ML17187A132

North Anna, Units 1 and 2 - Submittal of Topical Report VEP-NE-1, Revision 0.2-A, Relaxed Power Distribution Control Methodology and Associated Fq Surveillance Technical Specifications for Information Only

ML17187A132
Person / Time
Site:	North Anna
Issue date:	06/28/2017
From:	Sly C D Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
17-214 VEP-NE-1-A, Rev. 0, Minor Rev. 2
Download: ML17187A132 (44)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 June 28, 2017 U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 Serial No.17-214 NLOS/DEA RO Docket Nos.: 50-338/339 License Nos.: NPF-4n SUBMITTAL OF TOPICAL REPORT VEP-NE-1, REVISION 0.2-A, RELAXED POWER DISTRIBUTION CONTROL METHODOLOGY AND ASSOCIATED FQ SURVEILLANCE TECHNICAL SPECIFICATIONS FOR INFORMATION ONLY Enclosed is one copy of topical report VEP-NE-1, Revision 0.2-A, "Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications".

This is a minor revision to the topical report which modifies the FQ surveillance information in accordance with amendments 278 and 261 issued for North Anna Units 1 and 2, respectively, to address concerns identified in Westinghouse Nuclear Safety Advisory Letter (NSAL) 09-5, "Relaxed Axial Offset Control FQ Technical Specification Actions," Revision 1 and Westinghouse NSAL-15-1, "Heat Flux Hot Channel Factor Surveillance Requirements," Revision 0. The change bars in NE-1, Revision 0.2-A reflect differences from the previous approved version of topical report VEP-NE-1, Revision 0.1-A. This document is being submitted to the Nuclear Regulatory Commission for information only. Should you have any questions in regard to this submittal, please contact Ms. Diane E. Aitken at (804) 273-2694.

Craig D. Sly Acting Director Nuclear Regulatory Affairs Virginia Electric and Power Company

Attachment:

1. VEP-NE-1, Revision 0, Minor Revision 2, Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications, Revision 0.2-A Commitments made in this letter: None.

Serial No.17-214 Submittal of VEP-NE-1, Revision 0.2-A For Info. Only Page 2 of 2 cc: U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Ms. K. R. Cotton-Gross NRC Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 08 G-9A Rockville, MD 20852-2738 Mr. James R. Hall NRC Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 08 G-9A Rockville, MD 20852-2738 NRG Senior Resident Inspector North Anna Power Station Mr. J. E. Reasor, Jr. Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd. Suite 300 Glen Allen, VA 23060 Serial No.17-214 ATTACHMENT 1 TOPICAL REPORT VEP-NE-1 RELAXED POWER DISTRIBUTION CONTROL METHODOLOGY AND ASSOCIATED FQ SURVEILLANCE TECHNICAL SPECIFICATIONS REVISION 0.2-A Virginia Electric and Power Company, Inc. North Anna Power Station, Units 1, 2 and ISFSI VEP-NE-1-A, Revision 0, Minor Revision 2 RELAXED POWER DISTRIBUTION CONTROL METHODOLOGY AND ASSOCIATED FQ SURVEILLANCE TECHNICAL SPECIFICATIONS NUCLEAR ENGINEERING

& FUEL STAFF DOMINION RESOURCES SERVICES RICHMOND, VIRGINIA APRIL 2017 Review::_/j_L_._*

_:;_.

__ _ D. T. Smith [Nuclear Core Design I] Recommended .I\ . 1 (} For Approval By:

A. H. Nicholson

[Supervisor Nuclear Core Design I]

[iijec;Jng]

M.A am [Director Nuclear Engineering and Fuel] T. S. Psuik [Nuclear Core Design I] 'S.A. Luchau [Supervisor Nuclear Safety Analysis Design]

VEP-NE-1-A, Rev. 0, MRev. 2 Page 2 of 41 *UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 February 20, 1986 Serial # /..2 7 Rec'd. FE 8 2 6 1986 Mr. w. L. Stewart, Vice President Nuclear Operations Virginia Electric and Power Company Richmond, Virginia 23261

Dear Mr. Stewart:

Nuclear Operations Licensing Supervisor

SUBJECT:

ACCEPTANCE FOR REFERENCING OF LICENSING TOPICAL REPORT VEP-NE-1, "VEPCO RELAXED POWER DISTRIBUTION CONTROL METHODOLOGY AND ASSOCIATED FQ SURVEILLANCE TECHNICAL SPECIFICATIONS" We have completed our review of the subject topical report submitted by the Virginia Electric and Power Company (VEPCO) by letter dated December 10, 1984. We find the report to be acceptable for referencing in license applications to the extent specified and under the limitations delineated in the report and the associated NRC evaluation, which is enclosed.

The evaluation defines the basis for acceptance of the report. *

• We do **not intend to repeat our review of the matters described .in the* report and found acceptable when the report appears as a reference in license applications, except to* assure that the material is applicable to the specific plant involved.

Our acceptance applies only to the matters described in the report. In accordance with procedures established in NUREG-039.0, it is requested that VEPCO publish accepted versions of this report, proprietary and non-proprietary, within three months of receipt of this letter. The accepted versions shall incorporate this letter and the enclosed evaluation between the title page and the abstract.

The accepted versions shall include an -A (designating accepted) following the report identification symbol. Should our criteria or regulations change such that our conclusions as to the acceptability of the report are invalidated, VEPCO and/or the applicants referencing the topical report will be expected to revise and resubmit their respective documentation, or submit justification for the continued.

effective applicability of the topical report without revision of their respective documentation.

Enclosure:

As stated Sincerely, ( Hli:;.'tlfi.

Berkow, Director and Special Projects Directorate Division of PWR Licensing-B VEP*NE-1-A, Rev. 0, MRev. 2 Page 3of41 SAFETY EVALUATION REPORT Report Title: Vepco Relaxed Power Distribution Control Methodology and Associated F Q Surveillance Technical Report Number: VEP-NE-1 Report Date: October, 1984 INTRODUCTION The Virginia Electric and Power Company (Vepco) has developed the relaxed power I distribution control (RPDC) methodology to replace the constant axial offset control (CAOC) strategy currently employed at its Surry and North Anna reactors.

Associated with the RPDC methodology is direct monitoring of the maximum peaking factor (FQ) relative to plant limits; this replaces the present Fxy Technical Specifications." The analyses performed f n support of relaxed power distribution control, and sample generic FQ surveillance Technical Specifications described in the subject report. Additional information considered in this review is given in Ref. 1.

SUMMARY

OF TOPICAL REPORT -The constant axial offset control (CAOC) strategy currently employed by Vepco was developed by Westinghouse in order to meet power peaking limits imposed by loss of coolant accident (LOCA) analyses.

The CAOC requires the maintenance of the axial flux difference (AI) within a specified, constant band abqut a target axial offset defined at equilibrium conditions.

While ance of Al within these limits insures that the FQ is bounded by a specified limit, CACO is restrictive, particularly below full power where margin to peaking limits exits. These restrictive AI limits have a negative impact on operational flexibility, especially in the ability to return to full power quickly following a reactor trip near end-of-cycle (EOC). The development of the relaxed power distribution control approach by Vepco was motivated primarily by this limitation.

ENCLOSURE VEP-NE-1-A, Rev. 0, MRev. 2 Page 4 of 41 Under RPDC the AI vs. power operating domain is typically broader than that permitted under CAOC (even with band widening), with the width of the band increasing with decreasing power levels. (Similar variable width operating bands are employed by all three PWR vendors in their axial power tion control procedures).

