ML11194A290

License Amendment, Issuance of Amendments Regarding Administrative Changes to Technical Specifications 3.12 and 6.2

ML11194A290
Person / Time
Site:	Surry
Issue date:	07/28/2011
From:	Cotton K Plant Licensing Branch II
To:	Heacock D Virginia Electric & Power Co (VEPCO)
Cotton K, NRR/DORL/LPL2-1, 415-1438
References
TAC ME4349, TAC ME4350
Download: ML11194A290 (20)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 July 28, 2011 Mr. David A. Heacock President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING ADMINISTRATIVE CHANGES TO TECHNICAL SPECIFICATIONS 3.12 AND 6.2 (TAC NOS. ME4349 AND ME4350)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 275 to Renewed Facility Operating License No. DPR-32 and Amendment No. 275 to Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2, respectively. The amendments change the Technical Specifications (TSs) in response to your application dated July 12, 2010, (Agencywide Documents Access and Management System (ADAMS), Accession No. ML101950070).

These amendments revise administrative changes that: 1) correct an error in TS 3.12. E.5,

2) delete duplicative requirements in TS 3.12.E.2 and TS 3.12.EA, 3) relocate the shutdown margin value in TS 3.12 and the TS 3.12 Basis to the Core Operating Limits Report (COLR), and

4) expand the TS 6.2 list of parameters defined in the COLR.

A copy of the safety evaluation is also enclosed. The notice of issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Karen Cotton, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-280 and 50-281

Enclosures:

1. Amendment No. 275 to DPR-32

2. Amendment No. 275 to DPR-37

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO.1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 275 Renewed License No. DPR-32

1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated July 12, 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 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 0) 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 3.B of Renewed Facility Operating License No. DPR-32 is hereby amended to read as follows:

(B)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 275

, 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 within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. DPR-32 and the Technical Specifications Date of Issuance: July 28, 2011

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 275 Renewed License No. DPR-37

1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated July 12, 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 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 (ij) 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 3.B of Renewed Facility Operating License No. DPR-37 is hereby amended to read as follows:

(B)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 275

, 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 within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes License No. DPR-37 and the Technical Specifications Date of Issuance: July 28, 2011

ATIACHMENT TO LICENSE AMENDMENT NO. 275 RENEWED FACILITY OPERATING LICENSE NO. DPR-32 DOCKET NO. 50-280 AND TO LICENSE AMENDMENT NO. 275 RENEWED FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NO. 50-281 Replace the following pages of the licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License License No. DPR-32, page 3 License No. DPR-32, page 3 License No. DPR-37, page 3 License No. DPR-37, page 3 TSs TSs TS 3.12-2 TS 3.12-2 TS 3.12-11 TS 3.12-11 TS 3.12-12 TS 3.12-12 TS 3.12-13 TS 3.12-13 TS 6.2-1 TS 6.2-1

3. This renewed license shall be deemed to contain and is sUb)ectlo tt1e conditions specified in Ihe foHewing Commission regula1lons: '0 CFR Pan 20, Section 30;34 of 10 CFR Part 30, Sadlon 40.41 of 10 erR Part 40, Sedilns 50.54 and 50.59 of 10 CFR Pen SO, and Section 70.32 of 10 CFR Part 70: lind is subject to ali applicable provisions of tl'le ACI end Ike rules, regulations, and ortlers of the Commission row or heresfter in effect; and is 5ubjec1 to the additional ccndfllons speclfiad lJelow:

A. Maijmum ?ower Leve!

The IJoensee is authorized to operate the facility 81 steady sl81e reactor core power levels not In excess of 2587 megawatts (tl'lermal).

B. Technical Soegiflcations The Technical SPftcificalions c:ontaned In Appendix A, 85 revised through Amendment No *. 275 are hereby InC:Olllorsted in the ",newed license. Th~

Hcensee shaH oPerate the facIlity In accordance with 1he Technical SpecifIcations.

C. Re~

the licensee shall make certain reports in sct::Oroance witI'! the requirements of the Technica' Specifications.

D. Records The licensee shall keep facility operating records in accordance with the requirements of 1he Tech1"llCSI Specifications.

