ML20042F238
| ML20042F238 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 03/31/1990 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML19302E102 | List: |
| References | |
| DPC-NE-2011A, NUDOCS 9005080030 | |
| Download: ML20042F238 (80) | |
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DUKE POWER COMPANY NUCLEAR DESIGN METHODOLOGY for
-CORE OPERATING LIMITS of is WESTINGHOUSE REACTORS j.
DPC-NE-2011A MARCH 1990
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January 24, 1990
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-l Mr. P. B. Tucker, Vice President I-Nuclear Production W CO-Ouke Power Company
.f. htca ccMPLtANCE l P. 0.. Box 33189 Charlotte, NC 28242
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Dear Mr. Tucker:
SUBJECT:
ACCEPTANCE FOR REFERENCING OF TOPICAL REPORT DPC-NE-?011P, " DUKE POWER COMPANY NUCLEAR DESIGN METH000 LOGY FOR CORE OPERATIFG LIMITS OF WESTINGHOUSE REACTOR 5" i
The staff hrs completed its review of the Topical Report DPC-NE-2011P, " Duke Power Company Nuclear Pesign Methodology for Core Operating Limits of I
letter dated April 27, 1988.
Additional information was' submitted on bestinghouse Reactors" submitted for NRC review by the Duke Power Company by March 20, 1989. This topical report (DPC-fiE-2011P) provides information and -
(g justification for the operating limits on power distribution, control rod' e
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insertion and power distribution inputs to the overpower-delta-T and overtemperature-delta-T reactor protection system trip functions. These limits
'are the axial flux difference for a given power level, the rod insertion limits I
and the f(delta-I) function of the overpower-and overtemperature-delta-T.
These operating limits provide assurance that the peak local power is not greater than that assumed in the design basis transient and accident analyses.
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The limits are set such that the RPS will trip the reactor before fuel damage E
occurs. A three-dimensional reactor model power distribution is employed for the maneuvering analyses in several points in the core life. These power
- aj distributions' are based on a set of conservative xenon distributions to ensure g-that the predicted pcwer distributions are conservative with respect to those expected to occur.
These power distributions are augmented 'by appropriate.
uncertainty ~ factors.
We find the application of' DPC-NE-2011P to_ be acceptable for referencing in license applications to the extent specified, and under the limitations delineated, in DPC-NE-2011P and the associated NRC technical evaluation. The evaluation defines tne basis for acceptance of this topical report.
We do not intend to repeat our review of the matters found acceptable as described in DPC-NE-2011P when the report appears as a reference in license applications, except to assure that the n'aterial-presented is applicable to the specific plant. involved.
Our acceptance applies only to the matters oescribed in the' application of DPC-NE-2011P.
!n accordance with procedures established in NUREG-0390, it is requested that the Duke Power Company publish accepted versions of this topical report, proprietary and non-proprietary, within three months of receipt of this letter. The accepted versions shall includ,e an -A (designating accepted) following the report identification symbol.
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H. B. Tucker
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January 24, 1990-
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'Should our criteria'or regulations change so-that'our conclusions as to the acceptability of the' report are invalidated, Duke Power Company and/or the i
applicants referencing the topical report will be expected to revise and resubmit their respective documentation, or. submit justification for the continued effective applicability of the topical report without revision of their respective documentation.
Sincerely,
, h[ /M
[.N Ashok C. Thadani, Director 5-Division of Systems. Technology Office of Nuclear Reactor Regulation Enciosure:
DPC-NE-2011P Evaluation
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o ENCLOSURE SAFETY EVALUATION FOR THE TOPICAL REPORT DFC-NE-2011P
" DUKE POWER-COMPANY, NUCLEAR DESIGN METHODOLOGY FOR CORE OPERATING LIMITS OF WESTINGHOUSE REACTORS" 1.0' INTRODUCTION l
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' By letter dated April 27, 1988, the Duke Power Company submitted the Topical Report DPC-NE-2011P for NRC review (Ref. 1).
Additional information was submitted en March 28, 1989 (Ref. 2). This topical report provides information and justification for the operating limits on power distribution, control rod insertion and power distribution inputs to the overpower-delta-T and' overtemperature-delta-T reactor protection system-trip' functions. These-limits are the axial flux difference for_ a given power level, the rod insertion limits and the f(delta-I) function of the overpower-and overtemperature-delta-T. These operating limits provide assurance that the peak local' power-is not greater than that assumed in the design' basis transient and accident analyses. The lwMts are set such that the RPS will trip _the reactor before fuel damage occurs. A three-dimensional reactor model power distribution-is employed for the maneuvering analyses in several points in the, core life. These power distributions are based on a set of conservative xenon distributions to ensure that the predicted power distributions are conservative with respect to those expected to occur.
These
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power distributions are augmented by appropriate uncertainty factors.
<l The follow'ing evaluation incorporates our consultert's BNL, contribution to f
I this review.
Restrictions to be observed in the application of this topical f
report are listed in Section 3.5.
2.0 gMMARY OF THE TOPICAL REPORT At first the report describes the three-dimensional nodal power and xenon distribution generation method which is based on an NRC approved version of j
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'the'EPRI-NODE-P code.(Ref. 3).
The local radial factors are esticated using a I
pin-by-pin PDQ-07 model.
Power distributions are generated for different times in the cycle.
Limitino xenon dist-ibutions are cenerateo to assure conservatism. The power distribution is augmented by uncertainty factors which l
cccount for the (X-Y) power distribution calculation uncertainty, cuadrant tilt 1
and axial power distribution.
l The general methodology for the limiting condition of operation end the j
reactor protection. system limits is followed by the calculation of the LOCA rrargin and the estimation of the loss of flow DNB limits._ In addition IL the reactor protection system maroin, the centerline fuel melt margin, the 1
axial flux ~ difference power level limits and the control red insertion limits are calculated.
l The power distribution surveillance and their relation to the operation and
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transient limits are then estimated for the LOCA F limits, the Hss of flow j
q DNB, Fg, axial flux difference power level limits, control rod insertion.
limits, the heat flux Fot channel factor, the nuclear enthalpy rise hot
)j channel factor and the quadrant power tilt.
i
_ Appendix' A in the report gives a brief description of the computer codes used in the above calculations.
l 3.0 EVALUATION The proposed methodol)gy employs a three-dimensional reactor and cycle specific model in conjunction with xenon distributions obtained from a j
maneuvering analysis which sinulates severe xenon transients.
Bounding power 5
' distributions are then generated based on these severe xenon distributions,
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and various combinations of rod positior,s, inlet temperature, pcwer. level and cycle burnup. These power distributions are compared to operatino and safety i
thermal limits to define er validate the axial flux difference (AFD) power level openting space, the rod insertion limits and the f(delta-1) penalty function employed in the OPol and/or the OT& trip functions of the Peactor I
Protection System (RPS) such that power distributions that might exceed the
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3 respective thermal limits are prohibited.
In aedition to the xenon transient b6 sed power distributions, a number of anticipated transients (e.g., boron dilution, rod withdrawal, etc.) are analyzed in setting the RPS limits. A core monitoring / surveillance procedure which assures safe operation within the applicable limits is an integral part of the proposed methodology.
This approach is an alternative to the Relaxed Axial Offset Control (RAOC) i methodology (Ref. 4) currently in use at Duke Power Company's (DPC) McGuire i
i and Catawba Nuclear Stations, The present review considered the infomation provided in the topical report along with additional information provided by DPC in response to a request for i
additional infomation (RAI) (Pef. 5).
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The computer codes and associated nethodologies employed in the power distribution and peaking calculations have been previously reviewed by the NRC i
and found to be acceptable (Refs. 6 and 7). The shutdown margin and ejected rod analyses that enter into the setting of control rod insertion limits have also been approved by the NRC. A topical report describing the codes and methods to be used by DPC to generate the core thermal hydraulics (including hot rod) for Vestinghouse (W) reactors is presently under review (Ref. 9).
.in view of the above, and noting that the DPC methods for determinino maximum allowable LOCA peakirg and loss of flow accident (10FA) DNB based operating limits and maximum allowable DNB and linear heat rate based RPS limits have been appreved by the NRC, the acceptability of the proposed methodology hinoes on the following major issues.
3.1 Operatino~Soace AFD Limits Since the-proposed methodology represents a departure from currently accepted practice, any changes in limits relative to~ those obtained with the presently employed ano approved PAOC methodology that represent a reduction in conservatism must be justified.
DPC has indicated that the proposed methodology will yield operating space AFD limits that are a few percent wider (less conservative) than the current RA0C
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'imits; this is due primarily to the use of explicit three-dimensional (3-D)
I power distribution's as. opposed to the synthesired 3-D power aistributions on.
which RA00 is based.
The increase in the available margin. and consequently-the AFD operating space limits, is consistent with previous experience that supports a reduction in peaking when explicit 3-D power distributions are used as compared to synthesizing 3-D distributions from 1-0 and 2-D calculations.
