ML20140D889

VEPCO Relaxed Power Distribution Control Methodology & Associated Fq Surveillance Tech Specs

ML20140D889
Person / Time
Site:	Surry, North Anna, 05000000
Issue date:	10/31/1984
From:	Richard Anderson, Basehore K, Robins R VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18142A141	List: ML20140D889
References
VEP-NE-1, NUDOCS 8412190238
Download: ML20140D889 (63)

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VEP-NE-1 I VEPCO RELAXED POWER DISTRIBUTION CONTROL METHODOLOGY and ASSOCIATED F2 SURVEILLANCE TECHNICAL SPECIFICATIONS by K. L. Basehore R. T. Robins R. C. Anderson S. M. Bowman NUCLEAR INGINEERING GROUP ENGINEERING E CONSTRUCTION DEPARTMENT VIRGINIA ELECTRIC C POWER COMPANY RICHMOND, VIRGINIA OCTOBER 1984 l

l Recommended for Approval:

- d 'c.' n /

/ R. W. Cross l Supervisor, Nuclear Engineering hh p D-4 m ' '

I D. D=lados: "

Supervisor, Nuclear Engineering Approved by:

n /"h/,u r AA i

l R. M. Berryman Director, Nuclear E n g :.n e e r in g B412190238 841210 l

I PDR ADOCK 05000280 P PDR

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CLASSIyICATICM/DISCLAIMEp.

I The data, information, analytical techniques end conclusion: in this report have been prepared solely fe: use by the '.' i r g i n :

power I Electric appropriate and for Company (the use in situations othe: than those for which they Company), cnd they may not be were specifically prepared. The Company the:sfore caha: no cicar or Warranty whatsoever, express or implied, as to their :::urc:y, usefulness, or applicability. In pn=ticular. THE C C " P A M *. M A::E S MO

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s .e USAGE Oy T p. A D E , with :espect to this report or a n 't Of thi dt:L, information, analytical techniques, or conclusions in it. E mahing this report available, the Company dee: n0: Outhori:e it:

use by others, and any such use is expressly forbidden e :::1 p I

ul:1 the p:10: W:ltten approval cd the Comp:ny. An" such 2:lt:1:.

approval shall itself be deemed to incorpo:ste thc discl11:s:s of liability and disclcimers of uc::antles p:Ovided ha:aln. 1:. nc event shall the Company be liable, unde en" icycl nc::y t

whatsoever (whethe: cont:cc;, tort, ::::anty, or s;:le: 0: thiclu:1 i

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death, loss of use of p cparty, c: c hd: in. age :22nl:1ng d::: c:

arising out of the use, autho 1:e; or unauth0:1::i. 35 t i .10  : 2 '? ? : :

0 thG data, in5CImation. Cnd Onaly:1 cal techniqui ,  : O O .'. 01% C 1 : n:

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I ABSTRACT The Virginia Electric and power Company (VEpCC) has de/ eloped a methodology, called Relaxed power Distribution Centrol (R?DO),

for determining the maximum amount of axial power skeuin; permissible in its nuclear reacto:s. The EpDC methodology provadas a relaxation of the current delta-I operating limits by taking advantage of margin to the design bases criteria. This =cthedclegy establishes operating limits by sampling a ulde rance ci pctantial axial poue: profiles and determining the cenditiens ute:2 the design bases criteria are exceeded. These conditions define the limits of permissible operating space. poua: distribution:

resulting from both normal (Cendition :) and abnertal ( O c a d i c i s c.

!!) operation are analyzed.

I i VEPCO intends to use RpDC as the operational s t:cta r;' 5 0: 1:r 6

l nucien: units and to implement F2 Surveillance O c c h:.;; 11 Specifications that ecmpare the measured total pcchin- fact:- ( . e. , ,

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modified by a non-equilibrium operation .ultiplier, dirce:17 :: :.; s i

LOCA total peaking facto: limit.

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I PAGE 4 I ACKNOWLEDGEMENTS The authors would like to thank Mr. Noval Smith fer his assistance concerning the condition II transients, tir . Ca ' *arce for his assistance concerning reload design procedures, and Mrs.

Anna pegram for typing the draft and final manuscript. The authc:s

,I would also like to express their appreciation to the pec;1e uho reviewed and provided comments on this report.

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PAGE 5 I TABLE OF CONTENTS pc7e Classification / Disclaimer... ... ....................... ... 2 Abstract.................................................... 3 Aci:nowledgements............................................ u Table of Contents.......................... .. ............. 5 i List of Figures......................................... ... 6 List of Tables............................................... 6 I 1.0 Introduction.................. ...... .... ...... . ...

2.0 Analysis of 'xial Shapes Which Result from Normal Operation... . ..................... . ... .. iv 2.1 Axial Shape Generation.......... ............ ......... 'E 2.2 LOCA Delta-I Limit Formation......... . . .... ... . . 22 2.3 Loss of Flow Thermal / Hydraulic Evaluation.. .. ... .... 26 2.4 Final Normal Operation Delta-I Limit....... . . . . .