E. *Deleted by Amendment 65 F, DelieteO by Amendment 7:

G. Deleted by AmenOmcnt 227 H. Deleted by AmerKiment 227 I.

Firs Protection The licensee shari Implement and ma main in effec! the pmvisio'ns of the aporoved tire protection. program 8S tl8scrlbed In the Updated Ftnal Safety AnalysIs RePOM'snCl as approved 11'\\ t"18 SER oated September '9, '979, {and Suppiements dateo May 29,1960, O,*tober e, 19/3Q, December ~8. 1980,

ebfuary '3, 198', Decemoer~. 198', A.pril 27,,ga2, November 1e, 1982.

January n, 191)4, February 25, ~9BB, Bnd SURRY UNIT ~

Rtml!!wl!d Licen;e No D:;IP.*32 Amendmer.t

~Jo. 275

-3 E. Pursuant to the Act and W CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facUlty,

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regUIa1tons: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Pan 70; and is subject \\0 all applicable provisions of the Act and the rules, 'regulations; and orders of the Commission now or hereafter In effect; and is subject to the additional conditions specified below:

A. Maximum Power Level The licensee is authorized to ooorate the facility at steady state reactor core Dower levels not in excess of 2537 megawatts (thermalj.

B. Technical Specifica]ions The Technical Sf"I"r:ifications contained in Appendix A, as revised through Amendment No..275

,are hereby incorporated in this renewed license. The licensee shall operate the facility in accordance with the TeChnical Specifications.

C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specl*ncations.

D. Records The licensee shall keep facility operating recorOs Iii accordance with the requirements of the Technical Specifications.

E. Deleted by Amendment 54 F. Deleted by Amendment 59 and Amendment.65 G. Deleted by Amendment 227 H. Deleted by Amendment 227 SURRY - UNIT 2 Renewed License No. OPR*37 Amendment No.

275

TS 3.12-2

a. The sequence of withdrawal of the control banks, when going from zero to 100% power, is A, B, C, D.

b. An overlap of control banks, consistent with physics calculations and physics data obtained during unit startup and subsequent operation, will be permitted.

c. The shutdown margin with allowance for a stuck control rod assembly shall be within the limits specified in the CORE OPERATING LIMITS REPORT under all steady-state operation conditions, except for physics tests, from zero to full power, including effects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be subcritical at HOT SHUTDOWN (Tavg ~ 547°F) if all control rod assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or boron.

4. Whenever the reactor is subcritical, except for physics tests, the critical control rod assembly position, i.e., the control rod assembly position at which criticality would be achieved if the control rod assemblies were withdrawn in normal sequence with no other reactivity changes, shall not be lower than the insertion limit for zero power.

5. Insertion limits do not apply during physics tests or during periodic surveillance testing of control rod assemblies. However, the shutdown margin indicated above must be maintained except for the LOW POWER PHYSICS TEST to measure control and shutdown bank worth and shutdown margin. For this test the reactor may be critical with all but one full length control rod assembly, expected to have the highest worth, inserted.

6. With a maximum of one control or shutdown bank inserted beyond the insertion limit specified in Specification 3.12.A.2 during control rod assembly testing pursuant to Specification 4.1, and imraovable due to a failure of the Rod Control System, POWER OPERATION Amendment Nos.

275 and 275

TS 3.12-11 E. Rod Position Indication System and Bank Demand Position Indication System

1. From movement of control banks to achieve criticality and with the REACTOR CRITICAL, rod position indication shall be provided as follows:

a. Above 50% power, the Rod Position Indication System shall be OPERABLE and capable of determining the control rod assembly positions to within +/- 12 steps of their respective group step demand counter indications.

b. From movement of control banks to achieve criticality up to 50% power, the Rod Position Indication System shall be OPERABLE and capable of determining the control rod assembly positions to within +/- 24 steps of their respective group step demand counter indications for a maximum of one hour out of twenty-four, and to within +/- 12 steps otherwise.

c. From movement of control banks to achieve criticality and with the REACTOR CRITICAL, the Bank Demand Position Indication System shall be OPERABLE and capable of determining the group demand positions to within

+/- 2 steps.

