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Relaxed Power Distribution Control Methodology & Associated Fq Surveillance Tech Specs
	ML20203G548
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Site:	Surry, North Anna, 05000000
Issue date:	03/31/1986
From:	Richard Anderson, Bosehore K, Robins R; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
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PAGE 1 VEP-NE-1-A I VIRGINIA POWER I RELAXED POWER DISTRIBUTION CONTROL METHODOLOGY and ASSOCIATED FQ SURVEILLANCE TECHNICAL SPECIFICATIONS by K. L. Basehore R. T. Robins I R. C. Anderson S. M. Bownan NUCLEAR ENGINEERING ENGINEERING & CONSTRUCTION DEPARTMENT VIRGINIA POWER RICHMOND, VIRGINIA I

MARCH 1986 Reconsnended for Approval:

I Kf%am K. L. Basehore Supervisor, Nuclear Engineering I -

M

. D. Ditadosz O Supervisor, Nuclear Engineering Approved by:

I &R. M.&= = ,

I Eerryman' Director, Nuclear Engineering I

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/ ,'e, UNITED STATES

! 3 ,.y j NUCLEAR REGULATORY COMMISSION (I,- ; y WASHINGTON. D C. 20555 - 7

\, # February 20, 1986 Rec'd. FEB 2 6 1986 I Mr. W. L. Stewart, Vice President Nuclear Operations Nuclear Cperations l Virginia Electric and Power Company Ucensing Supervisor l l Richmond, Virginf>1 23261

Dear Mr. Stewar_t:

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SUBJECT:

ACCEPTANCE FOR REFERENCING OF LICENSING TOPICAL REPORT VEP-NE-1, "VEPC0 RELAXED POWER DISTRIBUTION CONTROL METHODOLOGY AND

550CIATE9 FQ SURVEILLANCE TECHNICAL SPECIFICATIONS" He have completed our review of the subject topical report submitted by the Virginia Elg.ctric and Power Company (VEPCO) by letter dated December 10, 1984.

I We find the report to be acceptable for referencing in license appidcations

'to the_ extent specified and under the limitations delir.eated in the report and the associated NRC evaluation, which is enclosed. The evaluation defines I the basis for acceptance of the report.

We to 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 l applications, except to assure that the material presented 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-0390, it is requested that VEPCO publish accepted versions of this report, proprietary and non-proprietary, l within three months of receipt of this letter. The accepted versions shall l incorporate this letter and the enclosed evaluation between the title page and l the abstract. The accepted versions shall include an -A (designating accepted) l following the report identification symbol.

Should our criteria or regulations change such that our conclusions as to the

acceptability of the report are invalidated, VEPC0 and/or the applicants I 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 i documenta tf or. -

lg5 l

Sincerely, f

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H rbert . Berkow, Director Standardization and Special Projects Directorate Division of PWR Licensing-B l

Enclosure:

As stated I

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FE.8 3 1M I SAFETY EVALUATION REPORT Report

Title:

Vepco Relaxed Power Distribution Control Methodology and Associated F Q Surveillance Technical Specifications l

Report Number: VEP-NE-1 Report Date: October, 1984 I INTRODUCTION The Virginia Electric and Power Company (Vepco) has developed the relaxed power distributien control (RPDC) methodology to replace the constant axial offset control (CAOC) strategy currently employed at its Surry and North Anna M ors. Associated with the RPDC methodology is direct monitoring of the I maximum peaking factorq (F ) relative to plant limits; this replaces the present Fxy Technical Specifications. The analyses performed in support of relaxed power distribution control, and sample generic Fq surveillance Technical I Specifications are described in the subject report. Additional information considered in this review is given in Ref.1.

I SUPNARY OF TOPICAL REPORT I -

"The constant axial offset control (CAOC) strategy currently employed by Vepco was developed by Westinghouse (W) in order to meet power peaking limits imposed by loss of coolant accident (LOCA) analyses. The CAOC procedure requires the maintenance of the axial flux difference (AI) within a specified, constant band about a target axial offset defined at equilibrium conditions. While mainten-ance of AI within these limits insures that the F qis bounded by a specified limit, CACO is unnecessarily restrictive, particularly below full power where significant margin to peaking ifnits exits. These restrictive AI limits have a negative impact on operational flexibility, especially in the ability to return I- 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.

I ENCLOSURE I

I 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 di,stribu-tion control procedures). The variable AI vs. power operating band takes advantage of the increased Fq limits permitted at reduced power by main-taining a roughly constant margin to design limits at all power levels (vs. an I 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 al operating band. The NOMAD analysis produces a spectrum of xenon distributions at selected I burnups via a free-oscillation 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 deoendent rod insertion limit curve at the selected burnups to produce a range of power distributions (and associated AI's) at power levels between 50% and full power.

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2. The axial power distributions from (1) are used in a 1D/2D/3D 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.

I 3. Comparison of the resultant Fg 's to limits prescribed by LOCA analyses defines a preliminary al vs. power operating domain.

4. The entire set of axial power distributions is also analyzed with the l 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 al power operating space that insures that the margin to the DNB design basis for LOFA is maintained.

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5. The most restrictive al power domain (based on LOCA and/or LOFA) defines the pemissible space for normal operation (Condition I). (For Vepco l plants, the LOCA based band is usually more restrictive). Maintenance of AI within this operating space, coupled with adherenck to control rod insertion limits, ensures that the margin to fuel centerline melt, DNB, and LOCA peak clad temperature design criteria are maintained during normal operation.

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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 (0PDT) and over-temperature delta-T (OTDT) trip setpoints have been conservatively calculated, and insures that required margins are maintained. The OPDT and OTDT trips I provide tr'ansient and steady-state protection against fuel center-line melt and DNS, 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).

3 7. The maximum linear power density for each resulting Condition II I ~

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 OPOT f(aI) 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 confirs that the OTDT trip function and its associated f(a!)

term remain valid.

I In conjunction with the implementation of the RPDC methodology, Vepco proposed to replace the current Fxy surveillance with direct monitoring of Fq (z).

In Fqsurveillance the measured Fqat equilibrium conditions is augmented by a factor, N(z), which accounts for the maximum potential increase in Fq (z) during normal opertion. The resultant augmented Fq (z) is compared to the plant LOCA Fq (z) Ifmits to determine acceptability, or to initiate remedial actions. Sample Technical Specifications to be used with F q surveillance are given.

