ML20056E381
| ML20056E381 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/05/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20056E378 | List: |
| References | |
| NUDOCS 9308230285 | |
| Download: ML20056E381 (3) | |
Text
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/b UNITED STATES E' 'L ! NUCLEAR REGULATORY COMMIS.ClON
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO TOPICAL REPORT RXE-89-002 "VIPRE-01 CORE THERMAL-HYDRAULIC ANALYSIS HETHODS FOR COMANCHE PEAK STEAM ELECTRIC STATION LICENSING APPLICATIONS" TEXAS UTILITIES ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATION. UNITS 1 AND 2 DOCKET NOS. 50-445 AND 50-446 l
1.0 INTRODUCTION
Texas Utilities Electric Company (TV Electric or licensee) submitted topical report RXE-89-002, "VIPRE-01 Core Thermal-Hydraulic Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications" in a letter dated June 30,1989 (Ref.1), as supplemented by letters of July 19, 1991 (Ref. 2) and October 15, 1991 (Ref. 3). This topical report and supplemental information were provided to document the modeling methodology for the VIPRE-01 computer code (Ref. 4) for Comanche Peak Steam Electric Station (CPSES) as required by the VIPRE-01 safety evaluation report (SER) (Ref. 5), and to demonstrate TU Electric's proficiency in using the VIPRE-01 computer code for the core thermal-hydraulic analyses as required by Generic Letter 83-11 (Ref.
6). The licensee stated that they intend to use the VIPRE-01 computer code to perform the departure from nucleate boiling (DNB) calculations for core safety analyses for normal operation and selected plant transients.
VIPRE-01 is an open channel code designed to evaluate departure from nucleate boiling and coolant state for steady state and transient core thermal-hydraulic analyses (Ref. 4). The VIPRE-01 computer code has been approved for licensing calculations for heat transfer regime up to critical heat flux (CHF) for pressurized water reactors (Ref. 5). The SER for VIPRE-01 requires each ,
user to submit for staff review documentation describing how they intend to -
use VIPRE and providing justification for its specific modeling assumptions, choices of particular models and correlations, and input values of plant specific data such as turbulent mixing coefficient and grid loss coefficient.
Generic Letter 83-11 requests that each licensee or vendor who intends to use ,
large, complex computer codes to perform their own safety analyses to demonstrate their proficiency to use the codes by submitting code verification performed by themselves (Ref. 6).
The licensee intends to use the TUE-1 CHF correlation to evaluate the steady state and transient departure from nucleate boiling ratios (DNBRs). The 95/95 DNBR limit of TUE-1 correlation has been determined to be 1.16. The licensee submitted a separate topical for staff review and approval for the use of the L
9308230285 930805 PDR P
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i TUE-1 CHF correlation with VIPRE-01 (Ref. 7). This report was approved in a !
staff safety evaluation (SE) of June 11, 1992. }
The licensee intends to use VIPRE-01 in conjunction with its RETRAN-02 based on TU Electric transient analysis methodology as described in the TU Electric topical report RXE-91-001 (Ref. 8). The TU Electric transient analysis methodology was found acceptable in a staff safety evaluation of July 16, i 1993. We find these methodologies compatible and acceptable for analyses of the Comanche Peak Steam Electric Station. l l
2.0 STAFF EVALUATION The staff review and evaluation of the topical report RXE-89-002 included- l (1) the acceptability of the VIPRE-01 modeling methodology with respect to the ;
intended licensing applications, (2) the acceptability of TU Electric's VIPRE- !
01 model, as described in the topical report and supplemental information, I with respect to the VIPRE-01 SE requirements, and (3) TU Electric's competence i in using the VIPRE-01 computer code as required by Generic Letter 83-11. !
The review was performed with technical assistance from International .
Technical Services, Incorporated (ITS) and its review findings are provided in j the attached Technical Evaluation Report (TER). The staff has reviewed the l TER and concurred with its findings, with clarifications as discussed in l Section 3.0. ;
3.0 CONCLUSION
S The staff has reviewed topical report RXE-89-003 and its supplemental i information. We find the-VIPRE-01 model for CPSES to be acceptable since the i selections of specific modeling assumptions, code models and correlations, and ,
input parameters were suitably justified by the licensee to provide adequate i assurances of conservative results. We also find the licensee to be competent i in using the VIPE-01 computer code. Therefore, we conclude that the subject topical is acceptable for referencing in the CPSES core thermal-hydraulic {
analyses for the DNB related licensing applications. s This acceptance includes the following clarifications of issues identified in f the TER. !
