coupling between the primary and secondary coils of the linear position system is small, and there is a small voltage induced in the secondary. As the control rod is raised by the magnetic jacks, the relatively high permeability of the lift rod causes an increase in magnetic coupling, producing an analog signal proportional to within plus or minus 3 percent of rod position. Direct, continuous readout of every RCCA position is presented to the operator by individual meter indications, without need for operator selection or switching to determine rod position. A deviation monitor alarm is actuated if an individual rod deviates from its group position by a preselected distance.

By letter dated October 24, 1997 (ADAMS Accession No. ML020560172), the NRC issued Amendment No. 176 to the licensee that approved HBR to convert from its custom TSs to the Improved Standard Technical Specifications (ISTSs). The ISTSs were based on NUREG-1431, "Standard Technical Specifications, Westinghouse Plants," Revision 0, dated September 1992, and included Technical Specification Task Force Travelers used in the issuance of NUREG-1431, Revision 1, dated April 1995.

The TS 3.1. 7 requirements ensure the operability of the control rod position indicators to determine control rod positions and thereby ensure compliance with control rod alignment and insertion limits.

The HBR TS Bases for TS 3.1.7 state that the RCCAs, or rods, are moved out of the core (up or withdrawn) or into the core (down or inserted) by their CRDMs. The TS Bases also state that the axial position of the shutdown and control rods is determined by two separate and independent systems: the Bank Demand Position Indication System (DPIS, commonly called group step counters) and the Analog Rod Position Indication System (ARPI). The limiting condition of operation (LCO) in HBR TS 3.1.7 is for the ARPI and the DPIS to be operable. The HBR TS Bases state that this LCO is required to ensure operability of the control rod position indicators' capability to determine control rod positions and thereby ensure compliance with control rod alignment and insertion limits. TS 3.1. 7 has four SRs associated with it:

SR 3.1.7.1 requires the licensee to perform a CHANNEL CHECK by comparing analog RPI and bank demand position indication. The frequency of this SR is every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following more than 6 inches of rod motion when the rod position deviation monitor is inoperable. This SR is required to be met for bank positions greater than or equal to 200 steps. The HBR TS Bases state that the CHANNEL CHECK of RPI is a comparison of the rod position indicated on ARPI channels and bank DPIS channels.

SR 3.1.7.2 requires the licensee to verify each ARPI is within 7.5 inches of the average of the individual ARP ls in the associated bank after moving each full length RCCA bank greater than or equal to 19 steps and returning the banks to their original positions. The frequency of this SR is every 31 days. This SR is required to be met for bank positions less than 200 steps.

SR 3.1. 7.3 requires the licensee to verify each ARPI is within 15 inches of the associated bank demand position after moving each full length RCCA bank greater than or equal to 19 steps and returning the banks to their original positions. The frequency of this SR is every 31 days. This SR is required to be met for bank positions of greater than or equal to 200 steps.

SR 3.1. 7.4 requires the licensee to perform a CHANNEL CALIBRATION of the ARPI system every 18 months.

The LCO for TS 3.1.4, "Rod Group Alignment Limits," is for all shutdown and control rods to be OPERABLE. The LCO also states that each individual indicated rod position shall be within 15 inches of its bank demand position for bank demand positions greater than or equal to 200 steps and within 7.5 inches of the average of the individual rod positions in the bank for bank demand positions less than 200 steps. The TS Bases for TS 3.1.4 state that the limits on control rod alignment and the monitoring of rod positions during power operation ensure that the power distribution and reactivity limits defined by the design power peaking and shutdown margin limits are preserved. The TS Bases state that control rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirements of a minimum shutdown margin. TS 3.1.4 has three SRs associated with it, two of which are described as follows:

SR 3.1.4.1 requires the licensee to verify individual rod positions within the alignment limit every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter when the rod position deviation monitor is inoperable.

SR 3.1.4.2 requires the licensee, every 92 days, to verify rod freedom movement (trippability) by moving each rod not fully inserted in the core greater than or equal to 10 steps in either direction.

2.2 Description of the Proposed Changes In its letter dated June 7, 2013, the licensee requested that TS SRs 3.1.7.1, 3.1.7.2, and 3.1.7.3 be deleted from TS 3.1.7. The licensee also proposed renumbering SR 3.1.7.4 as SR 3.1.7.1 upon deletion of the three SRs. The licensee stated that the changes would delete redundant requirements found in SR 3.1.4.1 and SR 3.1.4.2. The licensee also stated that the changes would eliminate a minimum of eight reactivity manipulations per year.

2.3 Regulatory Review The NRC staff reviewed the licensee's application to ensure that (1) the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance activities proposed will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public. The NRC staff considered the following regulatory requirements, guidance, and licensing and design-basis information during its review of the proposed changes.

The regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities," provide the regulatory requirements for the licensing of production and utilization facilities.

Section 50.92, "Issuance of amendment," of 10 CFR, paragraph (a) states that in determining whether an amendment to a license will be issued to the applicant, the Commission will be guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate.

Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the application for a license. These TSs are derived from the plants' safety analyses. The regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36, Technical specifications." Section 50.36 of 1 O CFR requires TSs to include the following categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) SRs; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports.

Section 50.36(c)(3) of 10 CFR requires that TSs include SRs, which are requirements relating to test, calibration, or inspection, to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

Appendix A, "General Design Criteria [GDC] for Nuclear Power Plants, to 10 CFR Part 50, establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety. The GDC applicable to HBR at the time its operating license was issued (July 1970) were based on the "Proposed Appendix A to 10 CFR 50, General Design Criteria for Nuclear Power Plants," which was published in the Federal Register on July 11, 1967 (32 FR 10213). These GDC predate those provided in Appendix A to 10 CFR Part 50. The HBR UFSAR, Sections 3.1.1 and 3.1.2, describe the GDC applicable to HBR. The HBR GDC applicable to instrumentation used to monitor the status of control system components, the analog RPI system, and the bank demand position indication system, are the following:

HBR GDC-7, "Suppression of Power Oscillations," in Section 3.1.2.7 of the UFSAR states, "The design of the reactor core with its related controls and protection systems shall ensure that power oscillations, the magnitude of which could cause damage in excess of acceptable fuel damage limits, are not possible or can be readily suppressed."

HBR GDC-12, "Instrumentation and Control Systems," in Section 3.1.2.12 of the UFSAR states, "Instrumentation and controls shall be provided as required to monitor and maintain within prescribed operating ranges for the essential reactor facility operating variables."

HBR GDC-13, "Fission Process Monitors and Controls," in Section 3.1.2.13 of the UFSAR states, "Means shall be provided for monitoring or otherwise measuring and maintaining control over the fission process throughout core life under all conditions that can reasonably be anticipated to cause variations in reactivity of the core."

HBR GDC-22, "Separation of Protection and Control Instrumentation," states, "The physical arrangement of the redundant elements of the protection system are such that the probability is reduced that a single physical event will impair the vital function of the system."

HBR GDC-27, "Redundancy of Reactivity Control," in Section 3.1.2.27 of the UFSAR states, "Two independent control systems, preferably of different principles, shall be provided."

HBR GDC-31, "Reactivity Control Systems Malfunction," in Section 3.1.2.31 of the UFSAR states, "The reactor protection systems shall be capable of protecting against any single malfunction of the reactivity control system, such as unplanned continuous withdrawal (not ejection or dropout) of a control rod, by limiting reactivity transients to avoid exceeding acceptable fuel damage limits."

HBR GDC-32, "Maximum Reactivity Worth of Control Rods," in Section 3.1.2.32 of the UFSAR states, "Limits, which include reasonable margin, shall be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot: a) Rupture the RCPB [reactor coolant pressure boundary], and b) Disrupt the core, its support structures, or other vessel internals sufficiently to lose capability of cooling the core."

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the licensee's proposed changes against the regulations, design-basis information, and guidance provided in Section 2 of this safety evaluation (SE). The NRC staff reviewed the acceptability of the licensee's proposed changes to the SRs by evaluating whether the changes provide reasonable assurance of public health and safety. The NRC staff also verified that the proposed changes to the SRs assured that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

3.1 Deletion of TS SR 3.1.7.1 In its letter dated June 7, 2013, the licensee requested that TS SR 3.1. 7.1 be deleted from the TSs. This SR requires the licensee to perform a CHANNEL CHECK by comparing ARPI and bank demand position indication. The frequency of this SR is every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following more than 6 inches of rod motion when the rod position deviation monitor is inoperable. This SR is required to be met for bank positions greater than or equal to 200 steps.

The HBR TS Bases for TS 3.1.7 state that the CHANNEL CHECK of RPI is a comparison of the rod position indicated on ARPI channels and bank DPIS channels.

In its letter dated June 7, 2013, the licensee stated that this proposed change would eliminate redundant SRs. The licensee stated that in TS 3.1.4, "Rod Group Alignment Limits," SR 3.1.4.1 already requires comparison of the analog position indication of all control rods, including those at bank positions greater than or equal to 200 steps.