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While the greatest benefit of relaxed power distribution control to Vepco is I- 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.

I SUpmARY 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).

I The approach used for generating bo' u nding 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 x.id 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 I 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 " bounding" nature of the approach, and the extreme distributions considered. This approach has been found acceptable for CE

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reac tors for many years, and is acceptable for RPDC.

The calculation of Fgvia a 10/20/30 synthesis is similar to accepted approaches. Uncertainties associated with the calculation of F are q 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 I uncertainty assigned to the calculated value of F gin the RPDC analyses is therefore acceptable.

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I Calculations of the radial xenon redistribution factor, Xe(z), component of the Fqsynthesis 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 I bounded all observed increases in 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 DN8 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 operation within the pemissible 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 I uncertainty permitted by the Technical Specifications. In addition, Vepco plans to further reduce the AI limits for the first-time analysis. The bounding j

j nature of the RPDC approach provides further conservatism.

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The Vepco RPDC methodology contains elements similar to those included in l

the W (Ref. 7) and CE variable-width AI band axial power distribution control strategies. Approved methods have been used in the analyses supporting RPDC I and justification has been provided for the uncertainties assigned. These analyses and uncertainties are consistent with currently approved methods and practices. In addition, the fapact 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.

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I I. The proposedqF surveillance is similar to the approach approved for W in conjunction 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 W and Exxon, respectively, and I 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 surveillance. This is acceptable because the Fq I surveillance is more appropriate for RPDC.

The sample Technical Specifications in the report acceptably implement RPOC with the following modifications:

I

Specification 3/4.2, page 3/4 2-1 The asterisk at the end of the APPLICA8ILITY line and the footnote should be deleted.

I Figure 3.2-1, 3/4 2-4 I .

This figure should be blank and contain the legend: "This curve is given in the Core Surveillance Report as per Specification 6.9.1.10."

Specification 3.2.2, page 3/4 2-5 -

I Parenthetical comments should be added to the final three lines of action a as follows: " subsequent POWER OPERATION may preceed provided the l Overpower AT Trip Setpoints (value of K4) have been reduced at least 1%

(in AT span) for each 1% F q (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 Commission by letter".

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I A complete set of these revisions will be approved for North Anna Unit 2, Cycle 4 and could be used as a model.

I CONCLUSION .

We find the subject report suitable for reference as support for use of RPDC in licensing applications.

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I REFERENCES Letter from W.L. Stewart (Vepca) to H.R. Denton (USNRC), " Virginia I 1.

Electric and Power Company Relaxed Power Distribution Control.(RPDC)

Supplemental Information," (Oct. 21, 1985).

2. S.M. Bowman, "The Vepco NOMAD Code and Model," VEP-NFE-1A, Virginia Electric arid Power Company (May 1985).

I 3. "C-E Setpoint N thodology," CENDP-199-NP Rev. 1-NP, Combustion Engi neering Inc. (March 1985).

4. W.C. Beck. "The Vepco FLAME Model," VEP-FRD-24A, Virginia Electric and Power Co. (July 1981).

I S. M.L. Smith, "The PDQ07 Discrete Model," 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," Vepco-FRD-33A, Virginia Electric and Power Co. (Oct.1983).

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7. R.W. Miller et al. , " Relaxation of Co,nstant Axial Offset Control,"

NS-EPR-2649 Part A, Westinghouse Electric Corp. (August 1982).

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PAGE 2 CLASSIFICATION / DISCLAIMER The data, information, analytical techniques and conclusions in this report have been prepared solely for use by the Virginia N Electric and' power Company (the Company), and they may not be b

appropriate for use in situations other than those for which they were specifically prepared. The company therefore makes no claim or warranty whatsoever, express or implied, as to their accuracy, usefulness, or applicability. In particular. THE COMPANY MAKES NO I WARRANTY OF MERCHANTABILITY OR FITNESS FOR A PARTICULAR PURPOSE, NOR SHALL ANY WARRANTY BE DEEMED TO ARISE FROM COURSE OF DEALING OR USAGE OF 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 authori=e 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 li .able , 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 unauthori=ed, of this report or the data, information, and analytical techniques, or conclusions in it.

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I PAGE 3 ABSTRACT The Virginia Electric and Power Company (VEPCO) has developed a methodology,,,

called Relaxed Power Distribution Control (RPDC),

for determining the maximum amount of axial power skewing permissible in its nuclear reactors. The RPDC methodology provides a relaxation of the current delta-I operating limits by taking advantage of margin to the design bases criteria. This methodology establishes operating limits by sampling a wide range of potential axial power profiles and determining the conditions where the design bases criteria are exceeded. These conditions define the I limits of permissible operating space. Power distributions resulting from both normal (Condition I) and abnormal (Condition II) operation are analyzed.

I VEPCO intends to use RPDC as the operational strategy for its nuclear units and to implement F2 Surveillance Technical Specifications that compare the measured total peaking factor (TO),

modified by a non-equilibrium operation multiplier, directly to the LOCA total peaking factor limit.

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I PAGE 4 ACKNOWLEDGEMENTS The authors would like to thank Mr. Noval Smith for his assistance concerning the Condition II transients. Mr. Cary Laroe for his assistance concerning reload design procedures, and Mrs.

Anna pegram for typing the draft and final manuscript. The authors would also like to express their appreciation to the people who reviewed and provided comments on this report.

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I PAGE 5 TABLE OF COMTENTS Page I Classification / Disclaimer.................................... 2 Abstract.....................................................