- 1. As discussed in Section 1.0, the acceptability of the TUE-1 CHF correlation which is used with VIPRE-01 is documented in the staff SE of- !
May 15, 1992. We find the use of the TUE-1 correlation with the TU !
Electric VIPRE-01 methodology acceptable; however, use of the TU Electric VIPRE-01 with other CHF correlations is not within the scope of this SER.
- 2. The acceptability of TU Electric's full core VIPRE model developed to '
simulate a main steamline break transient will be assessed in a separate ongoing review.
f
- 3. We find that the use of VIPRE-01 is compatible with the RETRAN-02 based ,
TV Electric transient analysis methodology (Ref. 8), and its use in the !
analysis of the Comanche Peak Steam Electric Station plants is ;
acceptable. i
4.0 REFERENCES
- 1. Letter from W.J. Cahill, Jr., (TUEC) to USNRC, " Comanche Peak Steam !
Electric Station (CPSES), Docket Nos. 50-445 and 50-446, Submittal of VIPRE-01 Core Thermal-Hydraulic Analysis Methods Topical Report RXE '
002", June 30, 1989.
- 2. Letter from W.J. Cahill, Jr., (TVEC) to USNRC, " Comanche Peak Steam Electric Station (CPSES) Unit 1 Docket No. 50-445, Request for Additional Information on RXE-89-002, VIPRE-01 Thermal-Hydraulic Analysis Methods," July 19, 1992.
- 3. Letter from W.J. Cahill, Jr., (TUEC) to USNRC, " Comanche Peak Steam ;
Electric Station (CPSES), Docket No. 50-445, Supplemental Information on RXE-89-002, VIPRE-01 Core Thermal-Hydraulic Analysis Methods," i October 15, 1991. i
- 4. "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores," EPRI NP-2511-CCM Revision 2, EPRI, July 1985.
- 5. 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-CCM, Vols.1-4," May 1,1986.
- 6. Licensing Qualification for Performing Safety Analyses in Support of Licensing Actions (Generic Letter No. 83-11), USNRC, February 8, 1983.
- 7. "TUE-1 Departure from Nucleate Boiling Correlation," RXE-88-102-P, 4 January 1989. (Approved - June 11,1992)
- 8. " Transient Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications," RXE-91-001, February 1991. (Approved - July 16, ,
1993)
Attachment:
Technical Evaluation Principal Contributor: Frank Orr, SRXB/NRR !
Date: August 5, 1993 i I
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ITS/NRC/91-9 December 1991 I
r TECHNICAL EVALUATION:
i VIPRE-01 Core Thermal-Hydraulic Analysis Methods l 1
fc - i Comanche Peak Steam Electric Station Licensing Applications ;
t Texas Utilities Electric Company !
Topical Report RXE-89-002 i
f i,
i Prepared for i Reactor Systems Branch !
Division of Systems Technology l Office of Nuclear Reactor Regulation !
U.S. Nuclear Regulatory Commission :
Washington, D.C. 20555 !
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TECHNICAL EVALUATION-VIPRE-01 CORE THERMAL-HYDRAULIC ANALYSIS METHODS FOR j l COMANCHE PEAK STEAM ELECTRIC STATION LICENSING APPLICATIONS ;
TOPICAL REPORT RXE-89-002 i FOR :
TEXAS UTILITIES ELECTRIC COMPANY i,
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1.0 INTRODUCTION
The topical report entitled "VIPRE-01 Core Thermal-Hydraulic Analysis Method
] for Comanche Peak Steam Electric Station Licensing Applications," RXE-89-002, j j dated June 1989 (Ref.1), was submitted by Texas Utilities Electric Company i;
- (TUEC) for NRC review and approval. Additional information was submitted on l July 19, 1991 (Ref. 2) and October 15, 1991 (Ref. 3). This topical report j documents the development of core thermal-hydraulic analysis methodology l using the VIPRE-01 computer code for the Comanche Peak Steam Electric Station f j (CPSES). l
~
VIPRE-01 has been previously reviewed and approved for application to t t
i with heat transfer regimes up to critical heat flux (CHF). The NRC safety l 1 i j evaluation report (SER) on VIPRE-01 (Ref. 4) includes conditions requiring i l each user to document and submit to the NRC for approval its procedure for f using VIPRE-01 and provide justification for its specific modeling :
! assumptions, choice of particular two-phase flow models and correlations, j heat transfer correlations, CHF correlation and DNBR limit and input values l of plant specific data such as turbulent mixing coefficient and grid loss [
I coefficient including defaults.
