ML16131A728

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Issuance of Amendment 268 Adopting Dominion Core Design and Safety Analysis Methods and Addressing the Issues Identified in Three Westinghouse Communication Documents
ML16131A728
Person / Time
Site: Millstone Dominion icon.png
Issue date: 07/28/2016
From: Richard Guzman
Plant Licensing Branch 1
To: Heacock D
Dominion Nuclear Connecticut
Guzman R
References
TAC MF6251
Download: ML16131A728 (69)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 28, 2016 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION, UNIT NO. 3 - ISSUANCE OF AMENDMENT ADOPTING DOMINION CORE DESIGN AND SAFETY ANALYSIS METHODS AND ADDRESSING THE ISSUES IDENTIFIED IN THREE WESTINGHOUSE COMMUNICATION DOCUMENTS (CAC NO. MF6251)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 268 to Renewed Facility Operating License No. NPF-49 for the Millstone Power Station, Unit No. 3. This amendment is in response to your application dated May 8, 2015, as supplemented by letters dated January 28, February 25, March 23, March 29, and May 2, 2016.

The amendment revises the Technical Specifications (TSs) to (1) allow the use of Dominion nuclear safety and reload core design methods; (2) allow the use of applicable departure from nucleate boiling ratio design limits for VIPRE-D; (3) update the approved reference methodologies cited in TS 6.9.1.6.b; (4) remove the base load mode of operation that is not a feature of the Dominion Relaxed Power Distribution Control power distribution control methodology; and (5) address the issues identified in Westinghouse Nuclear Safety Advisory Letter (NSAL-09-5), Rev. 1, NSAL-15-1, and Westinghouse Communication 06-IC-03.

Additionally, the amendment relocates certain equations, supporting descriptions and surveillance requirements from the TSs to licensee-controlled documents.

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

Sincerely, Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423

Enclosures:

1. Amendment No. 268 to NPF-49
2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DOMINION NUCLEAR CONNECTICUT, INC.

DOCKET NO. 50-423 MILLSTONE POWER STATION, UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 268 Renewed License No. NPF-49

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

A. The application for amendment by Dominion Nuclear Connecticut, Inc. (DNC) dated May 8, 2015, as supplemented by letters dated January 28, February 25, March 23, March 29, and May 2, 2016, 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.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-49 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 268 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DNC 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 issuance, and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULA TORY COMMISSION Travis L. Tate, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: July 28, 2016

ATTACHMENT TO LICENSE AMENDMENT NO. 268 RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove 4 4 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 Insert 1-7 1-7 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 3/4 2-9 3/4 2-9 3/42-10 3/42-10 3/42-11 3/4 2-11 3/4 2-19 3/4 2-19 3/4 2-20 3/4 2-20 6-19a 6-19a 6-20 6-20 6-20a 6-20a 6-20b

(2) Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 268 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DNC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) DNC shall not take any action that would cause Dominion Resources, Inc.

(ORI) or its parent companies to void, cancel, or diminish DNC's Commitment to have sufficient funds available to fund an extended plant shutdown as represented in the application for approval of the transfer of the licenses for MPS Unit No. 3.

(4) Immediately after the transfer of interests in MPS Unit No. 3 to DNC, the amount in the decommissioning trust fund for MPS Unit No. 3 must, with respect to the interest in MPS Unit No. 3, that DNC would then hold, be at a level no less than the formula amount under 10 CFR 50. 75.

(5) The decommissioning trust agreement for MPS Unit No. 3 at the time the transfer of the unit to DNC is effected and thereafter is subject to the following:

(a) The decommissioning trust agreement must be in a form acceptable to the NRC.

(b) With respect to the decommissioning trust fund, investments in the securities or other obligations of Dominion Resources, Inc. or its affiliates or subsidiaries, successors, or assigns are prohibited.

Except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.

(c) The decommissioning trust agreement for MPS Unit No. 3 must provide that no disbursements or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee until the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC.

(d) The decommissioning trust agreement must provide that the agreement cannot be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

Renewed License No. NPF-49 Amendment No. 268

DEFINITIONS VENTING 1.39 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

SPENT FUEL POOL STORAGE PATTERNS:

STORAGE PATTERN 1.40 STORAGE PATTERN refers to the blocked location in a Region 1 fuel storage rack and all adjacent and diagonal Region 1 (or Region 2) cell locations surrounding the blocked location. The blocked location is for criticality control.

3-0UT-OF-4 AND 4-0UT-OF-4 1.41 Region 1 spent fuel racks can store fuel in either of 2 ways:

(a) Areas of the Region 1 spent fuel racks with fuel allowed in every storage location are referred to as the 4-0UT-OF-4 Region 1 storage area.

(b) Areas of the Region 1 spent fuel racks which contain a cell blocking device in every 4th location for criticality control, are referred to as the 3-0UT-OF-4 Region 1 storage area. A STORAGE PATTERN is a subset of the 3-0UT-OF-4 Region 1 storage area.

CORE OPERATING LIMITS REPORT CCOLR) 1.42 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6. Unit Operation within these operating limits is addressed in individual specifications.

1.43 Deleted 1.44 Deleted MILLSTONE - UNIT 3 1-7 Amendment No. 39, 3{}, 69, :::;.+/-, +oo,

+89, 268

314.2 POWER DISTRIBUTION LIMITS 3/4.2. l AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

a. The limits specified in the CORE OPERATING LIMITS REPORT (COLR)
b. Deleted APPLICABILITY: MODE 1 above 50% RATED THERMAL POWER*.

ACTION:

a. With the indicated AFD outside of the applicable limits specified in the COLR,
1. Either restore the indicated AFD to within the COLR specified limits within 15 minutes, or
2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux--

High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. Deleted
c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.
  • See Special Test Exception 3.10.2 MILLSTONE - UNIT 3 314 2-1 Amendment No. -3{}, @, +/-++, 268

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.1.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel at the frequency specified in the Surveillance Frequency Control Program when the AFD Monitor Alarm is OPERABLE:
b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

4.2.1.1.2 The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside the limits.

4.2.1.1.3 Deleted 4.2.1.1.4 Deleted MILLSTONE - UNIT 3 314 2-2 Amendment No . .W, 6G, +/-§.&., 268

POWER DISTRIBUTION LIMITS 3/4.2.'J HEAT FLUX HOT CHANNEL FACTOR- FQ@

LIMITING CONDITION FOR OPERATION 3.2.2. l F 0 (Z), as approximated by F 0 M(Z), shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTION:

With F 0 (Z) exceeding its limit:

a. With Specification 4.2.2.1.2.b not being satisfied:

( l) Reduce THERMAL POWER at least 1% for each 1% F 0 (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower DT Trip setpoints have been reduced at least 1% for each 1% F 0 (Z) exceeds the limit, and (2) Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by item (1) above; THERMAL POWER may then be increased provided F 0 (Z) is demonstrated through incore mapping to be within its limits.

b. With Specification 4.2.2.1.2.c not being satisfied, all of the following ACTIONS shall be taken:

(1) a. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, control the AFD to within the new reduced AFD limits specified in the COLR that restores F0 (Z) to within its limits, and

b. Reduce the THERMAL POWER by the amount specified in the COLR that restores Fo(Z) to within its limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and
c. Reduce the Power Range Neutron Flux - High Trip Setpoints by 2:_1 % for each I% that the THERMAL POWER level is reduced within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and
d. Reduce the Overpower ~ T Trip Setpoints by 2:_ I% for each I% that the THERMAL POWER level is reduced within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and
e. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD Alarm Setpoints to the modified limits, and
f. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION b( I )b above; THERMAL POWER may then be increased pro~ided Fo(Z) is demonstrated through incore mapping to be within its hm1ts.

(2) Deleted

c. Deleted MILLSTONE - UNIT 3 314 2-5 Amendment No. W, 6G, 9-9, HG, +1G,

+/-H,ti-9-,268

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 3/4 2-6 Amendment No. 99, HG, -1{}, ~ 268

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

SURVEILLANCE REQUIREMENTS 4.2.2.1. I The provisions of Specification 4.0.4 are not applicable.

4.2.2.1.2 F 0 (Z) shall be evaluated to determine if F0 (Z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.
b. Evaluate the computed heat flux hot channel factor by performing both of the following:

(I) Determine the computed heat flux hot channel Factor, F QM(Z) by increasing the measured F 0 (Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties, and (2) Verify that FQM(z) satisfies the requirements of Specification 3 .2.2.1 for all core plane regions, i.e., 0-100% inclusive.

MILLSTONE - UNIT 3 3/4 2-7 Amendment No. 5G, @, 99, HG, 268

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

c. Verify FQ M(Z) satisfies the non-equilibrium limits specified in the COLR.
d. Measuring F Q M(Z) according to the following schedule:

( 1) Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which Fo(Z) was last determined,*** or (2) At the frequency specified in the Surveillance Frequency Control Program, whichever occurs first.

e. Compliance with the non-equilibrium limits shall be conservatively accounted for during intervals between F Q M(z) measurements by performing either of the following:

(1) Increase F0 M(z) by an appropriate factor specified in the COLR and verify that this value satisfies Specification 4.2.2.1.2.c, or (2) Verify F Q M(Z) satisfies its limits at least once per 7 Effective Full Power Days.

f. The limits specified in Specifications 4.2.2. l .2c and 4.2.2. I .2e above are not applicable in the core plane regions defined in the Bases.

4.2.2.1.3 Deleted 4.2.2.1.4 Deleted 4.2.2.1.5 When F o(Z) is measured for reasons other than meeting the requirements of Specifications 4.2.2.1.2, an overall measured Fo(Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

      • During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map outlined.

MILLSTONE - UNIT 3 314 2-8 Amendment No . .W, 6G, 99, -1:-+/-G, +G,

+/-+/-9,~268

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE- UNIT 3 314 2-9 Amendment No. -W, @, 99, 268

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 3/4 2-10 Amendment No. W, @, 99, -l--'.f.G, ;t+/-.9,

~268

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 3/4 2-11 Amendment No. ~' 69, 99, ~' .J...'.ffi, til),268

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RA TE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION N

3.2.3. l The indicated Reactor Coolant System (RCS) total flow rate and F t1H shall be maintained as follows:

a. RCS total flow rate 2 363,200 gpm and greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR), and N RTP
b. F t1H::; F t1H [ 1.0 +PF t1H( 1.0- P)]

Where:

p = THERMAL POWER 1)

RA TED THERMAL POWER' N N

2) F t1H = Measured values of F t1H obtained by using the movable incore detectors to obtain a power distribution map. The measured value of N

F t1H should be used since Specification 3.2.3.1 b. takes into consideration a measurement uncertainty of 4% for incore measurement,

3) F t1~TP =The F t1NH limit at RATED THERMAL POWER in the COLR, N
4) PF t1H =The power factor multiplier for F t1H provided in the COLR, and
5) The measured value of RCS total flow rate shall be used since uncertainties for flow measurement have been included in Specification 3.2.3.1 a.

APPLICABILITY: MODE 1.

ACTION:

N With the RCS total flow rate or F t1H outside the region of acceptable operation:

a. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:
1. Restore the RCS total flow rate to within the limits specified above and in N

the COLR and F t1H to within the above limit, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

314 2-19 Amendment No. R, W, ~, +14, +/-+-+,

MILLSTONE- UNIT 3

~,+/-4+/-,268

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION: (Continued)

b. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate that the RCS total flow rate is N

restored to within the limits specified above and in the COLR and F ~H is restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/orb., above; subsequent N

POWER OPERATION may proceed provided that F ~H and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation prior to exceeding the following THERMAL POWER levels:

1. A nominal 50% of RATED THERMAL POWER,
2. A nominal 75% of RATED THERMAL POWER, and
3. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining greater than or equal to 95% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.3.1.1 The provisions of Specification 4.0.4 are not applicable.

N 4.2.3.1.2 F ~H shall be determined to be within the acceptable range:

a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and
b. At the frequency specified in the Surveillance Frequency Control Program.

4.2.3.1.3 The RCS total flow rate shall be determined to be within the acceptable range by:

a. Verifying by precision heat balance that the RCS total flow rate is 2:: 363,200 gpm and greater than or equal to the limit specified in the COLR within 7 days after reaching 90% of RATED THERMAL POWER after each fuel loading, and MILLSTONE- UNIT 3 3/4 2-20 Amendment No. @, +9, +oo, ~'

+/-e,~268

ADMINISTRATIVE CONTROLS MONTHLY OPERA TING REPORTS 6.9.1.5 Deleted CORE OPERA TING LIMITS REPORT 6.9.1.6.a 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 for the following:

1. Reactor Core Safety Limit for Specification 2.1.1.
2. Overtemperature L1T and Overpower L1T setpoint parameters for Specification 2.2. l.
3. SHUTDOWN MARGIN for Specifications 3/4.1.1.1.1, 3/4.1.1.1.2, and 3/4.1.1.2.
4. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3.
5. Shutdown Rod Insertion Limit for Specification 3/4.1.3.5.
6. Control Rod Insertion Limits for Specification 3/4.1.3.6.
7. AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1.1.
8. Heat Flux Hot Channel Factor Limits for Specification 3/4.2.2.1.
9. RCS Total Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for Specification 3/4.2.3.1.
10. DNB Parameters for Specification 3/4.2.5.
11. Shutdown Margin Monitor minimum count rate for Specification 3/4.3 .5.
12. Boron Concentration for Specification 3/4.9.1.1.

MILLSTONE - UNIT 3 6-19a Amendment No. +/-4, ~' 69, &6, -l-8-8,

+/-+8, +/-H, +/-+/-9, B6; 268

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)

6.9.1.6.b The analytical methods used to determine the core operating limits in Specification 6.9.1.6.a shall be those previously reviewed and approved by the NRC and identified below. The CORE OPERATING LIMITS REPORT will contain the complete identification for each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT (i.e.,

report number, title, revision, date, and any supplements).

1. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," (W Proprietary). Methodology for Specifications:
  • 2.1.1 Reactor Core Safety Limits
  • 3.1.1.3 Moderator Temperature Coefficient
  • 3.1.3.5 Shutdown Rod Insertion Limit
  • 3.2.1.1 AXIAL FLUX DIFFERENCE
  • 3.2.2.1 Heat Flux Hot Channel Factor
  • 3.2.3.1 RCS Total Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor
  • 3.9.1.1 REFUELING Boron Concentration
  • 3 .2.5 DNB Parameters
2. Deleted
3. Deleted
4. WCAP-10216-P-A-RlA, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION,"

(W Proprietary). (Methodology for Specifications 3.2.1.1--AXIAL FLUX DIFFERENCE and 3.2.2.1--Heat Flux Hot Channel Factor)

5. WCAP-12945-P-A, "CODE QUALIFICATION DOCUMENT FOR BEST ESTIMATE LOCA ANALYSIS," (W Proprietary). (Methodology for Specification 3.2.2.1--Heat Flux Hot Channel Factor.)
6. WCAP-16009-P-A, "REALISTIC LARGE-BREAK LOCA EVALUATION METHODOLOGY USING THE AUTOMATED STATISTICAL TREATMENT OF UNCERTAINTY METHOD (ASTRUM)," (W Proprietary). (Methodology for Specification 3.2.2.1--Heat Flux Hot Channel Factor.)

MILLSTONE- UNIT 3 6-20 Amendment No. +/-4, ~, @, 69, 8+,

t+/-(}, +.+(}, ;H-8, ti9, +/-3(}, ~' ~ 268

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)

7. WCAP-11946, "Safety Evaluation Supporting a More Negative EOL Moderator Temperature Coefficient Technical Specification for the Millstone Nuclear Power Station Unit 3," (W Proprietary). Methodology for Specification:
  • 3 .1.1.3 - Moderator Temperature Coefficient
8. WCAP-10054-P-A, "WESTINGHOUSE SMALL BREAKECCS EVALUATION MODEL USING THE NOTRUMP CODE," (W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)
9. WCAP-10079-P-A, "NOTRUMP-A NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," (W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)
10. WCAP-12610, "VANTAGE+ Fuel Assembly Report," (W Proprietary).

(Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)

11. Deleted
12. Deleted
13. Deleted
14. Deleted
15. Deleted
16. WCAP-8301, "LOCTA-IV Program: Loss-of-Coolant Transient Analysis."

Methodology for Specification:

  • 3.2.2.1 - Heat Flux Hot Channel Factor
17. WCAP-10054-P-A, Addendum 2, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model." Methodology for Specification:
  • 3.2.2.1 - Heat Flux Hot Channel Factor
18. WCAP-8745-P-A, "Design Bases for the Thermal Overpower LiT and Thermal Overtemperature DT Trip Functions," (Westinghouse Proprietary Class 2).

(Methodology for Specifications 2.2.1 -- Overtemperature LiT and Overpower LiT Setpoints.)