The variable l>.I vs. power operating band takes advantage of the increased FQ limits permitted at reduced power by taining a roughly constant margin to design limits at all power levels (vs. an increasing margin with decreasing power in CAOC) ': The major elements of the RPDC methodology are: 1. *Axial power distributions are generated with the Vepco one-dimensional NOMAD (Ref. 2) code which bound the potential t>.I operating band. The* NOMAD analysis produces a spectrum of xenon distributions at selected burnups via a technique similar to that developed by Combustion Engineering (CE) (Ref. 3). The resulting xenon distributions are combined with rod insertions and power levels permitted by the power dependent rod insertion limit curve at the selected burnups to produce a range of power distributions (and associated t>.I's) at power levels between 50% and full power. 2. The axial power distributions from (1) are used in a 10/20/30 synthesis of FQ (z) based on values of Fxy(z) generated by the Vepco FLAME (Ref. 4) and PDQ07 (Ref. 5) models. The synthesis includes an axial height dependent radial xenon redistribution factor calculated by FLAME, and uncertainty factors which account for the calculational uncertainty, and manufacturing variabilities.

3. Comparison of the ,resultant Fq's to limits prescribed by LOCA analyses defines a preliminary l>.I vs. power operating domain. 4. The entire set of axial power distributions is als_o analyzed with the COBRA (Ref. 6) code relative to the 1.55 design axial power distribution for the loss of flow accident (LOFA). This analysis defines a second t>.I-power operating space that insures that the margin to the ONB design basis for LOFA is maintained.

2 VEP-NE-1-A, Rev. 0, MRev. 2 Page 5 of 41 5. The most restrictive domain (based on LOCA and/or LOFA) defines the permissible space for normal operation (Condition I). (For Vepco plants, the LOCA based band is usually more restrictive).

Maintenance of Al wf thin th.is operating space, coupled with adherence to cont.rol rod insertion limits, ensures that the margin to fuel centerline melt, DNB, and LOCA peak clad temperature design cdteria are maintained during normal operation.

6. Three abnormal operation (Condition II) events are also considered in the analyses supporting RPDC: uncontrolled rod withdrawal,*excessive heat removal, and erroneous boration/dilution.

The purpose of these analyses is to confirm that the over-power delta-T {OPDT) and over-temperature delta-T {OTDT) trip setpoints have been conservatively calculated, and insures that required margins are maintained.

The OPDT and OTDT trips provide transient and steady-state protection against fuel center-line melt and DNB, respectively.

The initial conditions for the analyses of these events. consist of the axial power distributions allowed by the A-I power operating domain determined in (5). 7. The maximum linear power density for each resulting Condition U distribution is determined by using the FQ {z) synthesis techniques (with an allowance for densification) and compared to the design basis for fuel centerline melt. The OPDT function is modified, if necessary, to insure that margin to the fuel center-line melt limit is maintained.

The axial power distributions from the Condition II analyses are also evaluated to confirm that the OTDT trip function and its associated f{Al) term remain valid. In conjunction with the implementation of the RPDC me.thodology, Vepco proposed to replace the current Fxy surveillance with direct monitoring of FQ (2). In FQ surveillance the measured Fq at equilibrium conditions is augmented by a factor, N(2), which accounts for the maximum potential increase in FQ (z) during normal opertion.

The resultant augmented FQ (2) is compared to the plant LOCA FQ (2) limits to determine acceptability, or to initiate remedial actions. Sample Technical Specifications to be used with FQ surveillance are given. 3 VEP-NE-1-A, Rev. 0, MRev. 2 Page 6 of 41 While the greatest benefit of relaxed power distribution control to Vepco is the ability to return to .power quickly following a trip near EOC, institution of this methodology with its wider operating band is expected to yield additional operational benefits including reduced control rod motion and coolant system.boration/dilution requirements.

SUMMARY

OF TECHNICAL EVALUATION All the analyses performed in support of RPDC employed codes which have been previously reviewed and approved by the staff (FLAME, PDQ07, NOMAD, COBRA). The approach used for generating bounding axial power distributions is based on the free xenon oscillation technique employed for a number of years by Combustion Engineering in their axial power distribution control methodology. (CE served as a consultant to Vepco in the implementation and application of this technique).

Vepco has determined that this approach results in axial power distributions that sufficiently span the AI-power domain to ensure .there is confidence that the most adverse conditions are available for subsequent analyses.

In addition, relevant analyses performed by CE show that the

• sensitivity of the results obtained employing the free xenon oscillation methodology to variations in the impacting parameters are small, and are more . . than compensated for by the 11 bounding 11 nature of the and the. extreme distributions considered.

This approach has been found acceptable for CE -reactors for many years, is acceptable for RPDC. The calculation of FQ via a 10/20/30 synthesis is similar to accepted approaches.

Uncertainties associated with the calculation of FQ are based on comparisons to measurements.

The measurements included situations where azimuthal tilts spanning the range permitted by the technical specification limits were present. The combination of the FNU and FGR components of the uncertainty given in the report is greater than the 95/95 upper tolerance limit determined on the basis of comparisons to measurements.

The magnitude of the uncertainty assigned to the calculated value of FQ in the RPDC analyses is therefore acceptable.

4 VEP-NE-1-A, Rev. 0, MRev. 2 Page 7of41 Calculations of the radial xenon redistribution factor, Xe(z); component of the FQ synthesis employed the FLAME code and considered a number of cycles, times in life and initiating conditions.

The final Xe(z) was chosen such that it bounded all observed increases fn Fxy(z). Even though this factor_is now less than the previously used axially uniform value of 1.03, the analyses performed to justify the lower values are adequate.

The LOFA analyses performed with COBRA, and the Condition II events considered are similar to those included in the Westinghouse relaxed axial offset control (RAOC) methodology.

The over-power and over-temperature AT trip functions will be evaluated on a reload basis_ to assure protection against fuel center-line melt and DNB design basis limits. Other accident analyses will be reevaluated on a reload basis to . insure that the assumption used in the RPDC analyses remain bounding.

Monitoring of adherence to oper.ation within the permissible AI-power domain is accomplished by reliance on the ex-core detectors.

The calculated AI domain will be reduced by 3% to accommodate the maximum excore detector calibration . . uncertainty permitted by the Technical Specifications.

In addition, Vepco plans to further reduce the AI limits for the first-time analysis.

The bounding nature of the RPDC approach provides further conservatism.

The Vepco RPDC methodology contains elements similar to those included in the (Ref. 7) and CE variable-width AI band axial power distribution control strategies.

Approved methods have been used in the analyses supporting RPDC and justification has been provided for the uncertainties assigned.

These analyses and uncertainties are consistent with currently approved methods and practices.

In addition, the impact of cycle specific variations on the AI .:. power domain, the over-power and over-temperature AT trip setpoints, and other safety analyses will be evaluated on a reload basis. Based on these considerations the RPDC approach represents an acceptable methodology for use with reload cores similar to those of the Surry and North Anna reactors.

5 VEP-NE-1-A, Rev. 0, MRev. 2 Page 8 of 41 The proposed FQ surveillance is similar to the approach approved fn . conjun.ction with RAOC. The N(z) factor by which the measured FQ (z) distribution is augmented to account for non-equilibrium normal operation is similar to the W(z) and V(z) functions used by and Exxon, and approved for use with RAOC and PD II power distribution control strategies.