Under the proposeo DPC methodology, if cperating limits are too restrictive for normal operation, a set of limits can be defined that may still allow l
operation at full pcwer..The resultino " base load" operation is typically
.used above 80 percent power and is similar to the widely used and accepted I
constant axial offset control (CAOC) approach.
The xenon distributions used in setting the limits in this case are restricted to a relatively narrow nperating band about a predicted AFD target.
l It is therefore concluded that the DPC approach is acceptable with respect to AFD limits.
3.2 Conservatism of Power Distributions 1
in order to have conficence in the cperating anc RPS limits obtained by the
'3 proposed methodolooy, there must be demonstrateo assurance that the power l:
distributions resulting from the DPC approach are conservative with respect to a
f those that might be reasonably expected to occur,. and that they sufficiently span the AFD/ rod-insertion power-level operating spaces to permit an accurate determination of limits.
- DPC has determined through sensitivity studies.that the power distributions j
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employed in setting the operating and RPS limits are conservative.
This is
-l due in part to the severity of the xenon transicists employed in the maneuvering analyses and conservative modelling ' assumptions.
In addition, since the limits are based on the analyses of almost 2000 three-dimensional power distributions (resulting from a matrix of power level / rod position / inlet temperature /burnup and xenon distribution statepoints), DPC is confident that the operating limits can be determined accurately, and any extrapolation would I
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be minimal.
A review of the statepoints (combinations of power level, rod insertion, etc.) and anticipated transients considered by DPC in generating bounding power distributions supports the conclusion that there is assurance that the power distributions assumed in the analyses of thermal limits are indeed conservative relative to the expected distributions, and this aspect of the DPC methodology is acceptable.
It should be noted that the matrix of tatepoints currently censidered in the analysis may be tredified as experience is accumulated.
However, any reductions in the number of statepoints considered should be implemented only if there are no concomitant adverse effects (e.g., excessive interpolations required to set limits).
3.3 Uncertainties and Parameters in Marcin and Monitorina Alcorithms I
The DPC methodology requires the determination of margins to linear heat rate and DNB thermal limits and the monitoring of the measured state to assure that operation is consistent with the DPC analyses performed to ensure that these limits will not be violated. Two linear heat rate related margins are
- I.
determined - an operating lim'it based on LOCA considerations and an RPS limit that protects against centerline fuel melt.
Similarly, two DNP related margins are also determined - an operating limit based on LOFA considerations crd an RPS limit.
In the core surveillance, precalculated factors based on Max the maneuvering analyses and the etailable marains are used to define an rg ax and F which are then ccmpared to meesured values to deternine whether the core is behaving as expected.
4 e equations used in the determination of the maratns, including the I
uncertainties, were reviewed and found to be acceptz ble.
The components of the margin equations used in the determination of linear heat rate ard DNB are justified, and the values of the uncertainties applied have been previously reviewed and approved by the NRC.
I Since only steady-state power distributions can be measured with reasonable accuracy, changes in the maroins to limits accompanying c'eviations from steaoy-state conditions must be determined on the basis of calculations.
The ax ax measured values of F and F are therefore compared to maximum g
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allowable values that account for the minimum margins determired in the I
maneuvering analysis to ensure that the limits on the reasured values will be met at the extremes of the AFD-power level operating limits.
If the measured ax ax values of F or F exceed their respective limits, then the AFD-power n
-level limits and the f(delta-1) function in the OPAT trip function are adjusted and/or the power level is reduced.
The trends in the margins to the limits are monitored from measurement-to-measurement, and the measurement frequency is increased or an additional penalty is incluced in the margins if f
increased peaking is expected. Monitoring in the case of base load operation
- s similar.
This monitoring philosophy is similar to that currently employed in connection with RA00.
The factors and uncertainties (and related methodologies) applied in the comparisons to measurements are justified, and
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the DPC rrethodology is acceptable.
'l 3.4-Evaluation Summary Based on the review of the topical report and the additional information provided, and recognizing that the NRC has reviewed and approved the computer codes and some components of the proposed methodology (e.g., the generation and use of DNB MATP curves), it is concluded that the DPC analysis represents an acceptable approach for determining and monitoring core operating ard RPS limits for the McGuire and CatcWba' Nuclear Stations.
The proposed methocology,
-however, should be confirmed by continued calculation-to-measurement comparisons, and monitoring of trends or any less of conservatism. While the L
application of the methodology to other four-loop, 193-assembly W PWRs is acceptable, the appropriate, plant specific reactor systems aspects must be considered and justified.
I l.
3.5 Restrictions
'he following restrictions are imposed on the use of the Nuclear Design Methodology described in DPC-NE-20ll:
(1) Application of this methodclogy is to be limited to the McGuire and i
Catawba nuclear power stations, I
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I (0) Applicatien to other Westinghouse 293-assembly plants would be acceptable I
provided that plant-specific differences be considered and justified, (3) Applict,tirm of this methooolocy is contingent upon NRC approval of the Pelead Design Thermal-Hydraulic Methodology DPC-NE-2004 (presently under j
NRC review) usina the VIPRE-01 code, and (4) Calculation of power and xenon distributions are limited to the use of
'i the EPRI-N0DE-P and the PDO-07 codes, j
4.0 REFEPENCES 1.
Letter from'H. D. Tucker Duke Power Company to USNRC, "f!uclear Design Methocology for Core Operating Limits of Vestinghouse Reactors," dated April 27,1988.
'I.
2.
Letter frem H. B. Tucker Duke Power Company to USNRC, " Nuclear Design-
]
Methodology for Core Operating Limits of Westinghouse Reactors - Response' to Request for Additional'Information," dated March 28, 1989.
I 3.
Letter from C. O. Thomas NRC, to H. B. Tucker Duke Power Company, dated March 13, 1985.
l 4
WCAP-10216-PA, " Relaxation of Constant Axial Offset Centrol, F(q)
Surveillance Technical Specification," June 1983.
j I
5.
Letter from H. B. Tucker (DPC) to NRC, " Nuclear Design Methodology for j
Core Operating Limits of Westinghouse Reactors - Response to Request for Additional Information," March 28, 1989.
6.
DPC-tlE-201CA, " Duke Power Company McGuire Nuclear Station, Catawba Nuclear Station, f:uclear Physics Methodology for Reload Design," i une l
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'1985.
7.
Letter from C. O. Thomas (NRC) to H. B. Tucker (DPC), March 13, 1985.
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8.
NFS-1001A', " Duke Power Company Oconee Nuclear Station, Reload Design o;
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i Methodology ^," April 1984
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9.
DPC-NE-2004, "Ouke Power Company, ficGuire and Catawba Nucle 6r Stations, I
Core Themal-Hydarulic Methodology Using VIPRE-01," January 1989.
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DUKE POWER COMPANY NUCLEAR' DESIGN METHODOLOGY for
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CORE OPERATING LIMITS of WESTINGHOUSE REACTORS i.
DPC-NE-2011A March 1990
.i Design Engineering Nuclear Design l
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Lt Table of. Contents Page 1.
Introduction 1
1.
Purpose 1
2.
Summary of the methods 1
t 31 3.
Applicability of this method 2
- g 4.
Definition of terms 3
2.
Generation of Power Distributions 5
1.
Descriptions of the models used 5-2.
Times in core life 5
3.
Generation of abnormal xenon distributions 6
l 4.
Generation of power distributions
'7 5.
Generation of' radial local factors 9
3.
Uncertainty Factors 16 1.
Power distribution 16 l
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_2.
Quadrant tilt 16
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Axial power distribution 17 i
j 4.
LCO and RPS Limits 18 1.
General methodology 18 i
2.
LOCA margin calculations 18 3.
LOFA DNB margin calculations 20 4.'
RPS DNB margin calculations 21 5.
Centerline fuel melt margin calculations 22 E
6.
. Determining the-AFD power level limits 23 lE-7.
Control rod insertion limits 25 l
1 5.
Base Load LCO Limits 30 6.
Power Distribution Surveillance
=31 1.
LOCA F -surveillance methodology 31 2.
LOFAOkBsurveillancemethodology 33 i
3.
Power distribution monitoring 34 7.
References 44 Appendix A,-Description of Computer Programs A-2 1
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m lY 1
List of Figures Page 1.
Flow of Data Through a Maneuvering Analysis 4
j 2.
Sample-Xenon Transients at Beginning of Life 12 AFD vs. Transient Time 3.
Sample Xenon Transients at Beginning of Life 13 Xenon Concentration vs. Transient Time i
4.
Sample Xenon Transients at Beginning of Life 14<
.l Xenon' Offset vs. Transient Time j'
5.
Sample Xeno'n Transients at Beginning of Life 15 j
g-Soluble Boron Concentration vs. Transient Time 6.
Typical LOCA Linear Heat Rate Limits vs. Core Height 27
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Sample LOCA Margin Plot 28
'l 8.
Sample LOFA DNB Margin Plot-29 9.
K(Z)'- Normalized F (Z) as a Function of Core Height 40 9
10.