3.0 Analysis of Axici Sha'ec p Rhich Result from Condition II Events.......... .... .... .... 2' 3.1 Determination of Accident pre-Conditions. ..... ..... 29 3.2 Condition II Accident Simulation. ......... . . . 27 3.3 overpower Limit Evaluation. ... . . ...... I:

3.4 DM3 Evaluation...... .. . . ... . . . 1:

4.0 Other Safety Analyses.. . . .. . . . ..  ?!

5.0 T2 Surveillance ... ... ... .. . .. .. 2 ""

6.0 Co nclus:.ons . . . ........ ........ .. .. .. . .. . . -0 References.. ... .... ... .......... . ...... .. . . ~2 Appendix A........... .......... . ....... ... . .. Le 1

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PAGE 6 I LIST OF TIGURES Section 1 1.0.1 Typical CAOC Limits for North-Anna and Surry......... 8 1.0.2 Typical Variable Axial Flux Difference Limits...... . 11 Section 2 i

-I 2.1.1 Typical RpDC BOC Xenon Oscillation................... 17 2.1.2 Typical RpDC EOC Xenon Oscillation................... 18 t.1.3 North Anna Rod Insertion Limits................ . .. 20 2.2.1 Typical LOCA Delta-I Limits.. ............... . . . . 25 Section 3 3.3.1 Maximum power Density Flyspeck................. . .. 3'4 Section 5 5.0.1 Typical N(=) Function.... .. ........ .. .. ... ... 39 I LIST OF TABLES I

F a .0 2 Section 2 2.1.1 Typical Conditions Analy=ed for Scr:rl Operaric.-

Under RpDC.. ...... . . . . . .. .. . '. 2

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1.0 Introduction

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In response to Loss-of-Coolant Accident (LOCA) Emergancy Cora Cooling System (ECCS) criteria that imposed new requirements on local power peaking, fJestinghouse developed the constant Axial Offset Control (CAOC) power distribution control procedure [1]. The CAOC strategy restricts axial power skeuing in the reactor core during normal operation to within a band of 25% delta-I around a target value, determined _ at all-rods-out equilibrium condirions.

Delta-I is defined as l I

delta-I (%) = 100 * (pt - ph) (1-1) where pt and pb are the fractions of rated full-core pouar in rhe I top and bottom halves of the core, r'espectively. This r5% limit on axial power skewing reduces the magnitude of axial xanen oscillations which, in turn, decreases the magnitude of any power peaking during abnormal operation. VEpCO's four nuclear units presently operate under the CAOC control strategy. A typical OAOC delta-I band for North Anna or Surry is shown in Figure 1.0.! Tha CAOC target value varies with burnup as tha all-rods-out equilibrium delta-I changes.

I Much of the lou pouer operational flex 1hilary of CAOC uns originally centered around the use of the part length rods as a means for axial power distribution control [1]. Tull length rods and boron were to be used mainly for reactivity control associated with changes in power. Since the requirement for r:moval cf part I

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length rods was imposed, full length rods have had to ha used to help control the axial power distributions. As a result, it becama more difficult to maintain the axial power distribution within the 25:: 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 deplation 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 dilut:.cn to allow the full length rod movement required to offset the buildup of xenon and, at the same time, maintain delta-I within its brnd.

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As a result, delta-I limits could be exceeded at,1cu powar levels.

requiring the plant to remain below 50*: power in order to mast tha "one hour in twenty-four"* requirement in the plant Technical Specifications.

I Some IJe s tinghouse CAOC plants with available full pcuer nr.rgin to their LOCA overall peaking Factor (p2) license limits hate transformed this margin into operating flexibility throu;h dalta-

" band widening." In the past (21, Surry had a delta-: Sand width of

+6, - 9 *: about the target value. This method of g a:.n:.nr o p e r . :. : n a l

> I 'The CAOC Technical Specifications impose no operational lima: ca delta-I while a plant operates below 5 0 *: power. However, :.n c .: d a r to ascend above 5 0:: power, the plant mu t not ha"e e >: c e e f e d the I delta-I bands for more than one penalty hour cf tha gravicus twenty-four.

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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 operational restriction on delta-I imposed by CAOC can be eased by the implementation of a variable detta-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 dalta-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 a=ounts of margin available at reduced power. For North Anna and Surry, which have LOCA-limited total peaking factors, this variable delta- hand would be selected such that the margin to the LOCA F 2 5 p * !' C ) limi-I would remain approximately constant for all power levels. An example of a variable delta-I band is given in Figure 1.3.2.

The principal benefits of a variable band delta-I cont s1 strategy over current CAOC operation are as follows:

, 1) The ability to return to power after a trip, p a r t i c u l . ::1.- at l

end-of-cycle, is enhanced;

2) Control rod motion necessary to compensate for tha prev sus I CAOC 15% delta-I band restrierions is now reduced to only cht:

, motion needed to maintain operation within a much uider hand:

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• ERCENT 4t! L FLJX C!FFE8ENCE 2C JC 40 50 FIGURE 1.0.2 - TYPICAL VARIABLE AXIAL FLUX DIFFERENCE LIMITS I

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3) The re ac to'r coolant system boration/ dilution requirements I,

g are decreased due, in part, to the reduced control rod motion:

I 4) The plant has enhanced operational flexibility.

I The concept of'uidened delta-I limits at reduced power lavels is not a new one. Combustion Engineering (31 and Babcock and Wilcox

[4] have supported increased axial skeuing at reduced power levals for their reload cores for several years. Westinghouse (51 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 Co:abus Bicn I .

Engineering methodology [3] with the current VIPCO analysis techniques (1,6]cto form an alternate methodology for variable h e.nd i

delta-I control. This methodology is called Re la:< e d pouc:

t Distribution Control (RpDC). The chapters that follou ulli discuss the VEpCO procedure for generating the variable width dalta-I cand.