TS LCO 3.1.4.a states that all shutdown and control rods shall be OPERABLE, and each individual indicated rod position shall be within 15 inches of its bank demand position for bank demand positions greater than or equal to 200 steps and within 7.5 inches of the average of the individual rod positions in the bank for bank demand positions less than 200 steps. SR 3.1.4.1 requires the licensee to verify that individual rod positions are within the alignment limit. The frequency of this SR is every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter when the rod position deviation monitor is inoperable. The HBR TS Bases for TS 3.1.4 state that the maximum uncertainty of the ARPI system is plus or minus 12 steps, which corresponds to plus or minus 7.5 inches. The TS Bases also state that with an indicated deviation of 12 steps between the group step counter and the ARPI, the maximum deviation between actual rod position and the demand position could be 24 steps, or 15 inches.

The NRC staff determined that the requirements in SR 3.1.4.1 encompass the requirements in SR 3.1. 7.1. SR 3.1.4.1 requires verification of the position of all rods rather than a subset of rods, as required by SR 3.1.7.1. In addition, the surveillance frequency of SR 3.1.4.1 is the same as (when the rod deviation monitor is operable) or more frequent than (when the rod position deviation monitor is inoperable) that of SR 3.1.7.1. The NRC staff concluded that if the requirements of SR 3.1.4.1 are met, then the requirements of SR 3.1.7.1 are met; therefore, SR 3.1. 7.1 is redundant to SR 3.1.4.1 and can be deleted from the TSs. Because the requirements in SR 3.1.4.1 encompass the requirements in SR 3.1. 7.1, the NRC staff also concludes that operation with the proposed change will continue to be in conformance with the GDC from the HBR UFSAR that are listed in Section 2.3 of this SE. Based on the NRC staff's review of the licensee's amendment request, the NRC staff concludes that deleting SR 3.1.7.1 from the TSs is acceptable.

3.2 Deletion of TS SR 3.1.7.2 and SR 3.1.7.3 In its letter dated June 7, 2013, the licensee requested that TS SR 3.1.7.2 be deleted from the TSs. This SR requires the licensee to verify each ARPI is within 7.5 inches of the average of the individual ARPls in the associated bank after moving each full length RCCA bank greater than or equal to 19 steps and returning the banks to their original positions. This SR is required to be met every 31 days for bank positions less than 200 steps.

In its letter dated June 7, 2013, the licensee requested that TS SR 3.1.7.3 be deleted from the TSs. This SR requires the licensee to verify each ARPI is within 15 inches of the associated bank demand position after moving each full length RCCA bank greater than or equal to 19 steps and returning the banks to their original positions. This SR is required to be met every 31 days for bank positions of greater than or equal to 200 steps.

In its letter dated June 7, 2013, the licensee stated that these proposed changes would result in less frequent required rod motion as well as less rod motion being required during performance of the surveillance to confirm ARPI system and DPIS capability to respond to rod motion. The licensee also stated that the deletion of these SRs would eliminate a monthly reactivity manipulation to meet the 19-step rod motion requirement.

TS LCO 3.1.4.a states that all shutdown and control rods shall be OPERABLE, and each individual indicated rod position shall be within 15 inches of its bank demand position for bank demand positions greater than or equal to 200 steps and within 7.5 inches of the average of the individual rod positions in the bank for bank demand positions less than 200 steps. SR 3.1.4.2 requires the licensee to verify rod freedom of movement (trippability) by moving each rod not fully inserted in the core greater than or equal to 10 steps in either direction every 92 days. The TS Bases state that exercising each individual control rod every 92 days provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit. The TS Bases also state that moving each control rod by 10 steps will not cause radial or axial power tilts, or oscillations, to occur. In its letter dated June 7, 2013, the licensee stated that moving each full length RCCA bank greater than or equal to 19 steps and returning the banks to their original positions is considered a reactivity management event and requires a reduction in reactor power to less than 98 percent of full power to perform SR 3.1.7.2.

To assist with its determination of the acceptability of having an SR that requires rod movement of at least 10 instead of 19 steps, the NRC staff requested, in its letter dated July 8, 2014, that the licensee confirm if the proposed changes had an effect on the accuracy of the RPI system.

In its response dated July 24, 2014, the licensee stated that the accuracy of the RPI system is based on the physical configuration associated with the coil stack and the drive shaft. The frequency of system calibrations will remain at 18 months. The licensee also stated that the proposed changes do not involve any physical change to the rod control system or the RPI system and thus will not impact the accuracy or performance of the RPI system.

The TS Bases state that exercising each individual control rod every 92 days provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit. The TS Bases also state that the 92-day frequency takes into consideration other information available to the operator in the control room and information from SR 3.1.4.1, which is performed more frequently and adds to the determination of the OPERABILITY of the rods. SR 3.1.4.1 requires verification that individual rod positions are within the alignment limits every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter when the rod position deviation monitor is inoperable.