Sample LOFA DNB MATP Curves for 100% Power 41
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- 11. - Sample AFD - Power Level Operating Space 42 l
12.
Sample Control Rod Insertion Limits vs. Thermal Power 43 j
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List of Tables
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Typical Reactor Conditions-During Xenon Transients 10 2.
Typical Power. Levels and Control Rod Positions for 11 Generating Power Distributions 3.
Typical RPS MATP Curve Conditions 26 I
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1.
Introduction 1.1. Purpose This report describes the methodology for performing a maneuvering-analysis for four-loop,193 fuel assembly Westinghouse reactors, such as McGuire and Catawba Nuclear Stations.
Duke. Power Company g
'W has developed this methodology as.an alternative to the existing Relaxed Axi.al Offset Control (RAOC) Methodology (1).
This maneuver-ing analysis results in several advantages:
more flexible and prompt engineering support for the operating stations,-consistency with the methods of Duke Power Company's nuclear design process, and potential increases in available margin through the use of three-dimensional monitoring techniques.
The increase in margin' occurs in limits on power distribution, control rod insertion, and power distribution inputs to the overpower AT (OPAT) and overtemperature AT (OTAT) roactor protection system trip functions.
Specifically, these limits are the axial flux difference (AFD) -
power level operating space, the rod insertion limits and the f(AI) function of either the OPAT or the OTAT trip functions of the RPS.
These limits are currently monitored via Technical Specifications (2,3).
The need for revisions to the Technical Specifications will be evaluated prior to implementing this methodology.
~1.2.
Summary of.the methods The operating limits define the AFD power. level space and rod insertion limits which provide assurance that the peak local power in y
the core is not greater than that assumed in the analysis of design basis accidents or transients (loss of coolant accident or loss of 4
flow accident).
Operating the reactor within the allowed AFD -
power level window and rod insertion limits satisfies the power peaking assumptions of the LOCA and LOFA analyses.
Page 1
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The RPS limits, among other functions, provide protection against l
fuel failure due to' fuel melting (CFM) or DNB during anticipated transients.
The relevant limits are set such that the RPS will trip the-reactor before fuel damage occurs.
.The maneuvering analysis uses a three dimensional nodal reactor model to calculate a set of power distributions at several points'.in coro life. These power distributions are based on a set of abnormal xenon-distributions to insure predict'ed power distributions are conservative with respect to those expected to occur.
The three dimensional power distribution is augmented by pin to assembly factors for the maximum pin power in each assembly and by appropriate uncertainty factors.
These pin to assembly factors are derived from a two dimensional fine mesh (pin by pin) model of the core.
The augmented power distribu-tions are then evaluated against the-various. thermal limits.
The operating limits and the f( AI) function of either the OPAT or the OTAT RPS trip functions are then set to exclude the power distribu-tions that exceed the respective thermal limits.
Figure I shows a I.
flow chart of the data as it goes through a maneuvering analysis.
1.3. Applicability of the method The maneuvering analysis presented in this report applies to Westinghouse four loop, 193 assembly reactors.
This method is intended to be used to set or validate the AFD power operating 1
limits, the control rod insertion limits and the RpS trip limits.
A system of computer-programs is used to implement this method.
A description of-the computer programs currently in use is in Appendix L
A.
Most of the programs have already been described in DPC's topical report on nuclear design-(4) and have been approved by the NRC (5).
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'1.4. Definition of terms
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.AFD Axial Flux Dif ference is the percent power in the top of the core minus the. percent-power in the bottom of the core.
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Radial Local Factors A Radial Local Factor is the peak rod power in an assembly divided by
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the average rod power in the same assembly.
QF is the local heat flux on a fuel rod surface divided by the core q
average fuel rod heat flux.
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,AH F
is the integral of linear power along a particular fuel rod divided AH by the average' integral of all of the fuel' rods.
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Generation 1of Power ~'
74butions 2.1. Description of the models used.
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The three dimensional nodal power and xenon distributions are gen-erated by a DPC version of EPRI-NODE-P (N0DE).
N0DE has an explicit xenon and iodine model that~. allows power and time dependent xenon transients..N0DE has a closed-channel thermal hydraulic feedback model to generate fuel and moderator temperature distributions that:
are used in the neutronics model.
The neutronics model accounts for fuel and moderator temperature, coolant flow, soluble boron concen-tration, lumped burnable absorbers, control rods, fuel burnup, and xenon and iodine distributions.
The N0DE model was approved by the.
NRC for use in reload design in reference 5.
The radial local factors are extracted from a quarter core, one pin per mesh PDQ07 model of the core.
PDQ calculations are run in two dimensions (X-Y) with a two dimensional thermal hydraulic feedback model.
The PDQ model was approved for use in reload core design in i5_
reference 5.
2.2. Times in core life The maneuvering analysis is typically performed at three times in
(
core 8ife:
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2.3. Generation of abnormal xenon distributions f.
The abnormal xenon distributions are generated with a set of limiting.
xenon transients at each point in core life that is to be analyzed.-
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} Table 1 shows the initial'and transient condi-tions of the reactor for each of the transients. [
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To add to the conservatism, these transients are modeled conserva-tively in several respects:
).Becauseofthesefactors,thexenontrans-ients in the reactor model will be more severe than could be reasonably expected to occur.
Each of the-xenon transients start with xenon in equilibrium with'the core at the initial conditions.
The initial conditions are different for each transient.1 Page 6
---_--__-__-_.-__________.m.____m_
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The control rod positions for the xenon transients were chosen to be
- W-at or near the expected rod insertion limits.
The final. control rod:
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' insertion limits may be different from the positions used in the xenon transients and the analysis will still.be valid.
This is because the xenon transients are so severe that the maneuvering
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analysis results are not sensitive to the control rod motions t. hat drive the xenon transients.
The xenon transients proceed until.
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Depending on the transient pcwer level, this usually takes aboutl,.j Lu hours.
Figures 2 through 5 show graphs of AFD, xenon offset, xenon concentration, and soluble boron concentration plotted against time for a typical. set of beginning of cycle xenon transients.
2.4. Generation -of power distributions Using the abnormal xenon distributions' from the xenon transients, three dimensional power distributions are generated.so that.-the operating and the RPS limits can be determined.
As shown on. Table-2,powerdistributionsaregeneratedwith[
]Theoperatinglimits-are pre-conditions that would prevent exceeding the peakilocal power in the core assumed in the loss of coolant accident (LOCA) analysis r
i or the loss of flow accident (LOFA, or a primary coolant pump. trip) analysis.
Because this is the normal operating mode of the reactor, control rod motion will be constrained by the power dependent. rod insertion limits.
i Jg, Power distributions for the operating limits are Page 7 l
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generated with these abnormalixer.on distributions with the reactor at nominal conditions.
I The, RPS ' limits protect the fuel against damage from DNB or fuel i
melting even if the reactor should go through any one of several l.
1 anticipated transients:
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The abnormal xenon distributions from the xenon transients are chosen
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I J
Table 2 shows the reactor conditions and range of control rod posi-tions.
Criticality in the reactor model is maintained by instan-l taneous changes in soluble boron concentrations.
~
2.5.' Generation of radial local factors-I-
The' radial local factor is the ratio of the maximum ' rod power in~ an assembly to the average rod power of the assembly.
Radial local factors are assembly and burnup dependent.
They are extracted from a 4
core specific fine mesh P0Q model that:has been depleted over the life of the. cycle.
The assembly average burnup, used as the indepen-dent variable to interpolate ~the radial local factors, is also extracted from the P0Q model.
The P0Q model has two neutron energy groups and one spatial mesh point per fuel-pin, Cross sections;are.
taken from the EPRI-CELL (6) system and the CASMO (7)' system.'
The P0Q I.
model is described more fully in reference 4.
1 1
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l Page 9
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Table 1 Typical Reactor Conditions During Xenon Transients l
Initial Conditions Transient Conditions Transient Name
% Power Control Rods
Power Control Rods I
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Page 10
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Table 2 Typical Power Levels and Control Rod Bank Positions for Generating Power Distributions 4
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m figure 2
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I Sanple Xenon Transients at Beginning of. Life i
AFD vs Transient Time I
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Figure 3 i
LI Sample Xenon Transients at Beginning of Lift Xenon Concentration vs Transient Time
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I Sample Xenon Transients at Beginning of Life j
Xenon Offset vs Transient Time I
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i Tigure 5 Sample Xenon Transients at Beginning of Life I
Soluble Boron Concentration vs Transient Time I
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Page 15 I
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3.
Uncertainty Factors 3.1. Power distribution The NRC SER (5) of the McGuire and Catawba design methodology report-(4) approved a total peak uncertainty factor of 1.073.
This includes an engineering f actor of 0.03 for Westinghouse fuel, a local peaking uncertainty of 0.02, and assembly axial peaking of 0.035.
The engineering f actor is specific to the fuel vendor and may be t
e changed for fuel from different vendors.