The sample Technical Specifications given in the subject report replace Fxy surveillance with FQ This is acceptable.because the FQ surveillance is more appropriate for RPDC. The sample Technical Specifications in the report acceptably implement RPOC with the following modifications:

0 0 0 Specification 3/4.2, page 3/4 2-1 The astertsk at the end of the APPLICABILITY line and the footnote should -. be deleted. Figure 3/4 2-4 This figure should be blank and contain the legend:* "This curve is given in the Core Surveillance Report as per Specification

6. 9. l.*10.11 Specification 3.2.2, page 3/4 2-5 Parenthetical comments should be added to the final three lines of action a as follows: 11 subsequent POWER OPERATION may preceed provided the Overpower llT Trip Setpoints (value of K 4) have been reduced at least 1% (in AT span) for each l% FQ(z) exceeds the limit. Specification 6.9.1.10 (page unnumbered)

After "initial criticality", add "unless otherwise approved by the Commission by letter", and change the end of the first paragraph to "approved by the Co1M1ission by letter". 6 VEP-NE-1-A, Rev. 0, MRev. 2 Page 9 of 41 . A complete set of. these revisions will be approved for North Anna Unit 2, Cycle 4 and could be used as a model. CONCLUSION We find the subject report suitable for reference as support for use of RPDC in licensing applications.

\. 7 VEP-NE-1-A, Rev. 0, MRev. 2 Page 10 of 41 REFERENCES

1. Letter from W.L: Stewart (Vepco) to H.R. Denton (USNRC), "Virginia Electric and Power Company Relaxed Power Distribution Control-(RPDC)

Supplemental Information," (Oct. 21, 1985). 2. S.M. Bowman, "The Vepco NOMAD Code and Model, 11 VEP-HFE-lA, Virginia Electric and Power Company (Hay 1985). 3. 11 C-E Setpoint Methodology," CENDP-199-NP Rev. 1-NP, Combustion Engi neering Inc. (March 1985). 4. W. C. "The Vepco FLAME. Model, 11 VEP-FRD-24A, Virginia Electric and Power Co. (July 1981). 5. M."L -Smith, 11:rhe PDQ07 Discrete Model , 11 VEP-FRD-19A, Virginia Electric *and Power Co., (July 1981). 6. F.W. Silz, "Vepco Reactor Core Thermal-Hydraulic Analysis Using the COBRA IIIC/MIT Computer Code, 11 Vepco-FRD-33A, Virginia Electric and Power Co. (Oct. 1983). 7. R.W. Miller et al., "Relaxation of Co.nstant Axial Offset .Coritrol

, 11 NS-EPR-2649 Part A, Westinghouse Electric Corp. (August 1982). 8 VEP-NE-1-A, Rev. 0, MRev. 2 Page 11of41 CLASSIFICATION/DISCLAIMER The data, information, analytical techniques, and conclusions in this report have been prepared solely for use by Dominion (the Company);

and they may not be appropriate for use in situations other than those for which they are specifically prepared.

The Company therefore makes no claim or warranty whatsoever, expressed or implied, as to their accuracy, usefulness, or applicability.

In particular, THE COMPANY MAKES NO WARRANTY OF MERCHANTABILITY OR FITNESS FOR A PARTICULAR PURPOSE, NOR SHALL ANY WARRANTY BE DEEMED TO ARISE FROM COURSE OF DEALING OR USAGE OR TRADE, with respect to this report or any of the data, information, analytical techniques, or conclusions in it. By making this report available, the Company does not authorize its use by others, and any such use is expressly forbidden except with the prior written approval of the Company. Any such written approval shall itself be deemed to incorporate the disclaimers of liability and disclaimers of warranties provided herein. In no event shall the Company be liable, under any legal theory whatsoever (whether contract, tort, warranty, or strict or absolute liability), for any property damage, mental or physical injury or death, loss of use of property, or other damage resulting from or arising out of the use, authorized or unauthorized, of this report or the data, information, and analytical techniques, or conclusions in it.

-

I\ VEP-NE-1-A, Rev. 0, MRev. 2 Page 12 of 41 PREFACE VEP-NE-1-A, Revision 0, Minor Revision 2: Modifies the FQ surveillance information in accordance with the approved Technical Specifications updates provided in NRC Correspondence Serial No.16-416 [18] to address issues noted in Westinghouse Nuclear Safety Advisory Letter (NSAL) 09-5, Rev. 1 and Westinghouse NSAL-15-1, Rev. 0. Updates Figure 5.0-1 to align with the expanded axial region included in Dominion's FQ Surveillance program. This is a minor revision; therefore, change bars have been included to reflect differences from the previous version (VEP-NE-1-A, Revision 0, Minor Revision 1 ). VEP-NE-1-A, Revision 0, Minor Revision 1: Presented a modified version of the Relaxed Power Distribution Control Methodology provided in VEP-NE-1-A, Rev. 0 published in March 1986. Updated the references to the current 3-D PDQ Two Zone and enhanced NOMAD models as well as outlined the use of the NRC approved Studsvik Core Management System in the RPDC methodology.

Referred to the COLR section of the plant Technical Specifications for the applicable thermal-hydraulic codes(s) and correlation(s) for DNB analyses.

VEP-NE-1-A, Rev. 0, MRev. 2 Page 13 of 41 TABLE OF CONTENTS Page Safety Evaluation Report .........................................................................................................................

2 Classification/Disclaimer

...............................................................................

.........................................

11 Preface ...................................................................................................................................................

12 Table of Contents ...................................................................................................................................

13 List of Figures .........................................................................................................................................

14 List.of Tables ......................................................

...................................................................................

14 Section 1 -Introduction

..................

........................................................................................................

15 Section 2 -Condition I Analysis ..............................................................................................................

19 2.0 Analysis of Axial Shapes Which Result from Normal Operation

............................................

19 2.1 Axial Shape Generation

.........................................................................................................

19 2.2 LOCA Delta-I Limit Formation

................................................................................................

26 2.3 Loss of Flow Thermal/Hydraulic Evaluation

.......................................

...................................

29 2.4 Final Normal Operation Delta-I Limit ......................................................................................

29 Section 3 -Condition II Analysis .............................................................................................................

30 3.0 Analysis of Axial Shapes Which Result from Condition II Events ...........................................

30 3. 1 Determination of Accident Pre-Conditions

.............................................................................

30 3.2 Condition 11 Accident Simulation

.............................................................................................

30 3.3 Overpower Limit Evaluation

...................................................................................................

32 3.4 DNB Evaluation

......................................................................................................................

34 Section 4 -Other Safety Analyses ..........................................................................................................

35 Section 5 -FQ Surveillance

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36 Section 6 -Conclusion

............................................................................................................................

39 Section 7 -References

...........................................................................................................................

40 VEP-NE-1-A, Rev. 0, MRev. 2 Page 14 of 41 LIST OF FIGURES Page Section 1 1.0-1 Typical CAOC Limits ...............................................................................................................

17 1.0-2 Typical Variable Axial Flux Difference Limits **l*********************************************************************

18 Section 2 I I I 2.1-1 Typical RPDC BOC Xenon Oscillation

.............

l ......................................................................

22 . I 2.1-2 Typical RPDC EOC Xenon Oscillation

.............

l ......................................................................

23 2.1-3 Typical Rod Insertion Limits ............................. ......................................................................

24 I I 2.2-1 Typical LOCA Delta-I Limits ............................. ......................................................................

28 i I Section 3 3.3-1 Typical Maximum Power Density Flyspeck ............................................................................

33 Section 5 5.0-1 Typical N(z) Function .............................................................................................................

38 LIST OF TABLES *Page Section 2 2.1-1 Typical Conditions Analyzed for Normal Operation Under RPDC ..........................................

