ML20086G598
| ML20086G598 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 10/31/1991 |
| From: | INTERNATIONAL TECHNICAL SERVICES, INC. |
| To: | NRC |
| Shared Package | |
| ML20086G581 | List: |
| References | |
| ITS-NRC-91-1, NUDOCS 9112050172 | |
| Download: ML20086G598 (19) | |
Text
.. _ _ _. _..
i-ITS/NRC/91-1 October 1991 i
TECHNICAL EVALVATION:
Core Thermal Hydraulic Methodology Using VIPRE-01 Topical Report DPC NE 2004 for Duke Power Company McGuire and Catawba Nuclear Stations i
Prepared for-Reactor Systems branch Division of Systems Technology Office of Nuclear-Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
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TECHNICAL EVALVAT10B 0F THE CDP.E THERMAL-H10RAyl[GJ{JHQQQLQGy USING Y1 PRE 01 10PICAL REPORT OfC NE 2003 fE E!LDELE0XLLLOBEAM MCGUIRE AND CATAWBA NUCLEAR SJATIONS 1.0 H(TRODUCTION t
DPC-NE 2004, -dated December 1988 (Ref. 1) and as revised (Ref. 2), was submitted by Duke Power Company (DPC) for NRC review and approval.
Additional information was submitted on September 14, 1990 (Ref. 3), November 29,1990 (Ref. 4), August 29, 1991 (Ref. 5) and October 25,1991 (Ref. 6) togather with the 10/7&B/91 NRC/ Duke meeting handouts (Ref. 7).
This topical report and.the aforesaid supplemental submittals document the development of core thermal hydraulic analysis based upon the statistical core design (SCO) methodology using the VIPRE 01 computer code for the McGuire and Catawba (M/C) Nuclear Stations.
The 500 method is a thermal-hydraulic - analysis 5
technique which computes DNB margin by statistically combining cora and fuel bundle uncertainties.
The submittal provides a-description and justification for t.pplylug _ uncertainties to the DPC DNBR limits calculations using a o
- statistical rather than a deterministic method.
DPC intends to replace the
-- Westinghouse Improved Thermal Design Procedure by the SCD methodology as part of the DNB design basis approach.
- The objective of-the subject topical report, therefore, is twofold: (i) to fulfill VIPRE 01 SER requirements (Ref.
8) by providing geometric L
representations of the core, and DPC's selection of thermal hydraulic models l
cnd -correlations for use -in analysis of M/C cores. in support of the SCD methodology, and (ii) to describe DPC developed SCD methodology.
The core I
thermal hydraulic metho'dology of this report is based on the BWCMV DNB
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l correlation which was previously approved for use with LYNXT and LYNX 2 for B&W and Westinghouse fuel designs (Refs. 9 and 10).
The purpose of this review, based upon a review of the submitted materials (Refs. 1-7), is to determine acceptobility of (i) conformity of the DPC topical report and supplemental information (Refs.17) to the VIPRE-01 SER requirements (Ref. 8),
(ii) use of the BWCMV critical heat flux (CHF) correlation with VIPRE 01, and (iii) DPC's SCD methodology.
Thereforo, this review of DPC's VIPRE 01 core model was conducted in the context of its use in support of the SCO methodology.
The SCD methodology was reviewed for acceptability of the generic methodology and its application to the M/C reload analyses.
McGuire and Catawba Nuclear Stations, each having two Westinghouse units, are assumed by DPC to be identical for the purpose of core thermal-hydraulic calculations. The analyser, presented in the submittals were performed for a core containing all B&W Mark BW fuel assemblies, except as noted for transition cores.
2.0
SUMMARY
OF TOPICAL.EP.DRT anj.10PPL EMENTS The topical report DPC-NE-2004 and its associated submitttis (refered to as OPC NE-2004) document descriptions of DPC's VIPRE-01 steady-state models for reload type analysis with all BAW Mark-BW fuel assemblies for McGuire and Catawba Nuclear Stations (all Westinghouse plants).
DFC's cbjective in submitting the topical report was twofold: to fulfili conditions required by the VIPRE-01 SER, and to documen+ DPC developed statistical core design mathodology as described below.
2.1 11 PRE-01 C m uter Crig VIPRE-01 has been previously reviwed and approved for application to pressurized water reactor (PWR) pia 9.s in steady state and transient analyses with heat transfer regimes up to itical heat flux (CHF).
