ML15264A617
| ML15264A617 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/19/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15264A616 | List: |
| References | |
| NUDOCS 8907240079 | |
| Download: ML15264A617 (16) | |
Text
jk~ REcl, UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ENCLOSURE 1 SAFETY EVALUATION BY THE OFFLCE OF NUCLEAR REACTOR REGULATION RELATING TO TOPICAL REPORT DPC-NE-2003, "CORE THERMAL-HYDRAULIC METHODOLOGY USING VIPRE-01" DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-269, 50-270 AND 50-287
1.0 INTRODUCTION
Duke Power Company (DPC) submitted Topical Report DPC-NE-2003, "Core Thermal Hydraulic Methodology Using VIPRE-01," for Nuclear Regulatory Commission staff review in a letter dated August 31, 1988 (Ref. 1) and amended by a letter of May 3, 1989 (Ref. 2).
This report documents DPC's use of the VIPRE-Ot code (Ref. 3) in lieu of the currently used codes, CHATA and TEMP (Refs. 4 and 5),
for Oconee Nuclear Station licensing core thermal-hydraulic methodology. The Oconee core thermal-hydraulic analyses are routinely performed for fuel reloads to ensure that the departure from nucleate boiling ratio (DNBR) limit will not be violated during steady state overpower condition and anticipated transients.
These analyses consist of (1) a steady state thermal hydraulic analysis to determine the allowable pressure-temperature operating limits and the power distribution limits, and (2) an analysis of the limiting two pump coastdown transient to determine a flux/flow reactor trip setpoint. Since the methodology of determining these safety and operating limits has been reviewed and approved (Ref. 6) previously, the staff review of the topical report concentrated on the use of the VIPRE-01 code in the core thermal hydraulic calculations.
VIPRE-O1 is an open-lattice subchannel core thermal-hydraulic code. In the open-lattice analysis, the reactor core or fuel bundle is divided into a number of quasi-one-dimensional channels that communicate laterally by diversion crossflow and turbulent mixing. This approach more realistically considers the 8072 4 6797019 P
OK05000269 F
2 flow redistribution effects in the open-lattice core of a pressurized water reactor (PWR) and results in less severe hot channel thermal hydraulic conditions than that obtained from the closed-channel approach used in CHATA.
VIPRE-01 was developed by Battelle Pacific Northwest Laboratories under the sponsorship of the Electric Power Research Institute. In December 1984, the Utility Group for Regulatory Application submitted the VIPRE-01 code for NRC staff review (Ref. 7).
In approving VIPRE-01 for PWR licensing applications (Ref. 8), the staff required each VIPRE-01 user to submit separate documentation describing its intended use of VIPRE-01 and providing justification for its specific modeling assumptions, choices of particular models and correlations, and input values of plant specific data.
In a letter of June 19, 1989 (Ref. 9), DPC.indicated that the VIPRE-01 models and methodology described in DPC-NE-2003 are related to the reload thermal hydraulic analyses, that the methodology of using VIPRE-01 model for predicting the minimum DNBRs resulting from FSAR Chapter 15 transients, except for the two-pump coastdown, are described in Topical Report DPC-NE-3000, and that the VIPRE-01 methodology for transient analyses may be different from that used in DPC-NE-2003. Therefore, the scope of the staff review of DPC-NE-2003 was limited to the application of VIPRE-01 in the steady state and two-pump coastdown analyses.
2.0 STAFF EVALUATION The staff review and evaluation of DPC-NE-2003 included: (1) the nodal sensitivity studies to determine the radial noding details and the axial node sizes, (2) the plant-specific core thermal-hydraulic parameters such as the crossflow parameters, grid loss coefficients, core bypass flow, inlet flow distribution and flow area reduction factor, power distributions, hot channel factors, (3) the selected two-phase flow, heat transfer models and correlations, (4) the validation of the BWC critical heat flux correlation (Ref. 10) and the DNER limit in conjunction with VIPRE-01, and (5) the fuel pin heat conduction parameters.
3 The review was performed with technical assistance from International Technical Services (ITS), and its review findings are contained in the technical evaluation report (TER) which is attaphed. The staff has reviewed the ITS TER and concurred with its findings.
3.0 CONCLUSION
The staff has reviewed the Topical Report DPC-NE-2003 and finds it acceptable for referencing in the Oconee reload thermal-hyaraulic analyses, subject to the following limitations:
(1) The validation analysis with limited CHF data has demonstrated that the approved DNBR limit of 1.18 for the BWC CHF correlation, which was derived with the LYNX2 thermal-hydraulic code, is conservative and acceptable for use with VIPRE-01. Acceptance of a DNBR limit less than 1.18 will require aralysis of broader CHF data base and detailed staff review.