To assist with its determination of the acceptability of having an SR that requires this check every 92 days, instead of 31 days, the NRC staff requested, in its letter dated July 8, 2014, that the licensee describe the other information available to the operator in the control room. In its letter dated June 7, 2013, the licensee stated that the rod control system includes a deviation monitor, which continuously compares the outputs of the bank (group) demand (step counter) position and analog rod position, and actuates an alarm when the two position indications differ by more than a preselected value. In its letter dated July 24, 2014, the licensee also stated that RPI is fed into the plant process computer and can be monitored for changes in individual control rod position. The licensee stated that these two real-time sources of information provide plant operators accurate and continuous indication of control rod deviation and position.

The NRC staff concluded that the combination of less frequent reactivity evolutions and smaller changes in reactivity reduces potential challenges to plant operation. The NRC staff determined that the requirements in SRs 3.1.4.1 and 3.1.4.2, along with the other indications available to the operators in the control room, ensure the operability of the RPI system and the trippability of the rods. The NRC staff determined that the requirements in TS 3.1.4 are sufficient for determining control rod position indicators' capability to determine control rod positions and rod movement, and thereby ensure compliance with control rod alignment and insertion limits. Thus, operation with the proposed changes will continue to be in conformance with the GDC from the HBR UFSAR that are listed in Section 2.3 of this SE. Based on the NRC staff's review of the licensee's amendment request, the NRC staff concludes that deleting SR 3.1.7.1, SR 3.1.7.2 and SR 3.1.7.3 from the TSs is acceptable.

3.3 Renumbering of TS SR 3.1.7.4 In its letter dated June 7, 2013, the licensee requested to renumber SR 3.1.7.4 as SR 3.1.7.1 upon deletion of SRs 3.1.7.1, 3.1.7.2, and 3.1.7.3. The licensee also requested to add text to TS page 3.1-19 to indicate that the page was intentionally left blank. The NRC staff concluded that deletion of SRs 3.1.7.1, 3.1.7.2, and 3.1.7.3 is acceptable. Renumbering SR 3.1.7.4 as SR 3.1.7.1, and denoting page 3.1-19 as intentionally left blank, are conforming editorial changes. No other changes to this SR or frequency are being made. Therefore, the NRC staff concludes it is acceptable to renumber SR 3.1.7.4 as SR 3.1.7.1 and to add "PAGE IS INTENTIONALLY LEFT BLANK" text to page 3.1-19.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, on February 20, 2015, the NRC staff notified the State of South Carolina official of the proposed issuance of the amendment. 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 and changes SRs.

The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration (78 FR 51222), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the afore-mentioned considerations, 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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Jennifer M. Whitman Gursharan Singh Date: May 27, 2015

Mr. Richard Michael Glover Site Vice President H. B. Robinson Steam Electric Plant Duke Energy 3581 West Entrance Road Hartsville, SC 29550 May 27, 2015

SUBJECT:

H.B. ROBINSON STEAM ELECTRIC PLANT UNIT NO. 2 - ISSUANCE OF AMENDMENT ON ROD POSITION INDICATION TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS (TAC NO. MF1966)

Dear Mr. Glover:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 241 to Renewed Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant Unit No. 2 (HBR). This amendment changes the HBR Technical Specifications (TSs) in response to Duke Energy Progress, lnc.'s application dated June 7, 2013, as supplemented by letter dated July 24, 2014.

This amendment deletes Surveillance Requirements (SRs) 3.1.7.1, 3.1.7.2, and 3.1.7.3 of HBR TS 3.1.7, "Rod Position Indication," and renumbers SR 3.1.7.4 as SR 3.1.7.1.

The NRC staff's related safety evaluation is enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Docket No. 50-261

Enclosures:

1. Amendment No. 241 to DPR-23

2. Safety Evaluation cc w/enclosures: Distribution via Listserv DISTRIBUTION:

Sincerely, IRA!

Martha Barillas, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation PUBLIC LPL2-2 R/F RidsNrrDeEicb JWhitman, NRR RidsNrrDorllpl2-2 LRonewicz, NRR RidsNrrPMRobinson GSingh, NRR RidsNrrDorlDpr RidsNrrLABClayton RidsRgn2MailCenter EEagle, NRR RidsACRS_MailCTR RidsNrrDssStsb RidsNrrDssSrxb RidsRgn2MailCenter D

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NAME AKlett LRonewicz BClayton JThorp DATE 3/23/15 3/13/15 5/27/15 3/23/15 OFFICE DSS/STSB OGC-NLO DORL/LPL2-2/BC DORL/LPL2-2/PM NAME RElliott (MChernoff for)

DRoth SHelton MBarillas DATE 4/3/15 4/20/15 5/8/15 5/27/15 OFFICIAL RECORD COPY DSS/SRXB CJackson 3/31/15