The NRC safety evaluation report (SER) on VIPRE r 1 (Ref. 8) includes conditions requiring 2
l L -__
each user to document and suk 4 to the NRC for approval its procedure for using VIPRE-01 and provide justification for its specific modeling assumptions, choice of particular two-phase flow models and correlations, heat transfer correlations, CHF correlation and DNBR limit and input values of plant specific data such as turbulent mixing coefficient and grid loss coefficient including defaults.
The conservative core model, use of certain thermal-hydraulic correlations, and other key input selections were justified through a series of sensitivity studies.
Since DPC's intended use of VIPRE-01 is to perform core thermal-hydraulic caletlations in support of the SCD methodology, wherever applicable, the core conditions used in the VIPRE-01 sensitivity analyses were selected from the set of conditions 'ned to develope the response surface for the SCD analysis.
Because the methodology is based on use of the BWCMV CHF correlation with VIPRE 01, DPC qualified use of BWCMV correlation with VIPRE-01 by predicting a set of data points from the original set of BWCMV CHF data base that cover the ranges of anticipated operation, obtaining a DNBR limit of 1.21.
BWCMV is approved for use with LYNXT and LYNX 2 computer codes with the DNBR limit of 1.21 (Refs. 9 and 10).
2.2 llath.tical_ fare Desian Ke_tAgjnl.2qy The traditional method for accounting for the design and modeling uncertainties that enter into the determination of a DNBR assumes that key input parameters to the core thermal-hydraulic code are simultaneously at their worst level of uncertainty.
The methodology described in the DPC Topical Report DFC-NE-2004 assumes that, while the input parameters are occasionally at their worst case values, the input uncertainties are independent and it is highly unlikely that all the input parameters will take on their worst case values simultaneously.
Tnerefore, the application of the SCD method differs from previous techniques in that the thermal-hydraulic limit analyses are performed by statistical analysis of a series of computations " perturbed" from a "zero" point computed at nominal plant 3
i
~._
.~ -
conditions.
DPC has applied the SCD method to simulate the direct computation of DNBR with VIPRE-01.
The SCD methodology statistically combines uncertainties associated with key r
paremeters used in determination of the DNBR.
In order to perform the required statistical combination of the various input uncertainties, the DPC SCD method employed a non-linear response surface model (RSM).
The response surface equation, variables, and the selection of the test cases which were used to determir.e the coefficients that appear in the response surface equation were-described. The topical report describes the process by which the DNBR is determined with the SCD methodology and ultimate'il* used in combination with other cperating limits.
The statistical DNBR 11tnit (SDL) to replace the traditional CHF correlation limit was determined for the cases analyzed in the submittal: the calculated SDL was found to be 1.40 for the M/C Mark-BW core.
.The range of applicability of the SCO method (therefore the RSM) is def'ned by the range of values from which the composite design points, used to determine the RSM equation, are selected.
For statepoints which fall outside of the -SCO range but which must nevertheless be analyzed for curtain transients, DPC developed a simplified methoi which used VIPRE 01 directly and avoided use of the RSM.
For these cases, values of the variables were generated by use of a Monte Carlo method according to the uncertainty d htribution for each state variable.
The DNB was computed for each such set using VIPRE-01.
Statistical analysis was performed of the set of DNBRs so computed and the computed DNBR is compared against the SOL for acceptability.
3.0 [yAUMJJQN L
3.1 VIpRE Hod.gl Description The M/C core models discussed in the topical report were assumed to contain all B&W Mark-BW fuel.
4
l 3.1.1 Core Nodalizati n In developing the core models, DPC assumed a 1/8 core symmetry with the hot assembly located in the center of the core.
The set of thermal-hydraulic models and correlations used by DPC m the nodalization sensitivity studies were those w':icc. aPC intends to use in future licensing analysis.
These models and correlations were found to yield acceptably conservative results.
3.1.1.1 Radial Nodina Susitivity A parametric study was performed to determine the sensitivity of predicted DNBR to the subchannel model si7.e.
The thermal-hydraulic calculations were performed for three different core subchannel models using steady-state conditions.
The coarse channel model was found to yield acceptably conservative MDNBRs.