2. If one rod position indicator per group for one or more groups is inoperable, the position of the control rod assembly shall be verified indirectly using the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Alternatively, reduce power to less than 50% of RATED POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During operations below 50% of RATED POWER, no special monitoring is required.

Amendment Nos. 275 and 275

TS 3.12-12

3. If more than one rod position indicator per group is inoperable, place the control rods under manual control immediately, monitor and record RCS Tavg once per hour, verify the position of the control rod assemblies indirectly using the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and restore inoperable position indicators to OPERABLE status such that a maximum of one position indicator per group is inoperable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4. If one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction since the last detennination of the rod's position, verify the position of the control rod assemblies indirectly using the movable incore detectors within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce power to less than SO% of RATED POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

S. If one group step demand counter per bank for one or more banks is inoperable, verify that all rod position indicators for the affected bank(s) are OPERABLE once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and verify that the most withdrawn rod and the least withdrawn rod of the affected bank(s) are less than or equal to 12 steps apart once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Alternatively, reduce power to less than 50% of RATED POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

6. If the requirements of Specification 3.12.E.2, 3.12.E.3, 3.12.E.4, or 3.12.E.5 are not satisfied. then the unit shall be placed in HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

F. DNB Parameters

1. The following DNB related parameters shall be maintained within their limits during POWER OPERA nON:

• Reactor Coolant System Tavg 5. 577.0°F

• Pressurizer Pressure ~ 220S psig

• Reactor Coolant System Total Flow Rate ~ 2 i3,000 gpm

a. The Reactor Coolant System Tavg and Pressurizer Pressure shall be verified to Amendment Nos.

275 and 275

TS 3.12-13 Basis The reactivity control concept assumed for operation is that reactivity changes accompanying changes in reactor power are compensated by control rod assembly motion. Reactivity changes associated with xenon, samarium, fuel depletion, and large changes in reactor coolant temperature (operating temperature to COLD SHUTDOWN) are compensated for by changes in the soluble boron concentration. During POWER OPERATION, the shutdown control rod assemblies are fully withdrawn and, control of power is by the control banks. A reactor trip occurring during POWER OPERATION will place the reactor into HOT SHUTDOWN. The control rod assembly insertion limits provide for achieving HOT SHUTDOWN by reactor trip at any, time, assuming the highest worth control rod assembly remains fully withdrawn, with sufficient margins to meet the assumptions used in the accident analysis. In addition, they provide a limit on the maximum inserted control rod assembly worth in the unlikely event of a hypothetical assembly ejection and provide for acceptable nuclear peaking factors. The limit may be detennined on the basis of unit startup and operating data to provide a more realistic limit which will allow for more flexibility in unit operation and still assure compliance with the shutdown requirement.

The maximum shutdown margin requirement occurs at end of core life and is based on the value used in the analyses of the hypothetical steam break accident. The control rod assembly insertion limits are based on end of core life conditions. The shutdown margin for the entire cycle length shall be within the limits specified in the CORE OPERATING LIMITS REPORT. Other accident analyses with the exception of the Chemical and Volume Control System malfunction analyses are based on 1 % reactivity shutdown margin. Relative positions of control banks are detennined by a specified control bank overlap. This overlap is based on the consideration of axial power shape control. The specified control rod assembly insertion limits have been established to limit the potential ejected control rod assembly worth in order to account for the effects of fuel densification. The various control rod assemblies (shutdown banks, control banks A, B, C, and D) are each to be moved as a bank; that is, with each assembly in the bank within one step (5/8 inch) of the bank position.

The axial position of shutdown rods and control rods are determined by two separate and independent systems: the Bank Demand Position Indication System (commonly called the group step demand counters) and the Rod Position Indication System.

The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one group step demand counter for each group of rods. Individual Amendment Nos. 275 and 275

TS 6.2-]

6.2 GENERAL NOTIFICATION AND REPORTING REOUIREMENTS Specification A. The following action shall be taken for Reportable Events:

A report shall be submitted pursuant to the requirements of Section 50.73 to 10 CFR.