This uncertainty factor will be applied to the LOCA margin calculation and the centerline fuel melt margin calculation as UCT in sections 4.2. and 4.5.
s The NRC SER also approved an uncertainty factor for fuel rod radial I
peaking of 1.047.
However, this value also includes the manufac-
[
turing tolerance of 0.03.
Since the DNB calculations (see section 4)
?
already account for this f actor it is not necessary to include it again.
Using the same values as the SER (radial assembly peaking of 0.03 and radial local peaking of 0.02), the uncertainty factor is:
I 1 + V(.03' +.023) = 1.036 This uncertainty factor will be applied to the radial pin power in the DNB margin calculations as UCR in sections 4.a and 4.4.
- i..
3.2. Quadrant tilt The excore detector system is used to monitor gross changes in the core power distribution.
The primary purpose of the excore detectors with respect to quadrant power tilts is to detect changes in tilt from the previous calibration.
Since the'iechnical Specifications
- I (2, 3) allow reactor operations with excore quadrant power tilts up to 2%, the relationship between excore quadrant power tilt and a penalty to apply to the thermal limits calculations had to be deter-mined.
I Page 16 1
m I
This relationship was determined by evaluating various tilt causing mechanisms for several reactor cores.
This analysis was performed with full core NODE models.
Theresultsshowedthata(
power peaking penalty is required to account for the allowed 2% excore quadrant power tilt.
This penalty will be applied as TILT to the I
LOCA, DNB and centerline fuel melt margin calculations in Section 4.
3.3 Axial power distribution The DNB calculations use an assembly normalized axial peak, which may be defined as F divided by F for an assembly.
The McGuire q
3g and Catawba design methodology report (4) does not provide an uncertainty value for this calculated parameter.
An Observed Nuclear Reliability Factor (ONRF) for the calculated assembly normalized peak can be computed from the McGuire and Sequoyah data that was used in Reference 4.
Using the.same methodology as Reference 4:
- b+
5 (6)
ONRF =
M E
where:
fl = mean value of (F /F9 3g) meas. = 1.251.
D = mean value of [(F /Fq 3g) meas. - (Fq AH) calc.) = 0.032.
-F S(U) = standard deviation of D = 0.015.
K = 95/95 one sided tolerance factor for 846 points = 1.7343.
846 points were selected from the full set of 1038 points by the criterion that both (F3g) meas and (FaH) calc. must be greater than unity.
The ONRF is 1.048.
This will be applied to the DNB calculation as UCA in Section 4.3.
I Page 17
m 4.
LCO and RPS Limits 4.1. General methodology i
i The power distributions are divided into two categories for the i
I thermal limits calculations.
The operating limits use power distri-butions that were calculated with nominal inlet temperature, with control rods inserted to the power dependent insertion limits and with power less than or equal to 100% power.
Control rod positions will bound insertion limits in order to set the insertion limits.
The RPS limits use power distributions with the power level up to and including 118% power, no administrative restriction on the control rod insertions and either nominal or low inlet temperature.
I The margin to the various limits is calculated in the following fashion:
MARGIN % = (ALLOWED PEAK - CALCULATED PEAKP100 / ALLOWED PEAK I
The calculated peak-is a synthesis of the three dimensional nodal power distribution and the radial local factors from the fine mesh two dimensional PDQ calculations.
Depending on the limit type, this equation may be in terms of a peaking factor or a linear heat rate.
Either the calculated peak or the allowed peak contain sufficient factors to account for the various uncertainties and tolerances.
AFD and control rod insertion limits for each limit type are set to exclude all power distributions with negative margins of the same limit type.
4.2. LOCA margin calculations Since the LOCA limits are used to define the operating limits of the core, the operating limits power distributions, as described in section 2.4, are used in this calculation.
The LOCA margin is calculated for each node in the core, but only the most limiting value is used in the determination of the AFD power level limits.
The equations below show how the LOCA margin, LOCAM, is calculated.
pa,e 1e c
5;
'I g
W LOCAM = Min ((LOCAMX(z) - LHR(x,y,z))
- 100 / LOCAMX(z))
Whers:
i LOCAMX(z)
= Axially dependent maximum allowable linear i
heat rite in kw/ft.
LHR(x,y,z)
= NP(x,y,z)
- AVGLHR
- RADLOC(x,y.e)
- TILT
- AMF I
NP(x,y,z)
= Nodal power from the power distribution calculation.
= Fraction of core power level, including 2%
uncertainty.
AVGLHR
= Total core power divided by the total length of fuel rods in the core, kw/ft, I
accounting for fuel densification and thermal expansion.
+
RADLOC(x,y,e) = Burnup (e) dependent maximum roc to assembly power factor.
s UCT
= Uncertainty factor on the calculated total peak, including manufacturing tolerances (see section 3.1).
TILT
= Factor to account for a peaking increase due to an allowed quadrant tilt (see section 3.2).
l RPF
= Factor to account for the power deposited i
in the fuel rod.
l-L AMF
= Additional: Margin Factor, optionally used to incorporate additional design margin.
Page 19 I
N r
The values for LOCAMX(z) are derived from the Technical Specification limits on F.
Typical limiting values are shown in Figure 6.
q i
The 2% uncertainty on power level and the factor to account for power deposited in the fuel will be used only if these factors were not accounted for in the limits on F.
q 4.3. LOFA DNB margin calculations The LOFA DNB limits are also used to define the operating limits, so the operating limits power distributions, as described in section 2.4, are used in this calculation.
The OtlB margin calculation is based on a set of Maximum Allowed Total Ptak (MATP) curves that are calculated in accordance with reference 14.
The MATP curves are I
determined for several power levels (e.g., 100, 75 and 50% power).
The input power distributions are selected to match the power level of each set of MATP curves.
Sample MATP curves for LOFA DNB are shown in Figure 10.
The DNB margin is computed for each assembly in /
the core, but only the minimum margin for (ach power distribution is used in the determination of the AFD power limits.
DNB margin, DNBM, is calculated as:
MARP(x,y) - RPP(x,y)
DNBM = Min {
} *100 MARP('x,y)
Where:
MARP(x,y)
= MATP(z,AP(x,y))/(AP(x,y)*UCA)
AP(x,y)
= Axial peak in an assembly, on an assembly normalized basis.
UCA
= Axial uncertainty factor (see Section 3.3).
MATP(z)
= Maximum allowed total peak, at the axial plane of the axial peak.
t Page 20
1 I
RNP(x,y)
- RADLOC(x,y,e)
- TILT
- UCR RPP(x,y)
=
i RNP(x,y)
= Normalized assembly power from the power distribution calculation.
RADLOC(x,y,e) = Burnup (e) dependent maximum rod to assembly power factor.
UCR
= Radial rod power uncertainty factor (see section 3.1).
I AMF
= Additional margin Factor, optionally used to incorporate additional design margin.
TILT
= Factor to account for a peaking increase due I
to an allowable quandrant tilt (see Section 3.2).
Theaxialuncertaintyfactorwillybe'includedonlyifithasnot been accounted for in the MATP curves.
I 4.4. RPS ONB margin calculations The rest of the DNB margin calculations are for the RPS limits, so the operating limits restrictions on power distributions are not applied.
The methodology for computing RPS ONB margin is the same as in section 4.3, however the FATP curves are different.
Table'3 lists the conditions at which the RPS MATP curves were generated and the 3
conditions of the power distributions that will be used for each set i.
of MATP curves.
I Page 21
l 4.5. Centerline fuel melt margin calculations Centerline fuel melt limit is also an RPS limit, so the operating I
limits restrictions on power distributions are not applied in the calculation.
Since there usually is a positive margin, for center-line fuel melting, only the power distributions at 118% power are i
used for the fuel melt margin calculations.
A por,itive margin at 118% power will preclude negative margins at lower power levels.
If the 118% power level results show negative margins, lower power-levels will be analyzed to fully define the AFD - power level limit.
The equations below show how the margin for centerline fuel telt is calculated.
Note that the linear heat rate is calculated similarly to the LOCA margin calculation.
Each node in the core model is analyzed, but only the minimum margin for a power distribution is I
used to determine the AFD power limits.
I' MAXLHR - LHR(x,y,z)
CFMM = Min {
} 100 MAXLHR I
Where:
MAXLHR
= Maximum allowable linear heat rate in kw/ft.
LHR(x,y,z)
= NP(x,y,z)
- AVGLHR
- RADLOC(x,y,e)
- TILT
- SC
- AMF NP(x,y,z)
= Nodal power-from the power distribution calculation, FP
= Fraction of core power level, including 2%-
p l
\\
uncertainty, l-J AVGLHR
= Total core power divided by the total length i.
of fuel rods in the core, kw/f t, accounting for fuel densification and thermal expansion.
i Page 22
v t, s
i RADLOC(x,y,e) = Burnup(e) dependent maximum rod to assembly power factor.
UCT
= Uncertainty factor on calculated power distribution, including manufacturing tolerances (see section 3.1),
TILT
= Factor to account for a peaking increase due to an allowable quadrant tilt (see section 3.2).