Therefore, DPC intends to used the coarse channel model for steady-state Mark-BW thermal-hydraulic analyses for McGuire and Catawba Nuclear Stations, 3.1.1.2 Axial Nodino Sensitivity A sensitivity aaalysis for axial node length was performed with the ccarse core channel model using two different axial node lengths under two different core conditions.
Both of these node lengths correspond to the range of code developer's recommended values.
The results indicated that the agreement was adequate enough that the large sizo noding is acceptably conservative.
3.1.2 VIPRE-01 Geometrical Inout Data DPC's approach to generation of input to the VIPRE 01 code was reviewed for acceptability.
No review was conducted of the input data in comparison to the actual physical geometry.
5
3.1.2.1
&qlive fuel Lenoth for B&W's low densification fuel, the amount of fuel densification is off-set by the fuel thermal expansion. OPC chose to use the cold nominal active fuel length for calculation since this is more conservative.
3.1.2.2 Spacer Grid Form Coefficienti The vendor determined coefficier,ts based on test data were used to account for the hydraulic loss caused by the sariation in flow area and turbulence at each spacer. grid.
3.1.2.3 Core Hyoass Flow Since the bypass flow depends on the number of control rod and burnable poison rod assemblies in the core, this is a cycle dependent parameter.
Therefore, the core bypass flow data used in the analysis should be based on a bounding value or an cycle specific data.
For the purpose of generating the response surface model using the SCD methodology presented in the t;,pical report, a 6% core bypass flow was assumed and thus the resulting core inlet flow was 04% of design flow of 382,000 gpm.
3.1.2.5 Inlet Flow Dist"ib.411RD CHF is decreased and the probability of DNB is enhanced if flowrate is reduced due to a flow ma1 distribution.
The use of 5% inlet flow maldistribution to the hot assembly, while the remaining flow was distributed between the outermost assemblies, yielded slightly more conservative DNBR prediction than did a uniform inlet flow distribution.
Prior to submittal of a Catawba reload report using the SCD methodology, the licensee will evaluate the impact of RCS flow anomaly observed at Catawba 6
)
that resulted in a more severe core inlet flow maldistribution.
3.1.2.7 hital.Poer Distribution The reference radial power distribution was used to calculated the core DNB limit, but the F(Al) portion of the OTDT i " function was determined to onsure DNO protection for any core power distribution.
The hot assembly power distribution was assumed to be relatively flat to minimize flow redistribution.
Thq, reference peak pin power factor, for the power levels below 100% power, was adjusted to determine the core DNB limits.
3.1.2.8 exiai Power Distributica A symmetric chopped cosine was used for the oxial power dl*tribution with a peaking factor of 1.55.
A routino has been added to the VIPRE-01 code to generate axial power shapes with inlet, symmetric, or outlet peaks.
3.1.2.9
-liot Channel Factor The hot channel ftetor F[H used for the McGuire/ Catawba analysis was 1.03 and j
accounted for the allowance on enthalpy rise to account for ' manufacturing tolerances.
3.1.2,10 fluegrical Solution Techningg The RECIRC solution method was used for the McGuire/ Catawba analyses presented in the subnittal and will be used for all of the steady-state core thermal hydraulic analyses.
3.1.3 VIPRE-01 Correlations VIPRE-01 requires empirical correlations for the following models:
L a,
turbulent mixing b.
two phase flow correlations (subcooled and saturated void, and 7
l 1
.,m
void quality relation) c.
critical heat flux 3.1.3.1 lurbulent Mixing The lateral momentun equation requires two parameters: a turbulent momentum factor (FTM) and a turbulent mixing coefficient, i
The turbulent momentum factor (FIM) describes the efficiency of the momentum mixing: 0.0 indicating that crossflow mixes enthalpy only; 1.0 indicating that crossflow mixes enthalpy and momentum at the same strength.
OPC selected a more realistic value for F1M; however, the MDNBR was found through a sensitivity study to be insensitive to this parameter.
Since the turbulent mixing coefficient determines the flow mixing rate, it is an important parameter.
A mixing coefficient was determined based upon tests performed by West irighouse.
Although the grid spacing in the test was different from that of the MARK BW fuels, since the mixing coefficient increases ss grid spacing decreased, the value obtained by Westinghouse and used by DPC is contarvative.
3.1.3.2 Two Phase Flow Correlut_qni for subcooled and bulk void correlations, a sensitivity study using three different combinations of subcooled void and bulk void correlations was perfomed for. two sets of steady-state core conditions.