(2) The studies provided in the topical report were performed with the Mark BZ fuel assembly design currently used in Oconee units. Though the approach described is acceptable for future fuel assembly designs, DPC should ensure that the selected correlations be used within their applicability ranges.
4.0 REFERENCES
- 1. Letter from H. B. Tucker (DPC) to USNRC Document Control Desk, "Oconee Nuclear Station, Docket Nos. 50-269, -270, -287, Oconee Nuclear Station Core Thermal-Hydraulic Methodology Using VIPRE-01, DPC-NE-2003," August 31, 1988.
- 2. Letter from H. B. Tucker (DPC) to USNRC Document Control Desk, "Oconee Nuclear Station, Docket Nos. 50-269, -270, -287, Topical Report DPC-NE-2003, 'Core Thermal-Hydraulic Methodology Using VIPRE-O'; Response to Request For Additional Information," May 3, 1989.
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- 3. EPRI-NP-2511-CCN-A, "VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores," 4 Volumes, Electric Power Research Institute.
- 4. BAW-10110, Rev. 1, "CHATA - Core Hydraulic and Thermal Analysis," May 1977.
- 5. BAW-10021, "TEMP - Thermal Enthalpy Mixing Program," April 1970.
Attachment:
Safety Evaluation Report on NFS-1001, "Oconee Nuclear Station Reload Design Methodology," July 29, 1981.
- 7. Letter from J. A. Blaisdell (Northeast Utilities Service Co.) to H. R.
Denton (USNRC), Subject related to UGRA submittal of the VIPRE-01 code, December 17, 1984.
- 8. Letter from C. E. Rossi (NRC) to J. A. Blaisdell, Chairman, UGRA Executive Committee, "Acceptance for Referencing of Licensing Topical Report, EPRI-NP-2511-CCM, 'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores,' Volumes 1, 2, 3 and 4," May 1, 1986.
- 9. Letter from H. B. Tucker (DPC) to USNRC Document Control Desk, "Oconee Nuclear Station, Docket Nos. 50-269, -270, -287, Response to Questions Regarding Differences Between Duke Topical Reports DPC-NE-2003 and DPC-NE-3000," June 19, 1989.
- 10. BAW-10143-A, "BWC Correlation of Critical Heat Flux," April 1985.
ATTACHME NT ITS/NRC/89-2 TECHNICAL EVALUATION OF THE CORE THERMAL-HYDRAULIC METHODOLOGY USING VIPRE-01 TECHNICAL REPORT DPC-NE-2003 FOR THE DUKE POWER COMPANY OCONEE NUCLEAR STATION
1.0 INTRODUCTION
In Duke Power Company (DPC) topical report DPC-NE-2003, dated August 1988 (Ref. 1),
DPC presented a description and qualification of their core thermal-hydraulic methodology using VIPRE-01 (Ref.
- 2) for steady-state and for two reactor coolant pump coastdown analyses of the Oconee Nuclear Station reload.
VIPRE-01 has been previously reviewed and approved for application to pressurized water reactor (PWR) plants in steady-state and transient analyses with heat transfer regimes up to critical heat flux. The NRC safety evaluation report (SER) on VIPRE-01 (Ref. 3) includes conditions requiring each user to document and submit 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, input values of plant specific data such as turbulent mixing coefficient and grid loss coefficient including defaults. This topical report was prepared to address these issues.
The purpose of this review was to assure conformity of the DPC topical report and supplemental information (Ref. 4, 5) to the VIPRE SER requirements, and to evaluate acceptability of DPC's intended use of the code as described in the report.
In the past DPC used (Ref.
- 6) CHATA, a closed-channel (no energy or mass interchange among assemblies) computer code for core-wide analysis, and TEMP 1
to determine the maximum permissible core power and distribution under various operating conditions for Oconee core thermal-hydraulic design and licensing analyses.
Although this approach was conservative, these codes were unable to realistically predict flow redistribution effects in an open lattice reactor core.
The VIPRE-01 computer code (Ref. 2) is an open-channel (permitting lateral communication among channels by diversion crossflow and turbulent mixing) thermal-hydraulic computer code developed to evaluate nuclear reactor core safety limits.
The code assumes the flow to be incompressible and homogeneous and incorporates models to reflect subcooled boiling and liquid/vapor slip.
The input data to the VIPRE-01 code are the geometry of the reactor core and coolant channels with thermal-hydraulic characteristics, and boundary conditions.