] The objective of the subject topical report is: (i) to fulfill VIPRE-01 SER }
l requirements by providing geometric representations of the core, and TUEC's l r
i l 1 l
, i
l 1 '
l' selection of thermal-hydraulic models and correlations, (ii) to describe the j
- TUEC VIPRE-01 model for licensing transient applications and (iii) to .
demonstrate TUEC staff competence in response to NRC Generic Letter 83-11 ;
(Ref. 5). [
The purpose of this review is to determine, based upon a review of the l j submitted materials (Refs.1 - 3), (i) conformity and sufficiency of the TUEC l topical report and supplemental information to satisfy the VIPRE-01 SER l
! requirements, and (ii) adequacy of demonstration of TUEC's VIPRE-01 model for j licensing applications. l Ceranche Peak Steam Electric Station consists of two Westinghouse 4-loop j PWRs, each with a rated thermal power of 3411 MW. The VIPRE-01 core model !
j described in the subject topical was developed to be generically applicable [
! _ to both units; however, where core specific data are needed, the model uses !
i the data associated with the CPSES Cycle 1 core which contains the ;
Westinghouse standard 17x17 R-grid fuel.
f l
The core thermal-hydraulic methodology of this report is based on the TUE-1 DNB correlation (Ref. 6) which is being reviewed separately by the NRC. In j
) order to incorporate the TUE-1 correlation and to produce additional output l
) convenient to the user, the VIPRE-01 computer code was modified. !
i i I
! 2.0
SUMMARY
OF TOPICAL REPORT l i The topical report RXE-89-002 documents descriptions of TUEC's VIPRE-01 l models for licensing applications for analysis of the Comanche Peak Nuclear j Stations. The topical report was prepared in response to the VIPRE-01 SER j j requirements and NRC Generic Letter 83-11. The report also describes the l TUEC VIPRE-01 model for licensing transient applications. l l The conservative core model, use of certain thermal-hydraulic correlations, j and other key input selections were justified through a series of sensitivity d analyses. Since TUEC's intended use of VIPRE-01 is to perform core thermal-hydraulic calculations in support of licensing, TUEC presented licensing i
l .
i
, analyses as demonstration of adequacy of the base model.
Because the methodology is based on use of the TUE-1 DNB correlation with :
VIPRE-01, it is necessary that the TUE-1 correlation be approved for use with VIPRE-01. TUEC submitted a separate topical report, RXE-88-102-P (Ref. 6), {
to the NRC for review and approval. The acceptability of this VIPRE-01 topical report is subject to the satisfactory approval of RXE-88-102-P. l
margin penalties are beyond the scope of the subject submittal and were not ;
reviewed. l l
3.0 EVALUATION l t
3.1 VIPRE Model Description The Comanche Peak core models discussed in the topical report were developed ,
to simulate a core containing Westinghouse 17x17 R grid fuel.
3.1.1 Core Nodalization i
i In developing the core models, TUEC assumed 1/8 core symmetry with the hot l assembly located in the center of the core by choosing a design power shape )
3 and an inlet flow distribution symmetric to the core center. The set of j thermal-hydraulic models and correlations used by TUEC in the nodalization sensitivity studies are those which TUEC intends to use in future licensing analysis. The parametric sensitivity studies were performed using two sets of conditions: steady-state nominal and loss of flow conditions. j 3.1.1.1 Radial Nodino Sensitivity l
A parametric study was performed to determine the sensitivity of predicted l DNBR to the subchannel model size as well as the degree of symmetry in the core. The thermal-hydraulic calculations were performed for three different core subchannel models (12, 16 and 40 channels) using both the nominal 3
l
1 j
steady-state and loss of flow conditions. In addition, a case with 1/4 core l symmetry was analyzed with similar subchannel modeling detail as the , f reference case which contained 1/8 core symmetry and 26 channels.
The results showed that the predicted MDNBRs were within 1% of each other and ;
therefore that the MDNBR's were relatively insensitive to the radial noding. !
The reference case resulted in acceptably conservative MDNBRs under conditions considered in this study. Therefore, TUEC's use of the 16 channel ;
model is acceptable.
I For asymmetric transients such as the main steam line break (MSLB) transient, TUEC developed a full core model to better simulate the asymmetric flow, core ;
inlet temperature, and power distribution in the core. The description of the full core model for use in the MSLB is presented in a separate topical ,
report and was not reviewed in this study. .