MILLSTONE - UNIT 3 6-20a Amendment No. 8-l-, -!-+(), +/-1-8, +/-+/-9,

+/-M),~268

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)

19. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO',"

(W Proprietary). (Methodology for Specification 3.2.2. l - Heat Flux Hot Channel Factor.)

20. VEP-FRD-42-A, "Reload Nuclear Design Methodology." Methodology for Specifications:
  • 2.1.1 Reactor Core Safety Limits
  • 3.1.1.3 Moderator Temperature Coefficient
  • 3.1.3.5 Shutdown Rod Insertion Limit
  • 3.2.2.1 Heat Flux Hot Channel Factor
  • 3.2.3.1 Nuclear Enthalpy Rise Hot Channel Factor
  • 3.9.1.l REFUELING Boron Concentration
21. VEP-NE-1-A, "Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications." Methodology for Specifications:
  • 3.2.1.1 AXIAL FLUX DIFFERENCE
  • 3.2.2.1 Heat Flux Hot Channel Factor
22. VEP-NE-2-A, "Statistical DNBR Evaluation Methodology." Methodology for Specifications:
  • 3.2.3.1 RCS Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor
  • 3.2.5 DNB Parameters
23. DOM-NAF-2-P-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix C, "Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code." Methodology for Specifications:
  • 3.2.3. l RCS Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor
  • 3 .2.5 DNB Parameters MILLSTONE - UNIT 3 6-20b Amendment No. 268

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 268 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.

DOCKET NO. 50-423 MILLSTONE POWER STATION, UNIT NO. 3

1.0 INTRODUCTION

By application dated May 8, 2015 (Reference 1), as supplemented on January 28, 2016 (Reference 2), February 25, 2016 (Reference 3), March 23, 2016 (Reference 4), March 29, 2016 (Reference 5), and May 2, 2016 (Reference 6), Dominion Nuclear Connecticut, Inc. (DNC, the licensee) submitted a license amendment request (LAR) to revise the technical specifications (TSs) for Millstone Power Station, Unit 3 (MPS3) that would allow use of Dominion nuclear safety and reload core design methods and address the issues identified in three Westinghouse communication documents. Specifically, the proposed TS changes would:

(1) allow the use of Dominion nuclear safety and reload core design methods; (2) allow the use of applicable departure from nucleate boiling ratio (DNBR) design limits for VIPRE-D; (3) update the approved reference methodologies cited in TS 6.9.1.6.b; (4) remove the base load mode of operation that is not a feature of the Dominion Relaxed Power Distribution Control (RPDC) power distribution control methodology; and (5) address the issues identified in Westinghouse Nuclear Safety Advisory Letter (NSAL-09-5), Rev. 1, NSAL-15-1, and Westinghouse Communication 06-IC-03. Additionally, the proposed changes would involve, in part, the relocation of certain equations, supporting descriptions and surveillance requirements from the TSs to licensee-controlled documents.

The Dominion reload methods documented in the following topical reports (TRs) were previously approved by the U.S. Nuclear Regulatory Commission (NRC) for use in the reload analysis and licensing applications for Dominion nuclear plants including North Anna Power Station (NAPS), Surry Power Station (SPS) and Kewaunee Power Station (KPS).

  • VEP-FRD-42-A, Reload Nuclear Design Methodology (Reference 7)
  • VEP-NE-1-A, Relaxed Power Distribution Control Methodology (Reference 8)
  • DOM-NAF-1-P-A, Core Management System (CMS) Reactor Physics Methods (Reference 9)
  • VEP-FRD-41-P-A, RETRAN NSSS Non-LOCA Analysis (Reference 10)

Enclosure 2

  • VEP-NE-2-A, Statistical DNBR Evaluation Methodology (Reference 11)
  • DOM-NAF-2-P-A, Core Thermal-Hydraulics Using VIPRE-D (Reference 12)

The supplements dated January 28, February 25, March 23, March 29, and May 2, 2016 provided additional information that clarified the application, did not expand the scope of the application, and did not change the NRC staffs original proposed no significant hazards consideration determination as originally noticed in the Federal Register (FR), 80 FR 52804 on September 1, 2015. A subsequent notice was published in the FR on June 13, 2016 (81 FR 38226), to include the added clarification that the proposed amendment involves the relocation of TS information either to the TS Bases or the Core Operating Limits Report (COLR) which are both licensee-controlled documents. There were no changes to the no significant hazards consideration determination as originally noticed.

2.0 REGULATORY EVALUATION

2.1 Applicable Regulatory Requirements The NRC used the following requirements and guidance documents in evaluating the licensee's amendment request:

In Title 10 of the Code of Federal Regulations (10 CFR), Section 50.34, "Contents of application; technical information," the NRC established its regulatory requirements that safety analysis reports analyze the design and performance of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.

As part of the core reload process, licensees perform reload safety evaluations to ensure that their safety analyses remain bounding for the design cycle. To confirm that the analyses remain bounding, licensees confirm that the inputs to the safety analyses are conservative with respect to the current design cycle. These inputs are checked using analytical models; and if key safety analysis parameters are not bounded, further analysis of the affected transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied.

In 10 CFR 50.36, the NRC established its regulatory requirements related to the content of TSs.

The regulation at 10 CFR 50.36(a)(1) states that a summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the TSs. Paragraph 10 CFR 50.36(b) requires that each license authorizing the operation of a facility will include TSs and will be derived from the analyses and evaluation included in the safety analysis report. The categories of items required to be in the TSs are provided 10 CFR 50.36(c). As required by 10 CFR 50.36(c)(2)(i),

the TSs will include limiting conditions for operation (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met. Paragraph 10 CFR 50.36(c)(2)(ii)(B) Criterion 2 requires that an LCO be established for:

"A process variable design feature or operating restriction that is an initial condition or a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier."

Paragraph 10 CFR 50.36(c)(3) requires TSs to include items in the category of surveillance requirements, 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.

Paragraph 10 CFR 50.36(c)(5), "Administrative controls," are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

In addition to the above regulatory requirements, the following guidance documents were considered during this review:

  • NUREG-0800, Standard Review Plan, Section 16, Revision 3.0, "Technical Specifications" (Reference 13)
  • NUREG-1431, Rev. 4.0, "Standard Technical Specifications for Westinghouse Plants" (Reference 15) 2.2 Background MPS3 is a four loop pressurized water reactor (PWR) of Westinghouse design with a subatmospheric reactor containment. As part of the design basis of the plant, thermal and hydraulic characteristics are incorporated in the core design. Therefore, when it is operated with consideration for mechanical and thermal limits, in combination with plant equipment characteristics, instrumentation, and the reactor protection system, no fuel damage will occur during normal operation or abnormal operating transients.

DNC proposed changes to the power distribution limit TSs. The purpose of the power distribution limit TSs (MPS3 TS Section 3/4.2) is explained by the following excerpt from the MPS3 TS bases document:

The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the minimum DNBR [departure from nucleate boiling ratio] in the core greater than or equal to the design limit during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200°F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

Fo(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; and FNt.H Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

The use of Axial Flux Difference limits (TS 3/4.2.1) is explained by the following TS bases excerpt:

The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fo(Z) upper bound envelope of the Fo limit specified in the CORE OPERATING LIMITS REPORT (COLR) times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions. The full-length [control] rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

Dominion methods involve the use of RPDC versus the relaxed axial offset control (RAOC) or constant axial offset control (CAOC) axial power distribution methodologies more frequently used at Westinghouse PWRs for establishment of operating power distribution limits. The Dominion method involves establishment of a variable axial flux difference (delta-I) band power distribution control strategy. As power decreases the allowed delta-I band increases" ...

maintaining an approximately constant analysis margin to the design bases limits at all power levels."

Currently MPS3 employs a method of operation when at power levels below the nuclear design allowed power level (APL ND) the limits on AFD are defined in the COLR consistent with RAOC.

At power levels greater than APL ND 2 modes of operation are allowed: (1) RAOC with the AFD limits defined in the COLR; or (2) base load operation which is defined as the maintenance of the AFD within the COLR specifications band about a target value.

The Dominion power distribution control strategy uses a variable AFD delta-I band. The delta-I band is a calculated analysis output. The objective of the RPDC analysis is to determine acceptable delta-I bands that maintain margin to all the applicable design basis criteria during normal operation, abnormal operating occurrences or analyzed accidents especially a LOCA or

loss of flow accident. The calculated delta-I bands will change depending on the specific core loading pattern for the cycle and core burnup; therefore, they will be located in the COLR.

In the LAR, DNC states that these changes accomplish three key objectives:

  • Accommodate the implementation of the Dominion RPDC,
  • Removal of base load operation, and
  • Provide resolution of issues documented in Westinghouse notification documents NSAL 05, Rev. 1, 06-IC-03, and NSAL-15-1.

The same power distribution control parameters of AFD, Heat Flux Hot Channel Factor, Reactor Coolant System (RCS) Flowrate and Nuclear Enthalpy Rise Hot Channel Factor are employed in TS for either strategy to protect the fuel. The licensee states that the proposed TS changes are structured in a manner that is independent of specific power distribution control methodology (RAOC or RPDC).

In the LAR, DNC has proposed the following changes to MPS3 TSs:

  • Remove TS 1.43 definition of minimum allowable nuclear design power level for base load operation (APL ND);
  • Remove TS 1.44 definition of maximum allowable power level when transitioning to base load operation, (APL BL);
  • Change TS 3/4.2.1.1, AFD, to support adoption of Dominion's RPDC methods;
  • Change TS 3/4.2.2.1, Heat Flux Hot Channel Factor, to support adoption of Dominion's RPDC methods;
  • Change TS 3/4.2.3.1, RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor, to remove a specific uncertainty and adjust the flow rate when the precision heat balance is done; and
  • Change TS 6.9.1.6, COLR, to support adoption of Dominion's RPDC methods.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the licensee's LAR in combination with the licensee's response to the NRC's requests for additional information (RAls) and the relevant NRG-approved Dominion TRs. The staff's evaluation of the LAR is discussed in Sections 3.1 and 3.2 below.

3.1 Dominion Core Reload and Safety Analysis Methodologies MPS3 became part of the Dominion nuclear fleet following DNC's acquisition of Millstone Power Station in 2001. Currently, the fuel supplier (Westinghouse) performs the reload analysis for MPS3, whereas the licensee performs the reload analysis using the Dominion methods for other Dominion nuclear plants including NAPS, SPS, and KPS. In its LAR, the licensee proposed to apply the Dominion reload methods to MPS3 for the analysis of the non-LOCA transients and accidents to support MPS3 reload applications. In support of its proposal, the licensee provided justifications for the application of Dominion methods to MPS3 in Attachment 4 of its LAR, "Application of Dominion Nuclear Core Design and Safety Analysis Methods," (Reference 16),

which includes bases for the use of the following methodologies: (1) the reload nuclear design methodology; (2) the RPDC methodology; (3) CMS reactor physics methodologies; (4) the methodology of the reactor system transient analyses using RETRAN; (5) the statistical DNBR evaluation methodology; and (6) the methodology of the reactor core thermal-hydraulics analysis using VIPRE-D computer code. All of the above methods are documented in the NRC-approved TRs for use in other Westinghouse-manufactured plants operated by Dominion, including NAPS, SPS, and KPS. Since MPS3 is also a Westinghouse-manufactured plant, the NRC staff's review discussed in the following Subsections 3.1.1 to 3.1.6 focuses on whether the licensee's proposed use of the Dominion safety analysis methodologies for MPS3 is in a manner complying with the conditions and limitations imposed by NRC safety evaluation reports (SERs) approving the relevant Dominion TRs.

3.1.1 Reload Nuclear Design Methodology (TR VEP-FRD-42, Revision 2.1-A)

The reload nuclear design methodology discussed in Dominion TR VEP-FRD-42, Revision 2.1-A, "Reload Nuclear Design Methodology" and Section 3.1 of Reference 16 consists of the following elements:

  • Analytical models including CMS models, VEPCO RETRAN models, and core thermal-hydraulics VIPRE-D models;
  • Analytical methods for core depletions, core reactivity parameters and coefficients, core reactivity control, safety analysis, and statistical DNB;
  • Reload design process for the core loading pattern design & optimization and key parameter treatment in nuclear design analyses; and
  • Reload safety evaluation process and nuclear design report.

The Dominion reload methodology is an iterative process that involves the determination of a core loading pattern that fulfills cycle energy requirements and the demonstration that the plant with the reload core satisfies the constraints of the plant design basis and safety analysis limits.

The Dominion reload methodology and the current MPS3 reload methodology use the same method discussed in Westinghouse TR, WCAP-9272-A, "Westinghouse Reload Safety Evaluation" (Reference 17). The reload method uses a bounding analysis concept in which key analysis parameters with limiting directions are identified such that, if all key analysis parameters are conservatively bounded, a reference safety analysis is applicable and no further analysis is necessary. If any values are not bounded, further analysis of the transient or accident in question is performed, the applicable safety analyses are revised, or changes are made in the operating requirements specified in the TSs or COLR to satisfy applicable safety analysis criteria. The safety analysis process typically consists of steady state nuclear calculations used to derive the core physics related key analysis parameters as well as a dynamic accident analysis that utilizes these parameters to determine the accident result.

While MPS3 has differences in the nuclear steam supply system (NSSS), reactor protection system (RPS), and fuel features, these differences can be modeled using the existing methodology and analytical methods, namely VEP-FRD-42, Revision 2.1-A, with the appropriate

selection of input variables. As indicated in Section 3.1.2 of Reference 16, VEP-FRD-42 SER limits the use of VEP-FRD-42, prohibiting its application to fuel types other than Westinghouse and Framatome ANP Advanced Mark-SW fuel. The restriction of the SER states that if the changes necessary to accommodate another fuel product require changes to the reload methodology of Dominion TR VEP-FRD-42-A, these proposed changes are required to be submitted for prior NRC review. The NRC staff finds that this SER restriction is met, since the MPS3 uses a Westinghouse fuel (robust fuel assembly with redesigned mid-grids fuel, RFA-2 which is the same as that of NAPS). As part of the implementation of Dominion methods, the licensee will verify the boration requirements for MPS3 on a reload basis using the same constituent equations utilized in TR WCAP-1441, which is currently used for MPS3 (as confirmed by the licensee in its response to NRC RAl-8 (SRXB) of Reference 2). Therefore, the NRC staff determines that the use of the Dominion reload methodology discussed in TR VEP-FRD-42, Revision 2.1-A and Section 3.1 of Reference 16 is acceptable to support licensing applications for MPS3.

3.1.2 Relaxed Power Distribution Control Methodology (TR VEP-NE-1, Rev. 0.1-A)

The RPDC methodology discussed in Dominion TR VEP-NE- 1, Rev. 0.1-A, "VEPCO Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications" and Section 3.2 of Reference 16 is a Dominion method for axial power distribution control with a delta-I band. This method provides an increasing delta-I band with decreasing power in order to maintain approximately constant analysis margin to the design bases limits at all power levels. The RPDC analysis process consists of: (1) the generation of power shapes that bound the delta-I range; (2) the selection of delta-I bands such that all bands satisfy the COLR height dependent heat flux hot channel factor, Fo(Z), limit with verification that the proposed delta-I bands satisfy LOCA FQ [total peaking factor] and loss of flow accident thermal-hydraulic evaluations; (3) the analysis of limiting Condition II events to ensure the power shapes within the final delta-I band are used as initial conditions; (4) the verification to confirm that over-power delta-temperature (OPLiT) and over-temperature delta-T (OTLiT) limits are conservative to ensure that margin to fuel design limits is maintained; and (5) the formulation of N(Z) functions [non-equilibrium power distribution multiplier] to support the implementation of FQ TS surveillance.

A number of similarities between the Dominion RPDC methods and the Westinghouse-RAOC methods currently used for MPS3 are shown in Table 3.2.1 of Reference 16. Section 3.2.3 of Reference 16 also indicates that the cooldown transient assumption of 30°F currently used for the Westinghouse method at MPS3 will be used unless a MPS3-specific analysis demonstrates that a plant trip will occur before reaching 30°F. The NRC SER approving Dominion TR VEP-NE-1, Rev. 0.1-A accepted the Dominion RPDC method for use at NAPS and SPS, and also allowed the RPDC method for use at plants with reload cores similar to those of NAPS and SPS. As previously discussed in this SE, MPS3, SPS, and NAPS are Westinghouse-manufactured plants, their NSSS, RPS, and fuel designs are similar such that its features are capable of being reflected via modeling inputs in the TR analytical methods without any changes to the methodology. Also, since both MPS3 and NAPS use Westinghouse RFA-2 fuel design (Robust Fuel Assembly with redesigned mid-grids}, their reload cores are essentially identical. Therefore, the NRC staff finds that MPS3 satisfies the SER restriction, limiting the use of the TR to the reload cores similar to those of SPS and NAPS, and therefore, determines that

the use of the RPDC method discussed in Dominion TR, VEP-NE-1, Rev 0.1-A, and Section 3.2 of Reference 16 is acceptable to support MPS3 licensing applications.