In addition, the user must select among certain correlations in the code for use in the particular analysis being performed.
The code calculates the core flow distributions, coolant conditions, fuel rod temperatures and the minimum departure frQm nucleate boiling ratio (MONBR).
The DPC submittal, in fulfillment of VIPRE SER (Ref. 3) conditions, contains DPC's geometric representation of the core, its selection of thermal hydraulic models and correlations, and a description of the methodology used for steady-state core reload design analysis and for a two-pump coastdown transient. These analyses are performed to determine the core thermal margin and acceptable safety and operating limits and to analyze a two-pump coastdown transient. It is not DPC's intent to use this methodology for FSAR Chapter 15 type licensing transient analysis.
2.0 EVALUATION Acceptability of DPC's application of the VIPRE-01 computer code for thermal hydraulic calculation of DNB for Oconee was evaluated with respect to the sensitivity of the computed steady-state operating conditions to input selection, nodalization, thermal-hydraulic modeling, and correlations, by examination of the overall conservatism in the results.
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2.1 CORE NODALIZATION 2.1.1 Radial Noding Sensitivity Since the VIPRE-01 code performs the thermal-hydraulic calculations simultaneously for all subchannels -(a single-pass approach) and permits flexibility in selection of channel sizes and shapes, a sensitivity study was performed to determine the sensitivity of predicted DNBR to the subchannel model sizes.
The modeling of the reactor core uses the 1/8-core symmetry in which the hot assembly is located at the center of the core.
The hot assembly includes the hot subchannel in which the minimum DNBR is expected to occur.
The thermal-hydraulic calculations were performed for three different core subchannel models; a 64 channel model, a 9 channel model, and an 8 channel model.
The 64 channel model consists of 36 subchannels representing the hot assembly and 28 subchannels individually modeling each of the remainder of assemblies in the 1/8-core segment.
In the 8 channel model, 6 subchannels around the hot subchannel in the hot assembly are modeled individually.
The rest of the subchannels in the hot assembly and the remaining 28 assemblies in the core are lumped into 2 individual subchannels (Channels 7 and 8).
The 9 channel model, developed for evaluation of transitional mixed core effects, includes an additional subchannel to account for the different fuel assembly designs in the transition core.
The nodalization sensitivity studies used the same thermal-hydraulic correlations and models which DPC intends to use in future reload licensing analysis. Review of the particular correlations and thermal-hydraulic models selected is provided in Section 2.2.
Steady-state and transient calculations using the previously approved RECIRC numerical solution option were performed using these three.different core models at four different operating conditions: the high temperature and the low pressure safety limits, and two different sets of initial conditions for pump coastdown transients including one representing the limiting MONBR case.
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Sensitivity to the core model size was studied by comparing the results of using the 64 and 8 channel models.
The 8 channel model was found to yield MDNBRs ranging from 0.44% to 2.2% lower than the 64 channel model.
We therefore find DPC's use of the 8 channel model acceptable for Oconee steady state and 2-pump coastdown reload thermal-hydraulic analysis.
Sensitivity of the core models to transitional mixed core effects was examined using the 9 and 64 channel models in both steady-state and 2-pump coastdown transient conditions.
For steady-state conditions, the 9 channel transition core model predicted 1.9% lower MDNBR than the 64 channel model.
For the transient analysis the MDNBR predicted by the 9 channel model was 1.6% lower than the 64 channel model.
Based upon these sensitivity studies, DPC intends to used the 9 channel model for steady-state and pump coastdown reload analyses involving transition cores of the Oconee Nuclear Station.
2.1.2 Axial Noding Sensitivity A steady-state sensitivity analysis for axial node length was performed with the 8 channel model using two sizes: a 3-inch node length applied uniformly and a 2-inch node length applied where DNB is expected to occur. The results indicated that the 3-inch axial nodes produced slightly more conservative MDNBR than did the 2-inch nodes.
We, therefore, find that use of 3-inch uniformly spaced axial nodes is acceptable for Oconee reload steady-state and pump coastdown thermal-hydraulic analyses.
2.2 VIPRE-01 Input 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.
2.2.1 Active Fuel Length Since power is distributed over the length of the active fuel, a shorter length yields higher power density, causing greater heat flux and is 4
therefore conservative. DPC's choice for the active fuel length as described in Section 5.2 of Ref. 1 is conservative when compared to hot conditions.
When a different assumption is used, DPC should justify its conservatism.
2.2.2 Centroid Distance and Effective Crossflow Gaps The centroidal distance is used in the crossflow momentum equation to determine the lateral pressure gradient.
The gap width is used in determination of the crossflow area.