E They will also discuss the methods used to ensure that ha nargin to the design bases criteria, such as Departure from Nu:le n.t a Boiling (DNB), fuel centerline melt and Loss of Coolant A c c i-l e n t (LOCA) peak clad temperature is maintained.

This report also discusses the formulation of F2 Curveillnac

, Technical Specifications. The current CAOC' radial paahing I nc t c,:

Fxy(=) surveillance is replaced by F2(=) monitoring, using ti. e measured value of F2(=) augmented by a non-equilib riuc- opa:atlan 5 -

I PAGE 13 multiplier, in order to verify compliance with the LCCA peaking factors. As will be seen in Chapter 5, F2 surveillance complements RpDC to form a consistent but more flexible plant monitoring scheme than that provided by the current CAOC methods.

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I CONDITION I ANALYSIS PAGI 14 I

2.0 Analysis of Axial Shapes Which Result from Normal Operation The objective of a RpDC analysis is to determine acceptchle 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 uhich adequately bound the potential delta-I range must be generazad.

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. VIpCO has developed the methodology of Section 2.1 to analyre the large number of axial shapes included in RpDC.

I After the axial power shapes have been created, ruo sagarare allowable delta-I limits for normal operation are established: one based on LOCA F2 considerations and the other one based on a Loss l

of Flow (the limiting DNB transient) thermal / hydraulic evcluction.

, The methods used are described in Sections 2.2 and 2.3, respectively. These tuo separate delta-I bands are combinad to form a composite delta-I limit as discussed in Section 2.4.

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I I 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

, 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 parameters. The following methof 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 c o .: a ' s operating history and, as a result, is consrantly changing. In order to analy=e a sufficient number of xenon distriburions to i

I ensure that all possible cases have been accounted for, a xeaon

" free oscillation" method similar to the one described in Referance 3 is used to form these distributions. By creating a divercent xenon power oscillation, axial xenon distributions can be obtainad that will be more severe than any experienced furing normal l operation, including load follow maneuvers.

To initiate a xenon-power oscillation, an e q u il r c r :.'.m 1-D i

l ,W model (7I of the reload cycle is perturbed. This parrurbati:n uill generally be in the form of a change in power, rod p o s :. t i o n , or both. However, since the core model may be inherently stable dua to the presence of feedback mechanisms, these me:hanisms .ust

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I CONDITION I ANALYSIS PAGI 16 either be modified or bypassed to obtain a divergent oscillation.

One way to accomplish this is to reduce the stability of the modal by reducing the amount of Doppler (i.e., fuel temperature) Ieedback 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 rha I expected operating limits.

The stability of the calculational model may vary with burnup or core loading. Therefore, the amount of perturbatica and l Wl feedback modification necessary to achieve a divergent -: e n o n l3 osc*illation may vary with cycle burnup or core loading. Ty?ical

<g examples are given in Figures 2.1.1 and 2.1.2 for beginning- and end-of-cycle, respectively. The VEpCO NOMAD [71 1D diffusion code was used to perform these calculations. These parricular

! oscillations were initiated by reducing power, deplatin7 der l several hours and then returning to full poua for an additional 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of depletion.

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I CONDITION I ANALYSIS PACE 19 I 2.1.2 power Level During Normal Operation For the normal operation analysis, power levels spanning the 5 00: 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 l require CAOC operation belou 50: of full power.* The power levels used for RpDC analysis are selected at increments within the 50:: to 1001 range which are small enough to ensure an adequate nunber of I power distributions are being analyred; 1.e.

safety-related effects due to the power level are accounted for.

that all 2.1.3 Control Bank position During Normal Operation I During normal operation, the control rod bank insertion is limited by the Technical Specification rod insertion li=its. 71gure 2.1.3 gives a set of typical rod insertion limits. The inserzion (I limits are a function of reactor power, and the rods may be jg anywhere between the fully withdrawn position and the variabla l g insertion limit. In order to adequately analyre tha various red positions allowed, control rod insertions versus power level ara selected which cover the range of rod insertions allowed for acch particular power.

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g FIGURE 2.1.3 - NORTH ANNA ROD INSERTION LIMITS I '

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I CONDITION I ANALYSIS PAGE 21 2.1.4 Cycle Burnup The RpDC analysis is performed at several times in cycle lifa in order to provide limiting delta-I bands for the entire cycle.

Typically, three cycle burnups, near beginning-of-cycle (3CC),

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

2.1.5 Combining Xenon Shapes, Rod position, power Level and 3urnup The final power distributions used in the RpDC normal operation analysis result from combining axial xenon chapes, power levels, rod insertions and cycle burnups. At each selected tina 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 calculatad axial power distribution is stored for use in the LOCA T2 and thermal / hydraulic evaluations discussed in Sections 2.2 and 2.2.

I I The combinations of burnups, power levels, rod configurations and xenon distributions typically evaluated on a reload baci: arc I summarired in Table 2.1.1. The conditions result in a d a l u-I range of approximately -60% to +50%, bounding the expected final delta-I envelope at all power levels. The combinations of rod l, insertions and power levels necessary for Surry and North Anna would be slightly different due to the difference in rod insertien limits between the two plants.