3.1.3 Core Management System Reactor Physics Methods (TR DOM-NAF-1-A)

The CMS methods discussed in Dominion TR DOM-NAF-1-A, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations," and Section 3.3 of Reference 16 involve two major computer codes, CASM0-4 and SIMULATE-3. CASM0-4 is a multi-group, two-dimensional transport theory code used for depletion and branch calculations for a single assembly. SIMULATE-3 is a two-group, three-dimensional diffusion theory code coupled with thermal-hydraulic and Doppler feedback.

The CMS methods model the core physics characteristics of the reload core including depletion/isotopic effects, reactivity, reactivity coefficients, power distribution, and shutdown margin. DNC uses the CMS methods in the analysis for RPDC, and licensing applications, including core reload design, core operation, and key core parameters for reload safety analyses.

The CMS benchmarking data provided in DOM-NAF-1-A is based on the 15x15 and 17x17 fuel designs used at SPS and NAPS, respectively, while MPS3 currently uses 17x17 fuel which is within the range of the CMS benchmarking data. In addition, the NRC SER approving TR DOM-NAF-1-A limits the TR use, prohibiting its application to "significantly different or new fuel designs." This restriction is met, since the current MPS3 fuel design bases on the Westinghouse RAF-2 fuel design, which is the same as that of NAPS. Therefore, the NRC staff determines that the use of the CMS methodology discussed in DOM-NAF-1-A and Section 3.3 of Reference 16 is acceptable to support the licensing applications for MPS3.

3.1.4 Reactor System Transient Analyses Using RETRAN (TR VEP-FRD-41-A, Rev. 0.2)

Dominion uses RETRAN discussed in TR VEP-FRD-41, Rev 0.2-A, "VEPCO Reactor System Transient Analyses Using the RETRAN Computer Code," and Section 3.4 of Reference 16 to perform the analyses for non-LOCA events presented in the Final Safety Analysis Report (FSAR) for Dominion's plants with Westinghouse-manufactured reactors including NAPS, SPS and KPS. RETRAN calculates general system parameters as a function of time and boundary conditions for input into more detailed calculations of DNB or other thermal and fuel performance margins. The licensee performs analyses for non-LOCA events to confirm the adherence of reload core design limits to the bounds established by the reference analysis of record (AOR) parameter values, as well as to verify that the core is acceptable from a safety operational point of view.

3.1.4.1 MPS3 RETRAN Model As discussed in Section 2.0 of Reference 19, the proposed nodal scheme of the RETRAN model for MPS3 is essentially identical to the NAPS and SPS models with the following differences:

1. The MPS3 model explicitly models the safety injection (SI) accumulators.
2. The MPS3 model has separate volumes for the steam generator (SG) inlet and outlet plenums.
3. The MPS3 model includes cooling paths between downcomer and upper head.
4. The MPS3 model includes a nodal scheme with a second parallel flow path through the active core from the lower plenum to the upper plenum for the analysis of the steam line break (SLB).

The Dominion RETRAN models also have some differences compared to the vendor RETRAN model that was used to perform the current FSAR Chapter 15 analyses. Table 2-1 of Reference 19 identifies the model differences, including differences in code versions, nodal schemes for the reactor vessel and steam generator, and the reactivity feedback models.

3.1.4.2 RETRAN Benchmarking Analysis The Dominion MPS3 RETRAN models have been benchmarked by selecting representative non-LOCA design-basis events and comparing the results of the MPS3 RETRAN models to the vendor RETRAN model that was used to perform the current FSAR analyses. This approach is similar to that discussed in TR VEP-FRD-41-A. The results of the MPS3 RETRAN benchmark are provided in Reference 19. Subsequent to the submittal of the results of the bench analysis, the licensee identified a discrepancy between the MPS3 RETRAN base model pressurizer shell heat conductor and the Dominion RETRAN TR. The MPS3 RETRAN base input deck models the pressurizer shell as a heat conductor, which differs from TR VEP-FRD-41-A, which states that "Dominion continues to model the non-equilibrium wall as an adiabatic surface." This model is not used in any AOR, so this discrepancy has no impact on plant licensing or operations. However, the RETRAN base model was used to benchmark/replicate Westinghouse AOR in support of the subject LAR for generic NRC approval of the Dominion application for reload design analysis methods to MPS3. The licensee repeated each of the benchmarking cases supporting the LAR with the needed model correction and discussed the results in Reference 3. In response to the NRC staff's RAI, the licensee also analyzed two additional cases: the feedwater line break (FLB) and steam generator tube rupture (SGTR) events. As part of the RAI response, the licensee provided an update to the RETRAN benchmarking information in Reference 5. The NRC staff has reviewed the information of the RETRAN benchmarking analyses in References 3, 5, and 19, and discusses its evaluation for each benchmark case in Subsections 3.1.4.2.1 through 3.1.4.2. 7 for MPS3 as follows:

3.1.4.2.1 Analysis of the Loss of Load/Turbine Trip Event Section 4.1 of Attachment 2 to Reference 5 discusses the updated benchmark analysis for the loss of load/turbine trip (LOL) event. The event is initiated from a complete loss-of-steam flow and turbine trip from full-power conditions. The loss-of-steam flow results in a rapid increase in

secondary system pressure and temperature, as well as a reduction of the heat transfer rate in the SGs, which, in turn, causes the reactor coolant system (RCS) pressure and temperature to rise. The licensee listed in Table 4.1-1 the initial plant conditions and the assumptions used in the LOL analysis and showed no differences in the key input and assumptions used in both the benchmark analysis and FSAR analysis. The results of the analysis is presented in Figure 4.1-1 to Figure 4.1-5. A comparison of the Dominion case with the FSAR case shows a comparable trend with small differences in magnitudes of key parameters during the LOL event.

The results of the pressure predictions in the LOL benchmark analysis show that: (1) the Dominion case trips slightly earlier than the FSAR case; and (2) the calculated peak RCS pressure for the Dominion case is lower than that of the FSAR case. During the review, the NRC staff requested the licensee to explain the causes for differences in the pressure response of the Dominion and FSAR cases. In its response to RAl-10 (SRXB) (Reference 3), the licensee indicated that the slightly earlier pressurization is attributed to differences in the SG primary-to-secondary heat transfer associated with the Dominion single-node SG (SNSG) model compared to the FSAR multi-node SG (MNSG) model. For the SNSG model, the secondary-side temperature corresponds to the saturation temperature for the secondary side pressure, and will increase with an increase in pressure. The MNSG model represents SG tube regions that may be either saturated or subcooled. It would predict higher heat transfer rates during transient conditions due to an increase in the nodal number and modeling of dynamic effects for the liquid/vapor flow through the tube bundle. These effects result in a slightly earlier heat-up for the SNSG model and associated increase in primary-side pressure. For the Dominion case, because the pressure increase starts earlier, the reactor trip on high pressurizer pressure occurs slightly earlier.

In addressing item 2 regarding differences in the peak pressure prediction, the licensee indicated that the peak RCS pressure, which occurs after the reactor trip, is closely related to the response of the pressurizer safety valves (PSV). Since the main steam safety valves (MSSV) actuate after the time of peak RCS pressure, they do not affect the calculated peak RCS pressure. As shown in Figure 4.1-1 in Attachment 2 to Reference 5, the peak pressurizer pressure varies over a small range for the FSAR case, achieving pressures that are slightly higher than the Dominion case which has a relatively flat pressure profile when the PSVs open.

Since the LOL event results in a very rapid pressure increase, small differences in PSV response (e.g., delays, opening profiles, etc.) can significantly affect the peak pressure. These differences are more pronounced in the RCS cold-leg and reactor vessel lower plenum where peak pressures exceed 2,700 psia [pounds per square inch absolute] and are affected by differences in loop response (RCS loop, reactor vessel, and surge line loss coefficients, reactor coolant pump head dynamics, etc.).

The difference in the over-all predicted RCS pressure between the Dominion and FSAR cases is attributed to a difference in the secondary safety valve models. Specifically, the Dominion model includes the modeling of blowdown in the main SG safety valves and the vendor model does not. The licensee clarified in an e-mail dated March 3, 2016 (Reference 18) that the valve blowdown applies to the closing phase of the valve and results in the valve not becoming fully closed until the steam pressure is less than the pressure at which the valve opened. As a result, the MSSV continues to provide relief flow at pressure below the opening pressure during the closing cycle for the valve.

The result comparison also shows that for the Dominion and FSAR cases, the vessel inlet temperature and RCS coolant average temperature agree in trend and rate of increase, with the Dominion case lagging the FSAR response before the inlet temperature peaks at a slightly lower value, which indicates that the FSAR SG heat transfer degrades sooner than that predicted by Dominion model. This difference in the temperature response is caused by the difference between the use of a MNSG in the FSAR model and the SNSG model employed in the Dominion model.

3.1.4.2.2 Analysis of the Locked Rotor Event Section 4.2 of Attachment 2 to Reference 5 discusses the updated benchmark analysis for the locked rotor (LR) event. For the LR event, flow through the affected reactor-loop drops rapidly, leading to a reactor trip on a low-flow signal. After the reactor trip, energy stored in the fuel rods continues to be transferred to the reactor coolant, causing the RCS temperature to increase and the coolant to expand. During the transient, heat transfer to the shell-side of the SGs drops because the reduced flow results in a decreased SG tube film coefficient. The rapid expansion of the coolant in the reactor core, combined with reduced heat transfer in the SGs, causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge into the pressurizer compresses the steam volume, actuates the automatic spray system, and opens the pressurizer safety valves. For the over-pressure analysis, the licensee assumed that the event was initiated from full-power conditions with one RCP speed set to zero, and credited the reactor coolant low loop flow reactor trip, with a setpoint of 85 percent of the initial flow. The licensee listed in Table 4.2-1 the initial plant conditions and the assumptions used in the LR analysis which showed no differences in the key input and assumptions used in both benchmark analysis and FSAR analysis. The licensee presented the results of the analysis in Figure 4.2-1 to Figure 4.2-7. A comparison of the Dominion analysis with the FSAR analysis shows that the responses are comparable in trend for the LR event, with the Dominion model predicting higher peak RCS pressures. As discussed in the response to RAl-11 (SRXB) (Reference 3), the differences are caused by RCS loop friction losses and fuel rod heat transfer model differences.

3.1.4.1.3 Analysis of the Loss Normal Feedwater Event Section 4.3 of Attachment 2 to Reference 5 discusses the benchmark analysis for the loss of normal feedwater (LONF) event. During a LONF event, the SG water inventory decreases as a consequence of continuous steam supply to the turbine. The mismatch between the steam flow to the turbine and the feedwater flow leads to the reactor trip on a low-low SG level signal, which actuates the auxiliary feedwater system. As the SG pressure increases following the trip, the SG safety valves open to remove the decay heat. Consistent with the FSAR approach, the licensee analyzed the event as an overpressure event. The licensee listed the initial plant conditions and the assumptions used in the LONF analysis in Table 4.3-1 and showed no differences in the key input and assumptions used in both benchmark analysis and FSAR analysis. The licensee presented the results of the analysis in Figure 4.3-1 to Figure 4.3-7 in Reference 5.

A comparison of the Dominion analysis with the FSAR analysis shows that the transient responses are similar, with a predicted higher peak pressure for the Dominion case during an

LONF event. The differences result mainly from the SG safety relief valve model, which includes the modeling of blowdown in the Dominion analysis and not in the FSAR analysis. This model difference results in higher steam releases and a subsequent increase in heat transfer following the reactor trip. The SG nodal scheme and related heat transfer models along with other modeling differences such as pressurizer spray also affect the transient response. These effects are cumulative, resulting in a higher pressurizer pressure peak compared to the FSAR results.

3.1.4.2.4 Analysis of the Main Steam Line Break Event Section 4.4 of Attachment 2 to Reference 5 discusses the updated benchmark analysis for the main steam line break (MSLB) event. During an MSLB event, the steam release causes a decrease in the RCS temperature and SG pressure. In the presence of a negative moderator temperature coefficient, the RCS temperature decrease results in an addition of positive reactivity with the potential of power increase. The licensee analyzed the MSLB analysis for the maximum peak power increase that determines a minimum margin to an acceptable fuel design limit. The MSLB analysis assumed that the event was initiated from an instantaneous, double-ended break at the nozzle of one SG from hot shutdown conditions with offsite power available.

The licensee listed the initial plant conditions and the assumptions used in the MSLB analysis in Table 4.4-1 of Attachment 2 to Reference 5 and identified three major differences used in the benchmark analysis and FSAR analysis: (1) the Dominion analysis used a heat transfer model that maximizes heat transfer coefficients for the faulted SG secondary side, while the FSAR analysis used a Westinghouse proprietary heat transfer formulation; (2) the Dominion analysis credited boron from the SI system, while the FSAR case did not; and (3) the Dominion analysis used only the Doppler power coefficient (DPC) while the FSAR cases credited the DPC plus the Doppler temperature coefficient (OTC) in the moderator density feedback. The licensee used the reactor vessel nodal scheme in Figure 2-2 of Reference 19 for the analysis of the MSLB event, which is an asymmetric response transient with lower temperature in the core next to the ruptured SG and higher temperature in the other side of the core. The reactor vessel model with a specification of mixing flow fractions was used to simulate conditions from complete to incomplete mixing for the flow from both the cold-side and hot-side of the core. The mixing flow fractions were based on scale model mixing tests performed by Westinghouse (the licensee's response to RAl-13 (SRXB), Reference 2). The assumption of imperfect mixing used in the MSLB analysis is consistent with the methodology documented in Topical Report, VEP-FRD-41-P-A.

The analysis shows that the peak power and heat flux based on the Dominion methods are higher and occur more quickly than the FSAR data. The differences are caused by the SNSG model employed in the Dominion model that calculates a higher steam rate, resulting in a greater cooling effect of the faulted SG on the RCS. They are also the results of differences in the nodal scheme and mixing at the core inlet and outlet between the Dominion case and the FSAR case. The power response for both models is not affected by the delivery of boron to the RCS. This is because the FSAR model does not credit boron, and in the Dominion model, boron does not reach the RCS from the SI system until after the termination of the transient.

3.1.4.2.5 Analysis of the Control Rod Bank Withdrawal at Power Event Section 4.5 of Attachment 2 to Reference 5 discusses the updated benchmark analysis for the control rod bank withdrawal at power event. The effect of this event is an increase in fuel and coolant temperature. The licensee listed the initial plant conditions and the assumptions used in the Control Rod Bank Withdrawal at Power (RWAP) analysis in Table 4.5-1 and showed no differences in the key input and assumptions used in both benchmark analysis and FSAR analysis. The licensee presented the results of the analysis in Figure 4.5-1 to Figure 4.5-6. A comparison of the Dominion case with the FSAR case shows a comparable trend, with small differences in magnitudes of key parameters during the RWAP event.

For the RWAP 1 percent millirho per second (pcm/sec) case, the core power response shows that its rate of increase for the Dominion model is greater than the FSAR data. The faster power increase rate leads to the Dominion modeling tripping on high neutron flux at about 74 seconds, and the lower power increase rate for the FSAR case results in a reactor trip on an overtemperature LlT (OTLlT) signal at about 93 seconds. In the RAl-14 (SRXB) response (Reference 3), the licensee indicated that the differences in the core power predictions are caused by the differences in the models for the moderator and Doppler reactivity feedback effects. The moderator reactivity feedback is assumed to be zero for both cases. For Doppler reactivity, the Dominion case uses a OTC while the FSAR case uses a DPC with minimum reactivity feedback conservatively assumed for both cases. The licensee also indicated that the reactor core model used in the FSAR case incorporates proprietary mechanisms to modify the removal of heat from the core. The difference in reactor trip mechanisms between the Dominion and FSAR cases reflects the breakpoint for switching between OTLlT and high flux as shown in FSAR Figure 15.4-10. The results of MPS3 FSAR Chapter 15 non-LOCA analyses indicates the RWAP event is the most limiting event in terms of the margin to the safety limit DNBR in the category of the anticipated operating occurrences (AOOs ). Since the licensee proposed to use the RETRAN and Dominion VIPRE-D method to perform DNBR calculations for assessing the fuel integrity during AOOs and accidents, the NRC staff requested the licensee to include in its benchmark analysis the results of the DNBR calculation by using the Dominion VIPRE-D method. In response to RAl-15 (SRXB) (Reference 3), the licensee provided calculated DNBRs as Figure 4.5-7 for the RWAP 1 pcm/sec case. A comparison of the Dominion case with the FSAR case shows a comparable trend with small differences in magnitudes of the predicted values of DNBR. As shown in Figure 4.5-1 of Attachment 2 to Reference 5, the core power rate of increase in the 1 pcm/sec case for the Dominion RETRAN model is greater than the FSAR data such that the reactor trip occurs approximately 20 seconds earlier. The inverse effect of power on DNB is observed in the transient DNBR plot shown in Figure 4.5-7 of Attachment 1 to Reference 3 and the minimum DNBR values for the Dominion and FSAR cases are comparable. In addition, the licensee confirmed that the thermal-hydraulic conditions of the RWAP transient analyzed are within the acceptable range of the NRG-approved DNBR correlations utilized in the VIPRE-D model (WRB-2M and ABB-NV) consistent with the limitations on the use of DOM-NAF-2-P-A.