25 I.

VEP-NE-1-A, Rev. 0, MRev. 2 Page 15 of 41 SECTION 1 -INTRODUCTION In response to Loss-of-Coolant Accident (LOCA) Emergency Core Cooling System (ECCS) criteria that imposed new requirements on local power peaking, Westinghouse developed the Constant Axial Offset Control (CAOC) power distribution control procedure

[1]. The CAOC strategy restricts axial power skewing in the reactor core during normal operation to within a band of +/-5% delta-I around a target value, determined at all-rods-out equilibrium conditions.

is defined as Delta -/(%) = 100 x (Pt -Pb) where Pt and Pb are the fractions of rated full-core po,wer in the top and bottom halves of the core, respectively.

This +/-5% limit on axial power skewing the magnitude of axial xenon oscillations which, in turn, decreases the magnitude of any power! peaking during abnormal operation.

A typical CAOC delta-I band is shown in Figure 1.0-1. The OAOC target value varies with burnup as the all-rods-out equilibrium delta-I changes. Much of the low power operational flexibility of CAOC was originally centered around the use of the part length rods as a means for axial power distribution control [1]. Full length rods and boron were to be used mainly for reactivity control associated with changes in power. Since the requirement for removal of part length rods was imposed, full length rods have rad to be used to help control the axial power distributions.

As a result, it became more difficult to the axial power distribution within the +/-5% delta-I band at low powers. This is especially true' near end-of-cycle when the soluble boron concentration has been reduced to a very low level to compensate for the effects of fuel depletion and fission product buildup. Should a trip occur during this portion of the cycle, a plant may not be able to return to full power easily because of difficulty in meeting the delta-I limits. There is insufficient reactivity available from boron dilution to allow the full length rod movement required to offset the buildup of I xenon and, at the same time, maintain delta-I within its band. As a result, delta-I limits could be exceeded at low power levels, requiring the plant to remain below 50% power in order to meet the "one hour in twenty-four" 1 requirement in the plant Technical

$pecifications.

Some Westinghouse CAOC plants with available full power margin to their LOCA Overall Peaking Factor (FQ) license limits have transformed this margin into operating flexibility through delta-I "band I widening." In the past [2], Surry had a delta-I band width: of +6, -9% about the target value. This method of gaining operational flexibility does provide some additional full power delta-I operating space, but ' offers only minimal relief for post-trip return to power at end-of-cycle conditions.

1 The CAOC Technical Specifications impose no operational limit on delta-I while a plant operates below 50% power. However, in order to ascend above 50% power, the plant must not have exceeded the delta-I bands for more than one penalty hour of the previous twenty-four.

'

VEP-NE-1-A, Rev. 0, MRev. 2 Page 16of41 This operational restriction on delta-I imposed by CAOC can be eased by the implementation of a variable delta-I band control strategy that takes credit for the full power delta-I margin available from standard band widening while also providing for an increasing delta-I band with decreasing power. The widened delta-I band is formed by maintaining an approximately constant analysis margin to the design bases limits at all power levels. This is in contrast to CAOC operation which has large amounts of margin available at reduced power. For North Anna and Surry, which have LOCA-limited total peaking factors, this variable delta-I would be selected sudh that the margin to the LOCA FQ*P*K(z) limit I would remain approximately constant for all power levels. An example of a variable delta-I band is given in Figure 1.0-2. The principal benefits of a variable band delta-I control strategy over CAOC operation are as follows: 1) The ability to return to power after a trip, particularly at end-of-cycle, is enhanced;

2) Control rod motion necessary to compensate for the CAOC +/-5% delta-I band restrictions is reduced to only that motion needed to maintain operation within a much wider band; 3) The reactor coolant system boration/dilution requirements are decreased, due, in part, to the reduced control rod motion; 4) The plant has enhanced operational flexibility.

The concept of widened delta-I limits at reduced power levels is not a new one. Combustion Engineering

[3] and Babcock and Wilcox [4] have supported increased axial skewing at reduced power levels for their reload cores for several years. Westinghouse

[5] has also developed and licensed a variable delta-I control strategy called RAOC (Relaxed Axial Offset Control) for application to reload cores. Dominion has combined some of the concepts from the Combustion Engineering methodology

[3] with the current Dominion analysis techniques

[1,6] to form an alternate methodology for variable band I delta-I control. This methodology is called Relaxed Power Distribution Control (RPDC). The Sections I . that follow will discuss the Dominion procedure for generating the variable width delta-I band. They will I also discuss the methods used to ensure that the margin to the design bases criteria, such as Departure from Nucleate Boiling (DNB), fuel centerline melt and Loss of Coolant Accident (LOCA) peak clad temperature is maintained.

This report also discusses the formulation of FQ Surveillance Technical Specifications.

The CAOC radial peaking factor Fxy(z) surveillance is replaced by FQ(z) monitoring, using the measured value of FQ(z) augmented by a non-equilibrium operation multiplier, in order to verify compliance with the LOCA I peaking factors. As will be seen in Section 5, FQ surveillance complements RPDC to form a consistent but more flexible plant monitoring scheme than that provided by the CAOC methods.

VEP-NE-1-A, Rev. 0, IVIRev. 2 FIGURE 1.0-1 -TYPICAL CAOC LIMITS p E R c E N T 0 F R A T E D p 0 w E R 80-60-40-:20-I I I I I I I I I I I I I I I I I I I

• I I I I I I

• I I I I I I I I I I I I I I I 1 I I I I I I I I I I I I I I I I I I I I I l I I I ' . I I . I I ' I I I I I I ' I ' ' I I I I I I I I t ' t I I I I ' I I I I I I t I t t I I t I I I I I I I I I I I I I I t I I I I ' I I I I I I I t I I l I I I I I I I I I t I I ' I l I ' I I ' ' I ' I t ' t I' I ; I I I I I I I I I " I I I

• 40 20 -10 a 10 20 30 40 50 PERCENT AX1AL FLUX DIFFERENCE Page 17 of 41 VEP-NE-1-A, Rev. 0, MRev. 2 Page 18 of 41 FIGURE 1.0-2-TYPICAL VARIABLE AXIAL FLUX DIFFERENCE LIMITS 100 90 80 p E R 70 RPDC c E Limits N T 60 0 F R A 50 T E 0 p 40 0 H E CAOC R Limits 30 40 20 -10. 0 I 10 20 30 40 50 PERCENT AXIAL FLUX 1 0JFFERENCE VEP-NE-1-A, Rev. 0, MRev. 2 Page 19 of 41 SECTION 2 -CONDITION I ANALYSIS 2.0 Analysis of Axial Shapes Which Result from Normal pperation The objective of a RPDC analysis is to determine acceptable delta-I band limits that will guarantee that margin to all the applicable design bases criteria has been maintained and, at the same time, will provide enhanced delta-I operating margin over CAOC. Because the RPDC delta-I band is an analysis output quantity rather than a fixed input limit, as in CAOC, axial shapes which adequately bound the potential delta-I range must be generated.

These axial shapes must include the effect of all potential combinations of the key parameters such as burnup, control rod position, xenon distribution, and power level. Dominion has developed the methodology of Section 2.1 to generate the large number of axial shapes included in RPDC. After the axial power shape's have been created, two separate allowable delta-I limits for normal operation are established:

one based on LOCA FQ considerations and the other one based on a Loss of Flow (the limiting DNB transient) thermal/hydraulic evaluation.

The methods used are described in Sections 2.2 and 2.3, respectively.

These two separate delta-I bands are combined to form a composite delta-I limit as discussed in Section 2.4. 2.1 Axial Shape Generation The axial power distributions encountered during normal operation (including load-follow) are primarily a function of four parameters:

the xenon distribution, power level, control rod bank position and burnup distribution.