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I I CONDITION I ANALYSIS PAGE 20 I

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TABLE 2.1.1 TYPICAL CONDITIONS ANALYZED FOR NORMAL OPERATION UNDER RPDC Cycle Burnups BOC, MOC, EOC I Xenon Shapes 100 f-r each time in life Power Level Range (%) 50-100 l

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= Rod Insertions Range Versus Power: See Figu:e 2.1.3 i

I (3 burnups) * (100 xenon shapes) * (30 power level /:cd l

l position combinations) = 9000 shapes I

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

I FSC=) = Fxy(=)

• p(=)

• Xe(=)

• FNU

• FSE

• FGR (2-2)

I where the following are non-dimensional parameters:

I Fxy(=) = Txy distribution calculated by FLAME [8] and pD207 ( 91; dependent upon burnup, core height, rod position and power level p(=) = Axial power shape function generated by MCdAD l71 Xe(=) = The radial xenon redistribution factor FMU = Nuclear uncertainty factor (11]

F2E = Engineering heat-flux hot-channel factor [11]

FGR = Grid correction factor (7)

I The axially varying radial xenon factor, Xe(=), companents:

for increases to F2(=) resulting from redistribution of th: xanoa l

in the radial plane due to rod movement. The  : dial  ::a ne n redistribution effect cannot be explicitly represented in a *0 cola and is therefore applied in the synthesis as an uncartainty fa: or.

i Xe(=) is calculated as follows:

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Xe(=) = ------------- (2-1)

F :< y ( = ) E where Fxy(=)T is the Fxy(=) calculated from a transient r e r ui t a r.g

! in xenon radial redistribution and Fxy(=)E is tha T::y( ) b2 cl u;cn l an equilibrium Menon distribution. Fxy( )T is calculated with the 1*

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CONDITION I ANALYSIS PAGE 2 '+

3D FLAME code by first pre-conditioning the radial  ::e non 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 tha 3D model. Fxy(=)T is calculated for each time step as this transient is followed in small time intervals. The maximum values of Fxy(=)T for the entire transient are used in equation (2-3) to determine Xe(=).

The synthesi=ed F2* power for each shape is compared to ths LOCA F2* power *K(=) limit at each power level to determina which axial shapes approach the LOCA limit, thereby establishing a preliminary allouable delta-I versus power band. This co.parison replaces the traditional CAOC FAC analysis [1I and ensures that the margin to the LOCA F2* power *KC ) envelope is maintained during the cycle as long as reactor operation remains within tha dalta-I limits. A typical LOCA delta-I limit is shown in Figura 2.2.1.

A sensitivity study to examine the impact of a chan;s .n F2 on the utdth of the LOCA delta-I limits determined that a chan;a of ;;;

increase in F2 results in less than a 1: decrease :. n i rs i t a - I ac constant power. This conclusion is based en the analyses of a D

range of F2 values for VIpCO plants using the - .s thc ls ;uzz i

described.

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CONDITION I ANALYSIS PAGE 26 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 apDC normal operation analysis are evaluated against the 1.55 cosine design axial power distribution for the Loss of Flou Accident analysis with the COBRA [10] 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.

These delta-I limits delineate the allouable oparating band which (l

uill ensure that the margin to the DNB design ba:e for LOFA is l

l3 maintained. The impact of RpDC on other DN3 transient events s l discussed in Chapter 3.

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CONDITION I ANALYSIS PAGE 27 2.4 Final Normal Operation Delta-I Limit Tha results of the LOTA 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 minimine future Technical Specification changes. These generic limits will be verified on a cycle-by-cycle basis using the ppDC serhods describad in this report.

The LOCA F2 based delta-I limits are generally scre restrictive than LOFA-based delta-I limits for VIpCO's plants. This will allow the plant Technical Specifications to t a:-;e advanrage of i

E the F2 versus delta-I sensitivity identified in Section 2.0 (see I

Appendix A.2).

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CONDITION II ANALYSIS PAGE 23 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 n o r. .al operation, and the control rods remain within the Technic 1

, Specification limits, then the margin to the design criteria of fuel centerline melt. DNB and LOCA peak clad temperature, utli ho maintained.

This chapter examines Condition II or Abnormal 0;arctica events, which may be the result of system malfunctions or Oparator errors and create reactor conditions that fall outside the bounds' analyzed in Section 2. The RpDC analysis examines the nore limiting of these Condition II events and confirns tha tha Overpower Delta-T (0pDT) and the Overtemperature Delta-! (070!}

setpoints* have been conservatively calculated and ensures th .t margin to the fuel design limits is maintained. Thera setpunts are verified on a cycle-by-cycle basis.

• The OpDT and OTDT setpoints were designed primarily to provtfa transient and steady ~ state protection against fuci cant;rline malt and DN3, respectively.

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I CONDITION II ANALYSIS PAGE 29 3.1 Determination of Accident Pre-Conditions Initial condition parameters for Condition II analysis are determined from the core conditions allowed by the normal oparation 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 theso conditions which produce an axial power distribution within the normal operation delta-I envelope established in Chapter 2 (T13urs 2.2.1) can be a potential starting point for a Condition II 1 accident. Each set of valid normal operation conditions is considered in the RpDC Condition II analyses.

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3.2 Condition II Accident Simulation Three categories of credible accidents bound tha ran<ja of abnormal operation events which must be considerad in tarms of their effect upon the axial power distribution or local pouar peaking. These three accidents are rod withdrawal. excessivt heat removal and erroneous boration/ dilution. Tha rod withdrausl .na horation/ dilution events (11 are the most limitin7 Ocnditi:n II events with respect to the impact of control rod pos;;1on on On axial power distribution or local power peaking. In the 6.< c a :: t v a heat removal event the impact of temperature is invest 17 tal.