= Factor to account for the pcwer generated in the fuel rod.
SC
= Statistical combin6 tion of UCT and RBOW.
From reference 5, the values are combined as :
(1 +
5+/(UCT-1-h5) + (RB0W-1)')
RBOW
= Peaking incre due to rM and assembly bow.
= Additional Margin Factor, opt M ally used to incorporate additional design i.eatn.
The 2% uncertainty on power level and the factor to account for power deposited in the fuel will be used only if these factors were not accounted for in the limiting heat generation rate.
The factors RBOW and UCT are statistically combined because they account for independent effects.
4.6. Determining the AFD power level limits L.
h The individual values of margin for each power distribution and margin calculation are collected into a database.
For each power level and margin calculation, the margin data is plotted against AFD.
I The data points are connected by drawing lines between points with an equal independent parameter.
Control rod position is usually 1
' ~
Page 23 2
m I
chosen as this independent parameter, which means that different points along these lines represent different xenon time steps.
The limit is taken from the most conservative zero margin intercept
+
of these lines.
Figures 7 and 8 shows an example plot of LOCA and LOFA DNB margin plotted against AFD, connected by equal rod position lines.
i The operating AFD limits are determined by selecting the limiting of either the LOCA margin results or the LOFA DNB margin results at the various power levels analyzed.
The AFD limits may be interpolated between rod position if the rod position chosen for the rod insertion limit was not explicitly modeled when the power distr'ibutions were generated.
The bounding AFD envelope is adjusted to account for measurement system (two segment power-range excore nuclear detectors) uncertainties.
The uncertainties account for the excore detector calibration error and drif t between calibrations.
The DNB margin calculations performed for the RPS OTAT AFD Trip penalty, f(al), provide AFD limits',
I
}Thepower-AFD penalty is determined by selecting the limiting breakpoints and slopesdefinedbythel
)Theuncer-tainty associated with the f(AI) function is combined with the uncertainties of the other OTAT function input parameters in deter-constant in the setpoint equation (reference f
mining the adjusted K3 2, 3), or the f(AI) function is adjusted to account for the AFD uncertainties.
The centerline fuel melt protection criterion is associated with the OPAT Trip f(AI) penalty function.
Since the OPAT f(al) function is I
usually zero, the check performed at 118% power is adequate to verify that the penalty is not required.
Should the centerline fuel melt margin calculations result in an AFD limit at 118% power, lower power levels would be analyzed in order to define the power - AFD penalty.
Page 24
N
,,I The penalty could then be incorporated into the OPAT trip function or the required protection could be provided by the OTAT function.
I 4.7. Control rod insertion limits The rod insertion limits are assumed when the operational AFD power level limits are set.
However, further iteration on the limits may be necessary depending on the rtssults of the shutdown margin and ejected rod analyses, which are performed as described in reference 4.
Adjustments are made to the rod insertion limits and AFD power l
1evel limits as necessary.
I I
t I'
I I
I Page 25
... -...-..... ~ ~
. -. -. -. -.. ~ - -.. - -..
. -. - - -. -. ~ - -.. -
l Table 3 Typical RPS MATP Curve Conditions and Conditions of the Pcwer Distributions used for each set of MATP Curves j
4 PMID 4
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t il
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Page 26
- -,,... ~ _..... -,,
I risure 6 Typical LOCA Linear Heat Rate Limits vs Core Heil,ht
=
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1
'g Page 27
-W
W Figure 7 Sample LOCA Margin Plot I
J i=
I
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./
i
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Page 28
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Figure 8 Sample LOFA Margin Plot I
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=
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NIOUVH %
Page 29
N 5.
Base Load LCO Limits If the operational limits for a particular fuel cycle are too restrictive I
for normal operation, then a set of base load limits can be defined that i
may allow power operation at 100% power.
Base load is defined as operat-g E
ing the reactor within a relatively narrow AFD band about a plant measured AFD target and within a limited power ranga.
By limiting the allowed AFD power level space, extra margin tan be gained in the power distribution monitoring factors (see section 6).
I Base load limits and monitoring f actors are computed the same as the operational limits, only the xenon transients will be re-defined so that they will be restricted to the base load operating band about a predicted AFD target.
The power level at which the plant will be allowed to enter base load will be greater than or equal to the power level of the xenon transients.
e I
I I
I I
I Page 30
i 6.
Power Distribution Surveillance 6.1. LOCA F surveillance methedology
)
g The AFD - power level limits are set to protect the fuel from damage during either a LOCA or LOFA when the power distribution i': skewed in the axial direction.
However, only steady siete power distributions l
I can be measured with reasonable accuracy.
Because of this, the limits on the measured F are reduced by pre-calculated factors that g
'I account for perturbations from steady state conditions to the operat-ing limits.
I The limit that must be met within the AFD - power level operating limits is currently specified in references 2 and 3 as:
?
'I g 5 2.32
- K(z) / P for P > 0.5 F
F 5 2.32
- K(z) / 0.5 for P 5 0.5 g
Where P is the relative thermal power and K(z) is shown in Figure 9.
Both K(z) and the value of 2.32 may change with future LOCA analyses.
Using definitions from section 4.2, the reduced limits for the measured F are specified as:
g M
F ( x, y, z )
- UMT
- TI LT 5 Where:
I MF (x,y,z) = The measured total peak in location x,y,z.
g UMT
= The measurement uncertainty on the total peak, taken as 1.05 in the current Technical Specifications (2,3).
= Manufacturing tolerance, taken as 1.03 in the current Technical Specifications (2,3).
TILT
= Factor to account for a peaking increase due to an allowable quandrant tilt (see Section 3.2).
f I.
Page 31
=
. -... ~.. -. -. - - - -. ~ -.... - -.... ~. -...... - -..
I I
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1 I.
I i
l l
I lI I
I l I I
I Page 32 P
-v
...-.mv
-w e - -
r,,,.-.-ty,,-.
,,m-wr,--.-
+..
--,w--wng
I 6.2. LOFA DNB F surveillance methodology AH The LOFA DNB F surveillance uses monitoring factors that are 3g similar to those used for the LOCA F surveillance. The power g
distribution surveillance limit on F that must apply anywhere 3g within the AFD - power level operating limits is given as Maximum A11ered Total Peak (MATP) curves.
Sample curves are given in I-Figure 10.
The limits for F must be reduced for the same reason as the F 3g n
limits are reduced (see Section 6,1).
Using definitions from Section 4.3, the reduced limit for monitoring F is given in the aH l-following relationship:
FM (x,y)
- UMR
- TILT s FUH *'#)
- HAH (*'Y) 5 aH Where:
F g(x,y)
= Measured value of FAH' UMR
= Uncertainty value for measured radial peaks, taken as 1.04 in the current Technical Specifications (2, 3).
TILT
= Factor to account for a peaking increase due to an
.g allowable quadrant tilt (see Section 3.2).
5 I
LI I
I
[ I-Page 33 1
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I I
I I
g I;
6.3 Monitoring of plant measured parameters During power operations, the power distribution is continuously monitored by the ex-core nuclear instrumentation.
The parameters of interest to power distribution' monitoring are the core power level, the AFD and the quadrant power tilt.
Limitations are imposed on these three. parameters by the maneuvering analysis.
The maneuvering analysis also imposes limits on control rod positions during power operations.
Tha power distribution is also measured periodically by Page 34
the in-core instrumentation system.
The results of these measurements are used to verify that the core is behaving as predicted by the maneuvering analysis or to adjust the AFD - power level limits if it I
is not.
The surveillance of these parameters is described below.
AFD - Dover Level Limits During normal operations, the combination of AFD and power level must be maintained within the operating limits that are provided by the maneuvering analysis.
Exsmple AFD - power level limits are shown in Figure 11.
Since the operating limits are a Limiting Condition of Operation (instead of a Limiting Safety System Setting),
I the plant would be allowed to operate outside of the operating AFD -
power level limits for short periods of time if necessary.
This I
allowance is meant to be used to increase the plant availability during transient situations and is not meant to be used for normal operation.
If the power distribution is unusually limiting (because of severe power peaking, for example), then base load operation may be used if it was provided for by the maneuvering analysis.
During base load operation, the measured AFD must be within a relatively small AFD band about a plant measured target AFD.
The size of the AFD band is specified by the maneuvering analysis.
Note that this target may or may not be within the AFD power level operating limits.
Base load may not be entered unless the plant has been relatively stable in AFD ano power level for a pericd of time.
The power level must be L
above the Allowed Power Level (APL - a value supplied by the maneuver-ing analysis) and the AFD must be within the AFD power level opera-ting limits.
The power level may then be increased to a maximum of I
100% rated thermal power or thE Maximum Base Load Power (MBLP - a value described below).
Control rod insertion limits I
The control rods must be maintained within the insertion limits that were determined by the maneuvering analysis.
Example limits are Page 35
I shown in Figure 12.