For RPDC, reasonable incremental variations that span the entire expected range of values must be considered for each of these The following method is used to create the axial power distributions needed for the development of the RPDC normal operation delta-I limits. 2.1.1 Axial Xenon Distributions During Normal Operation The axial xenon distribution is a function of the core's operating history and, as a result, is constantly changing.

In order to analyze a sufficient number of xenon distributions to ensure that all possible cases have been accounted for, a xenon "free oscillation" method similar to the one described in Reference 3 is used to form these distributions.

By creating a divergent xenon-power oscillation, axial xenon distributions can be obtained that will be more severe than any experienced during normal operation, including load follow maneuvers.

/

VEP-NE-1-A, Rev. 0, MRev. 2 Page 20 of 41 To initiate a xenon-power oscillation, an equilibrium 1-D model [7, 17] or 3-D model [16] of the reload cycle is perturbed.

This perturbation will generally be in the form of a change in power, rod position, or both. However, since the core model may be inherently stable due to the presence of feedback mechanisms, these mechanisms must either be modified or bypassed to obtain a divergent oscillation.

One way to accomplish this is to reduce the stability of the model by reducing the amount of Doppler (i.e., fuel temperature) feedback in the system. The divergent oscillation provides a spectrum of xenon distributions that will produce power distributions with delta-I values covering the expected delta-I range. The magnitude of the "free oscillations" should be such that the xenon distributions (when combined with normal operating conditions) produce axial power shapes with delta-I values that bound the expected operating limits. The stability of the calculational model may vary with burnup or core loading. Therefore, the amount of perturbation and feedback modification necessary to achieve a divergent xenon oscillation may vary with cycle burnup or core loading. Typical examples are given in Figures 2.1-1and2.1-2 for and end-of-cycle, respectively.

The Dominion NOMAD [7, 17] 1-D diffusion code was used to perform these examples.

These particular oscillations were initiated by reducing power, depleting for several hours and then returning to full power for an additional 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of depletion.

2.1.2 Power

Level During Normal Operation For the normal operation analysis, power levels spanning the 50% to 100% range are investigated to establish the RPDC delta-I limits. This range is consistent with the current CAOC Technical Specifications which do not impose axial flux difference limits or require CAOC operation below 50% of full power.2 The power levels used for RPDC analysis are selected at increments within the 50% to 100% range which are small enough to ensure an adequate number of power distributions are being analyzed; i.e. that all safety-related effects due to the power level are accounted for. 2.1.3 Control Bank Position During Normal Operation During normal operation, the control rod bank insertion is limited by the cycle-specific Core Operating Limits Report (COLR) rod insertion limits. Figure 2.1-3 gives a set of typical rod insertion limits. The insertion limits are a function of reactor power, and the rods may be anywhere*

between the fully withdrawn position and the variable insertion limit. In order to adequately analyze the various rod 2 The CAOC Technical Specifications impose no operational limit on delta-I while a plant operates below 50% power. However, in order to ascend above 50% power, the plant must not have exceeded the delta-I bands for more than one penalty hour of the previous twenty four.

VEP-NE-1-A, Rev. 0, MRev. 2 Page 21 of 41 positions allowed, control rod insertions versus power level are selected which cover the range of rod insertions allowed for each particular power. 2.1.4 Cycle Burnup The RPDC analysis is performed at several times in cycle life in order to provide limiting delta-I bands for the entire cycle. Typically, three cycle burnups, near beginning-of-cycle (BOC), middle-of-cycle (MOC) and end-of-cycle (EOG), are chosen for the RPDC analysis.

The MOC case is chosen to reflect the maximum middle-of-cycle radial peaking factors. 2.1.5 Combining Xenon Shapes, Rod Position, Power Level and Burnup The final power distributions used in the RPDC normal operation analysis result from combining axial xenon shapes, power levels, rod insertions and cycle burnups. *At each selected time in cycle life, the xenon shapes are combined with each power level and rod configuration.

A criticality search is then performed for each case using the NOMAD [7,17] or the SIMULATE [16] code with normal feedback.

Calculated axial power distributions are identified for use in the LOCA FQ and thermal/hydraulic evaluations discussed in Sections 2.2 and 2.3. The combinations of burnups, power levels, rod configurations and xenon distributions typically evaluated on a reload basis are summarized in Table 2.1-1. The conditions result in a delta-I range of approximately

-60% to +50%, bounding the expected final delta-I envelope at all power levels. The combinations of rod insertions and power levels necessary for Surry and North Anna would be slightly different due to the difference in rod insertion limits between the two plants.

VEP-NE-1-A, Rev. 0, MRev. 2 Page 22 of 41 p E R c t N T tl c l J A f" l ij '.'( 75-* ! so ' . ' : ' FIGURE 2.1-1 -TYPICAL RPDC BOC XENON OSCILLATION , ..

• l

• 0 Z-0 40 60 so lOO l? ttf t H1RIRS J VEP-NE-1-A, Rev. 0, MRev. 2 Page 23 of 41 FIGURE 2.1-2-TYPICAL RPDC EOC XENON OSCILLATION 75 so-p E R 25 c E N T 0 0 E L T A r -25 L lJ )( 0 20 40 60 BO 100 T ll'IE UHllJR:S I \

VEP-NE-1-A, Rev. 0, MRev. 2 Page 24 of 41 FIGURE 2.1-3 -TYPICAL ROD INSERTION LIMITS O:::FVLLY llNSERTED, Z26=FVLLY Tl'.l'm>RAWN 240 uo 20*0 HIO BANK c R. o. o: 160 0. R tl 140 u f> 5 13f\NK D T E p iOO f' o. 0 1 ao-T l 0 N' 60-4.(j..; 20 0 0 10 io 30 40 50 '60 10 9!0 100 FflACT JON Of' ltRTEO POWER VEP-NE-1-A, Rev. 0, MRev. 2 TABLE 2.1-1 TYPICAL CONDITIONS ANALYZED FOR NORMAL OPERATION UNDER RPDC Cycle Burnups Xenon Shapes Power Level Range(%) Rod Insertions Range Versus Power: BOC, MOC, EOC 100 for each time in life 50-100 See Figure 2.1-3 Page 25 of 41 (3 burnups) * (100 xenon shapes) * (30 power level/rod position combinations)

= 9000 shapes VEP-NE-1-A, Rev. 0, MRev. 2 Page 26 of 41 2.2 LOCA Delta-I Limit Formation The FQ*Power for each shape is compared to the LOCA FQ*Power*K(z) limit at each power level to determine which axial shapes approach the LOCA limit, thereby establishing a preliminary allowable delta-I versus power band. This comparison replaces the traditional CAOC FAC analysis [1] and ensures that the margin to the LOCA FQ*Power*K(z) envelope is maintained during the cycle as long as reactor operation remains within the delta-I limits. A typical LOCA delta-I limit is shown in Figure 2.2-1. Modification of the LOCA delta-I limits can be used as a means of reducing or increasing the allowable FQ at a constant power. Cycle specific analysis is required to determine the relationship between FQ and the LOCA delta-I limits and core power to determine the allowable operating space. 2.2.1 FQ Using Standard 1-D/3-D Synthesis The axial shapes created in Section 2.1 using NOMAD are combined with Fxy(Z) data using a standard 1-D/3-D FQ synthesis

[1, 7, 8, 17]: F 0 (z) = f°xy(z) x P(z) x Xe(z) x FNU x FQE x FGR where the following are non-dimensional parameters:

P(z) = Xe(z) = FNU = FQE = FGR = Fxy distribution calculated by 3-D PDQ Two Zone [8, 17], dependent upon burnup, core height and rod position and power level. Axial power shape function generated by NOMAD [7, 17] The radial xenon redistribution factor Nuclear uncertainty factor [7, 17] Engineering heat-flux hot-channel factor [9, 1 O] Grid correction factor [7, 17]

VEP-NE-1-A, Rev. 0, MRev. 2 Page 27 of 41 The axially varying radial xenon factor, Xe(z), compensates for increases to FQ(z) resulting from redistribution of the xenon in the radial plane due to rod movement.