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I I CONDITION II ANALYSIS PAGE 20 3.2.1 Uncontrolled Rod f4ithdrawal Event The rod withdrauel event [6] is an erroneous control rod withdrawal starting from a normal operation condition with the control banks operating in their normal overlap setuence. To Perform the analysis of this accident, the xenon distribution and boron concentration are fixed at values alloued by the normal operation analysis. The lead control bank is then withdrawn in increments from the fully inserted to the fully withdrawn pssitien.

After each incremental movement a criticality s a ar c h :.s performed with the No rt A D code (71 and the axial power distribution is caved for use in the Condition II evaluation of Sections 3.3 and 3.4 The analysis is limited to those cases producing power lovals hatusen I 50 of rated power and the high flux trip limit.

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I CONDITION II ANALYSIS PAGE 31 I

I 3.2.2 Excessive Heat Removal Event I The Excessive Heat Removal (or cooldown) event, like the rod withdrawal event, is an overpower accident. The accident assunas a decrease in the reactor core inlet temperature as a result of a sudden load increase, steam-dump valve opening, excessive feedwatar flow or a turbine valve opening ( 6 !. Since the control rods are assumed to be in manual control for this event, they will ramain at their original position, which allows the reactor p o t.:e r to increase.

To simulate this accident, allowable normal operation xenca I distributions, provided contr,o1 rod positions and boron concantraticn: :rc as input to the N0ft A D code (7!. The inlet tssperaturc is Leduced and a criticality search is performed. Tha a::ial pouar distribution from each case is saved for use in the Condit:.on :

evaluation of Sections 3.3 and 3.4. Reduction of the iniat temperature is limited to 30*F. which has been chot:n to bound the results of the above accidents in the Surry and North 1.nnr T C A p. ' s (13-131. Cases producing a power level greater :han the h:.ph flux

. trip limit are excluded from consideration.

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I I CONDITION II ANALYSIS PAGE 32 I

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. Tha manual mode of operation could result in an overpower transier.;

during a dilution incident. However, the consequances of this event are bounded by those of the rod withdrawal accident ( 61.

To perform the boration/ dilution analysis, NOMAD reads ench allowable xenon distribution from the Condition I analysis and r u.t s a series of cases inserting the rods from fully withd:r.un to the insertion limits in fixed incraments. At each step a criticality search is performed. Once the rods reach the insartion limits, a rod position search is performed to determine the amount of control rod insertion necessary to compensate for the reactivity associatad with a dilution of fifteen minutes. The rods are than starptd 1:1 from the insertion limits to the determined rod position, r.g . n performing criticality searches. All axial pouer dis tri::u ti ons from the boration/ dilution event are saved for the Condi;:on  :

evaluation of Sections 3.3 and 3.4 I

I CONDITION II ANALYSIS PAGE 33 I

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 F2 synthesis techniques as described in Section 2.2 (with the addition of the densification spike factor SC=)) to determine the maximum linear power density for each distribution. The results are generally 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 0?;T f(delta-I) function (which provides protection against this daciga limit) is modified to ensure that margin to the fuel cantarlina melt limit is maintained. If needed at all, this modific.tlon g would be required only for very large values ci delta-I. An i g alternative approach would be to maintain the margin to fuel centerline melt by restricting the OTDT f(delta-I) function beyond

the DNBR requirement, effectively eliminating the need for t!.e C?tT f(delta-I) function.

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1. ERCENT AX!AL FldX CIFFERENCE g FIGURE 3.3.1 - MAXIMUM POWER DENSITY FLYSPECK I

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I CONDITION II ANALYSIS PAGE 35 3.4 DNB Evaluation 5

The OTDT trip function and setpoints [141 provida DM3 protection for Condition II accidents. part of this function, the f(delta-I) term, responds to changes in the indicated dalta-I created by skewed axial power distributions. Ihe axial pouar distributions formed by the RpDC Condition II accident simulations are evaluated to confirm that the assumptions (141 used to form the f(delra-I) term and the rest of the OIDT trip function remain valid. If the RpDC power distributions for any subsequent relcad should be more limiting than those previously used to establish tha OTDT trip setpoints, the OTDT setpoints will be ref o rmulatad us:.ng standard techniques (141 and the appropriate F.p C C pouar distribution parameters.

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I OTHER SAFETY ANALYSES PAGE 36 4.0 Other Safety Analyses No changes will be required to the other safety analysis I methods described in Reference 6 to incorporate the effact of the widened delta-I hand resulting from the RpDC nethodology. The current CAOC methods used by VEPCO already employ a consarvative method for incorporating the effect of skewed axial pouar distributions. However, as is currently the practica uith C;0C.

the accident analyses will be evaluated on a reload basis for P.p C C to ensure that the key input parameters remain bound:.ng . Should an accident analysis be determined to be impacted by a reload famign, that accident will be re-evaluated or reanalyred, as a p p r o p r :.a t a .