These limits are a Limiting Condition of Opera-tion, so operation outside of these limits is allowed for short periods of time.
Heat flux hot channel factor - F (x,y,z)
The in-core instrumentation system is used periodically to me6s'ure M
g (x,y,z), which must always be within the limits specified by the F
LOCA analysis.
These limits are currently specified in references 2 and 3 as:
(x,y,z) < (2.32) (K(z)) for P > 0.5 F
i e
M l
Fq (x,y,z) < (4.64) (K(z)) for P 10.5 I
Thermal Power Where:
P=
, and Rated Thermal Power
- (') * '"* '"" "
'"*d '" * '" ' ' "
- 9'"*"
.RW core height (z).
M This limit on Fq (x,y,z) is a Limiting Condition of Operation, so operation outside of the limit is allowed for a short period of time to allow the operator to bring the reactor back within the limits without a reactor trip.
l
\\
i I.
M To q (x,y,z) is usually measured at or near nominal conditions.
F M
g~
g ensure that the limit on Fq (x,y,z) is met at the extremes of the AFD power level operating limits, the maneuvering analysis imposes the following limits on the measurements at nominal conditions:
Page 36 I
l u
M M
Max F
(x,y,z) < F (x,y,z) for nominal operation, or q
g M
Max, BL F
(x,y,z) for base load operation, q (x,y,z) < Fg Max Max, BL where F (x,y 2) and F (x,y,z) are supplied by the maneuver-q q
ing analysis.
These limits are not imposed on the top or bottom 15%
M of the core.
The limits on Fq (x,y,z) account for an appropriate mea-surement uncertainty, taken as 5% in the current Technical Specifica-M tions (2 and 3).
If the measured values of Fq (x,y,z) exceed these g
limits, then for normal operation the AFD power level limits must be adjusted by reducing the allowed AFD span (move the negative and positive AFD limits closer to the zero AFD point), so that positive margin would be maintained at the extremes of the AFD - power level operating limits.
A similar adjustment is made to the f(AI) function of the GPAT RPS trip function.
For base load operation, reactor power Max, BL must be reduced until the above limit on F (x,y,z) is satisfied.
q For base load operation, reactor thermal power may not exceed the Maximum Base Load Power (MBLP), which is defined as:
i (x,y,z)
- 10 3 MBLP = Min ver M
(x,y,z)
(x,y,z) pQ M
Note that this is equivalent to saying that Fq (x,y,z) may not exceed Max, BL F
(x,y,z) for base load operation.
q Page 37
'l 1
Nuclear enthalpy rise hot channel f actor FaH (x,y) k iI M
M' F3g (x,y) is measured at the same time that Fq (x,y,z) is measured with 3
M
_g the in-core instrumentation system.
FaH (x,y) must be within the Maxi-mum Allowed Tota'i Peak curves that were used in the maneuvering analysis (see Figure 10 for example curves).
This limit is a Limiting Condition of Operation, so operation outside of this limit is permitted for a period of time to allow the operator to bring the reactor back within the limit' without a reactor trip.
b M
FaH (x,y) is usually measured at or near nominci conditions.
To-M ensure that the limit on FAH (x,y) is met at the extremes of the AFD -
~
l power level operating limits, the maneuvering analysis imposes the following limits on the measured values at nominal conditions:
1 M
Max (x,y) f r n minal operation, or FaH (x,y) < FAH 3_
M Max, BL (x,y) for base -load operation.
3; FAH-(x,y) < F3g If the appropriate relationship is not satisfied, then the reactor-M power will be reduced until it is satisfied.
The limits on FaH (*'Y) account for an appropriate measurement uncertainty, taken as 4%'in the current Technical Specifications (2 and 3).
I Quadrant power tilt An allowance for a 2% quadrant power tilt was made in the AFD power Max Max BL (x,y,z),F (x,y,z),
level operating limits and in the values of F9 q
page 38
i
- h
' Max Max,.BL
'F (x,y) and F (x,y).
Thus, no act.f on is required for 'an 3g AH indicated quadrant power tilt of up to 2%.
A quadrant power tilt larger-than-2% is a t.imiting Condition of Operation, so operation.of the plant-is-
[-
allowed to continue for a period of time while the operator attempts to '
correct the coraition.
)
i
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,3 II Figure 9 K(Z) - Normalized F (z).as a Function of Core Height-'
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g Figure l'1 Sample AFD-Power Level Operating Space t
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1 Page 42
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l Figure 12 Control _ Rod Insertion Limits vs Thermal Power
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i Page 43
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7.
References 1.
" Relaxation of Constant Axial Offset Control, F(q) Surveillance Technical Specification", WCAP-10216-PA, June 1983.
2.
Technical-Specifications for McGuire Nuclear Station Units No.1 and 2, Docket Nos. 50-369/370.
3.
Technical Specifications, Catawba Nuclear Station,- Unit Nos.1 and 2, Docket Nos. 50-413 and 50-414.
4.
" Duke Power Company McGuire Nuclear Station, Catawba Nuclear Station, Nuclear Physics Methodology for Reload Design",
DPC-NF-2010A, June 1985.
5.
Letter from Cecil 0. Thomas, Chief Standardization and Special Projects Branch to H. B. Tucker, Vice President Nuclear Production, DPC, March 13, 1985.
I 6.
Cobb, W.
R., Eich, W. J.. Tivel, D.
E., "EPRI-CELL Code Description," EPRI-ARMP System Documentation, CCM-3, Part II, Chapter 5, October 1978.
t 7.
Studsvik Energitechnik AB,'"CASMO-2,.A' Fuel Assembly Burnup~ Program,"
1 Studsvik/NR-83/3,-1981.
8.
Rotbleder, B. M., Fisher, J. R., "EPRI-ARMP System Documentation".,
CCM-3, Part II, Chapter 14, September 1977.
9.
Delp, D. L., et al,'" FLARE - A Three Dimensional Boiling Water Reactor Simulator", GEAP-4598, July 1964.
L~
L 10.
User's manual for PDQ-XD, CE Computer Services, September 1986.
l-11.
" Computer Code Certification for PDQEDIT," DPC internal document.
1 Page 44 I.
l, 12.
" Computer Code Certification for MARGINS," DPC internal document.
13.
" Computer Code Certification for MARGINPLOT," DPC internal
.g document.
W..
14.
" Mark-BW Maximum Allowable Peaking Limits," DPC internal document.
I g
l gz
- I I;
IL
- \\
I 1
{jy 7 Page 45 6g/'
in
' g M ',
~-. -. - - -. -.
sm 4
1 l
APPENDIX A 1
I,.
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l Page A-1 p
- le-
--r.,
5 r
Appendix A Computer Program Descriptions EPRI-NODE-P N00E'(8) is a three dimensional nodal program that is derived from FLARE (9).
N0DE computes a three dimensional power distribution with' thermal hydraulic feedback, the core multiplication factor, the fuel burnup distribution and maintains a reactivity inventory.
The physics models within N0DE account for the presence of control rods, fuel and moderator temperatures, fixed burnable.
poisons, soluble boron, fuel depletion, and time dependent xenon and iodine.
The input to NODE is generated either from CASMO-2E (7) data or from EPRI-CELL (6) color set PDQ data.
PDQ07 1
PDQ07 (10) is an industry accepted multi group, multi-dimensional, neutron diffusion depletion program.
The Combustion Engineering version of PDQ that is used by DPC has been modified with a two dimensional thermal' hydraulic s
feedback model to account for fuel and moderator temperature distributions.
PDQ uses cross sections from either CASM0-2E or EPRI-CELL.
PDQEDIT PDQEDIT (11) is a utility program that reads the PDQ system files.
The-program has several abilities, one of which is to produce radial local power l
factors from the mesh average power file.
[
MARGINS l.
MARGINS (12) is a program written.by DPC that computes the margin to thermal
. limits for LOCA FQ, DNB and centerline fuel melt.
MARGINS requires.the radial local f actors from PDQEDIT and the three dimensional nodal power distributions from N0DE for input.
The output of MARGINS is a file that contains one entry per power distribu' ' :.. the entry contains the case and limit type identifiers, the core axial offset and the core margin to the thermal limit evaluated.
I Page A-2
'MARGINPLOT-1 1
MARGINPLOT (13) is a program written by DPC.that plots the. MARGINS data and.
'1
, l' computes the zero margin. intercepts for the thermal limits data, i
i i
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Page A-3 5
.......... _ __ _. ~._
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NRC Questions i
and 3
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DPC Answers t
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[/ja,P Ef 0tj '\\' i UNITED STATES h'
NUCLEAR REGULATORY COMMISSION wAsmNGTON, D. C. 20666
/
.!5 4,,,,,/
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Mr. Hal B. Tucker, Vice President Nuclear Production
-Duke Power Company P. 0. Box 33189 Charlotte, NC 28242 5
t
Dear Mr. Tucker:
t
SUBJECT:
RE00EST FOR ADDITIONAL INFORMATION REGARDING THE NUCLEAR DESIGN i-METHODOLOGY FOR CORE OPERATING LIMITS OF WESTINGHOUSE REACTORS, l
TOPICAL REPORT DPC-NE-2011P l- '
The Reactor Systems Branch has reviewed the subject topical report and has 3
concluded.that additional information is required for us to complete this-review.