The radial xenon redistribution effect cannot be explicitly represented in a 1-D code and is therefore applied in the synthesis as an uncertainty factor. Xe(z) is calculated as follows: Max Fxy(z)T Xe(z) = Fxy(z)E where Fxy(z) T is the Fxy(z) calculated from a transient resulting in xenon radial redistribution and Fxy(z)E is the Fxy(z) based upon an equilibrium xenon distribution.

Fxy(z)T is calculated with a 3-D code by first pre-conditioning the radial xenon distribution for several hours with the core at reduced power and the control rods inserted sufficiently to drive delta-I to the negative edge of the expected band. By withdrawing the rods and increasing power a xenon transient is created. This transient will cause the xenon to redistribute radially as well as axially in the 3-D model. Fxy(z)T is calculated for each time step as this transient is followed in small time intervals.

The maximum values of Fxy(z)T for the entire transient are used to determine Xe(z). 2.2.2 FQ Using 3-D Model The axial shapes created in Section 2.1 using SIMULATE are used directly [16]: FQ(z) = Fq(z) x FNU x FQE where the following are non-dimensional parameters:

Fq(z) = Fq distribution calculated by 3-D SIMULATE [16], dependent upon burnup, core height and rod position and power level. (Includes xenon redistribution and grid effects).

FNU = Nuclear uncertainty factor [16] FQE = Engineering heat-flux hot-channel factor [9, 10]

VEP-NE-1-A, Rev. 0, MRev. 2 Page 28 of 41 p E R c E N T a F R A T E D !' 0 H E R 100-SO-: 60-10.: 60-5'0-40-FIGURE 2.2-1 -TYPICAL LOCA DEL TA-I LIMITS : ' ' * . I ' I i ' ' * ' ' I ' ' ' f ' : ., ' . ' , J ' ' r ' ' ' ' I ' i * * * ,* ... ... _...,.., ... ..,, ... ..,,,. .... u .... '"' -_ .. _ -,._ ' l '* j ' \ * ' . ., ' I ' ' " " * ' \ . l . ' . I I * ' I . . . I I * * . ' ' \ ' * ' I 1

.* * * ) .

• l n4Q *30 10 0 \Q ZQ 30 40 so PERCENT AXJRL DlffERENCE VEP-NE-1-A, Rev. 0, MRev. 2 Page 29 of 41 2.3 Loss of Flow Thermal/Hydraulic Evaluation The Loss of Flow Accident (LOFA). represents the most limiting DNB transient not terminated by the Overtemperature Delta-T trip. In order to ensure the applicability of the current LOFA analysis, the entire set of axial power distributions formed by the RPDC normal operation analysis are evaluated against the 1.55 cosine design axial power distribution for the Loss of Flow Accident analysis with the applicable thermal-hydraulic code(s) and correlation(s) that are listed in the COLR section of the plant Technical Specification.

The thermal/hydraulic evaluation methods used in this LOFA evaluation are similar to those of the CAOC techniques.

As a result of this LOFA comparison, a second set of delta-I versus power limits is formed. These delta-I limits delineate the allowable operating band which will ensure that the margin to the DNB design base for LOFA is maintained.

The impact of RPDC on other DNB transient events is discussed in Section 3. 2.4 Final Normal Operation Delta-I Limit The results of the LOFA delta-I limit generation are combined with the LOCA delta-I limits (Figure 2.2-1) to produce a set of limits which will ensure that .the preconditions for both accidents are met. These generic limits will be verified on a cycle-by-cycle basis using the RPDC methods described in this report. The LOCA FQ based delta-I limits are generally more restrictive than LOFA-based delta-I limits for Dominion's plants. This will allow the plant cycle specific COLR to take advantage of the FQ versus delta-I relationship discussed in Section 2.2.

VEP-NE-1-A, Rev. 0, MRev. 2 Page 30 of 41 SECTION 3 -CONDITION II ANALYSIS 3.0 Analysis of Axial Shapes Which Result from Condition II Events One of the important features of any axial power distribution control strategy (RPDC, CAOC or any other) is the clear distinction between normal and accident conditions.

The delta-I limits established in Section 2 and the cycle specific COLR control rod insertion limits (see Figure 2.1-3) define conditions of normal operation.

If the axial power distribution (as measured by delta-I) remains inside the pre-established band during all normal operation, and the control rods remain within the cycle specific COLR limits, then the margin to the design criteria of fuel centerline melt, DNB and LOCA peak clad temperature, will be maintained.

This Section examines Condition II or Abnormal Operation events, which may be the result of system malfunctions or operator errors and create reactor conditions that fall outside the bounds analyzed in Section 2. The RPDC analysis examines the more limiting of these Condition II events and confirms that the Overpower Delta-T (OPDT) and the Overtemperature Delta-T (OTDT) setpoints 3 have been conservatively calculated and ensures that margin to the fuel design limits is maintained.

These setpoints are verified on a cycle-by-cycle basis. 3.1 Determination of Accident Pre-Conditions Initial condition parameters for Condition II analysis are determined from the core conditions allowed by the normal operation delta-I versus power envelope.

These conditions are a function of rod control cluster (RCC) position, boron concentration, xenon distribution, burnup and core power level. Any set of these conditions which produce an axial power distribution within the normal operation delta-I envelope established in Section 2 (Figure 2.2-1) can be a potential starting point for a Condition II accident.

Each set of valid normal operation conditions is considered in the RPDC Condition II analyses.

3.2 Condition

II Accident Simulation Three categories of credible accidents bound the range of abnormal operation events which must be considered in terms of their effect upon the axial power distribution or local power peaking. These three accidents are rod withdrawal, excessive heat removal and erroneous boration/dilution.

The rod withdrawal and boration/dilution events [1] are the most limiting Condition II events with respect to the 3 The OPDT and OTDT setpoints were designed primarily to provide transient and steady state protection against fuel centerline melt and DNB, respectively.

-

VEP-NE-1-A, Rev. 0, MRev. 2 Page 31 of 41 impact of control rod position on the axial power distribution or local power peaking. In the excessive heat removal event the impact of temperature is investigated.

3.2.1 Uncontrolled

Rod Withdrawal Event The rod withdrawal event [6] is an erroneous control rod withdrawal starting from a normal operation condition with the control banks operating in their normal overlap sequence.

To perform the analysis of this accident, the xenon distribution and boron conc.entration are fixed at values allowed by the normal operation analysis.

The lead control bank is then withdrawn in increments from the fully inserted to the fully withdrawn position.

After each incremental movement a criticality search is performed with either NOMAD [7, 17] or SIMULATE [16] and the axial power distribution is identified for use in the Condition II evaluation of Sections 3.3 and 3.4. The analysis is limited to those cases producing power levels between 50% of rated power and the high flux trip limit. 3.2.2 Excessive Heat Removal Event The Excessive Heat Removal (or cooldown) event, like the rod withdrawal event, is an overpower accident.