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FS SURVEILLANCE PAGE 37 l

5.0 TS Surveillance l

VEPCO proposes to institute TS Surveillance Iachnical Specifications as part of the RpDC procedures. Cample ganeric Technical Specifications (not specific to any VZpCO unit),

incorporating both TS Surveillance and RpDC, are enclosed in Appendix A. TS Surveillance Technical Specifications (15,16! are a convenient method for overall power distribution monitoring during plant operation to ensure compliance with the 572cifi11 :,0 C A TS*M(=) limit. In TS Surveillance, the current r a d :. 21 7 5..k :.n; factor Txy(=) surveillance is replaced by TS(=) sonatorin; which uses the measured equilibrium rq(=) augmented by a non-equ111braun operation multiplier and compares this value to the LOCA 11.it. " .c/

is implicitly included in the FS values. -he T2 r e l a- t o n s !.i p becomes:

1 TSL

• M(=)

TQM(=)sg(=) < ........--- for p > 0.3 (3-1) 1 r2L

• M(=)

TQM(=)*N(=) < ----------- for p 0,5 (3-2:

0.5 ,

where the nondimensional parameters are de.!:.ned as TQM(=) = the measured plant T2(=) at equilib rium c ondit:.cns TSL e the plant LOCA TQ limit

I T2 SURVEILLANCE PAGE 33 I

M(=) =

the normali=ed LOCA F2(=) limit p = the fraction of rated thermal power M(=) =

the maximum potential increase in TSMC=) resulting from non-equilibrium normal operation.

I N(=) is a factor that represents the largest possible incrensa in F2(=) that could result from changes in the pouer level and delta-I allowed during normal plant operation:

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

FS(=), equilibrium depletion The impact of control rod insertion and xenon transients, both axial and radial, are all included in N(=). The TS(=)'s in equat on (5-3) are formed by the standard T2 synthesis me thods dia ct.ssal previously in this report. N(=) similcr VC=) givan in I

is to l

Reference 16 and W(=) given in Reference 15. A typical Hi=)

function is given in rigure 5.0.1.

1 .

When TSM(=)*N(=) exceeds the LOCA F2*M(=) limit, tha delta -

versus T2 sensitivity discussed in Section 0,.2 permite coapensa non by means of a reduction in the normal operation delta-I bcnd. T h:. J provision and the other changes to tha p'.:n- Ta ::.n ic :1 Specifications resulting from T2 surveillance are shcun in th2 sample Technical Specifications given in Appendix A.

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I I i.5C TOP AND BOTTOM 15 PERCENT EXCLUDED AS PER TEC11NICAL SPECIFICATION 4.J.J.J.G 1.45-6 40i C L . 35-:

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FIGURE 5.0.1 - TYPICAL N(Z) FUNCTION

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! 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 uider delta-I limits at all pouers. The RpDC methodology may be summari=ed as follows:

1. A full range of normal-operation axial power shapas is cbtainsd by combining the key parameters upon which each shape is dependent: xenon distribution, boron concentration, core pouer I level and control rod position.

method is used to create the A

many xenon " free oscillarion' and varied axial nenon dis +<ibutions required for this analysis.

2. These axial power profiles are analyred to determine which I shapes result'in an approach to the LOCA and LOFA limits.

3. A final normal operation delta-I limit is esrablished by conservatively bounding both the LOCA and the LOTA limits.

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

5. The resultant transient shapes are analyred and the everpouar and overtemperature trip function /setpoints are spccified to ensure that margin to fuel design limits is maintained.

I 6. A F2's N(r) function is formulated based on calculared Cond1:1:n :

to support the implementation of 72 Surveillanca Ttch.ucs1 Specifications.

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

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PAGE 41 1

The RpDC methodology presented in this report will allow the VEpCO nuclear units to operate with additional operazional 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.

PAGE 42 l .

RETERENCES l

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

2. Letter from A. Schwencer (NRC) to W. L. Proffitt (VEPCO), dated I

April 4, 1978.

3. "C-E Setpoint Methodology," CENPD-199-NP Rev. 1-NP, Combustion I

l Engineering Inc., Windsor, CT (March 1982).

l

4. Hanson, G.E. " Normal Operating Controls." BAW-10122, lav. 1.,

Babcock E Wilcox, Lynchburg VA (April 1982).

5. Miller, R. W., Pogor:elski, M. A. and J. A. Vastov10h I " Relaxation A,

of Constant Axial Offser Control " NS-EPR-2549 Parr Westinghouse Electric Corp., Pittsburgh (August 1902).

I 6. Bordelon, F. M., et al.: " Westinghouse 2eload Safety Ivaluation Methodology," WCAP-9272, Westinghouse Electric Corp., Pittsburgh (March 1978).

I 7. Bouman, (September 1983).*

S. M.: "The Vepco NOMAD Code and.?fodel," VIP-NTI-1

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

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

10. Sli=, F. W. and K. L. Basehore: "Vepco Reactor ccre Thermal-Hydraulic Analysis Using the COBRA IIIC/MIT Cc putar Code," VEP-FRD---33-A (October 1983).

11. McFarlane, A. F.- " Topical Report - Power Paahing Facrcrs,"

I WCAP-7912-P-A, 1975).

Westinghouse Electric Corp., Pitrsburgh (Jcnuary

12. "Vepco North Anna Power Station Units 1 E 2 Updated T:nal Cadety Analysis Report," Virginia Electric and Peuer Company, Ra.. 1 (June 30, 1983).

I

• Currently under NRC revieu; approval is expected during 1?O .

l I

5

1 B

l PAGE 43 l

13. "Vepco Surry Power Station Units 1 C 2 Updated T nal Safety B Analysis Report," Virginia Electric and Power Company, Rav. 1 I (June 30, 1983).