. 3 Acknowledgements............................................. 4 Table of contents............................................ 5 List of Figures.............................................. 6 ListofTables............................................... 6 1.0 Introduction............................................. 7 2.0 Analysis of Axul Shapes Which Result from Normal Oper0 tion.................................... 14 2.1 Axial Shape Gener0 tion................................. 15 2.2 LOCA Delta-I Limit Fermation...... ................... 23 2.3 Loss of Flow Thermal / Hydraulic Evaluation.............. 26 2.4 Final Normal Operation Delta-I Limit................... 27 3.0 Analysis of Axial Shapes Which Result from Condition II Events................................. 28 3.1 Determination of Accident Pre-Conditions............... 29 3.2 Condition II Accident Simulation....................... 29 3.3 Overpower Limit Evaluation............................. 33 3.4 DMB Evaluation......................................... 35 4.0 Other Safety Analyses.................................... 36 5.0 FS Surveillance ......................................... 37 6.0 conclusions.............................................. 40 References................................................... 42 Appendix A................................................... 44 I

I I PAGE 6 LIST OF FIGURES I section 1 i Page 1.O.1 Typical CAOC Limits for Horth Anna and Surry......... 8 1.0.2 Typical Variable Axial Flux Difference Limits........ 11 Section 2 2.1.1 Typical RPDC BOC Xenon Oscillation................... 17 2.1.2 Typical RPDC EOC Xenon 0scillatJr,n................... 18 2.1.3 North Anna Rod Insertion Ilmits...................... 20 2.2.1 Typical LOCA Delta-I Li'2its.......................... 25 Section 3 3.3.1 Maximum Power Density Flyspeck....................... 34 Section 5 5.0.1 Typical N(2) Tunction................................ 39 I LIST OF TABLES Page Section 2 2.1.1 Typical Conditions Analyzed for Normal Operation Under RPDC........................................... 22 I

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PAGE 7 1.0 Introduction l

In response to Loss-of-Coolant Accident (LOCA) Emergency Core !

Cooling System (ECCS) criteria that imposed nou requirements on local power peaking, Westinghouse developed the Constant Asial Offset Control (CAOC) power distribution control procedure !11. The CAOC strategy restricts axial power skeuing in the reactor core I during target normal value, operation determined to within a band of 25% delta-I around a at all-rods-out equilibrium conditions.

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Delta-I is defined as delta-I (%) = 100 * (pt - pb) (1-1) where pt and Pb are the fractions of rated full-core power in the top and bottom halves of the core, respectively. This !5% limit on axial power skeuing reduces the magnitude of axial xenon oscillations which, in turn, decreases the magnitude cf any power Peaking during abnormal operation. VEpCO's four nuclear units Presently operate under the CAOC control strategy. A typical CAOC 3 delta-I hand for North Anna or Surry is shown in Figure 1.0.1. The CAOC target value varies with burnup as the all-rods-out l

i equilibrium delta-I changes, t

Much of the low power operational flexibility of CAOC uns originally centered around the use of the part length rods as a means for axial power distribution control [1]. Full length rods i

and boron were to be used mainly for reactivity control associated l

l with changes in power. Since the requirement for removal of part l

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40 -30 -20 -10 0 10 20 30 40 50 PERCENT AX]AL FLUX O!FFERENCE FIGURE 1.0.1 - TYPICAL CAOC LIMITS FOR NORTII ANNA AND SURRY I

I I PAGE 9 length rods was imposed, full length rods have had to be used to help control the axial power distributions. As a result, it became more difficult to maintain the axial power distribution within the 15% 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 lou 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 allou the full length rod movement required to offset the buildup of xenon and, at the same time, maintain delta-I within its band.

As a result, delta-I limits could be exceeded at lou power levels, requiring the plant to remain below 50% power in order to meet the "one hour in twenty-four"* requirement in the plant Technical Specifications.

Some Westinghouse CAOC plants with available full power mrrgin to their LOCA Overall peaking Factor (FQ) license limits have transformed this margin into operating flexibility through delta-I

" band widening." In the past [21, Surry had a delta-I band width of I +6, -9% about the target value. This method of gaining operational l

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

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I I PAGE 10 flexibility does provide some additional full power dalta-I operating space, but offers only minimal relief for post-trip return to power at end-of-cycle conditions.

This ope'r'ational 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 uidening while also providing for an increasing delta-I hand with decreasing power. The widened delta-I hand 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 band would be selected such that the margin to the LOCA FS*p*KC=) limit 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 l strategy over current CAOC operation are as follous a 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 previous CAOC 25% delta-I band restrictions is now reduced to only that

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3) The reactor coolant system boration/ dilution requirements are decreased, due, in part, to the reduced control rod motion; I 4) The plant has enhanced operational flexibility.

I The concept of widened delta-I limits at reduced power levels is not a new one. Combustion Engineering I31 and Babcock and Wilcox

[4] have supported increased axial skewing at reduced power levels for their reload cores for several years. Westinghouse [5I has also recently developed and licensed a variable delta-I control stategy called RAOC (Relaxed Axial Offset Control) for application to reload cores.

VEpCO has combined some of the concepts from the combustion Engineering methodology [3I with the current VEPCO analysis techniques [1,61 to form an alternate methodology for variable band delta-I control. This methodology is called Relaxed power Distribution Control (RpDC). The chapters that follow will discuss the VEPCO procedure for generating the variable width delta-I hand.

They will 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 FS Surveillance Technical Spacifications. The current CAOC radial peaking factor Fxy(z) surveillance is replaced by FS(=) monitoring, using the measured value of FS(=) augmented by a non-equilibrium operatio.s E

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I PAGE 13 multiplier, in order to verify compliance with the LOCA peaking factors. As will be seen in Chapter 5. FS surveillance complements I RPDC to form a consistent but more flexible plant monitoring scheme than that provided by the current CAOC methods. J I

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I CONDITION I ANALYSIS PAGE 14 2.0 Ana19 sis of Axial Shapes Which Rasult from Normal Operation The objective of a RpDC analysis is to determine acceptable delta-I hand 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 I 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 distributicn, and power level. VEpCO has developed the methodology of Section 2.1 to analy=e the large number of axial shapes included in RPDC.

After the axial power shapes have been created, two separate allowable delta-I limits for normal operation are established: one based on LOCA F2 considerations and the other one based on a Loss of Flou (the limiting DNB transient) thermal / hydraulic evaluation.

The methods used are described in Sections 2.2 and 2.3,

, respectively. These tuo separate deltn-I bands are combined to form a composite delta-I limit as discussed in Section 1.4.

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CONDITION I ANALYSIS PAGE 15 2.1 Axial shape Generation The axial power distributions encountered during normal operation (including load-follow) are primarily a function of four I

~

parameters

the xenon distribution, power level, control rod bank Position and bur nup distribution. For RpDC, reasonable incremental variations that span the entire expected range of values must be considered for euch of these parameters. The following method is used to create the axial power distributions needed for the development of the RpDC ncrmal 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.