3.1.4.2.6 Analysis of the Feedwater Line Break Event MPS3 FSAR Section 15.2.8 discusses the FLB analysis for both cases with and without offsite power available. FSAR Figures 15.2-13 and 15.2-19 indicates that a post-trip return-to-power

will occur for the case with offsite power available, and core will remain subcritical throughout the transient for the case without offsite power available. Also, FSAR page 15.2-16 indicates that the FLB is the most limiting event in the decrease in secondary removal category. The analysis of the FLB needs to use a broad scope of the models in RETRAN, including FLB break flow model, RC [reactor coolant] pumps coastdown model, SG heat transfer model, and reactivity feedback model. Although RETRAN is an NRC-approved code, it has not been applied to MPS3 for the FLB analysis. During the review, the NRC staff requested the licensee to provide an FLB benchmark analysis to demonstrate that the code produces acceptable results when applied to MPS3. In response, the licensee performed the RETRAN benchmark analysis for the FLB event for both FLB cases with and without offsite power available. The licensee listed the initial plant conditions and the assumptions used in the FLB analysis in Table 4.6-1 of Attachment 2 to Reference 5 and showed no differences in the key input and assumptions used in the Dominion analysis and FSAR analysis. It presented the results of the analysis in Figure 4.6-1 through Figure 4.6-8, and Figure 4.6-9 through Figure 4.6-15 of to Reference 5 for the FLB with offsite power case and the FLB without offsite power available case, respectively. A comparison of the Dominion analysis with the FSAR analysis shows that for the FLB with offsite power case, the transient responses are in good agreement, and for the FLB without offsite power available case, the transient responses are comparable in trend, with small differences observed early in the transient (for a period from 100 seconds to 1,000 seconds into the transient) for RCS temperatures. The RCS temperature differences are caused by the differences in the Dominion SNSG model and the FSAR MNSG model. These differences in the SG models have a negligible effect on the long-term primary side heat transfer and associated temperature response.

3.1.4.2.7 Analysis of the Steam Generator Tube Rupture Event MPS3 FSAR 15.6.3 discusses the SGTR analysis for two cases: (1) the SG overfill margin analysis that is used to validate the assumption of no water released from the affected SG to atmosphere; and (2) the mass release analysis that is used as input to a computer code for calculating the dose releases. This analysis involves simulation of the mitigating strategies directing operators to identify and isolate the ruptured SG, cooldown the RCS to establish subcooling margin, depressurize to restore RCS inventory, and terminate safety injection to stop primary-to-secondary leakage. Although RETRAN is an NRC-approved code, it has not been applied to MPS3 for the SGTR analysis. During the review, the NRC staff requested the licensee to provide an SGTR benchmark analysis to demonstrate that the code produces acceptable results when applied to MPS3. In response, the licensee performed the RETRAN benchmark analysis for the SGTR event for two cases: a mass releases case and a SG overfill case. The licensee presented the results in Section 4. 7 of Attachment 2 to Reference 5. The licensee listed the initial plant conditions and the assumptions used in the SGTR analysis in Table 4.7-1 of Attachment 2 to Reference 5 and showed no differences in the key input, operator actions, and assumptions used in the Dominion analysis and FSAR analysis. Although RETRAN was approved previously by NRC for use in the SGTR analysis, the NRC SER approving RETRAN limited its use in Limitation 38 (Reference 30), which indicates that the SGTR event should not be analyzed for two-phase flow conditions without further justification of two-phase slip models used in the analysis. In the response to NRC RAl-17 (SRXB)

(Reference 3) regarding compliance with the SER limitation, the licensee confirmed that the

RCS flow remains single-phase and subcooled throughout the entire STGR benchmark analysis, justifying that it meets the cited SER limitation.

Mass Releases Analysis The licensee presented the results of the mass releases analysis in Figure 4. 7-1 to Figure 4. 7-9.

A comparison of the mass releases analysis for the Dominion case and FSAR case shows good agreement in transient responses of the pressurizer pressure, SG pressure, intact-loop RCS temperature, primary to secondary break flow rate, ruptured SG mass release rate, intact SG mass release rate, and RCS flow. The differences of transient response of the pressurizer level and ruptured-loop cold-leg temperature are discussed below.

For the pressurizer level response, Figure 4.7-2 shows that the FSAR level decreases more than the Dominion level during the RCS cooldown phase (approximately 3,200-3,700 seconds).

The differences occur because the primary to secondary heat transfer is reduced for the Dominion case caused by the loss of natural circulation flow on the ruptured SG and during a period when the SI flow is increasing significantly due to the reduction in RCS pressure. After SI is isolated, the longer duration in break flow for the FSAR case is reflected in lower pressurizer level at the end of the transient. These divergences in pressurizer level occur late in the transient well after the flow path to atmosphere through the failed atmosphere dump valve (ADV) has been isolated and do not have a significant effect on the overall results of mass releases.

For the ruptured SGs, the predicted RCS temperatures {shown in Figure 4.7-5) are in good agreement between the Dominion and FSAR cases until about 3,600 seconds, at which time the Dominion cold-leg temperature trends below the FSAR results. This is caused by a small natural circulation flow rate that occurs on the ruptured loop as a result of the RCS cooldown.

With the small RCS loop flow rate, the SI flow with a low temperature has a more noticeable effect on cold-leg fluid temperature. The predicted low cold-leg temperature in the ruptured SG has a small effect on the overall results for the transient since most of the heat removal occurs through the intact SGs during this time and the ruptured SG has been previously isolated.

SG Overfill Analysis For the SG overfill analysis, the licensee presented the results in Figures 4.7-10 through Figure 4.7-17. A comparison of the SG overfill analysis for the Dominion case and AOR (stretch power uprate (SPU), Reference 29) case shows good agreement in transient responses of the pressurizer pressure, SG pressure, intact-loop RCS temperature, and RCS flow. The differences of transient response of the pressurizer level, ruptured-loop cold-leg temperature, primary-to-secondary break flow rate, and ruptured SG liquid volume are discussed below.

For the pressurizer level response, Figure 4. 7-11 shows similar trends between the Dominion case and SPU case, with the SPU level decreasing more than the Dominion level during the RCS cooldown phase (approximately 2200-3200 seconds). The differences occur because the primary to secondary heat transfer is reduced for the Dominion case caused by the loss of natural circulation flow on the ruptured SG and during a period when the SI flow is increasing significantly due to the reduction in RCS pressure. After the SI is isolated, the higher break flow

rates shown in Figure 4. 7-15 for the SPU case are reflected in lower pressurizer level at the end of the transient.

For the ruptured SGs, the predicted RCS cold-leg temperatures (shown in Figure 4. 7-14) are in good agreement between the Dominion and FSAR cases until about 2600 seconds when natural circulation flow is lost in the ruptured RCS loop and the cold-leg temperatures are more strongly affected by the cooler SI flow as discussed above for the mass release analysis after SI flow is terminated, the Dominion cold-leg temperature trends toward the SPU value. For the primary-to-secondary break flow rate, Figure 4. 7-15 shows there is good agreement between the Dominion and SPU cases until the period late in the transient after SI has been isolated and the break flow is trending towards zero. This is also observed for the ruptured SG liquid volume response shown on Figure 4.7-16 where the Dominion and SPU responses agree well, with the Dominion value stabilizing at a lower value near the end of the transient. The licensee indicated that the following factors could affect the final SG fluid volume: (1) any difference in the assumed decay heat profile resulting in a different amount of fluid boiled from the SG secondary and associated liquid volume; (2) any differences in the integrated SI fluid injection affecting the RCS fluid inventory available for release to the ruptured SG; (3) on the secondary side, differences in the integrated auxiliary feedwater (AFW) flow rates affecting the fluid delivered to the ruptured SG fluid volume as well as the energy removed by the intact SGs; (4) differences in SG relief valve flow rates affecting mass and energy removal from the system, and (5) any differences in the Dominion and SPU model nodal scheme and related assumptions affecting the differential pressure between the respective fluid levels in the RCS and SG secondary, which would affect the final equilibrium level and associated fluid volume.

Based on the discussion of the benchmark analysis in Subsection 3.1.4.2.1 through Subsection 3.1.4.2. 7 above, the NRC finds that: (1) the Dominion MPS3 RETRAN benchmarking analysis has included appropriate non-LOCA cases discussed in MPS3 FSAR; (2) the Dominion MPS3 RETRAN model compares reasonably well with the vendor RETRAN model in predicting the trend of the RCS response for the selected non-LOCA cases; (3) the differences in the magnitude of the RCS response can be explainable based on differences in nodal schemes, inputs, or modeling assumptions, and; (4) the use of the Dominion RETRAN method is within the NRG-accepted conditions. Therefore, the NRC staff determines that the RETRAN methodology, as discussed in VEP-FRD-41, Rev. 02, References 3, 5, 19, and Section 3.4 of Reference 16, is applicable to MPS3.

3.1.5 Statistical DNBR Evaluation Methodology (TR VEP-NE-2-A) 3.1.5.1 Introduction and Background This section describes plant-specific application of statistical DNBR methodology for MPS3 cores containing the Westinghouse 17x17 Robust Fuel Assembly (RFA-2) fuel product. This section provides technical basis and documentation for the application of NRG-approved Dominion Topical Report (TR), VEP-NE-2-A (Reference 11) to MPS3. This application employs VIPRE-D thermal-hydraulics (T-H) computer code (Reference 12) with the Westinghouse WRB-2M, ABB-NV, and WLOP Critical Heat Flux (CHF) correlations for the T-H analysis of Westinghouse 17x17 RFA-2 fuel products for MPS3. Attachment 6 of the LAR (Reference 20)

describes the development and implementation of the statistical DNBR limit evaluation methodology as applied to the MPS3 fuel design.

The licensee is seeking approval for the inclusion of TR VEP-NE-2-A and Fleet Report DOM-NAF-2-P-A, Appendix C and D (References 11 and 12) to the TS 6.9.1.6.b list of NRC-approved methodologies used to determine core operating limits and in the reference list of the COLR.

VIPRE-D is the Dominion version of the VIPRE computer code that was originally developed for Electric Power Research Institute (EPRI) by Pacific Northwest National Laboratory (PNNL) to predict the CHF and DNBR of reactors. The NRC-approved fleet report, DOM-NAF-2-P-A, Appendix C describes the verification and qualification of the WRB-2M CHF correlation and Appendix D describes the verification and qualification of the ABB-NV and WLOP CHF correlations. The WRB-2M CHF correlation is applicable to the DNBR evaluation of the Westinghouse 17x17 RFA-2 fuel design. The ABB-NV and WLOP CHF correlations are applicable to the DNBR evaluation of the Westinghouse 17x17 RFA-2 fuel product. RFA-2 fuel product for transients that leads to low primary system pressure. The statistical design limits (SDLs) obtained by this implementation are for the following applications:

1. Technical Specifications Change Request to add DOM-NAF-2-P-A and relevant Appendixes to the plant's COLR list,
2. SDL(s) for the relevant code/correlation(s),
3. Any TS changes related to thermal over-temperature LiT (OTLiT}, overpower LiT (OPLiT), axial power distribution (FLil), enthalpy rise factor (FLiH) or other reactor protection function, as well as revised Reactor Core Safety Limits (RCSLs), and
4. List of FSAR transients for which the code/correlations will be applied.

NRC-approved TR, VEP-NE-2-A describes a methodology for the statistical treatment of key uncertainties in core Thermal-Hydraulics (T-H) DNBR analysis and provides DNBR margin through statistical analysis rather than deterministic uncertainty treatment. This TR was approved by the NRC staff subject to the following conditions for its use:

1. The selection and justification of normal statepoints used for plant specific implementation,
2. Justification of the distribution, mean and standard deviation for all statistically treated parameters should be included in the submittal,
3. Justification of the value of model uncertainty must be included, and
4. For the relevant CHF correlations, justification of the 95/95 DNBR limit and the normality of the M/P distribution, its mean and standard deviation must be included in the submission.

3.1.5.2 Implementation of Statistical Methodology TR VEP-NE-2-A, "Statistical DNBR Evaluation Methodology" (Reference 11 ), describes Dominion's methodology for statistically treating several of the important uncertainties in the DNBR analysis. The methodology in TR VEP-NE-2-A is employed to develop SDLs for the VIPRE-D/WRB-2M and VIPRE-D/ABB-NV code/correlation pairs for Westinghouse RFA-2 fuel at MPS3. The VIPRE-D/WLOP code/correlation is not used for statistical analyses for RFA-2 fuel at MPS3. With the uncertainties accounted for in the statistical analyses, the new SOL is larger than the deterministic code/correlation pair design limit, however, it is advantageous to use the SOL since statistical methodology permits the use of nominal operating initial conditions instead of requiring the application of evaluated uncertainties to the initial conditions for statepoints and transient analysis.

SOL is a Monte Carlo type analysis where two-thousand (2,000) random statepoints are generated for each statepoint and supplied to the VIPRE-D code which calculated the minimum departure from nucleate boiling ratio (MDNBR) for each statepoint. Each MDNBR is randomized by a code correlation uncertainty described in TR VEP-NE-2-A using the 95 percent confidence limit on the VIPRE-D/WRB-2M and VIPRE-D/ABB-NV code/correlation pair measured-to-predicted (M/P) CHF ration standard deviation.

In a response to an NRC RAl-4 (SNPB) (Reference 2), the licensee responded that the randomized DNBR is consistent with the methodology described in the LAR and calculated as:

. d DNBR R an d om1ze =--..,(,.,.,M,,_)-----------

Calculated DNBR (1.o+s P .K(95).Normalized Random Number Where:

s(M/P) is the standard deviation of the code/correlation M/P database for the CHF correlation taken from DOM-NAF-2-P-A, and K (95) is a sample correction factor that depends on the size of the experimental database supporting the correlation, and is calculated based on the equation given in Statistical DNBR Evaluation Methodology and is equal to:

K(95) -- -v'[ (v2n-2 .(n-a) {R ef erence 11) 3 -i. 645 )" 2 The standard deviation of the resultant randomized DNBR distribution is increased by a correction factor to obtain a 95 percent upper confidence limit and then combined root-sum-square with code and model uncertainties to obtain a total DNBR standard deviation (s101a1) as:

SOL = 1 + 1.645

  • Stotal Where, the 1.645 multiplier is the z-values for the one-sided 95 percent probability of a normal distribution. The SOL provides peak fuel rod DNBR protection at greater than the "95/95 level"

(i.e., provides a 95% probability at a 95% confidence level (95/95) that the peak rod does not experience DNB).

Consistent with VEP-NE-2-A methodology, inlet temperature, pressurizer pressure, core thermal power, reactor vessel flow rate, core bypass flow, nuclear enthalpy rise factor (Ft:.HN, and the engineering enthalpy rise factor (Ft:.E) were selected parameters in the implementation of statistical analysis. These uncertainties are listed in Table 3.1.5-1.

Table 3.1.5-1 MPS3 Parameter Uncertainties Nominal Standard Parameter Uncertainty Distribution Uncertainty Description Value Deviation Pressure +/-58.8 psi at Uncertainty corresponds to 2250 30 psi Normal (psia) 2a two-sided 95% probability Temperature Two-sided, 95% probability (oF) 557.06 2.5 +/-4.9 at 2a Normal distribution Power Two-sided, 95% probability 3712 1.0% +/-1.96 at 2a Normal (MWt) distribution

+/-2.94% at Two-sided, 95% probability Flow (gpm) 379200 1.5% Normal 2a distribution Two-sided, 95% probability Ft:.HN 1.635 2.0% +/-4.0% at 2a Normal distribution VEP-NE-2-A (Reference 11)

Ft.HE 1.0 N/A +/- Uniform treats this uncertainty as a uniform probability distribution Monte Carlo analysis used a best estimate bypass flow of Bypass(%) 7.6 NIN +/-1.0% Uniform 7 .6% with an uncertainty of 1% and uniformly distributed 3.1.5.2.1 CHF Correlation Uncertainty Only the WRB-2M/ABB-NV/WLOP CHF correlations that are used for DNBR calculations for Westinghouse 17x17 RFA-2 fuel product and the WRB-2M and ABB-NV CHF correlations are applicable to operating conditions at which the statistical DNBR methodology is applied. The WLOP CHF correlation is used deterministically. The ABB-NV correlation is only used below the first mixing grid and the WLOP correlation is used when the operating conditions are outside of the range of validity of the WRB-2M and ABB-NV CHF correlations, such as the MSLB evaluation, where there is reduced temperature and pressure. Table 3.3-1 of the LAR lists the deterministic DNBR limit deterministic design limit (DDL) correlation data for VIPRE-D/WRB-2M and VIPRE-D/ABB-NV code/correlation pairs. Consistent with the methodology in VE-NE-2-A a 95 percent upper confidence limit (K(95) in Section 3.1.5.2) is applied to the calculated correlation statistics.