DPC calculates these parameters from channel geometry following the code's prescription.
2.2.3 Spacer Grid Form Coefficients Pressure losses across the spacer grids impact the axial pressure distribution and therefore the axial location of DNB.
The spacer grid form loss coefficients were obtained from a full size fuel assembly test conducted by the vendor (B&W).
To determine the. individual subchannel form loss coefficient, DPC stated (Ref. 4) in response to our question that the vendor used its computer code, GRIL.
The input data to the GRIL code are the individual subchannel geometry, drag areas and coefficients, and the coolant information.. From this input, the code calculates individual subchannel loss coefficients, an overall grid loss coefficient and subchannel velocities based on single-phase flow input data by a iterative process. The calculated overall grid loss coefficient is matched with the measured value by adjusting the velocity field in the subchannel until consistency between the measured and predicted values is achieved.
DPC has stated that the calculated velocity profiles were compared by the vendor with the experimental data and showed good agreement (Ref. 4).
2.2.5 Core Bypass Flow DNB is influenced by the aggregate flow rate past the location being
- examined, and therefore by the core bypass 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 5
flow data used in the analysis should be based on a bounding value or on a cycle specific data.
2.2.6 Inlet Flow Distribution CHF is decreased and the probability of DNB is enhanced if flowrate is reduced due to a flow maldistribution.
The use of 5% inlet flow maldistribution to the hot assembly with all four reactor coolant pumps operating yielded slightly more conservative results than a uniform inlet flow distribution. This value is supported by a B&W 1/6-scale Vessel Model Flow Test and was previously approved for Oconee reload analysis (Ref. 5).
For operation with less than four reactor coolant pumps operating, more restrictive flow reduction factors are applied.
2.2.7 Flow Area Reduction Factor DPC reduced the hot subchannel flow area..by a factor as stated in Section 5.12 of Ref. 1 to account for variations in as-built subchannel coolant flow area.
2.2.8 Reference Design Power Distribution The reference design power distribution was developed using a radial-local hot pin peak of 1.714 which has been.previously approved for Oconee reload analysis (Ref. 5, 6).
The corresponding assembly power was 1.6147.
2.2.9 Axial Power Distribution The axial power shape used in the analyses was a chopped cosine shape with a conservatively determined peaking factor. Although the axial power shape is cycle specific and transient dependent, the use of generic bounding axial power curves accounts for the effect on DNB of different axial shapes. This is discussed in Section 2.4.
DPC added an optional new routine to VIPRE-01 to generate the axial power 6
shapes using a generalized power function.
The currently defined function can generate both symmetric and skewed power shapes but cannot generate certain power shapes (such as double peaked) because of limitations of the generalized function used.
The axial power shapes calculated using this routine agreed with the symmetric axial shapes calculated using VIPRE-01 symmetric cosine routine for axial peaks of 1.2 and 1.5 (Ref. 4).
DPC intends to maintain two options for power shape generation: one is to use this routine and the other is to use a user specified table. The use of this routine is acceptable so long as the computed power shapes represent the true power shapes to be analyzed.
Although analyses in this report were performed using a higher axial peaking factor, DPC will continue to use the reference axial peaking factor consistent with the current FSAR Chapter 15 transient analysis in the reload licensing analysis (Ref. 5).
2.2.10 Hot Channel Factor The power factor, Fq, used to account for variations in average pin power caused by differences in the fuel loading per rod was selected to be 1.0107 which has been previously approved for Oconee reload analysis (Ref. 6).
The local heat flux factor, Fq" used to account for the uncertainty in the manufacturing tolerances was selected to be 1.0137. In the determination of the maximum allowable peaking limits, two additional factors were used to increase the limit to 1.0371.
These factors were 1.007 to account for power spikes occurring as a result of the flux depressions at the spacer grids, and 1.016 to account for axial nuclear uncertainty (Ref.
6).
All of these factors have been previously approved for Oconee reload analysis.
2.3 VIPRE-01 Correlations VIPRE-01 requires empirical correlations for the following models:
- a. turbulent mixing 7
- b. friction pressure loss
- c. two-phase flow correlations (subcooled and saturated void, and void-quality relation)
- d. single-phase forced convection
- e. nucleate boiling heat transfer
- f.
critical heat flux 2.3.1 Friction Pressure Loss, Subcooled Void, Single-Phase and Two-Phase Flow Correlations For-single-phase turbulent flow the Blasius smooth tube friction factor, a default option in VIPRE-01, will be used to calculate the friction pressure loss in the axial direction.
Crossflow resistance has a minimal effect on MDNBR in transients where axial flow dominates.