3.1.1.2 Axial Nodino Sensitivity i
The reference model contains 38 axial nodes with a nonuniform nodal length along the fuel rod. In the region where the MDNBR is expected, the nodal I length is 3 inches, which is within the range of the code developer's {
recommended values. Parametric sensitivity analyses for axial node -length were performed with coarser and finer noding and both uniform aad nonuniform distributions model. The results indicated that the reference noding was j adequate in predicting conservative MDNBR. l For the sensitivity study, axial nodalization reflected the fact that due to the power shape chosen the expected MDNBR was in the central region. TUEC, in its response to a NRC question, stated that a check will be made to f
determine the location of the DNB so that the model adequately represents the !
details of the event analyzed.
3.1.2 VIPRE-01 Geometrical Input Data TUEC's approach to generation of input to the VIPRE-01 code was reviewed for :
4 [
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i i . ;
acceptability. No review was conducted of the actual input data in comparison to the actual physical geometry.
3.1.2.1 Active Fuel Lenoth .
TUEC will select a value for active fuel length for licensing analyses which conservatively accounts for fuel densification and thermal expansion. TUEC >
. will obtain such data from the vendor.
l 3.1.2.2 Core Bvoass 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.
i Therefore, the core bypass flow data used in the analysis should be based on !
a bounding value or on cycle specific data. A nominal bypass flow of 5.8% .
was assumed, which is consistent with the current licensing analysis in CPSES :
1 FSAR.
3.1.2.3 Inlet Flow Distribution i
CHF is decreased and the probability of DNB is enhanced if flowrate is j reduced due to a flow maldistribution. The use of a 5% inlet flow reduction j to the hot assembly, while the flow to the assemblies in the outermost :
l channel was increased to conserve total flow, has been previously approved by [
- the NRC. _,
- 3.1.2.4 power Distributions e
d j 3.1.2.4.1 Pin Power Distribution i
TUEC's choice of the pin radial power distribution is based upon physics l calculations for power distribution in the peak pin assemblies and is the ;
flattest pin power distribution from such calculations and is therefore ,
acceptably conservative. Other distribution may be considered in future <
analyses if determined to be conservative from a DNB perspective. l 5
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3.1.2.4.2 Radial Power Distribution For the reference radial power distribution, the peak rod power is based upon l the peak rod power set to the Technical Specification limit of FN dH at full power. The current limit for CPSES-1 is 1.55.
i TUEC will select the peak rod power factor in the reference radial power {
distribution to ensure conservatism on a transient specific basis. This is i because, for some events, the radial power peak may exceed the Technical Specification peaking limit.
3.1.2.4.3 Axial Power Distribution A series of reactor physics calculations are performed to generate axial-power profiles. In order to determine the limiting axial power profile, each of these shapes is examined with respect to DNB implications. A symmetric chopped cosine with a peak-to-average factor of 1.55 is used as the reference axial power distribution. TUEC stated that although the selected reference
, axial shape is expected to be conservative with respect to the axial power ,
profiles encountered in normal plant operation, other axial power shapes may be more DNB limiting for some events. Therefore, the conservatism of the axial power shape in the CPSES VIPRE-01 model will be confirmed for each specific application. This approach is acceptable.
3.1.2.5 Crossflow Parameters The gap width, centroid length and axial nodel length characterize the crossflow between adjacent channels in the control volume. The gap width and centroid length are determined from the fuel assembly geometry.
3.1.2.5 Spacer Grid SDacina and Form Coefficients i
The grid spacing is obtained from the fuel vendor.
6 9
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= .- . . .
j The vendor determined form and grid loss coefficients based on fuel assembly ,
I hydraulic test data were used to account for the hydraulic losses caused by the variation in flow area and turbulence at a spacer grid. Bundle average l
' loss coefficients are converted to channel dependent loss coefficients to !
account for the different flow areas in the standard fuel channels and the guide thimble channels. ;
3.1.2.7 Enaineerina Uncertainties
- i l
The engineering enthalpy rise hot channel factor is obtained from the fuel ;
i vendor. l Although the current FSAR analysis performed by Westinghouse includes a penalty factor to account for the effect of the engineering heat flux hot ;
channel factor on the MDNBR, Westinghouse demonstrated that there is no DNB i
penalty for the relatively low intensity heat flux spikes caused by variation l in fabrication parameters. TUEC confirmed this by performing a sensitivity f study which indicated that no additional DNB p?nalty due to heat flux spikes !
was required. !