NRC issued Information Notice (IN) 2014-01 dated February 21, 2014 (Reference 22) which raised a concern that the DNBR safety limit generated from statistical methodologies may not properly account for a conservative bias that may be included in the NRC-approved CHF correlation limit as defined in the SER for VEP-NE-2-A. The IN further discussed the fact that the correction of this inconsistency may increase the statistically-based DNBR safety limit. The magnitude of the increase is dependent on the difference between the CHF correlation's 95/95 statistics and the NRC-approved CHF correlation limit.

The licensee stated in their LAR submittal that their implementation of SDL is consistent with the methodology of TR, VEP-NE-2-A. The acceptability of the use of the calculated standard deviations is based on the use of a 95 percent upper confidence factor that is essentially equivalent to the Owen's tables for ensuring a 95 percent probability at a 95 percent confidence limit. In section 3.6.1 of Attachment 6 of the LAR, the licensee provided supporting information and introduced the correction factor (SoNsR) to ensure that the SDL developed in accordance with the methodology of VEP-NE-2-A and using the calculated correlation statistics provides a 95 percent probability at a 95 percent confidence level (95/95) that the peak rod does not experience DNB. In its response to RAl-5 (SNPB) regarding the IN 2014-01 (Reference 2), the licensee reiterated that the SDL calculation was developed per the VEP-NE-2-A methodology, and there is no need to modify the licensee's calculations. The NRC staff verified the licensee's calculational procedures and methods and determined that their SDL calculations are acceptable.

3.1.5.2.2 Model Uncertainty Condition 3 of the SER for VEP-NE-2-A states that the licensee must provide justification of the value of model uncertainty (FM) and be included in the plant specific submittal. The VIPRE-D/WRB-2M and VIPRE-D/ABB-NV code/correlation pair SDLs for MPS3 were developed using the VIPRE-D 21-channel production model for MPS3 with the 17x17 RFA-2 fuel design. Since the production model that Dominion intends for the MPS3 evaluations are used to develop the SOL, the licensee determined there is no need for additional model uncertainty. Therefore, the model uncertainty is set to zero. The NRC staff finds this approach acceptable.

3.1.5.2.3 Code Uncertainty The NRC-approved VEP-NE-2-A methodology states that a code uncertainty (Fe) must be applied because of two factors: (i) the effect of analyzing a full core with a correlation which was based on steady state test bundle data and (ii) and the effect of performing the analyses with the Virginia Power (former licensee) COBRA code when the W-3 data were reduced by the use of a Westinghouse thermal-hydraulics code. These uncertainties are independent of the correlation. The code uncertainty was quantified at 5 percent; consistent with the factors specified for other thermal/hydraulic codes in VEP-NE-P-A and the basis of this uncertainty is described in the application of this methodology at the Surry power station (Reference 11 ). A one-sided 95 percent confidence level on the code uncertainty is then 3.04% (= (5.0%) /1.645).

The use of the 1.645 divisor (the one-sided 95 percent tolerance interval multiplier) is conservative since the NRC staff considers the 5 percent uncertainty to be a 2a value. Upon

review of the references and the methodology described above, the NRC staff has determined that the selection of the value of the code uncertainty is acceptable.

3.1.5.3 Monte Carlo Calculations For the Monte Carlo analysis, nine (9) nominal statepoints that cover full range of nominal operation and AOO transients are selected for both WRB-2M and ABB-NV CHF correlations.

These nine statepoints cover a range of conditions, such as pressure, temperature, etc. over which the statistical methodology is applied and also cover the ONB limiting range of the Reactor Core Safety Limits (RCSL) and within the validation range of applicability of the associated correlations. Tables 3.6-1 and 3.6-2 of the Attachment 6 of the LAR lists pressurizer pressure, inlet temperature, power, flow Ft.HN and MON BR for VIPRE-O/WRB-2M and VIPRE-0/ABB-NV, respectively, for the Westinghouse 17x17 RFA-2 fuel at MPS3.

The Monte Carlo calculations consisted of 2,000 calculations for each of the nine nominal statepoints for each CHF correlation. The ONBR standard deviation at each nominal statepoint was augmented by the code/correlation uncertainty, the small sample correction factor, and the code uncertainty to obtain a total ONBR standard deviation. Equation 3.3 of Attachment 6 of the LAR provides a relationship for the total ONBR standard deviation using the Root-Sum-Square method and the total standard deviation (srnTAL) is dependent on standard deviation of the randomized ONBR distribution, uncertainty in the standard deviation of the 2000 Monte Carlo simulations that provides a 95 percent upper confidence limit on standard deviation, the code uncertainty, and the model uncertainty.

2 + F,2c + FzM The limiting peak fuel rod SOL was calculated to be 1.225 for the VIPRE-O/WRB-2M code/correlation pair and 1.177 for the VIPRE-0/ABB-NV code/correlation pair. The Monte Carlo Statepoint analysis is summarized in Tables 3.6-3 and 3.6-4 of Attachment 6 of the LAR.

The use of the correction factor for the total ONBR standard deviation ensures that the SOL developed in accordance with the methodology of VEP-NE-2-A and using the calculated correlation statistics provides a 95 percent probability at a 95 percent confidence level (95/95) that the peak rod does not experience ONB. The NRC staff verified the use of the accepted methodology in the calculation of SOL and determined that the use of the calculated correlation statistics as input to SOL calculation is acceptable and reasonable.

3.1.5.4 Full Core ONB Probability Summation The data statistics are used to determine the number of rods expected in ONB. The ONB sensitivity is estimated as partial derivative of ONBR divided by partial derivative of (1/Fe.H), and are listed in Tables 3.7-1and3.7-2 of Attachment 6 of the LAR forWRB-2M and ABB-NV correlations, respectively and are denoted by 13. To ensure conservatism in the calculations, a one-sided tolerance limit of 13 is used. Variable 1/Fe.H is the most statistically significant

independent variable in the linear regression model, yielding a regression coefficient greater than 99 percent. Table 3. 7-3 of Attachment 6 of the LAR lists a representative rod census curve used for determining SOL; this table provides probable maximum limit Ft.H versus maximum%

of fuel rods in core. Tables 3.7-4 and 3.7-5 of Attachment 6 of the LAR provides full core ONB probability summations for VIPRE-O/WRB-2M and VIPRE-0/ABB-NV code/correlation pairs, respectively. After a review of the values presented in the above Tables, the NRC staff determined that the values listed in these tables are acceptable.

3.1.5.5 Verification of Nominal Statepoints Used in SOL Calculations Condition 1 of the SER for VEP-NE-2-A requires the nominal statepoints used in the SOL analysis must be justified in providing a bounding ONBR standard deviation for any set of conditions. This justification is performed by demonstrating that SroTAL is maximized for any real set of conditions at which the core approaches the SOL. For this, a regression analysis is performed using the unrandomized ONBR standard deviations at each statepoints.

Tables 3.8-1 and 3.8-2 of Attachment 6 of the LAR show the R2 linear regression coefficients verifying the nominal statepoints for WRB-2M and ABB-NV, thus validating the independence.

These values are listed in Table 3.1.5-2 Table 3.1.5-2 Regression Coefficients for the Verification of the Nominal Statepoints for MPS317x17 RFA-2 fuel with VIPRE-D/WRB-2M and VIPRE-D/ABB-NV Code Correlation Pairs Statepoints R2 Linear Regression R2 Linear Regression ForWRB-2M (%) For ABB-NV(%)

Pressure 2.75 23.62 Temperature 7.70 34.95 Flow Rate 1.27 24.3 Power 3.04 27.76 In a response to NRC RAl-6 (SNPB) (Reference 2), the licensee stated that the relatively large differences in R2 values between the two correlations are expected since the correlations used to evaluate the behavior are based on different equation forms and experimental databases.

The nominal statepoints at which the two CHF correlations are evaluated are also different (Tables 3.6-1 and 3.6-2 of Attachment 6 of the LAR). For a linear regression analysis that is performed for these two correlations, the analysis is expected to generate different values for R2 since different statepoints are used for the two correlations. Also, for a given correlation, the R2 values are similar and there is not a strong dependence on any one single input condition. The NRC staff has reviewed the statistical regression analysis as well as the inputs to the two correlation analyses and determined that the R2 values are acceptable.

3.1.5.6 Scope of Applicability Condition 4 of the SER for VEP-NE-2-A requires that for the relevant CHF correlations, justification of the 95/95 DNBR limit, and the normality of the M/P distribution, its mean and standard deviation must be included in the submission, unless there is an approved TR documenting these such as DOM-NAF-2-P-A.

Table 3.9-1 of Attachment 6 of the LAR lists the accidents to which the SOL methodology is applicable. These include all Condition I and II DNB events except the Rod Withdrawal from Subcritical (RWSC) and the complete Loss of Flow, the Locked Rotor Accident, the Single Rod Cluster Control Assembly Withdrawal at Power, and feedwater system pipe break. The Statistical DNBR Evaluation Methodology provides analytical margin by permitting transient analyses to be initiated from nominal operating conditions, and by allowing core thermal limits to be generated without the application of the bypass flow, Ft:.HN (measurement component) and hot channel uncertainties which are statistically calculated into the DNBR limit.

3.1.5.7 Application of VIPRE-D/WRB-2M/ABB-NV/WLOP to MPS3 The VIPRE-D/WRB-2M and VIPRE-D/ABB-NV code/correlation pairs, together with the Statistical DNBR Evaluation Methodology, will be applied to all Condition I and II DNB events (except RWSC), and to the Complete Loss of Flow event and the Locked Rotor Accident. The WRB-2M, ABB-NV and WLOP CHF correlations are used for the DNBR calculation for Westinghouse RFA-2 fuel product. The WLOP CHF correlation is used for operating conditions outside the range of applicability of WRB-2M and ABB-NV CHF correlations, namely, for the MSLB accident analysis.

Thermal margin analyses evaluates the design and safety analysis limits and these limits are used to define the available DNBR margins for each application. The difference between the safety analysis limit (SAL) and the design limit is the available retained DNBR margin. For deterministic DNB analyses, the design DNBR limit is set equal to the applicable code/correlation limit and it is termed the DDL. For statistical DNB analyses, the design DNBR limit is set equal to the applicable SOL.

Table 4.2-1 of Attachment 6 to the LAR lists the DDLs and SDLs for the three CHF correlations and reproduced here as Table 3.1.5-3:

Table 3.1.5-3 DNBR Limits for WRB-2M, ABB-NV and WLOP Correlations VI PRE-D/WRB-2M VIPRE-D/ABB- VIPRE-D/WLOP NV DDL 1.14 1.14 1.22 SOL 1.23 1.19 Deterministic SAL 1.50 1.50 1.50

The SDL limit provides a peak fuel rod DNB protection with at least 95 percent probability at a 95 percent confidence level and a 99.9 percent DNB protection for the full core. A deterministic and statistical SAL equal to 1.50 has been selected for 17x17 RFA-2 fuel at MPS3 with the VIPRE-D/WRB-2M, VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs. This SAL is applicable for all deterministic analyses for a maximum peaking factor Ft.HN equal to 1.65 and for all statistical analyses for a maximum peaking factor Ft.HN equal to 1.587.

The difference between SAL and the design limit is available as retained DNBR margin:

SAL-DDL Retained DNBR Margin = 100 * ( SAL

)

Retained DDNBR margins are listed in Table 3.1.5-4 Table 3.1.5-4 DNBR Limits and Retained DNBR Margins Deterministic DNB Applications DNB Correlation DDL SALoET Retained DNBR Margin(%)

WRB-2M 1.14 1.50 24.0 ABB-NV 1.14 1.50 24.0 WLOP 1.22 1.50 18.6 Statistical DNB Applications DNB Correlation SOL SALoET Retained DNBR Margin(%)

WRB-2M 1.23 1.50 18.0 ABB-NV 1.19 1.50 20.6 The NRC staff has reviewed processes that calculated the retained DNBR margins, deterministic safety analysis limits, for the code/correlation pairs and determined that the licensee has applied the approved methodology in the calculations and are therefore acceptable.

3.1.

5.8 NRC Staff Conclusion

- Statistical DNBR Evaluation Methodology The licensee has proposed to adopt the use of the Dominion methodology in TR VEP-NE-2-A for statistical DNBR evaluation for MPS3. Using a combination of Monte Carlo analysis using 2,000 random statepoints, standard deviation of randomized DNBR distributions which is the un-randomized standard deviation corrected for CHF correlation uncertainty, a combination of

Root Sum Square with code, and model uncertainty standard deviations, the licensee obtained a total DNBR standard deviation (Tables 3.6-3 and 3.6-4 of Attachment 6 of the LAR). The analysis resulted in the SD Ls of 1.23 for VI PRE-D/WRB-2M and 1.19 for VI PRE-0/ABB-NV.

The NRC staff reviewed Attachment 6 of the LAR that used the Dominion methodology to calculate the SDLs for MPS3. The NRC staff determined that the licensee appropriately used the approved methodology to determine the SOL and provided sufficient margin through the use of statistical rather than deterministic uncertainty treatment. The staff finds the licensee's analysis satisfied the conditions that were listed in the SER for the methodology TR VEP-NE A. Therefore, the NRC staff has determined that the SDLs that are listed in Table 3.1.5-3 of this SE constitute a design basis limit for a fission product barrier.

3.1.6 Reactor Core Thermal-Hydraulics using VIPRE-0 Computer Code (TR DOM-NAP A, Rev. 0.3)

The reactor core thermal-hydraulics code VIPRE-0 described in Dominion TR DOM-NAF-2-A, Rev. 0.3, and Section 3.6 of Reference 16 is a Dominion-modified version of the VIPRE-01.

VIPRE-0 is used to calculate reactor coolant conditions to verify that the DNBR design safety limit is maintained. It has been adapted to accommodate the various fuel designs used at Dominion nuclear power stations by incorporating vendor proprietary CHF correlations.

VIPRE-0 was approved by the NRC for PWR licensing calculations up to CHF using approved CHF correlations with the conditions and limitations listed in the SERs approving Dominion TR, DOM-NAF-2-A, and EPRI Report, NP-2511-CCM, "VIPRE-01: A Thermal Hydraulic Analysis Code for Reactor Cores," (Reference 34). The licensee showed in Section 3.6.3 of Reference 16 its compliance with each of the applicable conditions and limitations imposed in the NRC-SERs of TRs, DOM-NAF-2-A and NP-2511-CCM, for the use of VIPRE-0 at MPS3. The licensee also indicated that the use of VIPRE-0 for MPS3 would be in a manner consistent with the conditions and limitations relating to plant-specific and fuel-specific application discussed in Section 3.6.3 of Reference 16. In addition, as discussed in Section 3.1.5 above, the plant-specific and fuel-specific statistical design limit is determined within the context of the statistical DNBR evaluation methods. Therefore, the NRC staff determines that the use of VIPRE-0 discussed in Dominion TR DOM-NAF-2-A and Section 3.6 of Reference 16 is acceptable for the core thermal-hydraulics analysis to support licensing application for MPS3.

3.2 Technical Specifications Changes The proposed TS changes documented in Reference 21 intend to apply the Dominion reload methods to MPS3, and address the issues identified in Westinghouse NSAL-09-5, Rev. 1, NSAL-15-1 and Westinghouse Communication 06-IC-03. The NRC staff has reviewed the proposed TS changes and provided its evaluation as follows.

3.2.1 Deletion of TS 1.43 and 1.44 (ALLOWED POWER LEVEL) - Definitions The licensee proposed to delete the definitions of APL No and APL BL as entries 1.43 and 1.44 in the MPS3 TS. Both definitions would be replaced with the word "Deleted." The staff compared the proposed changes to the STS guidance as part of the review.

In the LAR (Reference 1), the licensee explains this as follows:

Definition 1.43 specifies APL ND as the minimum allowable nuclear design power level for base load operation. The value of APL ND is specified in the COLR. This definition is being deleted since the base load operation mode is not supported by the Dominion methods.

and, Definition 1.44 specifies APL BL as the maximum allowable power level when transitioning into base load operation. This definition is being deleted since the base load operation mode is not supported by Dominion methods.