The. results indicated that the use of DPC selected combination of correlations in conjunction.with Columbit/EPRI two phase friction multiplier predicted conservatively computed DNBR relative to other combinations of correlations DPC intends to_use this combination in McGuire and Catawba analysis.
This is consistent-with the VIPRE-01 SER findings.
8
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!.l.3.3 Friction Pressure lost Selection of (i) the axial friction factor in the Blasius friction pressure factor and (ii) the EPRI two phase friction multiplier was based on sensitivity studies which resulted in conservative prediction of MDNBRs.
)
3.1.3.3 BWCMV Critical Heat Flux Correlalign Use vf BWLMV CHF correlation with the LYNX 2 and LYNXT codes hss been approved by the NRC with a DNBR limit of 1.21.
DPC provided qualification of its use with the -VIPRE 01 code based upon prediction of CHF data points from each test section included in the BWCMV data base to cover the range of anticipated operation.
DPC obtained comparable results from these calculations when compared with LYNX 2 results and determined a DNBR of 1.21.
Therefore use of-BWCMV CHF correlation is acceptable for use with VIPRE-01 with a DNBR limit of 1.21.
3.2 McGuire/ Catawba Core Thermal-Hydraulic Analyses
- . " thermal margin and allowable operating limits define the safe operating i
e,. on (which DPC defines as preventing excessive power, coolant temperature e4 pressure, or any combination thereof).
The primary objective of DPC's
[
use of this approach is to assure the ability to maintain coolability of the l
core _ to prevent fuel damage during Condition I and 11 events.
Three core l
protection design bases were used:
l 1.
Departure from Nucleate Boiling (DNB) design basis; 2.
Fuel temperature design basis; 3.
Hot leg boiling limit.
l The DNBR limit determined by use of SCD method is a statistical DNBR limit-(SDL). The design DNBR limit (DDL) incorporates additional margin to the SDL to allow for adver:e impacts on DNB from other parameters, such as the RCS r
flow anomaly and transition core effects.
The core DNB limits are_ ths 9
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i e--r.+-,
e r e e -
w---+
r+-3.-r
,-m-sw e-m.--e,
---e
.--rarm--e+r wer w-n,----
=y-
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-~wv v
v
l combinations of power, coolant inlet temperature and pressure at which the MDNBR equals the design DNBR limit. Due to the range of applicability of the BWCMV correlation, the quality at the point of MDNBR is also limited to +?2%.
The core DNb limits were calculated using the coarse channel model referred to in a previous se d ion of this report.
The use of the statistical core design (SCD) methodology as part of the DNB design basis approach was proposed by DPC to replace the Westinghouse improved Thermal Design Procedure which considers core T/H analysis parameter uncertainties statistically.
The description of the SCD method is given in the next section.
i 3.3 Statistical Core Defian Methodoloav The traditional method for accounting for the design and modeling uncertainties that enter into the determination of a DNBR assumes that key input parameters to the core thermal hydraulic code are siir.ultaneously at I
their worst level of uncertainty.
The proposed DPC methodology assumes that, while the input parameters are occasionally. at their worst case values, the i
input uncertainties are independent and it is highly unlikely that all the input parameters will take on their worst case values simultaneously.
Therefore, the application of the 500 method differs from traditional-techniques in that the thermal-hydraulic limit analyses are performed by perturbation analyses from a "zero" point computed at nominal plant conditions.
DPC has applied the SCD method to simulated direct computation of DNBR with VIPRE-01.
The SCD methodology-statistically combines uncertainties associated with key paramaters used in' determination of ' the DNBR.
In order to perform the l-required. statistical combination of the varions input uncertainties the DPC 1
SCD method employed a non-linear response surface model (RSM).
l' The response surface-equation is an equation for MDNBR as a function of l
seven state variables.
'A probability distribution of MDh3R as a function of l
10 1
l.
~ _. _. _ _
those variables can be obtained if one knows the distribution of probabilities of the seven variables.
A probability distribution is established for each of the seven variables with the nominal state conditions as the center and with normal distributions for core power, core flow, radial power peaking factor and axial power peak and bounded uniform distributions j
for core outlet pressure, axial location of the axial peak and core inlet temperature.
The core flow uncertainty is comprised of two parts:
measurement and bypass flow uncertainties.