Please submit the responses to the questions in the enclosure within 45 ' days
-.E of the receipt of this letter to enable the staff =to complete its review.
If E'
you need any clarification, please contact Lambros Lois of iny staff at 301-492-0890.
Sincerely,
{
W M.-Wayne Hodges, Chief
- g Reactor Systems Branch l
5:
Division of Engineering & Systems Technolooy j
Enclosure:
As stated
- I I
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m 1
REQUEST FOR ADDITI0t:AL INFORMATION' DPC-NE-2011P u
1.
How do operating limits obtained via this methodology compare to limits
[
based on the use of the present RA00 nethodology?
~ 2.
Is the potential increase in the available margin associated with the subject methodology due solely to the use of three-dimensional analyses / monitoring, or do other aspects contribute?
3.
There is no indication in that the methodology employed in generating and using the LOFA DilB MATP curves has been-reviewed and accepted by the NRC.
4 The procedure for. generating power distributions appears to involve' running two xenon transients at each of three times in a cycle, followed by using xenon distributions from each transient / time-in-life to calculate instantar,eous power distributions associated with various combinations of power level, inlet temperature and control rod bank position, as well as those occurring during the course of several l
, anticipated transients.
It appears that only four xenon distributions from each transient at each
.g W..
time in life are used along with the statepoint configurations given in Table 2.
Please clarify / elaborate as to how many powir level / inlet temperature / control rod / xenon statepoints are evaluated 5t each time in life.
'5..
Is there demonstrated assurance that the power distributions resulting i
from the.above analyses are indeed conservative with respect to those that might occur, and that they sufficiently span the AFD/ rod insertion power level operating spaces to permit an accurate determination of operating limits?
1
.;)
I 6.
What is the basis for the 15 minute limit assumed in the analysis of the-boron oilution accident?
^
I:
L -
7.
Radial local factors appear to be obtained from a nominal all-rods-out depletion calculation for the cycle and are, therefore, only functions.of assembly type and burnup.
However, local peaking should also be affected by transient xenon, control presence, etc. What is the basis for not accounting for these effects?
I
-8.
What are the other components of UCT in addition to those specifically l
mentioned in 3.17 i
How are the axial peaking due to grid spacers and densification spike 9.
l' effects accounted for in the margin calculations?
1-10.
Please explain the basis for the use of SC in the CFMM calculation, and its form.
L
- 11. -Please explain why the uncertainties considered in the linear Nat rate-equation for the CFMM calculation are different from those used in 1
obtaining LOCAll given that they refer to the same basic quantity.
.g The definition of the TILT factor varies while its value appears to be 12.
L N
constant.
Please explain / elaborate.
~!
13.
Since the maneuvering analysis involves two xenon transients at three timesincorelifetherearesixFhandsixFAH design distributions D
available for comparisons to measurements.
Pave the errors introduced by the subsequent interpolation on cycle burnup and power level been quantified and included in the analysis? How are mismatches between the measured and design data associated with AFD and control rod position differences accounted for?
I; py 4%-
.i 7
3 14 Are P and M minimum values over the cycle?
0 g
15.
How are possible increases in peaking between measurements due to o g E'
mechanisms other than tilt (e.g. burnup) accounted for in the F and F g
6H surveillance?
E
- 16. What are the similarities /differneces between base loao operation and l.
CAOC7
)
b 17.
Under what conditions would the AFD target and operating band for base I
I load operation not fall within the normal AFD-power level operating l'g limits?
,g WhyaretheuncertaintiesassociatedwithFhandFfgin6.2different?
18.
I I
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8-v a
- m.
mm-.aa..'f a'mA% *
}
I Duuz POWER GOMPANY
(
P.O. BOX 33189 L
cMARLOTTE. N.C. 28949 RAL B. Tt*CKER teLzessone m........,
po43 oro.4aai :
.m u.....u no.
March 28, 1989 p
-?-r h'
U. S, Nuclear Regulatory Commission
.[
Attention: Document Control Desk Washington, D.' C.
20555 l
Subject:
McGuire Nuclear Station Docket Numbers 50-369 and -370
-3' Catawba Nuclear' Station l
Jgu Docket Numbers 50-413 and -414 Topical Report DPC-NE-2011P -
" Nuclear Design Methodology for Core 1
' Operating Limits of Westinghouse Reactors";
Response to Request for Additional Information j
E' Attached are responses to questions regarding the subject topical-report, which--
5 3-were transmitted by. letter dated March 3.'1989.
t
- 3-Please note that the proprietary nature of the original topical report, as i
'g' identified in my April 27, 1988 transmittal letter and accompanying' affidavit, is Therefore, they should'be maintained in the responses to these questions.
withheld from-public disclosure.
1 1.
O Very truly yours, fI A
H.- B. Tucker
.l 1
SAG 154/lcs xc
.Mr. Darl S._ Hood, Project Manager Mr. W. T. Orde'rs l' l, i
Office of Nuclear Reactor Regulation NRC Resident Inspector H : 5' U. S. Nuclear Regulatory Commission Catawba Nuclear Station 3_
l1 Washington, D. C.
20555 Mr. P. K. Van Doorn L 3:
ig' Dr. Kahtan Jabbour, Project Manager:.
NRC Resident Inspector _
.0ffice:of' Nuclear Reactor Regulation McGuire Nuclear Station' U. S.LNuclear Regulatory Commission.
Washington, D.:C.
20555
.Mr.~S. D. Ebneter, Regional Administrator i-l U. S. Nuclear Regulatory Commission,
- E Region II 101 Marietta Street, NW, Suite 2900 u
^'
""ta, e rgia 3 323 i.B3 lh gc
$ m2 4'
1 i
I'
' U..S. Nuclect Regulatory Commission'
--Page Two
. March 28.-1989
'bxe s. R. L. Gill Jr.
'P.- G. LeRoy I.
-J. S.:-Warren R. H. Clark-T. C. Geer l
G. D. Seeburger LR. O. Sharpe - MNS R.:M... Glover - CNS File:
MC, CN-801.01
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l Q1 How do operating limits obtained via this methodology compare to limits based on the use of the present RAOC methodology?
Al The operating space AFD limits from this method are enpacted to be a few percent vider than the current RAOC limits.
Q2 Is the potential increase in the available margin associated with the subject methodology due solely to the use of three-dimensional analyses / monitoring, or do other aspects contribute?
A2 The margin increase is due primarily to analysis of three-dimensional
,g power distributions, as opposed to the ID/2D synthesized power; g
distribution that the RAOC limits are based on.
Q3 There is no indication in that the methodology employed in generating and using the LOFA DNB MATP curves has been reviewed and accepted by the NRC.
I The general methodology for generating DNB MATP curves has previously A3 been approved by the NRC as applied to Oconee Nuclear Station in the SER for the topical report, " Duke Power Company, Oconee Nuclear Station.
Reload Design Methodology," NFS-1001A, April 1984. A topical report describing the codes and methods used by Duke Power for generating DNB MATP limits specifically for Westinghouse reactors was submitted to the NRC in January 1989 under the title, " Duke Power Company McGuire and
,I'-
VIPRE-01," DPC-NE-2004. When approved, this methodology will be used to Catawba Nuclear Stations, Core Thermal-Hydraulic Methodology Using generate DNB MATP curves for setting the core limits.
Q4 The prceedure for generating power distributions appears to involve reing two xenon transients at each of three times in a cycle, followed by using xenon distributions from each transient / time-in-life to calculate instantaneous power distributions associated with various combications: of power. level, inlet temperature and control rod bank'
.I position, as well as those occurring during the course of several anticipated transients.
It appears that only four zenon distributions from each transient at each.
time in life are used along with the statapoint configurations given in Table 2.
Please clarify / elaborate as to how many power level / inlet temperature / control rod / xenon statapoints are evaluated at each time in -
life.
A4 TN4 matrix of statspoints shown below will be used as an initial guide and may be modified as experience is accumulated.
A power d$stribution will be analyzed for each s'tatapoint in the matriz below for g
) That is,asetof(
Jhree-dimensionalpowerdistributionswill be analyzed to set limits.
I; li l
1 Q1 How do operating limits obtained via this methodology compare to limits based on the use of the present RAOC methodology?
A1 - The operating space AFD limits from this method are expected to be a few
~.
percent wider than-the current RAOC limits.
Q2 Is-the potential increase in the available margin associated with the subject methodology due solely to the use of three-dimensional analyses / monitoring, or do other aspects contribute?
A2 The margin increase is due primarily to analysis of three-dimensional j
power distributions, as opposed-to the 1D/2D synthesized power distribution that the RAOC limits are based on.
Q3 There is no indication in that the methodology employed in generating and using the LOFA DNB MATP curves has been reviewed and accepted by the NRC.