I To initiate a xenon-power oscillation, an equilibrium 1-D model (71 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 I

I . . - -- - --

I I CONDITION I ANALYSIS PAGE 16 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 I 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 burnu, 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 2.1.1 and 2.1.2 for beginning- and end-of-cycle, respectively. The VEpCO NOMAD [7] 1D diffusion code was used to perform these calcula6 ions. 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 0.0278 hours 1.653439e-4 weeks 3.805e-5 months of depletion.

I I

I

I I

I

,ee g

I -4 I 50-I ,

n 2,:

I i i h"'

f l [ -2,i I .

I I I -100' 0 20 40 60 80 100 TIME (HOURSI I FIGURE 2.1.1 - TYPICAL RPDC BOC XENON OSCILLATION I

I I

I 100 15f I 50.

I .

25-I

~

i T

iR j F 8 I -502

!I I i

I -100' h 2h 40 60 80 100 I

TIME (HOUR $1 I FIGURE 2.1.2 - TYPICAL RPDC E0C XENON OSCILLATION lI l

'I I - --

I CONDITION I ANALYSIS PAGE 19 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.* 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 nunbar of Power distributions are being analy=ed; i.e. that all safety-related eifects due to the power level are accounted for.

I 2.1.3 Control Bank position During Normal Operation During normal operation, the control rod bank insertion is limited by the Technical Specification 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 th. variable insertion limit. In order to adequately analy=e the various rod Positions allowed, control rod insertions versus power level are selected which cover the range of rod insertions allowed for each particular power.

I

- Se. the footnote on p.g. ..

I I -

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I 0=FU11Y INSERTED. 228= FULLY WITIIDRAWN 240-220-'

200- '

I 180-I R 0

0 160-0 0 140-U P

y120- BANK D E

P 100-I P O

S I 80-I T I

O N 60-I 40-20-0 0 50 20 30 40 50 60 70 80 90 100 I FRRCTION OF RATED POWER FIGURE 2.1.3 - NORTH ANNA ROD INSERTION LIMITS I

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CONDITION I ANALYSIS PAGE 21 2 . 1 . 84 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 (BCC),

middle-of-cycle (MOC) und end-of-cycle (EOC), are chosen for the RpDC analysis. The MOC case is chosen to reflect the maximum middle-of-cycle radial peaking factors.

I 2.1.5 combining Xenon Shapes, Rod position, Power Level and Burnup The final pouar 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 code with normal feedback. Each calculated axial power distribution is stored for use in the LOCA FS and thermal /hydrauli'c 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 ar.d power levels necessary for Surry and North Anna would be slightly different due to the difference in rod insertion limits between the two plants.

v I /

I CONDITION I ANALYSIS PAGE 22 I

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TABLE 2.1.1 TYPICAL CONDITIONS ANALYZED FOR I NORMAL OPERATION UNDER RPDC I Cycle Burnups BOC, MOC, EOC I Xenon Shapes 100 for each time in life Power Level Range (X) 50-100 Rod Inst.rtions Range Versus Power See Figure 2.1.3 I (3 burnups) * (100 P.enon shapes) * (30 power level / rod position combinations) = 9000 shapes I

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l I I . - -- - -- - . - .

I CONDITION I ANALYSIS PAGE 23 2.2 LOCA Delta-I Limit Formation The axial shapes created in Section 2.1 are combined with Fxy(=) data using a standard 1D/2D/3D FS synthesis [ 1,71 :

I FS(=) = Fxy(=)

p(=)

Xe(=)

FMU

FSE

FGR (2-2) where the following are non-dimensional parameters:

Fxy(=) = Fxy distribution calculated by FLAME l81 and I PD207 [ 9 ]; dependent upon burnup, c o r.e height, rod position and power level p(=) = Axial power shape function generated by NOMAD [71 Xe(=) = The radial xenon redistribution factor FNU = Nuclear uncertainty factor i11]

FSE = Engineering heat-flux hot-channel factor i111 I FGR = Grid correction factor [7]

I The axially varying radial xenon factor, Xe(=), compensates for increases to FQ(=) resulting from redistribution of the xenon in the radial plane due to rod movement. The radial xenon redistribution effect cannot be e xpli citly represented in a 1D code and is therefore applied in the synthesis as an uncertainty factor.

Xe(=) is calculated as follows:

max Fxy(=)T Xe(=) = ------------- (2-3)

Fxy(=)E where Fxy(=)T is the /xy(=) calculated from a transient resulting in xenon radial redistribution and Fxy(=)E is the Fxy(=) based upon an equilibrium xenon distribution. Fxy(=)T is calculated with the I

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i CONDITION I ANALYSIS PAGE 24 3D FLAME code by first pre-conditioning the radial xenon distribution for several hours with the core at reduced pouer 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 3D model. Fxy(=)T is calculated for each time step as this transient is followed in small time intervals. The maximum values of Fxy(s)T for the entire transient are used in equation (2-3) to determine Xe(=).

The synthesized FS* power for each shape is compared to the LOCA FS* power *K(=) limit at each power level to determine which axial shapes approach the LOCA limit, thereby establishing a preliminary allouable delta-I versus power band. This comparison replaces the traditional CAOC FAC analysis [1] and ensures that the margin to the LOCA FS* Power *KC=) envelope is maintained during the cycle as long as reactor operation remains uithin the delta-I limits. A typical LOCA delta-I limit is shown in Figure 2.2.1.

A sensitivity study to examine the impact of a change in FS on the width of the LOCA delta-I limits determined that a change of 1%

increase in FS results in less than a 1% decrease in delta-I at constant power. This conclusion is based on the analyses of a range of FS values for VEpCO plants using the nothods just I described.

I I

I 110 I

1004 ,.------------------- .,

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I "*

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l \

E P

[

l \ '

E l \

l E'i E l i t.

N l '

T l '

O

^

F l 1, R [ '.

A 50- ! '

E D

1 I P 0

N 40- '

E I R 30i E 08 10f I O' 0 hD hD -[0 h th 2h 3h 4h 50 PERCENT AX1AL FLUX O!FFERENCE FIGURE 2.2.1 - TYPICAL LOCA DELTA-I LIMITS I

I I - - -. . - _ - . - _ _ .