Similarly three components contribute to determi1ation of the radial power peaking factor.
In addition three other variables are assumed to impact computation of DNBR (code /model uncertainty, CHf correlation uncertainty, and error associated with the fit of the response surface equation to the VIPRE-01 computations used to develop the RSM equation).
The uncertainty due to code /model includes a difference between VIPRE 01 and LYNX 2 BWCMV results and the difference in the DNBR results due to the difference between sizes of the core channel models.
A 11onte Carlo computation is used to select sets of values of each of the seven state variables at random (weighted by the distribution functiont)- and a resultant MDNDR is computed from the response surface equation.
The SCD method is used to determine an overall DNBR uncertainty.
The core thermal hydraulic analyses are performed using nominal valves for the parameters that are treated statistically; all other input parameters are assumed at their conservative values.
)
1 3.3.1 Selection of Parameters Seven parameters were identified as those which would significantly impact the calculation of DNBR: (1) core power, (2) core flow. (3) core outlet pressure, (4) core inlet temperature, (5) peak radial power factor, (6) axial peak, and (7) axial locatien of the axial peak.
These variables, assumed independent, define a core statepoint on the response surface.
Since these parameters vary during reactor operation, parameter ranges were developed which wnuld bound the values that would be expected to be encountered in typical M/C T/H analyses.
These ranges are-plant and transient set i
dependent.
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3.3.2 figsoonse Surface Model Assuming independence of the seven selected variables, the equation of the response surface -is defined as a quadratic function of the statepoint variables including cross-terms.
The coefficients in the RSM equation were determined by performing a least-squares fit of the response surface equation to a set of values of MDNBR actually calculated with VIPRE-01 using a DPC eight channel eighth-core symmetric VIPRE-01 model. The basic statepoints at which MONBR was calculated for the purpose of computation of coefficients were determined by a Central Composite Design technique.
Nearly one half of the 143 possible points in a full central composite design were selected to include all extreme points but one plus a series of other points to enable the response surface to cover a wi<ter range of parameter variations including those expected to produce the worst MDNBR.
The standard error between the RSM predicted and VIPRE-01 calculated MDNBR values for the set of statepoints considered was found to be acceptable.
3.3.3 S t a t i s t i c al l y T re a t eILVp e e rt a i n t i e s In order to statistically combine the effect of the uncertainties of the parameters, DPC determined the uncertt.inties, uncertainty distributior; and the uncertainty standard deviation.
A probability distribution was established for each of the seven variables with the nominal state conditions as the center and with normal distributions for core power, core flow, radial power peaking factor and axial power peak, and with bounded uniform distributions for core outlet pressure, axial location of the axial peak and core inlet temperature. The core flow uncertainty is covrised of two parts:
measurement and bypass flow uncertainties.
Similarly three components contribute to determination of the radial power peaking factor.
DPC's rational for assignment of uncertainty distribution was that a normal distribution was assumed when the uncertainty was due either to measurement uncertainty or a known statistical uncertainty distribution.
Whenever such assumption could not be reasonably made, DPC chose the conservative approach 12
i of assuming a uniform distribution with estimated reasonable upper and lower bounds.
The licensee stated in the topical report that the uncertainties and distributions will be justified on a plant-specific basis in the reload ieport for the first appilcation of this methodology.
3.3.4 Propaaatiqa of Uncertaintjits.
In order to combine tha uncertainties to compute an overall DNBR uncertainty, a Monte Carlo-method analysis is performed using the distribution of uncertainties defined with each variables.
A probability distribution is established for each of the seven variables with the nominal state (loss of flow) conditions as the center.
A Monte-Carlo computation is used to select sets of values at random (waighted by the aistribution functions) and a resultant MDNBR is computed for each such set from the response surface equation.
Statistical analysis is then performed on the set of MDNBRs so generated.
This process is repeated for ten core statepoints that cover the range of conditions considered when determining core DNB limits and the limiting Condition 11 DNB transient in order to maximize the coefficient of variation resulting from the propagation of uncertainties.
In addition three other factors are assumed to have the dominant impact on computation of DNBR (code /model uncertainty, CHF correlation uncertainty, and error associated with the fit of the response surf ace ege' tion to the VIPRE-01 computations used to develop the RSM equation).