Similarly, the use of hot channel pitch reduction factor was dropped from the f current CPSES VIPRE-01 inodel . This effect will be compensated by the rod bow penalty to be calculated for CPSES.
3.1.3 VIPRE-01 Models and Correlations VIPRE-01 requires empirical correlations for the following models: ,
- a. turbulent mixing; j
- b. two-phase flow correlations (subcooled and saturated void, and !
void-quality relation)-
l
- c. heat transfer correlations; and
- d. critical heat flux. .
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In addition, the VIPRE user has an option to model fuel rods as dummy rods or
{
conduction rods. Selection of these models are application (therefore l transient) dependent. l 3.1.3.1 Turbulent Mixina t
The lateral momentum equation requires two parameters: a turbulent momentum f factor (FTM) and a turbulent mixing coefficient.
The turbulent momentum factor (FTM) 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. TUEC l selected 0.0 as a conservative value for FTM. !
i Since the turbulent mixing coefficient, ABETA, determines the flow mixing i rate, it is an important parameter. A best-estimate mixing coefficient of f 0.059 was determined based upon interchannel mixing tests performed by l Westinghouse for this fuel type with 26" grid spacing. The value used in the i reference CPSES model is 0.038, therefore, is conservatively low.
e Since this coefficient is a fuel dependent value, TUEC will use values for ;
other grid types or fuel types based upon vendor test data. In the absence l
of any test data, TUEC states that conservative values of ABETA will be used "
! and justified on the basis of grid type and comparison to other similar grid
- types for which test data do exist. This approach is acceptable. ;
i t
3.1.3.2 Two-Phase Flow Correlations ;
i i
In order to model the two-phase flow effect, subcooled void correlations, j bulk void models and two-phase friction multipliers are used in VIPRE-01. A !
rensitivity study was performed using three options for modeling subcooled l void, four bulk void models and three two-phase friction multipliers. The f results indicated that the use of- the EPRI subcooled void and bulk void i correlations, and the Columbia /EPRI two-phase friction multiplier l conservatively predicted DNBR relative to other combinations of correlations.
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TUEC will use this combination in future analyses. '
t 3.1.3.3 Heat Transfer Correlations j Several heat transfer correlations are available in the VIPRE-01 code. TUEC performed a sensitivity study with a transient model using a rod withdrawal event. Two single-phase convection correlations, four subcooled boiling models and three saturated boiling model s were selected in various ;
- combinations. The results indicated that the heat transfer package contained j in the CPSES reference model conservatively computed DNBR relative to other {
combinations of correlations.
3.1.3.4 Friction Pressure loss The Blasius correlation is used in the CPSES VIPRE-01 model. A sensitivity !
i study was performed to examine the impact of axial friction factor on the l computed MDNBR. TUEC used a constant friction factor, Blasius and McAdam l smooth tube correlations in the study. Although the differences between the l j computed MDNBRs are small, the use of the Blasius correlation resulted in a conservative prediction of the MDNBR for both nominal and the loss of flow !
conditions. _,
i :
3.1.3.5 TUE-1 Critical Heat Flux Correlation l The VIPRE-01 computer code was modified to incorporate the TUEC developed j TUE-1 DNB correlation. Use of TUE-1 DNB correlation with the VIPRE-01 code j j has been under review by the NRC. TUEC provided qualification of its use with the VIPRE-01 code based upon prediction of CHF data points from 19 Westinghouse test sections consisting of 934 points selected from the l Columbia University data bank. The 95/95 DNBR limit was determined to be l 1.16. Acceptability of TUEC's TUE-1 DNB correlation as part of the CPSES j
- VIPRE-01 model, however, is subject to the review outcome of the separate topical report which documents the development and basis of the TUE-1 l
- correlation.
9 i l ,
d.
_ _ _ - _ . . _ . ~ . _ , , _
3.1.4 Fuel Rod Modelina With VIPRE-01, fuel rods can be modeled either as dummy rods or conduction rods. For steady-state application, TUEC will use the dummy rod model by which the rod surface heat flux is specified as an input parameter and the heat conduction or the temperature distribution within the fuel rods will not be computed. The rod surface heat flux is calculated by RETRAN-02.