According to the applicable STS for MPS3, NUREG-1431, Rev. 4, in Section 1.1 Definitions, "The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications." The NRC staff reviewed the licensee's reasons for deletion of these definitions. With the use of Dominion's methodology, definition of these terms is no longer needed because they will no longer appear elsewhere in the TS. The proposed changes are administrative in nature and consistent with use of defined terms in the NUREG-1431, Rev. 4, guidance; therefore, the NRC staff finds them to be acceptable and in accordance with 10 CFR 50.36. Additionally, the licensee proposed to use Dominion RPDC methods to replace the RAOC methods for future cycles of MPS3. Dominion RPDC methods do not support the base load operation mode. The proposed deletion is consistent the approved Dominion methods discussed in Section 3.1.2 of this safety evaluation (SE), and therefore, is acceptable.

3.2.2 TS 3.2.1.1, SR 4.2.1.1 - Axial Flux Difference and TS 3.2.2.1, SR 4.2.2.1 - Heat Flux Hot Channel Factor - Fa(Z)

The proposed changes involve additions, deletions and revisions to existing content in the TS that are associated with TS 3.2.1.1, TS 3.2.2.1, SR 4.2.1.1, and SR 4.2.2.1. These changes accomplish three key objectives: (1) accommodate implementation of the Dominion RPDC method; (2) removal of base load operation; and (3) provide resolution of issues documented in Westinghouse letters NSAL-09-5, Rev. 1 (Reference 23), 06-IC-03 (Reference 24) and NSAL-15-1 (Reference 25). The NRC staff provided its evaluation of the specific proposed changes as follows:

3.2.2.1 TS LCO 3.2.1.1

  • TS LCO 3.2.1.1.a states that 'The limits specified in the CORE OPERATING LIMITS REPORT (COLR) for Relaxed Axial Offset Control (RAOC) operation, or"
  • TS LCO 3.2.1.1 Action a. states that "For RAOC operation with the indicated AFD outside of the applicable limits specified in the COLR."

The proposed changes to above TS 3.2.1.1.a and TS 3.2.1.1 Action a. deletes a reference to RAOC operation. The changes are to revise the TS to be consistent with the licensee's intent to make the TS more general regarding specific axial power distribution control methodology.

Existing LCO 3.2.1.1 Action a. would be modified by removing the words "For Relaxed Axial Offset Control (RAOC) operation" since RAOC operation will no longer be employed. These words distinguish maintenance of the AFD within the limits specified in the COLR from optional part b which was to maintain the AFD within a target band about the flux difference during base load operation. Additionally the letter "w" in the word "With ... " is capitalized to begin revised Action a.

Since the RAOC methods are replaced by the approved Dominion RPDC methods discussed in Section 3.1.2 of this SE, the deletion of a reference to RAOC is acceptable.

  • TS LCO 3.2.1.1.b states that "Within the target band about the target flux difference during base load operation, specified in the COLR."
  • TS LCO 3.2.1.1 Action b.1 and b.2 states that "For base load operation above APL No with the indicted AFD outside of the applicable target band about the target flux differences:
1. Either restore the indicated AFD to within the COLR specified target band within 15 minutes, or
2. Reduce THERMAL POWER to less than APLN° of RATED THERMAL POWER and discontinue base load operation within 30 minutes."

The above TS 3.2.1.1.b, TS 3.2.1.1 Action b.1 and Action b.2 are being deleted, and replaced for each with the word "Deleted". The TSs are associated with the base load mode of operation, which is supported by the RAOC methods for the current cycle. The approved Dominion methods to replace the RAOC methods, as discussed in Section 3.1.2 for MPS3 future cycles, do not support the base load of operation. Therefore, the proposed TS deletions are acceptable.

3.2.2.2 Surveillance Requirements 4.2.1.1.3 and 4.2.1.1.4 SR 4.2.1.1.3 states that "When in base load operation, the target flux difference of each OPERABLE excore channel shall be determined by measurement at the frequency specified in the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable."

SR 4.2.1.1.4 states that "When in base load operation, the target flux difference shall be updated at the frequency specified in the Surveillance Frequency Control Program by either determining the target flux difference in conjunction with the surveillance requirements of Specification 4.2.1.1.3 or by linear interpolation between the most recently measured value and the calculated value at the end of cycle life. The provisions of Specification 4.0.4 are not applicable."

The above SR 4.2.1.1.3 and SR 4.2.1.1.4 are being deleted and replaced with the word "Deleted". The SRs are related to the base load mode of operation, which is supported by the RAOC methods for the current cycle. The approved Dominion methods to replace the RAOC methods, as discussed in Section 3.1.2 above for MPS3 future cycles, do not support the base load of operation. The NRC staff has reviewed the proposed deletion of the above SRs and determined that they are consistent with the change to Dominion's RPDC methods and sufficient to continue to comply with 10 CFR 50.36(c)(3). Therefore, the proposed deletions of the SRs are acceptable.

3.2.2.3 TS LCO 3.2.2.1 TS LCO 3.2.2.1 is modified from:

Fo(Z) shall be limited by the following relationships:

Fo(Z) :5 (FoRTP/P) K(Z) for P > 0.5 Fo(Z) ::; (FoRTP/0.5) K(Z) for P ::; 0.5 F0 RTP = the Fa limit at RA TED THERMAL POWER (RTP) provided in the CORE OPERATING LIMITS REPORT (COLR).

Where: P =THERMAL POWER I RATED THERMAL POWER, and K(Z) = the normalized Fo(Z) as a function of core height specified in the COLR.

To:

Fo(Z), as approximated by FoM(Z), shall be within the limits specified in the COLR.

Use of a relationship where Fo(Z) is approximated by FoM(Z) is consistent with the approved Dominion RPDC methodology and agrees with the guidance of NUREG-1431, Revision 4 where a similar relationship is used in RAOC-W(Z) Methodology or CAOC-W(Z) Methodology. Limits for the specification vary with each specific core reload, and are therefore, located in the COLR.

The NRC staff finds that this is consistent with current technical specifications (CTS) and the NUREG-1431, Revision 4 guidance.

The relationships for Fo(Z) for P > 0.5 or P ::; 0.5 and a supporting description are relocated to the TS bases. The basic requirement of this LCO is that Fo(Z) remain within appropriate limits to prevent fuel damage. This requirement will remain in the TS as an LCO. The detail of the limits that vary with each specific core reload will be located in the COLR. The detail of the specific relationships was determined to not be needed in TS during formulation of the TS improvement program that resulted in this detail being relocated to documents under licensee control for creation of the improved TS of NUREG-1431. Therefore, these proposed changes are consistent with the guidance of NUREG-1431, Revision 4 and are acceptable.

LCO 3.2.2.1 Action a. is modified from:

a. For RAOC operation with Specification 4.2.2.1.2.b not being satisfied or for base load operation with Specification 4.2.2.1.4.b not being satisfied:

To:

a. With Specification 4.2.2.1.2.b not being satisfied:

LCO 3.2.2.1 Action a. is modified by removing the text associated with RAOC and base load operation since these power distribution control methods will no longer be used. The modified action statement is more generic and also applies to Dominion's RPDC methods; therefore, the changes are acceptable.

Existing LCO 3.2.2.1 Action b. states:

b. For RAOC operation with Specification 4.2.2.1.2.c not being satisfied, one of the following ACTIONS shall be taken:

( 1) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in the CORE OPERATING LIMITS REPORT by at least 1% AFD for each percent Fa(Z) exceeds its limits.

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm setpoints to these modified limits, or (2) Verify that the requirements of Specification 4.2.2.1.3 for base load operation are satisfied and enter base load operation.

Where it is necessary to calculate the percent that Fa(Z) exceeds the limits for item (1) above, it shall be calculated as the maximum percent over the core height (Z),

consistent with Specification 4.2.2.1.2.f, that Fa(Z) exceeds its limits by the following expression: [equation not shown]

In LCO 3.2.2.1 Action b., the text "For RAOC operation ... " is removed from the beginning of the action statement. This is acceptable because RAOC will no longer be used. The "W" in the word "with" is capitalized and the action continues to apply when F0 M(Z) does not meet the equilibrium limits. Additionally, existing required actions in LCO 3.2.2.1 Action b(1) is split among new sub-actions b(1 )a, b(1 )b, b(1 )c, b(1 )d, b(1 )e, and b(1 )f. The equation for determining the percent by which Fa(Z) exceeds its limits in LCO 3.2.2.1 Action b. is relocated to the TS bases. This is acceptable because the detail of the specific relationships was determined to not be needed in the TSs during formulation of the TS improvement program that resulted in this detail being relocated to documents under licensee control for creation of the improved TS of NUREG-1431. The change is consistent with the NUREG-1431, Revision 4 guidance, and is therefore, acceptable.

LCO 3.2.2.1 Action b(1 ). would be revised as follows:

a. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, control the AFD to within the new reduced AFD limits specified in the COLR that restores Fa(Z) to within its limits, and
b. Reduce the THERMAL POWER by the amount specified in the COLR that restores Fa(Z) to within its limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and
c. Reduce the Power Range Neutron Flux - High Trip Setpoints by ~ 1% for each 1% that the THERMAL POWER level is reduced within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and
d. Reduce the Overpower~ T Trip Setpoints by ~1 % for each 1% that the THERMAL POWER level is reduced within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and
e. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD Alarm Setpoints to the modified limits, and
f. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION b(1)(b) above; THERMAL POWER may then be increased provided Fa(Z) is demonstrated through incore mapping to be within its limits.

The licensee states in Reference 1 that the proposed changes in TS LCO 3.2.2.1 Action b. will

" ... incorporate a modified version of the interim actions identified in NSAL-09-5, Rev. 1

[Reference 23), in the event that SR 4.2.2.1.2.c is not satisfied." Additionally, the licensee asserts the following:

This approach was determined by Dominion analysis to most appropriately address the issues in NSAL-09-5, Rev. 1 for MPS3. The allowable operating space that applies to TS 3.2.2.1 - ACTION, step b, is relocated to the COLR. A new table, entitled "Required Operating Space Reductions for Fa(Z) Exceeding its Non-Equilibrium Limits," will be added to the COLR to quantify the required THERMAL POWER and AFD limits associated with different amounts of non-equilibrium Fa(Z) margin improvement (1 %, 2%, etc.). If TS 3.2.2.1 - ACTION, step b is entered, the operating space as defined in the new COLR table will ensure that sufficient margin exists. Including the numerical specification of the operating space in the COLR provides greater assurance that the recommended actions are acceptable without regard to the specific power distribution control methodology. The proposed change can be applied under either the Westinghouse (RAOC) or Dominion (RPDC) power distribution control methodologies for a given reload cycle.

For Action b(1 )a, the former action to control AFD to within the new AFD limits was originally proposed with a completion time of 15 minutes and was to be retained as new Action b(1 )a except that with the change, the limit relationship would now be specified in the COLR. The relocation of the relationship to the COLR is allowed since the Dominion RPDC strategy uses a calculated AFD (delta-I) band that varies with each core reload. This is acceptable because the detail of the specific relationships was determined to not be needed in the TSs during

formulation of the TS improvement program that resulted in this detail being relocated to documents under licensee control for creation of the improved TS of NUREG-1431. Therefore, this is consistent with the NUREG-1431, Revision 4 guidance. The original completion time of 15 minutes for control of AFD to within limits was proposed to be changed in the licensee's response to RAls (Reference 4 ).

Action b(1 )bis a newly proposed action to reduce thermal power to the amount specified in the COLR that restores Fa(Z) to within its limits. The licensee determined that reduction of thermal power to the amount specified in the COLR that restores Fa(Z) to within its limits appropriately addresses the issues in NSAL-09-5, Rev. 1. However, a 15 minute time is allowed for reduction of Fa(Z) elsewhere in the CTS (e.g. LCO 3.2.2.1, Action a(1 )). Proposed Actions b(1) differ from associated with proposed Action a(1 ). The proposed Actions b(1) are for Fa(Z) exceeding its non-equilibrium limits and the proposed Actions a(1) are for Fa(Z) exceeding its equilibrium limits. The NRC staff issued RAI #1 (Reference 26) dated February 24, 2016, requesting additional justification for the 4-hour completion time proposed for LCO 3.2.2.1 Action b(1 )(b).

The licensee responded via letter dated March 23, 2016 (Reference 4), stating that the augmentation of the Fa(Z) by the cycle dependent function is mathematically equivalent to the Faw(z) nomenclature described in NUREG-1431 for LCO 3.2.1 B, Action Band its associated technical basis. DNC responded that the augmentation of Fa(Z) by the cycle dependent function was mathematically equivalent to the Faw(z) nomenclature in NUREG-1431 for LCO 3.2.1 B, Action B and its basis. DNC supplemented its response by letter dated May 2, 2016, (Reference 6). In the supplement, the licensee provided the below technical justification for the use of the 4-hour completion time instead of the 15-minute completion time:

The technical justification for a 4-hour completion time for Action b to LCO 3.2.2.1 Action b( 1), instead of the 15-minute completion time in LCO 3.2.2.1 Action a(1 ), can be explained through a comparison of the different scenarios under which the LCOs are entered:

Action a(1) of LCO 3.2.2.1 is entered when surveillance requirement 4.2.2.1.2.b is not met. This surveillance requirement is to address an active violation of Fa(Z) limits. When measured Fa(Z) is above its limit, a 15-minute action time is appropriate to return FQ(Z) within the limit as quickly as possible.

In contrast, Action b(1) is entered when surveillance requirement 4.2.2.1.2.c is not met. This surveillance requirement is to address the condition when the non-equilibrium (or transient) Fa(Z) limit has not been met. In this case, measured Fa(Z) is not currently above its limit but could exceed its limit if a normal operation transient occurs. A 4-hour completion time is appropriate because a normal operation transient would occur based upon fission product (Xe) time scales and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is sufficient time to restrict [AFD] limits and thermal power so that core peaking factors are not exceeded.

In addition, reducing power and controlling/reducing AFD to be within new limits (and any resultant actions such as insertion of control rods) within a 15-minute time frame could lead to the initiation of a normal operation transient and make it

more likely that core peaking factors could be violated. A 4-hour completion time allows for deliberate operator actions to minimize the initiation of a normal operation transient.

The NRC staff reviewed the proposed changes and determined that the licensee's justification for the 4-hour completion time for proposed Actions b(1 )a and b(1 )bis acceptable. If a power reduction is necessary as proposed in new Action b(1 )b then proposed Actions b(1 )c and b(1 )d reasonably follow to reduce the power range neutron flux - high trip setpoint and overpower L1T trip setpoint respectively. It is reasonable to reduce these RPS trip setpoints to the new allowable power level so that automatic trip can occur at that lower level if necessary to protect the fuel. For proposed Action b(1 )d, the completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to reduce thermal power agrees with the completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed in CTS action a(1) for evaluation of Fo(Z).

The requirement to reduce the power range neutron flux high trip setpoints is additionally equivalent to that in CTS action a(1 ); however, only a 4-hour completion time is allowed versus the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> proposed.

The NRC staff issued RAI #2 dated February 24, 2016 (Reference 26) requesting additional technical justification for the 72-hour completion time in new proposed LCO 3.2.2.1 Action b(1)(c). In its response (Reference 4), as supplemented by letter dated May 2, 2016, (Reference 6), the licensee stated that a 72-hour completion time is appropriate for this action because of the very low probability of a severe accident occurring during this time as opposed to a normal operational transient and because Action b(1 )a (AFD limit reduction) and Action b(1 )b (thermal power reduction) will be performed under a 4-hour completion time which reduces possible initial conditions that form the starting point for a severe accident. Additionally, minimizing or reducing possible initial conditions that form the starting point for a severe accident increases the likelihood that achievable power shapes that could occur during a severe accident have already been considered in the safety analysis calculation. The NRC staff has reviewed the proposed changes and the licensee's RAI responses and finds the 72-hour completion time for proposed LCO 3.2.2.1 Action b(1 )c is acceptable.

Proposed Action b(1 )e relocates part of the existing Action b(1) to reduce the AFD alarm setpoint within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The proposed change provides for administrative relocation of the applicable requirement and is editorial in nature; the NRC staff finds this proposed change acceptable.

Proposed Action b(1 )(f) is a restorative action to correct the cause of the out of limit condition prior to increasing thermal power. This action applies when heat flux hot channel factor non-equilibrium limits are exceeded versus the equilibrium limits of Action a. The proposed change is a restorative action and consistent with the CTS; the NRC staff finds this proposed change acceptable.

Existing LCO 3.2.2.1 Action b(2) is deleted because it applies only to base load operation which will no longer be used. The proposed change provides is administrative in nature; the NRC staff finds it acceptable.