I C0KDITION I ANALYSIS PAGE 26 I

2.3 Loss of Flou Thermal / Hydraulic Evaluation The Loss of Flou Accident (LOTA) 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 uith the COBRA I10) code. The thermal / hydraulic evaluation methods used in this LOFA evaluation are similar to those of the present CAOC techniques. As a result of this LOTA comparison, a second set of delta-I versus power limits is formed.

Thaise delta-I limits delineate the allouable operating band which will ensure that the margin to the DHB design base for LOTA is maintained. The impact of RPDC on other DNB transient events is discussed in Chapter 3.

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I CONDITION I ANALYSIS PAGE 27 I

2 . t4 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. This set of composite cycle-specific delta-I limits will be made more restrictive than necessary for the first-time analyses in order to bound upcoming reload cycles and minimi=e future Technical Specification changes. These generic limits will be verified on a cycle-by-cycle basis using the RpDC methods described in this report.

I The LOCA FQ based delta-I limits are generally more restrictive than LOTA-based delta-I limits for VEpCO's plants. This will allow the plant Technical Specifications to take advantage of the FS versus delta-I sensitivity identified in Section 2.2 (see Appendix A.2).

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I I CONDITION II ANALYSIS PAGE 28 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 Chapter 2 and the Technical Specification 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 nor nal operation, and the control rods remain within the Technical Specification limits, then the margin to the design criteria of fuel centerline melt, DNB and LOCA peak clad temperature, will be maintained.

I This chapter 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 analy=ed in Section 2. The RpDC analysis examines the more limiting of these condition II events and confirms that the Overpower Delte.-T (0pDT) and the Overtemperature Delta-T (OTDT) setpoints* 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.

I I

The OpDT transient and and OTDT setpoints were designed primarily to provide steady state protection against fuel centerline melt and DNB, respectively.

I I - -- - --

I I I CONDITION II ANALYSIS PAGE 29 I 3.1 Determination of Accident pre-Conditions I Initial condition parameters for condition II analysis are determined f roin 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 Chapter 2 (rigure 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.

I 3.2 Condition II Accident Simulation I 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 pouer Peaking. These three accidents are rod withdrawal, excessive heat removal and erroneous horation/ dilution. The rod withdrawal and boration/ dilution events (1) are the most limiting condition II events with respect to the impact of control rod position on the axial power distribution or local power peaking. In the excessiva heat removal event the impact of temperature is investigated.

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I I CONDITION II ANALYSIS PAGE 30 3.2.1 U'. controlled Rod Withdrawal Event The rod withdrawal event [6] is an erroneous control rod withdrawal starting from a normal operation condition with the control bank's' operating in their normal overlap sequence. To Perform the analysis of this accident, the xenon distribution and boron concentration 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 the NOMAD code (7] and the axial power distribution is saved for use in the Condition II evaluation of Sections 3.3 and 3 . f4 . The analysis is limited to those cases producing power levels between 5 0 Y. of rated power and the high flux trip limit.

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I I CONDITION II ANALYSIS PAGE 31 I 3.2.2 Excessive Heat Removal Event I The Excessive Heat Removal (or cooldoun) event, like the rod withdrawal event, is an overpower accident. The accident assumes a decrease in the reactor c' ore inlet temperature as a result of a sudden load increase, steam-dump valve opening, excessive feedwater flow or a turbine valve opening [61. 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.

I To simulate this accident, allowable normal operation xenon distributions, control rod positions and boron concentrations are provided as input to the NOMAD code (7). The inlet temperature is reduced and a criticality search is performed. The axial power distribution from each case is saved for use in the condition II I 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 FSAR's

[12-131. Cases producing a power level greater than the high flux trip limit are excluded from consideration.

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I I CONDITION II ANALYSIS PAGE 32 I 3.2.3 Boration/ Dilution I 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 dilution incident. However, the consequences of this event are bounded by those of the rod withdrawal accident [61.

I To perform the boration/ dilution analysis, NOMAD reads each allouable 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. 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 saved for the Condition II evaluation of Sections 3.3 and 3 . 84 .

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I I CONDITION II ANALYSIS PAGE 33 3.3 overpower Limit Evaluation The axial power distributions and power levels produced by the condition II accident simulations are combined with calculated Fxy(=) data using the FQ synthesis techniques as described in Section 2.2 (with the addition of the densification spike factor S(=)) to determine the maximum linear power density for each distribution. The results are generally plotted in the " flyspeck" I 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-I) 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 centerlin1t melt by restricting the OTDT f(delta-I) function beyond the DNBR requirement, effectively eliminating the need for the opDT f(delta-I) function.

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-_m - m. _ m- _ _ mn 4.-w.m _ L ab __ _ .__.m..

m m___;Aam ,- . - -na _ - . _-..____.,a_ _ _ _ _ _ . , , ,

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c-I !

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-60 -40 -20 0 20 40 60 PERCENT AX1AL FLUX OlFFERENCE FIGURE 3.3.1 - MAXIMUM POWER DENSITY FLYSPECK I l I _ __

I I CONDITION II ANALYSIS PAGE 35 3.4 DNB Evaluation The OTDT trip function and setpoints [14I provide DNB protection for condition II accidents. part of this function, the f(delta-I) t'e r m , 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 [141 used to form the f(delta-I) 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 !141 and the appropriate RpDC power distribution parameters.

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I OTHER SAFETY ANALYSES PAGE 36 14.0 other Safety Analyses No changes will be required to the other safety analysis methods described in Reference 6 to incorporate the effect of the widened delta-I band resulting from the RpDC methodology. The current CAOC methods used by VEpCO already employ a conservative method for incorporating the effect of skewed axial power distributions. However, as is currently 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 reanaly=ed, as appropriate.