^
A3 The general methodology for generating DNB MATP curves has previously been approved by the NRC as applied to Oconee Nuclear St.ation in the SER
.l for the topical report, " Duke Power Company, Oconee Nuclear Station, Reload Design Methodology," NFS-1001A, April 1984.
A topical report describing the code e and methods used by Duka Power for generating DNB MATP limits specifically for Westinghouse reactors was= submitted to the 3
NRC in January 1989 under the title, "Duka Power Company, McGuire and
- g.
Catsuba Nuclear Stations, Core Thermal-Hydraulic Methodology Using VIPRE-01," DPC-NE-2004. ' When approved, this methodology will be used to generate DNB MATP curves for setting the core limits.
'I Q4 The procedure for generating power distributions appears to involve-1 running two xenon transients at each of three times in a cycle, followed by using xenon distributions from each transient / time-in-life to calculate instantaneous power distributions associated with various
. combinations of power level, inlet temperature and control rod bank position, as well as those occurring during the course of several anticipated transients.
L It appears that only four zenon distributions from each tranaient at each time in life are used along with the statopoint configurations given in Table 2.
Please clarify / elaborate as to how many power level / inlet temperature / control rod / xenon statapoints are evaluated at each time in life.
A4 The matrix of statapoints shown below will be used as an initial guide l '.
will be analyzed for each statapoint in the matrix below for(L and may be modified as experience is accumulated. A power d stribution
{
conditions frpm both tenon trJpsients at-three points in core life) That I-1s, a set of three-dimensional power distributions will be analyzed t'o set limits.
)
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.q i
List of State Points
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Q5:
Is there demonstrated assurance that'the power distributions resulting L
from the above analyses are indeed conservative with respect to those-l; that.might occur, and that they suf ficiently span the AFD/ rod. insertion
[
power level operating-spaces to permit an accurate determination of operating limits?
Yes.{
A5 L:
.i As 'shown in the response to question 4, the statepoint conditions will I
3 span' the allowable rod insertion limits.and the accident condition red L,g insertions as described in section 2.4 of-the report.. The-power
. distributions will generally span the AFD space, although some -
extrapolation on AFD may be required at times. Therefore, the AFD/ rod i
insertion space will be sufficiently analyzed to accurately determine the L
operating limits.
i L
cQ6 What is the basis for the 15 minute' limit assumed in the analysis of the l.-
boron dilution accident?
A6 The 15 minute limit is based on the operator action time acceptance I
P criteria of the Standard Review Plan, section 15.4.6-U.
m l'I S
C
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Q7 Radial local factors appear to be obtained from a nominal all-rods-cut depletion calculation for the cycle and are, therefore, only functions of assembly-type and burnup. However, local peaking should also be affected by transient zenon, control presence, etc. What is the basis for not accounting for these effects?
^E A7 Duke Power has examined the effects of control rods and transient, xenon 5
on local peakins factors using both two-dimensional and three-dimensional models.
In general, it has been observed that the limiting nodes in a specific _ case are located away from the inserted control rods. That is, e
the peak nodal power occurs in an unrodded plane and/or an assembly removed from the rodded assemblins by several assembly pitches.
Therefore, the intra-assembly flux distribution of the limiting node is relatively unaffected by the flux gradients induced locally near the rodded assembly. Similarly, the transient xenon distributions, while significantly skewed globally, do not cause significant changes in local-power distributions.
Q8 What are the other components of UCT in addition to those specifically I,
mentioned in 3.17 A8 UCT is defined in Reference 4 of the report to be
_I
,1 + (.031/1.375) + d(.03) 2 + (.035) 2 + (.02)2 = 1.073.
The term (.031/1.375) accounts for a small bias in the calculated power distributions.
I Q9 How are the axial peaking due to grik spacers and densification spike effects accounted for in the margin calculations?
- A9 In the development of the observed reliability factors the calculated peaks did not include any grid ef fects while the measured data did.
Therefore, the effects of the grid on peaking are inherently included in I.:
the observed reliability factors which are applied to the calculated '
values.
Current fuel designs used by Duke Power specify fuel pellet density greater than or equal to 95% of theoretical density.
Results of hot cell and genea scan measurements on fuel rods containing pellets of these densities have not shown any significant gap formation. Thus, no power Q10 Pla e e lai te asis f r the use of Ci he C ca e lation, and its form.
A10 A rod bow penalty is applied to the. calculated peak when computing CDO(.
However, since rod bow is considered to be independent of the calculational uncertainty, it is statistically combined with the engineering and power distribution factors in the equation for UCT found
.I l5 in Reference 4 of the report.. The algebraic derivation is shewn g
below:
UCT = 1 +.031/1.375 + M(.03)2 + (.035)2 + (.02)2 h
= 1 +.031/1.375 + h(.03) 2 + (.035)2 + (.02)2 + (RBOW-1)2 SC
~
=
h(.03) + ( 035)2 + (.02)2 = UCT.031/1.3 75.
t, SC = 1 +.035/1.375 + d (UCT - 1
.031/1.375 2 + (RB0W-1)2 U
Q11 Please explain why the uncertainties considered in the linear heat rate equation for the CDO( calculation are different from those used in L..
obtaining LOCAM given that they refer to the same basic. quantity.-
p t' -
All ' The only difference is that the. rod bow penalty is not applied to the L
LOCA limits, since any increase in peaking vill be compensated for by the
) ~..-
increased coolant flow.
r-3 L
-Q12 The definition of the TILT f actor varies while its value appears to be
'i constan::. Please explain / elaborate.
i A12 The magnitude of the tilt factor is the same in all sections and the correct definition in all sections is " peaking increase due to allowable l,
quadrant tilt."
Q13 Since the maneuvering analysis i lves two non transients at.three t
times in core life there are six and six design distributions
.H available for comparisons to measurements. Hive the errors introduced by 1-the subsequent interpolation on cycle burnup and power level been quantified:and. included in the analysis? How are mismatches between the-L measured and design data associated with AFD and control rod position differences accounted for?
U A13 The values of Pq.and F H fr a the design power distributions are not the I.
values that are compared to measurements.
l L
This is very similar to the current monitoring methods which apply.
burnup-dependent W(Z) transient peaking factors to the measured peaks.
~g
I I
J ne impact on peaking of differences between the measured and design data for AFD are inherently included in the uncertainty factors which are Li applied to the predicted peaks.
D e uncertainty factor used is an observed nuclear reliability factor developed.by matching reactor power and rod positions between predicted and measured st.atepoints, no.
l-calculated AFD was allowed to vary from the measure 4d value in these i
l calculations, although these dif forences are generally within 2%..ne impact of control rod position differences between measured and design data is considered negligible since power distribution maps are usually taken at nearly all-rods-ouc conditions.
Q14 Are M and M1H "i"i'"" **1"** *** Eh* "I'1*t f
g A14 1
!.g Q15 ' How are possible increases in peaking between mea surements due to I f mechanisms other than tilt (e.g., burnup) accounted for in the F and F q
g surveillance?
l A15 If Ff is greater than' T
, then the AFD power irrel space is reduced by
+
j.-
anappropriateamountsuchthatFf,atthenewAFDlimit,willbewithin the LOCA limits.
If F is greater than T then power level will'be reduced until the limitismet.
~
'If the margins to the limits are found to be decreasing.over successive measurements, than either the measurement frequen:y will be increased or the margine will be reevaluated with an additional penalty to account for the expected peaking increase to the next measuretnant.
Q16 What are the similarities / differences between base load operation and
'CAOC7 A16 no only significant difference is the power level at which the mode of I
operation may be entered.
Base load operation is typically entered at 80% power af ter stabilizing the plant at the target AFD. CAOC is used for the full range of power operation.
I r
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l' I
Q17 Under what conditions would the AFD target and operating band for base load operation not fall within the normal AFD-power level operating limits?
A17 This condition is not expected to occur since the AFD-power level limits will be set each cycle with a cycle specific three-dimensional core I
model. However, operating for a significant period of time at reduced power may cause the AFD target to be outside of the operating AFD space.
if this condition should occur, the surveillance of the measured peaking -
will ensure that the allowable limits are not ascoeded and tighter AFD limits would be used to minimise potential transient peaking.
D Q18 Why are the uncertainties associated with P and F in 6.2 difforent?
aH 6H A typograph2 cal error was made in the equation for Ff. The -equation A18 I
~H that was intended ist I
However, in further research it was discovered that a cod bow penalty does not need t., be applied to a limit that is related to DNB. This approach has previously been approved by the NRC in the SER to "Duka -
Power Company Oconee, Nuclear Station Reload Design Method 1ogy II,y J
DPC-NE-1002A, October 1985. Thus, the uncerta19 ties in r'H and F3g should be the same. The correct equation for. F 188 aH
=
Also,. the rod bow penalty should be removed from the calculation of DNBM.
I In section 4.3 of the report, the equation for RPP(x,y) should bet -
l I
i