The proposed changes to LCO 3.2.2.1, Actions a. and Action b. are acceptable and consistent with the corresponding LCO 3.2.1 B, Action B, of NUREG-1431, Rev. 4, guidance. Therefore,

the NRC staff finds that LCO 3.2.2.1 as proposed will meet the requirements of 10 CFR 50.36(c)(2).

3.2.2.4 SR 4.2.2.1.2, SR 4.2.2.1.3, SR 4.2.2.1.4, and SR 4.2.2.1.5 SR 4.2.2.1.2 is modified by deleting the words "For RAOC operation" from the beginning of the paragraph. This is an administrative change due to MPS3 changing to Dominion methodology therefore this is acceptable.

SR 4.2.2.1.2.c is modified by inserting "Verify FaM(Z) satisfies the non-equilibrium limits specified in the COLR." The added text more adequately and succinctly describes the intent of the SR. The existing text is removed and the relationship equations are relocated to the TS bases. The licensee states that the limit equations and associated description are already described in the COLR making the information in the TS redundant. The NRC staff finds that these proposed changes adequately retain the requirements of the existing SR while more closely aligning with NUREG-1431, Rev. 4, and are therefore, acceptable.

SR 4.2.2.1.2.e is modified by replacing the existing language with "Compliance with the non-equilibrium limits shall be conservatively accounted for during intervals between FaM(Z) measurements by performing either of the following." Westinghouse NSAL-15-1 identified certain conditions in which the required actions associated with this SR may not provide assurance that the non-equilibrium Fa(Z) LCO limit will be met between the performance of the required surveillance intervals. The NRC staff finds that the proposed change is applicable to MPS3 SR 4.2.2.1.2.e, conservatively addresses the deficiency identified in Westinghouse NSAL-15-1, and is therefore, acceptable.

SR 4.2.2.1.2.f is modified: (1) so that the definition of the non-applicable core regions are moved to the bases, and (2) to narrow the regions to satisfy the Westinghouse guidance of 06-IC-03. The licensee stated that "The intent of 06-1 C-03 was to inform utilities that it is probable for the minimum Fa margin to occur near the top or bottom of the core. In response to the information in 06-1 C-03, the proposed change increases the core plane regions for which the limits apply and reduces the "not applicable" portion at the top and bottom to 8 percent.

Reload methodology confirms transient Fa margin over the entire core height and the 'not applicable' region can be adjusted larger or smaller, as necessary, to ensure peak transient Fa is not in this region. Moving the 'not applicable' region to the Bases allows this adjustment.

Relocating the defined core planar regions (where SRs 4.2.2.1.2.c and 4.2.2.1.2.e are applicable) to the bases matches the guidance in NUREG-1431, Rev. 4, and is therefore, acceptable.

SRs 4.2.2.1.3 and 4.2.2.1.4 will be deleted since they only applied to base load operations which will no longer be used. SR 4.2.2.1.5 will be revised to delete the reference to SR 4.2.2.1.4. The NRC staff has reviewed the proposed changes to the above SRs and determined that they are consistent with the change to Dominion's RPDC methods and sufficient to continue to comply with 10 CFR 50.36(c)(3). Therefore, the proposed changes to the SRs are acceptable.

3.2.3 TS 3.2.3.1.b(5) - Deletion of the Value of the RCS Flow Rate Measurement Uncertainty TS 3.2.3.1.b(5) currently states that "The measured value of RCS total flow rate shall be used since uncertainties of 2.4 percent for flow measurement have been included in Specification 3.2.3.1.a". The licensee proposed to delete the value (2.4 percent) of the flow measurement uncertainty from TS 3.2.3.1.b(5).

The licensee justified the proposed deletion by stating that the value of the RCS flow uncertainty does not meet any of the following categories defined in 10 CFR 50.36 for items required to be in TS. TS categories are: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; (5) administrative controls; (6) decommissioning; (7) Initial notification; or (8) written reports.

Paragraph (c)(2)(ii) of 10 CFR 50.36 states that a TS must be established for each item meeting one or more of the following criteria:

  • Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
  • Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
  • Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
  • Criterion 4: A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

RCS total flow rate, which is already part of TS 3.2.3.1.a, meets Criterion 2. In the FSAR Chapter 15 analyses, the RCS total flow rate is used as an input value based on the nominal conditions (for statistical events) with consideration of the RCS flow uncertainty, while the thermal design flow rate is used as an input value based on the deterministic conditions (nominally biased for uncertainty). The difference between the RCS total flow rate and the thermal design flow rate bounds the RCS flow uncertainty. This approach is consistent with reload evaluation methods of VEP-FRD-42-A (Reference 7) and WCAP-9272-P-A (Reference 17), as well as the statistical DNB methodologies of VEP-NE-2-A (Reference 11) and WCAP-11397-P-A (Reference 28).

The NRC staff finds that the value of the flow measurement uncertainty does not meet:

(1) criterion 1 discussed above since it is not used to detect radiological releases; (2) criterion 2 since it is not used as an initial condition in a design basis accident and or transient analysis; (3) criterion 3 since is not part of the primary success path used for mitigating a design basis accident or transient; and (4) criterion 4 since it is not a component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. Therefore, NRC staff determines that the proposed deletion of the value (2.4 percent) of the flow measurement uncertainty from the TSs meets the requirements of Paragraph (c)(2)(ii) in 10 CFR 50.36. Also, the NRC staff finds that the proposed TS deletion is

consistent with NUREG-1431 (Reference 15), since the STSs do not contain values of the RCS flow uncertainties.

Based on the above discussion, the NRC staff determines that the proposed deletion of the value (2.4 percent) of the flow measurement uncertainty from the TS meets the 10 CFR 50.36 requirements and is consistent with Westinghouse STS documented in NUREG-1431.

Therefore, the NRC staff determined that the proposed TS change is acceptable.

3.2.4 SR 4.2.3.1.3.a - RCS Flow Rate Measurement Time Current SR 4.2.3.1.3.a states that "Verifying by precision heat balance that the RCS total flow rate is> 363,200 gallons per minute (gpm) and greater than or equal to the limit specified in the COLR within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 90 percent of RATED THERMAL POWER after each fuel loading, and ... " The proposed changes to SR 4.2.3.1.3.a relax the time requirement to perform the precision heat balance from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 7 days after reaching 90 percent of the rated thermal power. The licensee justified the completion time relaxation by stating that 7 days will allow the establishment of stable operating conditions, installation of the test equipment, performance of the test, and completion of the analysis. The proposed 7-day period is consistent with that approved for SPS. The NRC SER (Reference 27) approving the flow rate measurement time for SPS provided a basis for acceptance stating that the 7-day period is adequate to complete an accurate flow rate measurement, which increases the quality of confirmation available to the licensee that it is operating its plants within safety analysis limits.

Both SPS and MPS3 are owned and operated by Dominion; since their nuclear steam supply systems were designed and manufactured by Westinghouse, the differences in the design features of 4-RCS loops for MPS3 and 3-RCS loops for SPS will not affect the basis for establishing the 7-day flow rate measurement time. The NRC staff determines that the basis of its acceptance of the flow rate measurement time for SPS is applicable to MPS3, and thus, the proposed measurement time of 7 days is acceptable for MPS3.

3.2.5 TS 6.9.1.6.a - Core Operating Limits Report (COLR)

TS 6.9.1.6.a lists 12 cycle-specific core operating limits to be included in the COLR. The proposed changes to Items 7 and 8 in the core operating limits list of TS 6.9.1.6.a are to delete the "target band, and APL ND" from Item 7 and "K(z), W(z), and APL ND" from Item 8. The proposed deletions are terminologies associated with specific power distribution control methodology and base load operation. The revised Items 7 and 8 are as follows:

7. AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1.1.
8. Heat Flux Hot Channel Factor Limits for Specification 3/4.2.2.1.

These proposed changes are acceptable, since they are consistent with the changes described in Section 2.2 of Reference 21 and the NRC staffs evaluation discussed in above Section 3.2.2 of the SER, where the proposed changes to TS 3/4.2.1.1 and 3/4.2.2.1 eliminate base load operation and relocate from TS the equations and terminology for either RAOC or RPDC transient multiplication factors.

The proposed change to Item 9 is also acceptable, since the change, renumbering TS 3/4.2.3 to TS 3/4.2.3.1 to align the associated TS subsection with that specified in the COLR, is an editorial change and does not change the technical content of the TS requirements.

3.2.6 TS 6.9.1.6.b - NRG-Approved Methods Referenced in COLR 3.2.6.1 Statement Preceding TS 6.9.1.6.b Reference List The licensee proposed changes to the statement preceding the reference list in TS 6.9.1.6.b.

The added wording requires that the cycle-specific COLR identify the full reference citation of the TRs used to support that cycle, including citations of the report number, title, revision, date, and any supplements. The NRC staff determines that the change is acceptable, since it is consistent with guidance of GL 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," (Reference 14). Specifically, the GL states, in part, that the analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC. The number, title, and date of the TRs documenting the methodologies used for determining the core operating limits should be identified. The TS change is also consistent with the "Reviewer's Note" in TS 5.6.3.b of NUREG-1431 (Reference 15), which provides clarification that the reference methodologies used for a reload core should be specifically identified in the cycle-specific COLR.

3.2.6.2 Addition of Approved Dominion Methodologies to the TS 6.9.1.6.b Reference List MPS3 TS 6.9.1.6.b currently states that: "The analytical methods used to determine core operating limits shall be those previously reviewed and approved by the NRC." This TS also provides a list of NRG-approved analytical methods for MPS3. In order to use the analytical methods described in Reference 16, the licensee proposed to add the NRG-approved methodologies documented in TR VEP-FRD-42, TR VEP-NE-1, TR VEP-NE-2, and TR DOM-NAF-2 to the list in TS 6.9.1.6.b as References 20 through 23, respectively. The revised Items 20 through 23 are as follows:

20. VEP-FRD-42-A, "Reload Nuclear Design Methodology." Methodology for Specifications
  • 2.1.1 Reactor Core Safety Limits
  • 3.1.1.1.2 SHUTDOWN MARGIN- MODES 3, 4 and 5 Loops Filled
  • 3.1.1.3 Moderator Temperature Coefficient
  • 3.1.3.5 Shutdown Rod Insertion Limit
  • 3.2.2.1 Heat Flux Hot Channel Factor
  • 3.2.3.1 Nuclear Enthalpy Rise Hot Channel Factor
  • 3.9.1.1 REFUELING Boron Concentration
21. VEP-NE-1-A, "Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications." Methodology for Specifications:
  • 3.2.1.1 AXIAL FLUX DIFFERENCE
  • 3.2.2.1 Heat Flux Hot Channel Factor
22. VEP-NE-2-A, "Statistical DNBR Evaluation Methodology." Methodology for Specifications:
  • 3.2.3.1 RCS Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor
  • 3.2.5 DNB Parameters
23. DOM-NAF-2-P-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix C, "Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code." Methodology for Specifications:
  • 3.2.3.1 RCS Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor
  • 3.2.5 DNB Parameters As discussed in Sections 3.1 above, the added TRs are acceptable for use in the analysis supporting licensing applications for MPS3. Therefore, the NRC staff determines that TS 6.9.1.6.b with the added TRs is acceptable.

TRs (a) and (b) below also document the methodologies that are acceptable for use in the reload analysis at MPS3.

(a) DOM-NAF-1-P-A, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations."

(b) VEP-FRD-41-P-A, "VEPCO Reactor System Transient Analyses Using the RETRAN Computer Code."

Dominion does not include the above two TRs in TS 6.9.1.6.b. In response to NRC RAl-18 (SRXB) (Reference 2) regarding the adequacy of Dominion's approach, Dominion indicated that its approach of not including TR (a) and TR (b) above in TS 6.9.1.6.b is consistent with the application of Dominion's approved reload methods for North Anna (TS 5.6.5.b) and Surry (TS 6.2.C). Also, the MPS3 currently does not contain the Westinghouse equivalent references to above TR (a) and TR (b).

In addition, Dominion indicated (in its response to RAl-18 (SRXB), Reference 2) that the added Reference 10, VEP-FRD-42-A, "Reload Nuclear Design Methodology", contains in Section 2.2 the methodology (discussed in TR (a), DOM-NAF-1-P-A) for calculating reload core physical parameters. Sections 2.1.3 and 3.3 contains the methodology (discussed in TR (b),

VEP-FRD-41-P-A) for use of RETRAN in the reload analysis. Section 2.3 and Appendix B of the added Reference 10, VEP-FRD-42-A, discusses the process by which either analytical models

or methods could achieve approved status for use in Dominion's reload methodology. The process is based on the following NRC requirements and industry guidance:

  • GL 88-16, "Removal of Cycle-Specific Parameter Limits From Technical Specifications" (Reference 14)
  • GL 83-11, Supplement 1, "Licensee Qualifications for Performing Safety Analyses" (Reference 33)
  • 10 CFR 50.59, "Changes, tests and experiments." and in particular, 10 CFR 50.59(c)(2)(viii) states, in part, that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.

Based on the above discussion, the NRC staff finds that: (1) the approach of not including TR (a) and TR (b) listed above is consistent with the application of Dominion's approved reload methods to North Anna (TS 5.6.5.b) and Surry (TS 6.2.C); (2) MPS3 currently does not contain the Westinghouse equivalent references to the above TR (a) and TR (b); and (3) the Dominion existing process of changes to approved methodologies is based on the NRC and industry guidance, including the 10 CFR 50.59(c)(2)(viii) requirements. Therefore, the NRC staff determines that the licensee's approach of not including TR DOM-NAF-1-P-A and TR VEP-FRD-41-P-A is acceptable.

3.2.6.3 Changes of the TS 6.9.1.6.b Reference List The licensee proposed various changes to the TS 6.9.1.6.b Reference List. The NRC staff has reviewed the changes and provides its evaluation as follows:

3.2.6.3.1 Changes for Readability Improvement The licensee proposed to reformat the specifications listed under each reference in TS 6.9.1.6.b using bullets to improve readability. Also, it made minor changes to these specifications to reflect conformance to the usage that is appropriate for either Westinghouse or Dominion references. Since the proposed changes are editorial and do not change the technical content of the TS requirements, the NRC staff determines the proposed changes are acceptable.

3.2.6.3.2 Modifications to Reference 1 of TS 6.9.1.6.b The licensee proposed Reference 1 (WCAP-9272-P-A) of TS 6.9.1.6.b that uses TS 2.1.1 in the list to substitute TS 2.1.1.1, "Departure from Nuclear Boiling Ratio", and 2.1.1.2, "Peak Fuel Centerline Temperature". The proposed change is to align with Item 1 in TS 6.9.1.6.a for which the reference is applicable. In addition, the licensee proposed to add TS 3.2.5, "DNB Parameter", and 3.3.5, "Shutdown Margin Monitor", to the list since both TSs did not previously appear in Reference 1 of TS 6.9.1.6.b. The NRC staff finds that the proposed changes provide

additional clarification for the appropriate use of the cited Reference 1, and therefore, are acceptable.

3.2.6.3.3 Retaining of References 1 and 4 in TS 6.9.1.6.b The licensee proposed to retain the following References 1 and 4 in TS 6.9.1.6.b:

1. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," (W Proprietary). Methodology for Specifications:
  • 2.1.1 Reactor Core Safety Limits
  • 3.1.1.2 SHUTDOWN MARGIN-Cold Shutdown - Loops Not Filled
  • 3.1.1.3 Moderator Temperature Coefficient
  • 3.1.3.5 Shutdown Rod Insertion Limit
  • 3.2.1.1 AXIAL FLUX DIFFERENCE
  • 3.2.2.1 Heat Flux Hot Channel Factor
  • 3.2.3.1 RCS Total Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor
  • 3.9.1.1 REFUELING Boron Concentration
  • 3.2.5 DNB Parameters
4. WCAP-10216-P-A-R1A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION,"

(W Proprietary). (Methodology for Specifications 3.2.1.1--AXIAL FLUX DIFFERENCE and 3.2.2.1--Heat Flux Hot Channel Factor)

Reference items 1 and 4 of TS 6.9.1.6.b pertain to Westinghouse reload methods. The proposed list of reference methodologies in TS 6.9.1.6.b would contain both Westinghouse and Dominion references. As stated by the licensee in the LAR, the two reference items listed above "are being retained in TS 6.9.1.6.b since these methodologies are applicable to Westinghouse for establishing core operating limits and may be used for a specific core during the transition to Dominion methods." Additionally, the licensee states "the references used in a cycle-specific COLR will be a subset of the TS 6.9.1.6.b methodologies that are applicable to the specific reload cycle. If Westinghouse reload methods are used, then Westinghouse reload methods shall be listed in the cycle-specific COLR. If Dominion methods are used, then Dominion reload methods shall be listed."