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

15. Miller, R. W., et al.: "The T2 Surveillance Technical Specification," NS-EPR-2649 Part B. Westinghouse Electric Corp.

Pittsburgh sSeptember 1982).

16. Holm, J. S., and R. J. Burnsidu, " Exxon Nuclear .'o. e r Distribution Control for Pressurized Water Reactors - Phase II,"

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

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I APPENDIX A SAMPLE TECHNICAL SPECIFICATIONS 3 .

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E PAGE 45 I

A.1 CHANGES TO TECHNICAL SPECIFICATION 3/4.2.1 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 apDC 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.

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. __ _ . _ _ _ _ . _ _ _ . - _ _ _ . _ . _ _ _ _ _ _ _ - _ - _ _ . _ _ _ _ _ _ . - _ _ __ ._ . _ _ O

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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.

APPLICABILITY: MODE 1 AB0VE 50% RATED THERMAL POWER

• ACTION:

I a. With the indicated AXIAL FLUX DIFFERErlCE outside of the Figure 3.2-1 limits, 1.) Either restore the indicated AFD to within the Figure 3.2-1 limits within 15 minutes, or 2.) Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range fleutron Flux -

High Trip setpoints to less than or eoual to 55 percent of

' RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,

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

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"See Special Test Exception 3.10.2.

3/4 2-1 I -

5 .

I POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its li:aits during POWER OPERATION above 50". of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1. At least once per 7 days when the AFD ffonitor Alarm is I OPERABLE, and

2. At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after I restoring the AFD Monitor Alarm to OPERABLE status.

b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE I for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the 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.

. l 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 OPERATION -

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I AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER (TYPICAL EXAMPLE) l 3/4 2-4 I

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PAGE 49 il A.2 CHANGES TO TECHNICAL SPECIFICATION 3/4.2.2

g The Surveillance Requirements given in TS 4.2.2.2 have been modified to incorporate F9 Surveillance Technical Specifications as discussed in Chapter 5 of this report. The measured overall peaking factor FSM(=), formed by increasing the full core flux map F2(2) by 3% for manufacturing tolerances and 5% for measurement uncertainties, is used to confirm that the plant is operating within the LOCA F2(=)

limit. The top and bottom 15% of the core are not considered in the FQ(=) evaluation due to difficulty in, obtaining flux measurements and the small likelihood of obtaining a limiting F2 in these core cones. Since FSM(c) is based on equilibrium conditions, the LOCA FSC=) limit is modified by the N(=) factor defined in Chapter 5 of this report.

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FQ Surveillance is required at least once every 31 effective full power days. If any two consecutive F2 measurements show an increase in peak FSM(=), as sometimes occurs near beginning-of-cycle, more {

frequent mapping (every 7 effective full power days) is necessary to accurately determine F9M(=). As an alternative, TS 4.2.2.2e l provides for a 2% penalty to be applied to F2M(c), allowing 31 day mapping to continue. A review of recent VEpCO plant cycles has shown this penalty to conservatively bound any expected increase in FSM(=).

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I PAGE 50 Should the actual plant F2 measurements indicate that there is not adequate F2 margin to the limit to allou utilisation of the entire RPDC AFD band, the AFD limits can be reduced 1% for every 13: in F2 violation. This action is based on the F2 versus delta-I sensitivity study described in Section 2.2 of this report.

Sample Technical Specifications are attached.

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lI POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-F g (Z)

I LIMITING CONDITION FOR OPERATION I

3.2.2 Fg (Z) shall be limited by the following relationships *:

F(Z)3[Fg][K(Z)]forP' g

> 0. 5 I "

Fg (Z) 3 [Fj] [K(Z)] for P < 0.5 0.5 where P = THERMAL POWER RAit0 inERMAL F0WER and K(Z) is the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1.

ACTION:

I With F0 (Z) exceeding its limit:

a. Reduce THERMAL POWER at least l'. for each l'; F (Z) n exceeds the I limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsecuent I POWER OPERATION may proceed provided the Overpower :T Trip Setpoints have been reduced at least l'; for each l', Fg(Z) exceeds the limit.

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

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• Foranactuaiplantsubr.ittal,Ffwouldbereplacedwiththeplantspecific value for the F gLOCA limit.  ;

3/4 2-5 I

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I POWER DISTRIBUTION LIMITS

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

4.2.2.2 F g

bh(:Z)shallbeevaluatedtodetermineifF(Z)iswithinitslimit

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 I measurement uncertainties.

c. Satisfying the following relationship:

for P > 0. 5 Fh't(Z) FjxK(Z)

P x N(Z)

F"(Z) 1 FjxK(Z) for P 1 0.5 N(Z) x 0.5 where FM (Z) is the measured F (Z) increased by allcwances for I manufac9uring tolerancas and beasurement uncertainty, F is the Fn limit, K(Z) is given in Figure 3.2-2, P is the relat ve THERMAL POWER, and N(Z) is the cycle dependent function that I accounts for non-equilibrium power distribution effects encountered during normal operation. This function is given in the Core Surveillance Report as per Specification 6.9.1.10.

M

d. Measuring F g (Z) according to the following schedule:

1. Upon achieving eouilibrium conditices after exceeding by 10 I percent or more of RATED THERMAL POWER, the THERitAL POWER at which Fg (Z) was last determined,* or

• 0uring power escalation at the beginning of each cycle, Ocwer level may be increased until a pcwer level for extended operation has been acnieved and a pcwer distribution map obtained.