Finally, the statistical DNBR limit (SDL) to replace the traditional CHF correlation limit is determined.
The statistical design. limit is determined l
from the largest coefficient of variation based on the DNBRs computed by the l
Monte Carlo. computations referred to above which avoid DNB at a 95%
probability /95% confidence level.
For the cases analyzed in the submittal the calculated SDL is 1.396 for the M/C Mark-BW core.
The resulting SDL is 1.40.
l 13 l
L.
3.3.5 Eirtp1Eigi SCD Methodolqgy The range of applicability of the 500 method (therefore the RSM) is defined by the range of values from which the composite design points are selected.
Therefore, for this methodc.ogy to be generally applicable, the composite design points must be selected so that values expected during any expected transient would fall within the envelope defined by these variables, for the statepoints which fall outside of the SCD range but which must nevertheless be analyzed for certain transients, OPC developed a simplified method which used VIPRE 01 directly and avoided use of the RSM.
For example, the lowest value of RCS flow used in the composite design point is 80% of nominal core inlet flow which is 359,080 gpm.
However, during at least three licensing type transients (loss of forced flow, feedwater line break and uncontrolled control rod assembly withdrawal from a suberitical or low-power condition), the flow is expected to be less than 80% when HDNBR occurs. Therefore, for these cases, the RSM as developed is inapplicable.
The DPC developed simplified method is generally usable and will be applied to conditions such as:
(a) a significant change is made in the fuel assembly design; (b) a new or revised CHF correlation is developed; and (c) operating conditions falling outside of the range of conditions listed in revised Table 8 (Ref. 5).
The simplified method bypasses the RSM by directly computing DNBR with the VIPRE-01 code based on the values for the seven variables generated by the propagation of uncertainties through the use of the Monte Carlo method.
An SCD limit is determined for each case as before and compared against the SDL.
DPC stated (Ref 5) that a submittal will be made to the NRC for review if a new SDL is required to be calculated under the following conditions:
14 I
i
l..
1)
A completely new fuel design would require development of a new l
SDL.
However, development of a new design feature, such as a new j
spacer grid design, may only require propagation of the most limiting statepoint to validate the existing limit.
The NRC would only be notified if the design changes result in the calculation of.
t i
l-an SDL > 1.40.
l 2)
Development of a new version of the BWCMV correlation or use of a different CHF correlation would require calculation of a new SDL and a submittal to the NRC.
3)
Computation of the SDL for any statopoint outside of the range j
given in revised Table 8 would be propagated using VIPRE 01 directly as was done for FSAR transient 15.4.1.
A submittal would l
be made to the NRC only if this propagation results in an SDL >
1.40.
3.3.6 DM R Penal u m The SDL for McGuire/ Catawba was computed to be 1.40.
The design DNBR limit l
(DDL) was selected to be 1.55, allowing 10.7% DNBR margin.
l Against this margin, several penalties were applied:
L a.
a transition core penalty against OFA fuel of 3.8%;
b.-
a total instrumentation DNBR penalty of ).9% and l
c.
a rod bowing penalty of 3.5%.
L Therefore, the total DNBR. penalty against 0FA fuel is 9.2% leaving a 1.5%
L margin to DDL.
l 3.3.6.1 Iransition Corn Analvsh The transition core effect was analyzed using VIPRE-1 by modeling both thr.
l.
15 I
... =
~ _. _ _ _ _. _ _. _ _._
i Optimized fuel assemblies (OFA) and Mark BW fuel.
The BWCMV CHF correlation was used fer this analysis.
Although the data base on which the development of the correlation is based cover the range of parameters of 0FA fuel, it does not specifically include OFA data.
The application of the BWCMV correlation for analysis of 0FA fuel has been approved by the NRC (Ref.11) and was further validated by comparison of VIPRE/BWCHV to LYNX 2/BWCMV results.
The power peaking associated with each fuel type is calculated and compared against MAP limits, i
Using the simplified SCD method, the SDL was computed for OFA fuel and Mark-BW fuel and_was bounded by the SOL of 1.4.
4.0 Conclusioni We find-that the subject topical report, together with DPC responses, contains sufficient information to satisfy the VIPRE-01 SLR requirement that 4
each VIPRE-01 user submit a document describing proposed use, sources of input variables, and selection and justification of correlations as it relates to use by DPC for development of statistical core design methodology.