The accident.

assumes a decrease in the reactor core inlet temperature as a result of a sudden load increase, steam-dump valve opening, excessive feedwater flow or a turbine valve opening [6]. Since the control rods are assumed to be in manual control for this event, they will remain at their original position, which allows the reactor power to increase.

To simulate this accident, allowable normal operation xenon distributions, control rod positions and boron concentrations are provided as input to the NOMAD [7, 17] or SIMULATE [16] code. The inlet temperature is reduced and a criticality search is performed.

The axial power distribution from each case is identified for use in the Condition II evaluation of Sections 3.3 and 3.4. Reduction of the inlet temperature is limited to 30°F, which has been shown to bound the results of the above accidents in the Surry and North Anna UFSAR's [11, 12]. Cases producing a power level greater than the high flux trip limit are excluded from consideration.

3.2.3 Boration/Dilution The Boration/Dilution event causes a movement in the control rods to compensate for the reactivity changes due to a change in soluble boron concentration as a result of inadvertent boration or dilution.

In this analysis the control banks are assumed to be in automatic mode and to operate in a normal overlap sequence.

The manual mode of operation could result in an overpower transient during a VEP-NE-1-A, Rev. 0, MRev. 2 Page 32 of 41 dilution incident.

However, the consequences of this event are bounded by those of the rod withdrawal accident [6]. To perform the boration/dilution analysis, NOMAD reads each allowable xenon distribution from the ' Condition I analysis and runs a series of cases inserting the rods from fully withdrawn to the insertion limits in fixed increments.

For SIMULATE, a restart case is read for each allowed xenon distribution from the Condition I analysis.

For both NOMAD and SIMULATE, at each step a criticality search is performed.

Once the rods reach the insertion limits, a rod position search is performed to determine the amount of control rod insertion necessary to compensate for the reactivity associated with a dilution of fifteen minutes. The rods are then stepped in from the insertion limits to the determined rod position, again performing criticality searches.

All axial power distributions from the boration/dilution event are identified for the Condition II evaluation of Sections 3.3 and 3.4. 3.3 Overpower Limit Evaluation The maximum linear power density for each distribution produced by the Condition II accident simulations is determined using the 1-D/3-D FQ synthesis or the 3-D model techniques as described in Section 2.2 (with the addition of the densification spike factor S(z)). The results may be plotted in the "flyspeck" format shown in Figure 3.3-1, which shows typical results for the three limiting Condition II accidents described in Section 3.2. The peak power density "flyspeck" is compared to the design basis limit for fuel centerline melt. If necessary, the OPDT f(delta-1) function (which provides protection against this design limit) is modified to ensure that margin to the fuel centerline melt limit is maintained.

If needed at all, this modification would be required only for very large values of delta-I. An alternative approach would be to maintain the margin to fuel centerline melt by restricting the OTDT f(delta-1) function beyond the DNBR requirement, effectively eliminating the need for the OPDT f(delta-1) function.

VEP-NE-1-A, Rev. 0, MRev. 2 r E A l 4 K l. J N E R R 0 w E' R 5 K lO M I F T FIGURE 3.3-1 -TYPICAL MAXIMUM POWER DENSITY FLYSPECK 40 -20 0 z:o 40 PERCENT AX!AL FLUX O!FFERENCE Page 33 of 41 VEP-NE-1-A, Rev. 0, MRev. 2 3.4 DNB Evaluation Page 34 of 41 The OTDT trip function and setpoints

[13] provide DNB protection for Condition II accidents.

Part of this function, the f(delta-1) term, responds to changes in the indicated delta-I created by skewed axial power distributions.

The axial power distributions formed by the RPDC Condition II accident simulations are evaluated to confirm that the assumptions

[13] used to form the f(delta-1) term and the rest of the OTDT trip function remain valid. If the RPDC power distributions for any subsequent reload should be more limiting than those previously used to establish the OTDT trip setpoints, the OTDT setpoints will be reformulated using standard techniques

[13] and the appropriate RPDC power distribution parameters.

VEP-NE-1-A, Rev. 0, MRev. 2 Page 35 of 41 SECTION 4 -OTHER SAFETY ANALYSES No changes are required to the other safety analysis f!lethods described in Reference 6 to incorporate the effect of the widened delta-I band resulting from the RPDC methodology.

The CAOC methods used by Dominion employ a conservative method for incorporating the effect of skewed axial power distributions.

However, as is the practice with CAOC, the accident analyses will be evaluated on a reload basis for RPDC to ensure that the key input parameters remain bounding.

Should an accident analysis be determined to be impacted by a reload design, that accident will be re-evaluated or reanalyzed, as appropriate.

VEP-NE-1-A, Rev. 0, MRev. 2 Page 36 of 41 SECTION 5 -FQ SURVEILLANCE Dominion instituted FQ Surveillance Technical Specifications (TS) as part of the RPDC implementation process. FQ Surveillance Technical Specifications

[14, 15, 18, 19] are a convenient method for overall power distribution monitoring during plant operation to ensure compliance with the specified LOCA FQ*K(z) limit. The Heat Flux Hot Channel Factor, F Q(z), shall be limited by the following relationships:

• ( ) CFQ x K(z) FQ z :::; p for P > 0.5 [5 -1] ( CFQ x K(z)

• FQ z):::; 0.5 for P:::; 0.5 [5 -2] where the non-dimensional parameters are defined as: CFQ = the plant LOCA FQ limit K(z) = the normalized LOCA FQ(z) limit as a function of core height P = the fraction of rated thermal power For TS surveillance and compliance, FQ(z) is approximated by F3(z) [Equilibrium F 0 (z)] and FJ(z) [Transient F 0 (z)]. Thus both Fg (z) and FJ (z) must meet the preceding limits on FQ (z) and have separate required actions if the measurements are not within the limits. Fg (z) is an excellent approximation for FQ (z) when the reactor is at the steady-state power at which the incore flux map was taken and is calculated by taking the measured FQ(z) (from incore) and accounting for manufacturing tolerances and measurement uncertainties.

F8(z) = FQ(z) x FNU x FQE [5 -3] where the non-dimensional parameters are defined as FQ(z) = the measured plant FQ(z) FNU = Nuclear uncertainty factor [7, 17] FQE = Engineering heat-flux hot-channel factor [9, 10] FJ (z) represents the maximum potential FQ (z) by accounting for increases in localized power due to non-equilibrium normal operation within the allowable Condition I space (core power, delta-I limits, control rod positions, etc.). FJ (z) is calculated by applying a cycle dependent function, N(z), to Fg (z) . The N(z) function represents the maximum possible increase in F 0 (z) that could result from normal VEP-NE-1-A, Rev. 0, MRev. 2 Page 37 of 41 operation within a defined period, power range, and delta-I operating space. The impact of control rod insertion and xenon transients, both axial and radial, are included in N(z). The expression for FJ (z) is: FJ(z) = F§(z) x N(z) [5 -4] ( ) _ Fq(z),Maximum Condition I N z -Fq(z), Equilibrium Condition

[5 -5] The Fq(z)'s in equ'ation

[5 -5] are formed by the standard 1-D/3-D FQ synthesis or the 3-D model techniques as described in Section 2.2. N(z) is similar to V(z) given in Reference 15 and W(z) given in Reference

14. A typical N(z) function is given in Figure 5.0-1. When FQ(z) exceeds the LOCA FQ*K(z) limits defined in equations

[5 -1] and [5 -2], compensatory measures are needed in order to permit continued operation.

There are separate required actions and completion times associated with F§ (z) and FJ (z) not meeting the limits, due to the severity of the condition and the amount of margin needed to restore compliance.