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I F2 SURVEILLANCE PAGE 37 5.0 FQ Surveillance I VEPCO proposes to institute F2 Surveillance Tachnical Specifications, as part of the RpDC procedures. Sample generic Technical Specifications (not specific to any VEPCO unit),

incorporating both FS Surveillance and RpDC, are enclosed in Appendix A. F2 Surveillance Technical Specifications (15,161 are a convenient method for overall power distribution monitoring during plant operation to ensure compliance with the specified LOCA FS*K(=) limit. In FQ Surveillance, the current radial peaking factor Fxy(=) surveillance is replaced by FQ(=) monitoring uhich uses the measured equilibrium Fq(=) augmented by a non-equilibrium operation multiplier and compares this value to the LOCA limit. Fxy is implicitly included in the F2 values. The F2 relationship becomes:

I FSL

M(=)

FEM (=)*H(=) < ----------- for p > 0.5 (5-1)

I '

FQL

K(=)

FQM(=)*H(=) < ----------- for p 5 0.5 (5-2) 0.5 I where the nondimensional parameters are defined as I FSM(=) = the measured plant F2(=) at equilibrium conditions

. the p1 ant tocA r 11mit g r2L I

I - -

I FS SURVEILLANCE PAGE 38 KC=) = the normali=ed LOCA FS(=) limit p = the fraction of rated thermal power N(=) = the maximum potential increase in FSrt(=) resulting from non-equilibrium normal operation.

I M(=) is a factor that represents the largest possible increase in FS(=) that could result from changes in the pouer level and delta-I allowed during normal plant operation:

I FS(=), max Condition I M(=) = -------------------------- -- . (5-3)

FS(=), equilibrium depletion The impact of control rod insertion ant xenon transients, both axial and radial, are all included in M(=). The FS(=)'s in equation (5-3) are formed by the standard FS synthesis methods discussed previously in this report. M(=) is similar to V(=) given in Reference 16 and W(=) given in Reference 15. A typical M(=)

, function is given in Figure 5.0.1.

When FStt(=)*M(=) exceeds the LOCA FS*M(=) limit, the delta-I versus TS sensitivity discussed in Section 2.2 permits compensation by means of a reduction in the normal operation delta-I band. This Provision and the other changes to the plant Technical Specifications resulting from FS surveillance are shown in the sample Technical Specifications given in Appendix A.

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TOP AND BOTTOM 15 PERCENT EXCLUDED AS PER TEcl1NICAL SPECIFICATION 4.2.2.2.G

.50 I 1.45f I

1.40i N

O I.35-I N

E G

U I a.J0-I L *** .,

1 .- ~.

B '

R ,'

I 1.25' I

l 0 ,

l n .

l n I 1.20' I

U '

/

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T I I ',.

p L i.ISi

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I c8 I I.00' r- -,- ,-. ._,-- . . - ' ' - -

-l' 4 l ncicar irecri I FIGURE 5.0.1 - TYPICAL N(Z) FUNCTION I

I

I PAGE 40 6.0 Conclusions 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 pouers. The RpDC methodology may be summari=ed as follous

1. A full range of normal-operation axial power shapes is obtained by combining the key parameters upon which each shape is dependents xenon distribution, boron concentration, core power I 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.

I 2. These axial power profiles are analy=ed 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 LOTA limits.

4. Conditions which yield shapes within the final delta-I limit are used as initial conditions for the bounding condition II accident simulations.

I 5. The resultant transient shapes are analy=ed and the overponer and overtemperature trip function /setpoints are specified to ensure that margin to fuel design limita is maintained.

I H(z)

6. A function is formulated based on calculated condition I I FS's to support the implementation of TQ Surveillance Technical Specifications.

I All neutronics and thermal / hydraulic calculations are performed with HRC-approved codes (7-101.

I

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PAGE 41 The RPDC methodology presented in this report will allow the VEPCO 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. The Technical Specification changes proposed in Appendix A provide the mechanism by which the RPDC methodology can be properly implemented.

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PAGE 42 REFERENCES I 1. Morita, T., et al.: " Topical Report -Power Distribution Control and Load Following Procedures," WCAP-8385, Westinghouse Electric Corporation, Pittsburgh (September 1974).

2. Letter from A. Schwencer (HRC) to W. L. Proffitt (VEPCO), dated April 4, 1978.

3. "C-E Setpoint Methodology," CEMPD-199-NP Rev. 1-NP, combustion Engineering Inc., Windsor, CT (March 1982).

4. Hanson, G.E. "Hormal Operating Controls." BAW-10122, Rev. 1.,

Babcock & Wilcox, Lynchburg VA (April 1982).

5. Miller, R. W., Pogor=elski, H. A. and J. A. Vestovicht

" Relaxation of Constant Axial Offset Control," HS-EPR-2649 Part A. Westinghouse Electric Corp., Pittsburgh (August 1982).

6. Bordelon, T. M., et al.: " Westinghouse Reload Safety Evaluation Methodology," WCAP-9272. Westinghouse Electric Corp., Pittsburgh (March 1978).

7. Bowman, S. M.: "The Vepco NOMAD Code and Model," VEp-HTE-1 (September 1983).*

8. Beck, W. C.: "The Vepco FLAME Model," VEP-FRD-20A (July 1981).

9. Smith, M. L.: "The PD207 Discrete Model," VEP-TRD-19A (July 1981).

10. Sli=, T. W. and K. L. Basehore: "Vapco Reactor Core Thermal-Hydraulic Analysis Using the COBRA IIIC/MIT Computer Code," VEP-FRD-33-A (October 1983).

11. McFarlane, A. T.: " Topical Report - Power Peaking factors "

WCAP-7912-P-A, Westinghouse Electric Corp., Pittsburgh (January 1975).

12. "Vapco North Anna Power Station Units 1 C 2 Updated Final Safety Analysis Report," Virginia Electric and Power Company, Rev. 1 (June 30, 1983).

Currently under NRC reviews approval is expected during 1984 I

I

PAGE 43 I 13. "Vepco Analysis Surry Report "

(June 30, 1983).

Power Station Units 1 E 2 Updated Final Safety Virginia Electric and Power Company, Rev. 1

14. Ellenberger, S. L., et al.: " Design Bases for the Thermal Overpower delta-T and Thermal Overtemperature delta-T Trip Functions," WCAP-8745 (March 1977).

15. Miller, R. W., et al.t "The F9 Surveillance Tachnical Specification," NS-EPR-2649 Part B. Westinghouse Electric Corp.

Pittsburgh (September 1982).

16. Holm, J. S., and R. J. Burnside, " Exxon Nuclear Power Distribution Control for Pressuri=ed Water Reactors -

Phase II,"

XH-NF-77-57(A), Exxon Nuclear Co., Bellevue WA (May 1981).