For transient applications where fuel stored energy and thermal inertia effects are important, TUEC will use the conduction rod model. With this model, the surface heat flux is computed. A uniform radial power shape is assumed within the fuel rods. A constant gap conductance, provided that the [
appropriate bounding value is selected, is conservative with respect to the DNB calculation. TUEC will select a bounding value for the gap conductance ;
on a transient specific basis based upon a detailed fuel performance :
analysis. ;
3.1.5 Numerical Solution Technioue [
r The direct UPFLOW solution method was used for the Comanche Peak analyses !
presented in the submittal and will be used for the core thermal-hydraulic l analyses. 'The convergence limits and damping factors are adjusted to ensure l solution convergence within a reasonable degree of accuracy. In the event, i however, convergence is not obtained, the RECIRC solution method is j recommended. I l
3.2 Demonstration Analyses ;
Demonstration DNBR calculations were performed, using the CPSES VIPRE-01 ;
model described areviously, for the following four transients: (1) nominal I stnady-state operating conditions, (2) the complete loss of flow transient, (3) the uncontrolled fast rod withdrawal transient, and (4)_ the uncontrolled i slow rod withdrawal transient. Results were compared with those presented in the current FSAR performed by Westinghouse using the THINC computer code and the W-3 DNB correlation.
10 l
i i
i The purpose of performing demonstration analyses is to illustrate the DNB l analysis methodology which TUEC intends to use for CPSES licensing
- applications. i j
I For this demonstration, the input boundary conditions were obtained from the !
- current CPSES-1 FSAR. In future licensing applications, the input operating ;
conditions will be computed by the RETRAN-02 code. The rod conduction model [
was used in the set of transient calculations. In addition, several ,
assumptions were made with respect to the power distributions and the hot channel: (1) the power related factors were held constant with time, and (2) hot channel related data were held constant with time.
- Significant increases in a range of 22 to 29% in MDNBR were predicted with !
the CPSES VIPRE-01 model over the current FSAR MDNBRs. The differences were I
- principally attributed to (1) the use of TUE-1 DNB correlation in the VIPRE-01 model instead of the W-3 correlation by Westinghouse, (2) more accurate
- modeling of the grid spacing by TUEC and (3) the absence of an engineering .
heat flux hot channel factor and a hot channel pitch reduction factor in the VIPRE model. The same set of calculations were re-performed after the CPSES l
, VIPRE-01 model was modified to closely match the Westinghouse THINC model ,
l used in the FSAR analyses. The re-analyses agreed well. l l
4.0 Corclusions }
We find that the subject topical report, together with TUEC responses, j l contains sufficient information to satisfy the VIPRE-01 SER requirement that l each VIPRE-01 user submit a document describing proposed use, sources of j input variables, and selection and justification of correlations by TUEC for !
1 use in CPSES core thermal-hydraulic analyses.
I Acceptability of the TUEC VIPRE-01 model for steady-state and transient l application to analysis of CPSES is based upon selection of
- models/ correlations supported by the sensitivity study results submitted. !
a !
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0 l
We further find that the manner in which the code if to be used for such !
analyses, selection of nodalization, models, and correlations provides, !
j except as listed below, adequate assurances of conservative results and is l therefore acceptable.
Although in the current Westinghouse DNB methodology certain DNBR penalties were eliminated, this may have been done in light of the overall conservatism
) of the methodology. Therefore, when the desigu DNBR limit is determined by TUEC, the acceptability of eliminating these factors must be re-addressed to !
assure that there is no reduction in safety margin. At the same time, review ;
4 of DNB penalties, which were not addressed in this topical report, must be '
made to assure that an adequate margin' is incorporated in the DNBR design ;
, i l limit.
l l
5.0 REFERENCES
l
- 1. "VIPRE-01 Core Thermal-Hydraulic Analysis Method for Comanche Peak Steam l Electric Station Licensing Applications," RXE-89-002, June 1989. !
I
- 2. Letter from W.J. Cahill, Jr. (TUEC) to USNRC, Attachment, " Responses to l NRC Questions on RXE-89-002," July 19,1991. j
- 3. Letter from W.J. Cahill (TUEC) to USNRC, Supplemental Information on l
, RXE-89-002 VIPRE-01 Core Thermal-Hydraulic Analysis Methods" October 15, }
j 1991. !
- 4. 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-CCM, Vols. 1-4," May 1, 1986. !
I
! 5. " Licensee Qualification for Performing Safety Analyses in Support of '
l Licensing Action (Generic Letter 83-11)," USNRC, February 8, 1983. t 1 l "TUE-1 Departure from Nucleate Boiling Correlation," RXE-88-102-P, t
6.
g January 1989.
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