A cycle-specific analysis is performed to determine the relationship between FJ (z) and core power and LOCA delta-I limits. Results from this analysis are presented in the cycle-specific Core Operating Limits Report (COLR) in the form of operating space reductions (thermal power limit and delta-I band limits) that are the TS required actions to address the condition in which the FJ (z) is not within the limit. In addition, a cycle-specific analysis is performed to determine the maximum change in FQ(z) over the required TS surveillance intervals.

Results from this analysis are presented in the cycle-specific COLR in the form of FQ(z) surveillance penalties to be applied to the FQ(z) assessment for any of the following conditions:

FE(z) 1. Increase in measured maximum _Q_ from the previous surveillance, K(z) FT(z) 2. Increase in measured maximum :(z) from the previous surveillance, FE(z) 3. Increase in predicted maximum .!l__) over the next surveillance period, K(z FT(z) 4. Increase in predicted maximum over the next surveillance period. K(z)

VEP-NE-1-A, Rev. 0, MRev. 2 Page 38 of 41' FIGURE 5.0-1 -TYPICAL N(z) FUNCTION ,.... QJ 1.20 0. E :l :? E :l *;:: 1.15 .c *-:l O" w I c 0 z 1.10 0 1 2 3 4 5 6 7 8 9 10 11 12 Height (Feet) N(z)'s IN THE AXIAL EXCLUSION ZONES AT THE TOP AND BOTTOM OF THE CORE ARE OMMITTED FROM FQ SURVEILLANCE PER TECHNICAL SPECIFICATION BASES , I VEP-NE-1-A, Rev. 0, MRev. 2 Page 39 of 41 SECTION 6 -CONCLUSION The RPDC methodology takes advantage of the large amounts of margin to the design bases limits available at reduced power levels in CAOC and forms wider delta-I limits at all powers. The RPDC methodology may be summarized as follows: 1. A full range of normal-operation axial power shapes is obtained by combining the key parameters upon which each shape is dependent:

xenon distribution, boron concentration, core power level and control rod position.

A xenon "free oscillation" method is used to create the many and varied axial xenon distributions required for this analysis.

2. These axial power profiles are analyzed to determine which shapes result in an approach to the LOCA and LOFA limits. 3. A final normal operation delta-I limit is established by conservatively bounding both the LOCA and the LOFA limits. 4. Conditions which yield shapes within the final normal operation delta-I limit are used as initial conditions for the bounding Condition II accident simulations.

5. The resultant transient shapes are analyzed and the overpower and overtemperature trip function/setpoints are specified to ensure that margin to fuel design limits is maintained.

6. A N(z) function is formulated based on calculated Condition I Fq's to support the implementation of Fq Surveillance Technical Specifications.

All neutronics calculations are performed with NRC approved codes. All DNBR calculations are performed using the applicable thermal-hydraulic code(s) and correlation(s) that are listed in the COLR section of the plant Technical Specification.

The RPDC methodology presented in this report allows the Dominion nuclear units to operate with additional operational flexibility while at the same time ensuring that the design bases limits are met with an appropriate margin.

VEP-NE-1-A, Rev. 0, MRev. 2 Page 40 of 41 SECTION 7 -REFERENCES

1. Morita, T., et al.: "Topical Report -Power Distribution Control and Load Following Procedures," WCAP-8385, Westinghouse Electric Corporation, Pittsburgh, PA (September 1974). 2. Letter from A. Schwencer (NRC) to W. L. Proffitt (VEPCO), dated April 4, 1978. 3. "C-E Setpoint Methodology," CENPD-199-NP Rev. 1-NP, Combustion Engineering Inc., Windsor, CT (March 1982). 4. Hanson, G.E.: "Normal Operating Controls," BAW-10122, Rev. 1., Babcock & Wilcox, Lynchburg, VA (April 1982). 5. Miller, R. W., Pogorzelski, N. A. and J. A. Vestovich: "Relaxation of Constant Axial Offset Control," NS-EPR-2649 Part A, Westinghouse Electric Corp., Pittsburgh, PA (August 1982). 6. Bordelon, F. M., et al.: "Westinghouse Reload Safety Evaluation Methodology," WCAP-9272, Westinghouse Electric Corp., Pittsburgh, PA (March 1978). 7. "The VEPCO NOMAD Code and Model," VEP-NFE-1-A, Rev. 0 (May 1985) and Supplement 1 (September 1996). 8. "The PDQ Two Zone Model," VEP-NAF-1, Rev. 0 (July 1990). 9. Letter from Virginia Electric and Power Company (L.N. Hartz) to U. S. Nuclear Regulatory Commission, "Virginia Electric And Power Company, North Anna Power Station Units 1 And 2, Proposed Technical Specifications Changes And Exemption Request Use Of Framatome ANP Advanced Mark-BW Fuel", Serial No.02-167, March 28, 2002. 10. Letter from C. 0. Thomas (NRC) to E. P. Rahe, Jr. (W), "Acceptance for Referencing of Licensing Topical Report WCAP-8691 (P) /WCAP-8692 (NP)," dated December 29, 1982. 11. North Anna Power Station Units 1 and 2 UFSAR. 12. Surry Power Station Units 1 and 2 UFSAR. 13. Ellenberger, S. L., et al.: "Design Bases for the Thermal Overpower delta-T and Thermal Overtemperature delta-T Trip Functions," WCAP-8745 (March 1977). 14. Miller, R. W., et al.: "The FQ Surveillance Technical Specification," NS-EPR-2649 Part B, Westinghouse Electric Corp., Pittsburgh, PA (September

.1982).

/ VEP-NE-1-A, Rev. 0, MRev. 2 Page 41of41 15. Holm, J. S., and R. J. Burnside, "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors -Phase II," XN-NF-77-57(A), Exxon Nuclear Co., Bellevue, WA (May 1981) 16. "Qualification of the Studsvik Core Management System Reactor Physics Methods for Application to North Anna and Surry Power Stations," DOM-NAF-1-P-A, Rev. 0 (June 2003). 17. Letter from United States Nuclear Regulatory Commission (Scott Moore) to Virginia Electric and Power Company (David A. Christian), "Virginia Electric and Power Company -Acceptance of Topical Report VEP-FRD-42, Revision 2, "Reload Nuclear Design Methodology," North Anna and Surry Power Stations, Units 1 and 2 (TAC NOS. MB3141, MB3142, MB3151, and MB3152)", June 11 J 2003. 18. Letter from United States Nuclear Regulatory Commission (V. Sreenivas) to Virginia Electric and Power Company (David A. Heacock), "North Anna Power Station, Unit Nos. 1 and 2-Issuance of Amendments to Revise Technical Specifications to Address Issues Identified in Westinghouse NSAL-09-5, Revision 1, and NSAL-15-1, Revision O (CAC. Nos. MF7186 and MF7187)", October 17, 2016. 19. Letter from Virginia Electric and Power Company (M. Sartain) to United States Nuclear Regulatory Commission, "Virginia Electric and Power Company (Dominion)

Nort.h Anna Power Station Units 1 and 2 License Amendment Request to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1 and NSAL-15-1", December 10, 2015. 20. Letter from Virginia Electric and Power Company (M. Sartain) to United States Nuclear Regulatory Commission, "Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2 Response to NRC Audit Regarding License Amendment Request to Address the Issues Identified in Westinghouse Documents NSAL-09-5, Rev. 1 and NSAL-15-1", June 15, 2016. Note: Current applicable revision of Dominion Topical Reports is maintained in the site-specific SAR.