I **F' will be replaced with the plant specific value for the F LOCA limit in an a tual plant submittal.

g 3/4 2-6 I

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I POWER DISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENT (Continued)

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 Fg (Z) either of the following actions shall be taken:

1. F"(Z) shall be increased by 2 percent over that specified i 4.2.2.2.c, or I 2. F' (Z) shall be measured at least once per 7 effective full power days until 2 successive maps indicate that maximum F0 (Z) is not increasing, over Z K(Z) j

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

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

maximum F'(Z)

O x N(Z) -1 x 100 for P 1 0.5 I

I (overZ FhxK(Z) s P .

)

f 7 I l 41 maximum over Z F'q(Z) x M(Z) 0

-1

{

x 100 for P < 0.5 FfxK(Z)

- J

2. Either of the following actions shall be taken:

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

I:

POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued)

I b. Comply with the requirements of Specification 3.2.2 for F0 (Z) exceeding its limit by the percent calculated aDoVe.

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. Lower 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 ofSpe9ification4.2.2.2anoverallmeasuredF(Z)shallbeobtained from a power distribution map and increased by9 3 percent to account I for manufacturing tolerances and further increased by 5 percent to account for measurement uncertainty.

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-l Figure 3.2-2 K(Z) - Nonnalized Fq (Z) as a Function of Core Height 3/4 2-8A I

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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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B 3/4.2 POWER DISTRIBUTION LIMITS BASES The Condition specifications I (Normal of this section Operation) andprovide assurance II (Incidents of fuelFrecuency of Moderate integrity)during events by: (a) maintaining the minimum DNBR in the core 2.1.30 during normal 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:

Fg (Z) Heat Flux Hot Channel Factor, is defined as the maxinum local heat flux on the surface of a fuel red at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances en I F,H N

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 pcwer to the average rod power. .

3/4.2.1 AXIAL FLUX DIFFERENCE (AFC)

The limits on AXIAL FLUX DIFFERENCE assure that the Fg (Z) upper bound enveloce, as given in Specification 3.2.2, is not exceeced during 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 Alarm. The computer determines I the one minute average of each of the OPERABLE excore detector autouts and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allcwed iI-pcwer operating space and the THERMAL POWER is greater than SOS of RATED THERMAL POWER.

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A.4 CHANGES TO TECHNICAL 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 describe the N(=) function and allow for its update through the core Surveillance Report. Sample Technical Specifications are attached.

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I POWER DISTRIBUTION LIMITS BASES I

3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS F(Z)andF$H I g 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.

Each of these hot channel factors are measurable but will normally only be I 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 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.

I c.

The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained,

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

TherelaxationinF5'g asafunctionofTHERMALPOWERallowschaqgesinthe radial power shape Tor all permissible rod insertion limits. F. will be maintained within its limits provided conditions a thru d above~gare maintained.

When a F n measurement is taken, both experimental error and manufacturing toleraned must be allowed for. 55 is the acpropriate allcwance for a full core I cap taken with the incore detector flux map::ing system and 3" is the appropriate allowance for manufacturing tolerance.

..h en F',t is measured, experimental error must be allowed for and C' is the I aoproplg1 ate allowance for a full core map taker with the incore detection system. The specified limit for F g also contains an SS allowance for uncertainties which mean that normai operation will result in Plg' ; 1.55/1.08.

The 8'. allowance is based on the following censicerations:

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I POWER DISTRIBUTION LIMITS BASES lg

a. abnonnal perturbationsq in the radial power shape, such as from rod misalignment,effectFbH more directly than Fg ,

b. although rod movement has a direct influence upon limiting Fg to within its limit, such control is not readily available to ifmit F Ng, and

c. errors in prediction for control power shape detected during startup physics tests can be compensated forgg in F by restricting axial flux 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 powbZfactor, N(Z), to provide assurance that the limit on I the hot channel factor, F effectsofnormaloperatibn(Z),ismet.

N(Z) accounts for the ncn-equilibrium transients and was determined 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 OUADRANT 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 pcwer distribution measurements are made during startup testing and periodically during power operation.

The limit of 1.02 at which corrective action is reouired provides Of:B and I linear heat generation rate protection with x-y plane power tilts.

The two hour time allowance for operation with a tilt condition areater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned roc. In the event such acticn does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the power by 3 percent for each percent of til9 in excess of 1.0.

For purposes of monitoring CUADRANT POWER TILT RATIC when one excore detector is incperable, the moveable incore cetectors are used to confirm that the nor-alized symmetric power distribution is consistent with the CL'ACRANT PCWER TILT RATIO. The incore detector monitoring is done with a full inccre flux map or two sets of 4 symmetric thimbles. The two sets of a syrretric thimoles are a unicue set of 8 detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-ll, and N-8.

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i A.5 TECHNICAL SPECIFICATION 6/9.1.10 i

Technical Specification 6/9.1.10 gives a description of the core Surveillance Report which is to be provided to the NRC for every

[ cycle.

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I CORE SURVEILLANCE REPORT 6.9.1.10 The N(Z) function for normal operation shall be provided to the Regional Administrator, Region II, with a copy to:

Director, Office of Nuclear Reactor Regulation Attention: Chief, Core Performance Branch U. S. Nuclear Regulatory Commission Washington, D.C. 20555 at least 60 day prior to cycle initial criticality. 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 exempted by the Commission.

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

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