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I I PAGE 4:4 1

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I APPENDIX A

i r

1 I SAMPLE TECHNICAL SPECIFICATIONS 1I l

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l PAGE 45 A.1 CHANGES TO TECHNICAL SPECITICATIOh 3/4.2.1 I The Actions and Surveillance Requirements relating to the Constant Axial Offset Control delta-I band have been removed from Technical Specification (TS) 3/4.2.1 and replaced with the RPDC requirements.

The Axial Flux Difference (AFD) limit in Figure 3.2-1 is replaced with the RPDC delta-I limits derived in Section 2.4 of this report.

The modified TS 3/4.2.1 requires that delta-I be maintained within the AFD limit or thermal power be reduced. Sample Technical Specifications are attached.

I I

I 3/4.2 POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION I 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the allowed operational space defined by Figure 3.2-1.

I APPLICABILITY: MODE 1 AB0VE 50% RATED THERMAL POWER ACTION:

a. With the indicated AXIAL FLUX DIFFERENCE outside of the Figure 3.2-1 limits, I 1.) Either restore the indicated AFD to within the Figure 3.2-1 limits within 15 minutes, or I 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 percent of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. THERMAL POWER shall not be increased above 50% of RATFD THERMAL POWER unless the indicated AFD is within the Figure 3.2-1 limits.

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I POWER DISTRIBUTION LIMITS I SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE 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:

I 1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after I 2.

restoring the AFD Monitor Alarm to OPERABLE status,

b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is I inoperable. The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

I 4.2.1.2 The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside of the limit shown in Figure 3.2-1.

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I This curve is given in the Core Surveillance Report as per Specification 6.9.1.10.

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, FIGURE 3.2-1 AXIAL FLUX OIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER I

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I PAGE 49 A.2 CHANGES TO TECHNICAL SPECIFICATION 3/4.2.2 The Surveillance Requirements given in TS 4.2.2.2 have been modified to< incorporate FS Surveillance Technical Specifications as discussed in Chapter 5 of this report. The measured overall peaking factor FSM(z), formed by increasing the full core flux map F9(2) by 3% for manufacturing and tolerances 5% for measurement uncertainties, is used to confirm that the plant is operating within the LOCA FR(=)

limit. The top and bottom 15% of the core are not considered in the FQ(z) evaluation due to difficulty in obtaining flux measurements and the small likelihood of obtaining a limiting TS in these core zones. Since FSM(z) is based on equilibrium conditions, the LOCA FSC2) limit is modified by the H(2) factor defined in Chapter 5 of this report.

FS Surveillance is required at least once every 31 effective full power days. If any two consecutive F2 measurements show an increase in peak FSM(z), as' sometimes occurs near beginning-of-cycle, more frequent mapping (every 7 effective full power days) is necessary to accurately determine FSM(z). As an alternative, TS 4.2.2.2e provides for a 2% penalty to be applied to FSM(=), allowing 31 day I mapping to continue. A review of recent VEpCO plant cycles has l

I shown FSM(z).

this penalty to conservatively bound any expected increase in I

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I PAGE 50 Should the actual plant FS measurements indicate that there is not adequate F2 margin to the limit to allow utilization of the entire RPDC AFD band, the AFD limits can be reduced 1% for every 1% in F2 l

violation. This action is based on the FS versus delta-I sensitivity study described in Section 2.2 of this report.

I Sample Technical Specifications are attached.

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I I POWER DISTRIBUTION LIMITS I HEAT FLUX HOT CHANNEL FACTOR-F LIMITING CONDITION FOR OPERATION g (Z)

I 3.2.2 Fq (Z) sball be limited by the following relationships *:

F9 (Z) 1 [Fp [K(Z)] for P > 0.5 I "

Fq (Z) >[Fp[K(Z)]forP 1 0.5 0.5 where P = THERMAL POWER RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1.

ACTION:

With Fq (Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1%nF (Z) exceeds the I limit within 15 minutes and similarly reduce tMe Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent I POWER OPERATION may proceed provided the Overpower AT Trip Setpoints (value of K4) have been reduced at least 1% (in AT span) for each 1% F g (Z) exceeds the limit.

b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a, above; THERMAL POWER may then be increased I provided F (Z) is demonstrated through incore mapping to be within its 0limit.

I *For an actual plant submittal, F*q would be replaced with the plant specific value for the FgLOCA limit.

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I POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS c

4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 Fn (Z) shall be evaluated to determine if9 F (Z) is within its limit by:**

a. Using the moveable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5 percent of RATED THERMAL POWER.

b. Increasing the measured F n (Z) component of the power distribution map by 3 percent to accoutit for manufacturing tolerances and further increasing the value by 5 percent to account for measurement uncertainties, I

c. Satisfying the following relationship:

F"(Z) g < Fg x K(Z) for P > 0.5 P x N(Z)

M Fg (Z) 1 FhxK(Z) for P 1 0.5 N(Z) x 0.5 where F (Z) is the measured Fn (Z) increased by allowances for n is the I manufac Fn limit,K(Zuring) tolerances 1s given in Figureand measurement 3.2-2, uncertainty, Ft P is the relatYve TMERMAL POWER, and N(Z) is the cycle dependent function that accounts for non-equilibrium power dis >ribution effects encountered during normal operation. Thi- function is given in I the Core Surveillance Report as per Specifiu. tion 6.9.1.10.

d. MeasuringFkZ)accordingtothefollowingschedule:

g

1. Upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at I which F (Z) was last determined,* or 9

I *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 a power distribution map obtained.

- FE will be replaced with the plant specific value for the Fg LOCA limit in an ahtualplantsubmittal.