B. Immediate notifications shall be made in accordance with Section 50.72 to 10 CFR.

C. CORE OPERATING LIMITS REPORT Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.

Parameter limits for the following Technical Specifications are defined in the CORE OPERATING LIMITS REPORT:

]. TS 3.1.E - Moderator Temperature Coefficient

2. TS 3.12.A.l, TS 3.12.A.2 and TS 3.12.A.3 - Control Bank Insertion Limits

3. TS 3.12.B.1 and TS 3.12.B.2 - Power Distribution Limits

4. TS 3.12.A.1.a, TS 3.12.A.2.a, TS 3.12.A.3.cand TS 3.12.0 - Shutdown Margin Amendment Nos. 275 and 275

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 275 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 275 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281

1.0 INTRODUCTION

By letter dated July 12, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML101950070), Virginia Electric and Power Company (the licensee) submitted a request for changes to the Surry Power Station, Unit Nos. 1 and 2 (Surry 1/2),

Technical Specifications (TSs).

The proposed changes would 1) correct an error in TS 3.12.E.5, 2) delete duplicative requirements in TS 3.12.E.2 and TS 3.12.E.4, 3) relocate the shutdown margin value in TS 3.12 and the TS 3.12 Basis to the Core Operating Limits Report (COLR), and 4) expand the TS 6.2 list of parameters defined in the COLR.

2.0 REGULATORY EVALUATION

2.1 Rod Position Indication Reactivity changes accompanying changes in reactor power at Surry Power Station are compensated by, among other things, control rod assembly motion. The axial position of control rods are determined by two separate and independent systems: the bank demand position indication system (commonly called the group step demand counters) and the rod position indication system. The bank demand position indication system counts the pulses from the rod control system that move the rods. There is one group step demand counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step demand counter for that group.

The rod position indication system provides an accurate indication of actual rod position. This system is based on inductive analog signals from a series of coils spaced along a hollow tube.

The requirements on the rod position indicators and the group step demand counters are only

-2 applicable from the movement of control banks to achieve criticality and with the reactor critical, because these are the only conditions in which the rods can affect core power distribution and in which the rods are relied upon to provide required shutdown margin.

Group demand step counters and the rod position indication system read the axial position of control rods. It is important for them to remain operable so that the current position can be read and is being read accurately. If a group demand step counter or rod position indication system is not reading accurately. core power distribution and, therefore, shutdown margin could be affected.

In association with these systems, the licensee proposes two revisions. First, the licensee proposes to decrease the number of inoperable group step demand counters per bank needed before verifying the operability of all rod position indicators. Second, to relax the time frame associated with verifying control rod position if a rod's position has changed greater than 24 steps and the rod has an inoperable rod position indicator.

Section 182a. of the Atomic Energy Act of 1954, as amended, requires all applicants for nuclear power plant licenses to include technical specifications as part of the license. Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 36, "Technical Specifications," implements section 182a. of the Atomic Energy Act.

As originally codified in 1968, TSs requirements were placed on two general classes of technical matters: (1) those related to prevention of accidents, and (2) those related to mitigation of the consequences of accidents (Reference 1). TSs revisions must meet the regulatory framework outlined in 10 CFR 50.36(c)(2)(i) to be considered acceptable. This requirement states:

Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

2.2 Core Operating Limits Report Relocation The shutdown margin (SDM) is the amount of negative reactivity by which a reactor is maintained in a subcritical state at hot zero power (HZP) conditions after a reactor trip. Shutdown margin is calculated by determining the amount of negative reactivity available (control and shutdown bank worth) and finding the excess available once the positive reactivity associated with going from hot full power (HFP) to HZP conditions has been overcome. This amendment request proposes to relocate the value for shutdown margin from the TSs to the COLR and expand the list of parameters defined in the COLR TSs.

NRC Generic Letter 88-16, "Removal of Cycle Specific Parameter Limits from Technical Specifications," (Reference 2) provides guidance for licensees, allowing relocation of cycle dependant variables from the TSs provided that the values of these variables are included in a COLR and are determined with NRC-approved methodologies contained in a reporting requirement in the TSs. The variables removed from the TS and relocated to the COLR can then be changed via the appropriate regulatory change mechanisms, principally 10 CFR 50.59, thereby avoiding the need to frequently change the TSs.