Acceptability of the DPC VIPRE-01 model for steady state application to analysis of McGuire and Catawba Nuclear Stations is based upon selection of models/ correlations supported by the sensitivity study results submitted.
Should DPC change any of these itemt, OPC should submit justification for the change to the NRC for approval.
Use of the BWCMV CHF correlation with VIPRE 01 is found acceptable with a ONBR limit of 1.21.
.We further find that the manner in which the code is to-be used for such analyses, selection of noda11 ration, models, and correlations provides, except as listed below, ade.: Late assurances of conservative results and is therefore acceptable.
16
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-,y---r--*m e
w--r-,- -
-,',wr-e
,--y<y,,rw
-ary-w.-r,-
wwwt,.--,-
-w-
+<,-se--%.-
=
+
The following lindtations and restrictions are recommended regarding the use of DPC's VIPRE-01 model and its associated statistical core design methodology presented in DPC-NE-2004 and its supplemental materials for analysis of McGuire and Catawba Nuclear Stations:
1.
The DPC developed statistical core design methodology, as described in the submittal, is a generic methodology and is conceptually acceptable and generally appitcable to other PWR plants; however, the approval we recommend at this time is for only McGuire and Catawba Nuclear Stations due to DPC's use of specific uncertainties and distributions based upon plant data and its selection of statepoints used for generating the statistical design limit.
2.
Whenever conditions provided in response to NRC question 4 in Reference 5 are present, cither the response surface must be re-evaluated or the "simpilfied method" must be used.
The licensee is further required to make a submittal to the NRC for review if a new SDL is calculated as a result of conditions outside the ranse of parameters set forth in revised Table 8 of Reference 5.
l 3.
Core bypass flow is cycle dependent.
DPC will veri fy, in future applications, that its use of a particular core flowrate resulting from a bypass flowrate for that cycle is bounded by the range of values used in the subject topical report.
Otherwise, OPC will reassess tne need for regeneration of a new response surface.
I
5.0 REFERENCES
l 1.
" Duke Power Company - McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodclogy Using V.' PRE-01," DPC-NE-2004, December 1988.
2.
Letter from H.B. Tucker (DPC) to USNRC, "McGuire and Catawba Nuc'. ear Stations Core Thermal Hydraulic Methodology Uising. VIPRE-01," February 22, 1990, 3.
Letter from-H.B. Tucker (DPC) to USNRC, ' Topical Report OPC-NE-2004,"
17 i
l
~.
_ ~ _ _.
i September 14, 1990.
4.
Letter from H.S. Tuckman (DPC) to USNRC, " Topical Report DPC NE-2004,"
November 29, 1990.
5.
Letter from M.S. Tuckman (DPC) to USNRC, " Supplemental Information to Assist in Review of Topical Reports DPC NE-3000 and OPC NE-2004," August 29, 1991.
6.
Letter from H.C.1991 Meet {ng with NRC Staff and Contract Reviewers,*
Tucker 0PC) to USNRC, " Handouts Presented in the October 7 & 8, October 16, 1991.
~
7.
Letter from H.B. Tucker (DPC) to USNRC, " Final Response to Questions Regartling the Topical Reports Associated with the MIC8 Reload Package,"
November 5, 1991.
D.
Letter from C.E. Rossi (NRC) to J.A. Blaisdell (UGRA), " Acceptance for Referencing of Licensing Topical Report VIPRE-01: A Thermal Hydraulic Code for Reactor Cores, EPRI NP-2511-CCH, Vols. 1 4," May 1, 1986.
9.
Letter from A. Thadani (USNRC) to J.H. Taylor (B&W), ' Acceptance for i
Referencing of Augmented Topical Report BAW 10159P "BWCMV Correlation of.
Crf tical Heat Flux in Mixing Vane Grid fuel Assemblies" Hay 1986," May 22, 1989.
10.
Letter from A.
Thsdani (USNRC) to J.H. Taylor (R&W), " Acceptance for Referencing of Topical Report BAW-10159P "BWCMV Correlation of Critical
-Heat Flux in Mixing Vane Grid Fuel Asuniblies" May 1986," February 17, 1989.
11.
Letter from USNRC to N.B. Tucker (DPC), " Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to Topical Report BAW-10173P, Revision 2, Mark BW Reload Safety Analysis for Catawon and McGuire,"
February 20, 1991.
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