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I POWER DISTRIPUTION LIMITS SURVEILLANCE REQUIREMENT (Continued)

I I 2. At least once per 31 effective full power days, whichever occurs first.

e. With measurements indicating I maximum F"(Z) over Z K(Z)

(

has increased since the previous determination of F (Z) either 9

of the following actions shall be taken:

1. (Z) shall be increased by 2 percent over that specified i 4.2.2.2.c, or I 2. (Z) shall be measured at least once per 7 effective full power days until 2 successive maps indicate that maximum (Z) is not increasing.

over Z (K(Z)

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j With the relationships specified in 4.2.2.2.c above not being satisfied:

1. Calculate the percent Fq (Z) exceeds its limit by the following expression:

maximum F"O(Z) x N(Z) \-1 x 100 for P > 0. 5 over Z FhxK(Z) r .

y h maximum F0 (Z) x N(Z) -1 1x 100 for P < 0.5 over Z x K(Z) b ,P , s

2. Either of the following actions shall be taken:

a. Power operation may continue provided the AFD limits of Figure 3.2-1 are reduced 1% AFD for each percent F (Z) 0 exceeded its limit, or 3/4 2-7 I

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I POWER DISTRIBUTION LIMITS I SURVEILLANCE REQUIREMENTS (Continued)

I b. Comply with the requirements of Specification 3.2.2 for F

above.(Z)exceedingitslimitbythepercentcalculated

g. The limits specified in 4.2.2.2.c, 4.2.2.2.e, and 4.2.2.2.f above are not applicable in the following core plane regions:

1. Lcwer core region 0 to 15 percent inclusive.

2. Upper core region 85 to 100 percent inclusive.

4.2.2.3 When F (Z) is measured for reasons other than meeting the requirements ofSpe0ification4.2.2.2anoverallmeasuredF(Z)shallbeobtained from a power distribution map and increased by93 percent to account I for manufacturing tolerances and further increased by 5 percent to account for measurement uncertainty.

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it , , . . t i , . . . . , , , . . , . . ,

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g 0 2 4 8 3 :D 12 IA I C0RE *E,Ca? t FT)

I Figure 3.2-2 K(Z) - Nonnalized Fg(Z) as a Function of Core Height

~

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I PAGE 56 A.3 CHANGES TO TECHNICAL SPECIFICATION B 3/4.2.1 I The Bases in TS B 3/4.2.1 have been modified to remove references to the CAOC target flux difference. Sample Technical Specifications are attached.

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I 3/4.2 POWER DISTRIBUTION LIMITS I

BATES I The specifications Condition I (Normal of this section Operation) andprovide assurance II (Incidents of fuelFrequency of Moderate integrity)during events ,

by: (a) maintaining the minimum DNBR in the core 2.1.30 during normal I

l operation and in short term transients, and (b) 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 I 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:

I Fg (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 I F N

AH fuel pellets and rods.

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.

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper bound envelope, as given in Specification 3.2.2, is not exceeded dur10g either normal operation or in the event of xenon redistribution following power changes.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alann. The computer determines I the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message imediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowed al-power operating space and I the THERMAL POWER is greater than 50% of RATED THERMAL POWER.

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I I PAGE 58

, A.4 Cl3tHGES TO TECHMICAL SPECIFICATIONS B 3/4.2.2 and 3/4.2.3 I The Bases of TS B 3/4.2.2 and B 3/4.2.3 have been modified to I describe the MCb) function and allow for its update through the Core Surveillance Report. Sample Technical Specifications are attached.

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I POWER DISTRIBUTION LIMITS I

BASES I 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY H0T CHANNEL FACTORS Fg(Z) and F $H ~

The limits on heat flux and nuclear enthalpy hot channel factors ensure that

1) the design limits on peak local power density and minimum DNBR are not I exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.

I Each of these hot channel factors are measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the hot channel factor l

limits are maintained provided:

a. Control rod in a single group move together with no individual rod insertion differing by more than + 12 steps from the group demand position.

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6.

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c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained. l

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

The relaxation in F N as a function of THERMAL POWER allows cha ges in the radial power shape h$r all permissible rod insertion limits. F will be maintainedwithinitslimitsprovidedconditionsathrudabove$remaintained.

When a F n measurement is taken, both experimental error and manufacturing tolerance must be allowed for. 5% is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3% is the I appropriate allowance for manufacturing tolerance.

is measured, experimental error must be allowed for and 4% is the WhenF"$ateallowanceforafullcoremaptakenwiththeincoredetection I appropr 3

system. The specified limit for F g alsocontainsan8%allowagcefor uncertainties which mean that nor m operc+ ion will result in F aH 1 1.55/1.08.

The 8% allowance is based on tb:: following considerations:

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I I POWER DISTRIBUTION LIMITS I BASES I a. abnonnal perturbationsN in the radial power shape, such as from rod misalignment, effect F3H m re directly than F g, l

b. although rod movement has a direct influence upon limiting F to withinitslimit,suchcontrolisnotreadilyavailabletoIkmitF AH' N I

and 1 I c. errors in prediction for control power shape detected during startup physics tests can be compensated forNg in F by restricting axial flux l

distributions. This compensation for F AH is less readily available.

The hot channel factor F M is measured periodically and increased by a cycle and height dependent powbactor, N(Z), to provide assurance that the limit on the hot channel factor, Fn (Z), is met. N(Z) accounts for the non-equilibrium effects of normal operatiUn transients and was detennined from expected power control maneuvers over the full range of burnup conditions in the core. The I N(Z) function for normal operation is provided in the Core Surveillance Report per Specification 6.9.1.10.

3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial I power distribution measurements are made during startup testing and periodically during power operation.

The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts.

The two hour time allowance for operation with a tilt condition greater than I 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on Fn is reinstated by reducing the power by 3 l percent for each percent of tilt in excess of 1.0.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of 4 symmetric thimbles. The two sets of 4 symetric thimbles are I a unique set of 8 detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.

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1 PAGE 61 A.5 TECHNICAL SPECIFICATION 6/9.1.10 ;

I Technical Specification 6/9.1.10 gives a description of the core l

I Surveillance ' Report which is to be provided to the NRC for every cycle.

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CORE SURVEILLANCE REPORT 6.9.1.10 The N(Z) function for normal operation and the Axial Flux Difference limits (T.S. Figure 3.2-1) shall be provided to the Regional Administrator, Region II, with a copy to:

Director, Office of Nuclear Reactor Regulation Attention: Chief, Core Performance Branch I 0.~ S. Nuclear Regulatory Commission Washington, D.C. 20555 at least 60 day prior to cycle initial criticality, unless otherwise approved I by the Commission by letter. In the event that the limits would be submitted at some other time during core life, they shall be submitted 60 days prior to the date the limits would become effective unless otherwise approved by the Commission by letter.

Any information needed to support N(Z) and/or the Axial Flux Difference limits will be by request from the NRC and need not be included in this report.

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ML20203G548

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ML20203G548

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Dear Mr. Stewar_t:

SUBJECT:

Enclosure:

Title:

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