As stated in the licensee's LAR, Section 3.2 (page 10 of 20), "the proposed changes are structured in a manner that is independent of specific power distribution control methodology (RAOC or RPDC). Relocating the specific equations associated with either the Westinghouse or Dominion power distribution control methodologies to the Bases is consistent with the guidance contained in NUREG-1431, Rev. 4 (Reference 3).

The NRC staff notes that the proposed changes to the affected LCOs and associated SRs involved replacement of terminology that applies only to the current Westinghouse-based methodology (RAOC or RAOC) with language that is not associated with a particular methodology. Additionally, specific equations containing terms that are unique to either methodology would be relocated to the Bases. This makes the final resultant TS LCOs independent of the particular methodology, so that either the Westinghouse RAOC or Dominion RPDC may be used. Also, this TS structure is not dependent upon any required or implied timeframe in which to transition to Dominion methods. Therefore, no issues or unintended consequence would be created from a postulated delay in transition to Dominion methods.

The listed references document the NRG-approved Westinghouse methodologies that are applicable to MPS3 for establishing core operating limits and are used for a specific core during the transition to Dominion methods; therefore, the NRC staff determines that the proposed TS 6.9.1.6.b is acceptable during the transition to Dominion methods.

Additionally, in Reference Item 4, the specification number 3.2.1 would be revised to 3.2.1.1; the text "[Relaxed Axial Offset Control)" is removed, specification number 3.2.2 is revised to 3.2.2.1, and the text "[W(z) surveillance requirements for FQ Methodology)" is removed. The proposed change to Reference item 4 is acceptable, since the revision provides additional clarity by adding the correct specification to which the methodology applies.

3.2.6.3.4 Retaining of References 4 through 10, 16, 17, 18 and 19 in TS 6.9.1.6.b The licensee proposed to retain and revise for clarity of specification the following References in TS 6.9.1.6.b.: 4, 5, 6, 7, 8, 9, 10, 16, 17, 18 and 19:

4. WCAP-10216-P-A-R1A, "RELAXATION OF CONSTANT AXIAL OFFSETCONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION,"

(W Proprietary). (Methodology for Specifications 3.2.1.1--AXIAL FLUXDIFFERENCE and 3.2.2.1--Heat Flux Hot Channel Factor)

5. WCAP-12945-P-A, "CODE QUALIFICATION DOCUMENT FOR BEST ESTIMATE LOCA ANALYSIS," (W Proprietary). (Methodology for Specification 3.2.2.1--Heat Flux Hot Channel Factor.)
6. WCAP-16009-P-A, "REALISTIC LARGE-BREAK LOCA EVALUATION METHODOLOGY USING THE AUTOMATED STATISTICAL TREATMENT OF UNCERTAINTY METHOD (ASTRUM)," (W Proprietary). (Methodology for Specification 3.2.2.1--Heat Flux Hot Channel Factor.)
7. WCAP-11946, "Safety Evaluation Supporting a More Negative EOL Moderator Temperature Coefficient Technical Specification for the Millstone Nuclear Power Station Unit 3," (W Proprietary). Methodology for Specification:
  • 3.1.1.3 - Moderator Temperature Coefficient
8. WCAP-10054-P-A, "WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE," (W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)
9. WCAP-10079-P-A, "NOTRUMP -A NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," (W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)
10. WCAP-12610, "VANTAGE+ Fuel Assembly Report," (W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)
16. WCAP-8301, "LOCTA-IV Program: Loss-of-Coolant Transient Analysis."

Methodology for Specification:

  • 3.2.2.1 - Heat Flux Hot Channel Factor
17. WCAP-10054-P-A, Addendum 2, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model." Methodology for Specification:
  • 3.2.2.1 - Heat Flux Hot Channel Factor
18. WCAP-87 45-P-A, "Design Bases for the Thermal Overpower LiT and Thermal Overtemperature DT Trip Functions," (Westinghouse Proprietary Class 2)

(Methodology for Specifications 2.2.1 - Overtemperature LlT and Overpower LlT Setpoints.)

19. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO',"

(W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)

In Items 5, 6, 8, 9, and 10, and 19 specification number 3.2.2 is replaced with 3.2.2.1. In Item 7 the text: "Methodology for Specification:" and bulleted item "3.1.1.3 - Moderator Temperature Coefficient" are added to the end of the reference. For Items 15 and 16, the specification 3.2.2.1 is included as the applicable methodology. For Item 17, the applicable specification for 2.2.1, "Overtemperature LlT and Overpower LlT Setpoints" is added to the end of the reference.

The licensee states in its LAR that references 5, 6, 7, 8, 9, 10, 16, 17, 18 and 19, are Westinghouse TRs which document methodologies that are independent of the scope for nuclear safety analysis and core design methods which Dominion is applying. Dominion asserts that these references remain applicable to either Westinghouse or Dominion reload core design methods.

The NRC finds that the above references document Westinghouse methodologies are unrelated to the proposed application of Dominion reload methods to MPS3. Therefore, these references

remain applicable for use with either Westinghouse or Dominion reload methods. The proposed changes to the above items are also acceptable, since the revisions provide additional clarity by adding the correct specification to which the methodology applies, is editorial in nature, and does not involve any changes to the technical content of the TS requirements.

3.2.6.3.5 Deletion of References from TS 6.9.1.6.b Reference List The licensee proposed to delete the following References 2, 3, 11, 12, 13, 14 and 15 from TS 6.9.1.6.b and replace each with the word "deleted". :

2. T. M. Anderson to K. Kniel (Chief of Core performance Branch, NRC), January 31, 1980 -

Attachment:

Operation and Safety Analysis Aspects of Improved Load Follow Package.

3. NUREG-800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, "Nuclear Design, July 1981 Branch Technical Position CPB-4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Revision 2, July 1981.
11. Letter from V. L. Rooney (USNRC) to J. F. Opeka, "Safety Evaluation for Topical Report, NUSC0-152, Addendum 4, 'Physics Methodology for PWR Reload Design,'

TAC No. M91815," July 18, 1995.

12. Letter from E. J. Mroczka to the USNRC, "Proposed Changes to Technical Specifications, Cycle 4 Reload Submittal - Boron Dilution Analysis," B13678, December 4, 1990.
13. Letter from D. H. Jaffe (USNRC) to E. J. Mroczka, "Issuance of Amendment (TAC No. 77924)," March 11, 1991.
14. Letter from M. H. Brothers to the USNRC, "Proposed Revision to Technical Specification, SHUTDOWN MARGIN Requirements and Shutdown Margin Monitor OPERABILITY for MODES 3, 4, and 5 (PTSCR 3-16-97), B16447, May 9, 1997.
15. Letter from J. W. Anderson (USN RC) to M. L. Bowling (NNECO), "Issuance of Amendment - Millstone Nuclear Power Station, Unit No.3 (TAC No. M98699),"

October 21, 1998.

The above references were proposed to be deleted from TS 6.9.1.6.b, because they do not describe a methodology that establishes core operating limits. The licensee clarified the proposed deletion by stating that a methodology in the COLR reference list is to satisfy two conditions: (1) the methodology is used to determine core operating limits; and (2) it has been previously approved by the NRC. The NRC staff finds that the licensee's clarification is consistent with the guidance discussed in GL 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications" (Reference 14), which states, in part, that: "generally, the methodology for determining cycle-specific parameter limits is documented in an NRG-approved Topical Report or in a plant-specific submittal." Additionally, the NRC staff agrees with the

licensee's reasoning that these items contain guidance or other information but are not analytical methods used in determining core operating limits, and therefore, may be removed.

Therefore, the NRC determines the proposed deletion of the cited references is acceptable.

3.5 NRC Staff Conclusion

The NRC staff has reviewed the licensee's submittals and supporting documentation and finds that the proposed use of Dominion nuclear core design and safety analysis methods discussed in Section 3.1 and the proposed TS changes discussed in Section 3.2 are acceptable for use in licensing applications at MPS3.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Connecticut State official was notified on May 2, 2016, of the proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register (FR) on June 13, 2016 (81 FR 38226). 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.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) there is reasonable assurance that such activities will be conducted in compliance with the 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.

8.0 REFERENCES

1. Dominion Nuclear Connecticut, Inc. (DNC) Letter, Mark D. Sartain to NRC, "Dominion Nuclear Connecticut Inc., Millstone Power Station Unit 3, License Amendment Request to Adopt Dominion Core Design and Safety Analysis Methods and to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1, NSAL-15-1, and 06-IC-03,"

May 8, 2015, (ADAMS Accession No. ML15134A244).

2. DNC Letter, Mark D. Sartain to NRC, "Dominion Nuclear Connecticut Inc., Millstone Power Station Unit 3, Response to Request for Additional Information Regarding License Amendment Request to Adopt Dominion Core Design and Safety Analysis Methods and to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1, NSAL-15-1, and 06-IC-03 (CAC No. MF6521)," January 28, 2016, (ADAMS Accession No. ML16034A216).
3. DNC Letter, Mark D. Sartain to NRC, "Dominion Nuclear Connecticut Inc., Millstone Power Station Unit 3, Response to Request for Additional Information Regarding License Amendment Request to Adopt Dominion Core Design and Safety Analysis Methods and to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1, NSAL-15-1, and 06-IC-03 (CAC No. MF6521 )," February 25, 2016, {ADAMS Accession No. ML16057A812).
4. DNC Letter, Mark D. Sartain to NRC, "Dominion Nuclear Connecticut Inc., Millstone Power Station Unit 3, Response to Request for Additional Information Regarding License Amendment Request to Adopt Dominion Core Design and Safety Analysis Methods and to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1, NSAL-15-1, and 06-IC-03 (CAC No. MF6521)," March 23, 2016, (ADAMS Accession No. ML16088A140).
5. DNC Letter with two Attachments, Daniel G. Stoddard to NRC, "Dominion Nuclear Connecticut Inc., Millstone Power Station Unit 3, Response to Request for Additional Information Regarding License Amendment Request to Adopt Dominion Core Design and Safety Analysis Methods and to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1, NSAL-15-1, and 06-IC-03 (CAC No. MF6521 ),"

March 29, 2016, (ADAMS Accession No. ML16095A233).

6. DNC Letter, Mark D. Sartain to NRC, "Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 3, Supplement to Response to Request for Additional Information Regarding License Amendment Request to Adopt Dominion Core Design and Safety Analysis Methods and to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1, NSAL-15-1, and 06-IC-03 (CAC No. MF6521 )," May 2, 2016, (ADAMS Accession No. ML16130A563).
7. Topical Report, VEP-FRD-42, Rev. 2.1-A, "Reload Nuclear Design Methodology,"

August 2003, (ADAMS Accession No. ML15313A149).

8. Topical Report, VEP-NE-1, Rev. 0.1-A, "VEPCO Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications," August 2003, (ADAMS Accession No. ML15313A154 ).
9. Topical Report, DOM-NAF-1-P-A, "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations," June 2003, (ADAMS Accession No. ML031690108).
10. Topical Report, VEP-FRD-41-P-A, Rev. 0.2, "VEPCO Reactor System Transient Analyses Using the RETRAN Computer Code," March 2015, (ADAMS Accession No. ML15313A141).
11. Topical Report, VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," June 1987 (Proprietary), (ADAMS Accession No. ML101330527).
12. Fleet Report, DOM-NAF-2-P-A, Rev. 0.3, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," Appendix D & Attachments, September 2014 (Proprietary),

(ADAMS Accession Nos. ML14294A517 for the Fleet Report and ML14294A516 for the SER approving the Report).

13. NUREG-0800, Standard Review Plan, Section 16, Revision 3.0, "Technical Specification," dated March 2010, (ADAMS Accession No. ML100351425).
14. Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," dated October 4, 1988, (ADAMS Accession No. ML031200485).
15. NUREG-1431, Revision 4, Vol. 1 and 2, "Standard Technical Specifications -

Westinghouse Plants," (ADAMS Accession No. ML12100A222).

16. Attachment 4 to Reference 1, "Application of Dominion Nuclear Core Design and Safety Analysis Methods," (ADAMS Accession No. ML15134A244).
17. Westinghouse Topical Report WCAP-9272-P-A, 'Westinghouse Reload Safety Evaluation Methodology," (Proprietary), March 1978, (ADAMS Accession No. ML051390150).
18. Dominion Nuclear Connecticut, Inc. (DNC) e-mail, Wanda D. Craft to NRC, "RAl-12 Supplement RAI Response Adopt Core Design and Safety Analysis Methods - MPS3,"

dated March 3, 2016, (ADAMS Accession No. ML16069A250).

19. Attachment 5 to Reference 1, "RETRAN Benchmarking Information," (ADAMS Accession No. ML15134A244).
20. Attachment 6 to Reference 1, "Development of Statistical Design Limits," (ADAMS Accession No. ML15134A244).
21. Attachment 1 to Reference 1, "Evaluation of Technical Specifications Changes", and Attachment 2 to Reference 1, "Marked-up Technical Specifications pages," (ADAMS Accession No. ML15134A244).
22. USNRC Information Notice (IN) 2014-01, "NRC Information Notice 2014-01: Fuel Safety Limit Calculations Inputs Were Inconsistent With NRG-Approved Fuel Design," dated February 21, 2014, (ADAMS Accession No. ML13325A966)
23. Westinghouse Nuclear Safety Advisory Letter, NSAL-09-5, Rev. 1, "Relaxed Axial Offset Control FQ Technical Specification Actions," September 23, 2009.

24 Westinghouse Notice "06-IC-03, Fa and Fxy Surveillance Zone Issue," February 21, 2006.

25. Westinghouse Nuclear Safety Advisory Letter, NSAL-15-1, "Heat Flux Hot Channel Factor Technical Specification Surveillance," February 3, 2015, (ADAMS Accession No. ML15105A102).
26. E-mail from R.Guzman to W.Craft, Re: Request for Additional Information - LAR to Adopt Dominion Core Design and Safety Analysis Methods (MF6251 ). Millstone Power Station, Unit 3 RAI #2 dated February 24, 2016, (ADAMS Accession No, ML16055A530).
27. Letter from K. Cotton (USNRC) to D. A. Heacock (Dominion), "Surry Power Station, Unit Nos. 1 and 2, Issuance of Amendments Regarding Request for Technical Specification Revisions Related to the Core Operating Limits Report (TAC Nos. ME2591 and ME2592)," dated October 19, 2010 (ADAMS Accession No. ML102530115); corrected by Letter from K. Cotton (USNRC) to D. A. Heacock (Dominion), "Surry Power Station, Unit Nos. 1 and 2, Correction to Amendments Regarding Technical Specification Revisions Related to the Core Operating Limits Report (TAC Nos. ME2591 and ME2592)," October 21, 2010, (ADAMS Accession No. ML102930565).
28. WCAP-11397-A, "Revised Thermal Design Procedure" Proprietary, (ADAMS Accession No. ML080630437).
29. Attachment 5 of DNC Letter 07-0450, "Dominion Nuclear Connecticut, Inc., Millstone Power Station Unit 3 License Amendment Request Stretch Power Uprate," July 13, 2007, (ADAMS Accession No. ML072000386).
30. Letter from S. A. Richard (NRC) to G. L. Vine (EPRI}, "Safety Evaluation Report on EPRI Topical Report NP-7450(P), Revision 4, 'RETRAN-30 -A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems,"' January 25, 2001, (TAC No. MA4311 ). (ADAMS Accession No. ML010470342).
31. Nuclear Energy Institute (NEI) 96-07, Revision 1, "Guidelines for 10 CFR 50.59 Implementation," (ADAMS Accession No. ML003771157).
32. USN RC Regulatory Guide (RG) 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments" (endorses NEI 96-07, Rev. 1), (ADAMS Accession No. ML003759710).
33. GL 83-11, Supplement 1, "Licensee Qualifications for Performing Safety Analyses" (NUDOCS Accession Number 9906210103).
34. Electric Power Research Institute (EPRI), "VIPRE-01: A Thermal Hydraulic Code for Reactor Cores," NP-2511-CCM-A, Revision 4, Palo Alto, CA, June 2007 (ADAMS Accession Nos. ML102090545, ML102090544, ML102090543, and ML102070202; Non-Publicly Available).

Principal Contributors: Summer Sun Pete Snyder Mathew M. Panicker Date: July 28, 2016

  • .. ML16131A728 *SE memo dated OFFICE NRR/DORULPLl-1 /PM NRR/DORULPLl-1 /LA DSS/SRXB/BC DSS/STSB/BC NAME RGuzman KGoldstein EOesterle* AKlein*

DATE 5/23/2016 5/27/2016 5/02/2016 5/11/2016 OFFICE DSS/SNPB/BC OGC DORULPLl-1 /BC DORL/LPLl-1 /PM NAME JDean* BHarris TT ate RGuzman DATE 5/10/2016 7/18/2016 7/28/2016 7/28/2016

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