- 3 Regulations in 10 CFR 50.36 specify the categories and criteria for information that must be included in the TSs. These include the following: (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. Along with meeting guidelines of Generic Letter 88-16, cycle dependant variables may be removed from the TSs and put into the COLR, if appropriate safety limits are maintained in the TSs so that the regulatory requirements of 10 CFR 50.36 are met. Changes to the COLR parameters are reported to the NRC so that they can be monitored and trended.

3.0 TECHNICAL EVALUATION

The changes associated with this license amendment request (LAR) are described below.

3.1 Correction of Text Error The licensee proposes to correct an error by eliminating the words "more than" in TS 3.12.E.5.

TS 3.12, "Control Rod Assemblies and Power Distribution Limits," contains the requirements for the operation of control rod assemblies and power distribution limits to ensure core subcriticality after a reactor trip, a limit on potential reactivity insertions from hypothetical control rod assembly ejection, and an acceptable core power distribution during power operation. The text of TS 3.12.E.5 currently reads:

If one group demand counter per bank for more than one or more banks is inoperable, verify that all rod position indicators for the affected bank(s) are OPERABLE once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and verify that the most withdrawn rod and the least withdrawn rod of the affected bank:(s) are less than or equal to 12 steps apart once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Alternatively, reduce power to less than 50% [percent] of RATED POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The proposed change would allow it to read:

If one group demand counter per bank for one or more banks is inoperable, verify that all rod position indicators for the affected bank(s) are OPERABLE once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and verify that the most withdrawn rod and the least withdrawn rod of the affected bank: (s) are less than or equal to 12 steps apart once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Alternatively, reduce power to less than 50%

of RATED POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Interpretation of the phrase before the revision is such that "more than one or more banks" means action must be taken when two or more banks are inoperable, whereas interpretation of the phrase after the revision is such that "one or more banks" means action is taken when just one bank is inoperable. The revision creates a more restrictive requirement, therefore, the NRC staff concludes that this change is conservative, has no adverse impact on plant safety, and is acceptable.

3.2 Revision of Conflicting TS Requirements The licensee proposes to revise conflicting TS requirements pertaining to TS 3.12.E.2 and TS 3.12. E.4 by specifically, deleting the phrase "and immediately after any motion of the

-4 non-indicating control rod assembly exceeding 24 steps" from TS 3.12.E.2. TS 3.12.E.2, which currently states:

If one rod position indicator per group for one or more groups is inoperable, the position of the control rod assembly shall be verified indirectly using the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and immediately after any motion of the non-indicating control rod assembly exceeding 24 steps. Alternatively, reduce power to less than 50% of RATED POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During operations below 50% of RATED POWER, no speCial monitoring is required.

will be revised to state:

If one rod position indicator per group for one or more groups is inoperable, the position of the control rod assembly shall be verified indirectly using the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Alternatively, reduce power to less than 50% of RATED POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During operations below 50% of RATED POWER, no special monitoring is required.

This would eliminate conflicting requirements with TS 3.12. EA, which reads:

If one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction since the last determination of the rod's position, verify the position of the control rod assemblies indirectly using the movable incore detectors within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce power to less than 50% of RATED POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Verifying the position of the control rod assembly immediately if any motion of the control rod assembly exceeds 24 steps is being changed to requiring this action be taken within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as stated in TS 3.12.E.4. Deleting the specified phrase from TS 3.12.E.2 creates a less restrictive requirement. Even with this change, TS 3.12.E.2 and TS 3.12.EA retain an alternative action of reducing rated power to less than 50 percent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, which remains the same. With the deletion made in TS 3.12.E.2, the first action in TS 3.12.E.4 of verifying the position of the control rod assembly within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if any motion of the control rod assembly exceeds 24 steps is still less restrictive than reducing rated power to 50 percent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Even though the first action is creating a less restrictive time requirement, the required action is still overall more restrictive than the unchanged alternative action of reducing rated power to 50% within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The intent of LCOs and required actions is to ensure an acceptable level of safety if an LCO is not met. Because the proposed change to the LCO ultimately retains an unchanged least restrictive action (reduce power to 50 percent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />), the staff concludes that the proposed LCO revision still retains an acceptable level of safety, consistent with 10 CFR 50.36(c)(2)(i). This revision eliminates conflicting requirements, clarifies required actions, assures that appropriate safety limits are maintained in the TS and is, therefore, considered acceptable from a regulatory standing.

- 5 3.3 Relocation of the Shutdown Margin Value from TS to the COLR This licensee proposes to relocate the value for shutdown margin from the TS to the COLR. The method used by Surry Generating Power to determine shutdown margin is VEP-FRD-42-A, "Reload Nuclear Design Methodology" (Reference 3). The staff reviewed the reference to confirm shutdown margin is, in fact, calculated in this methodology. The shutdown margin is a cycle specific parameter determined by analysis performed in accordance with NRC approved methodologies meeting the guidance in Generic Letter 88-16. Therefore, the staff finds this change acceptable.

3.4 Expansion of the List of Parameters Defined in the COLR TS In order to relocate parameters from the TSs to the COLR, the licensee must (1) use NRC approved methodology, and (2) list what parameters are being determined under administrative control in the TSs. Section 3.3, "Relocation of the Shutdown Margin Value from the TSs to the COLR," describes how the licensee has used NRC approved methodology to relocate parameters from TS to the COLR. Parameters being put under administrative control are described in this section.

The licensee proposes to add technical specifications related to Shutdown Margin as item "4" in TS 6.2.C "Core Operating Limits Report." These TS include TS 3.12.A.1.a, TS 3.12.A.2.a, TS 3.12.A.3.c, and TS 3.12.G. It also proposes to expand the list of TS for Control Bank Insertion Limits by adding TS 3.12.A.1 to item "2" in TS 6.2.C. Generic Letter 88-16 outlines certain TS limits that could be removed from TS and placed into licensee controlled documents. Section 6.2.C of Surry Power Station Units 1 and 2 TS provides a listing of core operating limits that are required to be documented in the COLR. Adding the control bank insertion limits and shutdown margin parameters to TS 6.2.C will be consistent with the other core operating limits currently specified. This change is administrative in nature, makes the TS more conservative by adding additional references to the core operating limits parameter listing, will have no impact on plant safety, and is, therefore, acceptable.

3.5 CONCLUSION

The NRC Staff has reviewed the proposed changes to TS 3.12.E.5, TS 3.12.E.2, TS 3.12.A.3.c, and TS 6.2.C, "Core Operating Limits Report," based on applicable regulatory requirements outlined in 10 CFR 50.36 and Generic Letter 88-16. Based on the considerations discussed above, the NRC staff finds that the requested changes remain compliant with 10 CFR 50.36 and are adherent to Generic Letter 88-16 guidance. Therefore, the NRC staff finds that the changes made in this amendment will have no impact on plant safety and are, therefore, acceptable.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements 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

- 6 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 (May 17, 2011; 76 FR 28477). The amendments also relate to changes in recordkeeping, reporting, or administrative procedures or requirements.

Accordingly, the amendments meet the eligibility criteria for categorical exclusions set forth in 10 CFR 51.22(c)(9) and (c)(10). 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) 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.

7.0 REFERENCES

1.

U.S. Nuclear Regulatory Commission, "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (10 CFR Part 50)," Federal Register, Vol. 58, No. 139, pp. 39133-39139.

2.

U.S. Nuclear Regulatory Commission, Generic Letter 88-16, "Removal of Cycle Specific Parameter Limits from Technical Specifications," ADAMS Accession No. ML031130447.

3.

Virginia Electric and Power Company, "Reload Nuclear Design Methodology,"

VEP-FRD-42-A, August 29, 2003. ADAMS Accession No. ML032680720.

Principal Contributor: Ashley Guzzetta Date: July 28, 2011

ML11194A290 NRR/LPL2-1/LA SS/SRXB/BC OGC (NLO)

NRR/LPL2-1/BC NRR/LPL2-1/PM KCotton 07/14/2011 07/20/2011 07/28/11 07/28/11