DPC's selection therefore has an inherent assumption of axial flow dominance.
This choice is acceptable since we agree that in the analyses to be performed in the context of this topical report, the flows are expected to be axially dominant.
For two-phase flow, subcooled and bulk void correlations, a sensitivity study using six different combinations of three subcooled and bulk void correlations was performed for two operating conditions.
The results indicated that the use of Levy subcooled void and Zuber-Findlay bulk void correlations, in conjunction with EPRI two-phase friction multiplier results in conservatively predicted DNBR relative to other combinations of correlations. DPC intends to use this combination in Oconee steady-state and pump coastdown reload analysis.
This is consistent with the VIPRE-01 SER findings.
2.3.2 Turbulent Mixing The lateral momentum equation requires two parameters: a turbulent momentum factor and a turbulent mixing coefficient.
The turbulent momentum factor (FTM) describes the efficiency of the momentum 8
mixing: 0.0 indicating that crossflow mixes enthalpy only; 1.0 indicating that crossflow mixes enthalpy and momentum at the same strength.
A sensitivity study using the 8 channel model was performed for two operating conditions and for three different values of FTM of 0.0, 0.8, and 1.0 and found little sensitivity in DNBR by different values of FTM.
Conservative DNBR's were obtained with zero (Table 5-4 in Ref.
1).
However, in reality there will be always some momentum mixing.
An FTM of 0.8 has been recommended by the VIPRE-01 code developer.
Since the turbulent mixing coefficient determines the flow mixing rate, it is an important parameter. Based upon tests using a 5x5 heated bundle conducted by B&W, where the subchannel exit temperatures were measured, a mixing coefficient was conservatively determined for B&W Mark-B fuel (Ref. 4).
This will be used in the Oconee core steady-state and pump coastdown reload thermal-hydraulic analysis (Ref. 1),
2.3.3 Single-Phased Forced Convection, Nucleate Boiling Heat Transfer DPC will use (for its steady-state and pump coastdown analyses) the default EPRI single-phased forced convection correlation and Thom subcooled and saturated nucleate boiling correlations, both of which were found to result in conservative MDNBR for the two-pump coastdown transient.
2.3.4 BWC Critical Heat Flux Correlation The BWC correlation (Ref. 7) was originally developed for 17x17 Mark-C fuel, and later used for 15x15 Zr grid Mark-BZ fuel.
The use of BWC correlation with the LYNX2 code (Ref. 8) for 15x15 Zr grid Mark-BZ fuel was previously approved by NRC with a design limit of 1.18 (Ref. 8, 9).
All Oconee thermal-hydraulic analyses using VIPRE-01 and the BWC correlation will use a design limit of 1.18. Since the BWC correlation is now being used with VIPRE-01, it is necessary for DPC to demonstrate that the DNBR limit of 1.18 for BWC CHF correlation used in VIPRE-01 can predict its date base of DNB occurrence with at least a 95% probability and a 95% confidence level.
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In Section 5.13 of the topical report, DPC performed validation using more than 200 data point.
Results show a 95%/95% limit of 1.16.
Therefore use with VIPRE-01 of the previously approved (with LYNX2) value of 1.18 is conservative and acceptable. DPC agreed that when a lower DNBR limit becomes desirable with use of BWC CHF correlation with VIPRE-01, it will submit a separate topical report documenting analysis based on a broader CHF database for detailed NRC review and approval.
2.4 Oconee Thermal-Hydraulic Analyses Using the input, assumptions, and-thermal-hydraulic correlations selected and justified in the subject topical report, DPC discussed its methodology to perform steady-state and generic two-pump coastdown analyses necessary to define the core thermal margin or safety limits and acceptable operating 1 imits-.
The core safety limits that provide DNB protection are pressure - temperature (P-T) envelope and power - power imbalance limits. The P-T envelope defines a region of allowable operation in terms of reactor coolant system;.pressure and coolant temperature (Ref. 6).
To ensure that the P-T envelope provides adequate DNB protection, P-T curves are determined for different numbers of RC pump operation.
P-T curves are the combinations of RCS pressure and vessel outlet temperature that yield the design DNBR limit or the BWC correlation quality limit.
The P-T envelope must be more restrictive than the most limiting P-T conditions. VIPRE-01 was used to generate the generic P-T curves using the 8 channel model.
The following are input to the code for generation of P-T curves:
- 1. a symmetric chopped cosine with a conservative axial peaking factor;
- 2. 112 % of full power for 4-pump operation, and the power level for other modes of pump operation are based on trip setpoint plus margin for uncertainties;
- 3. 104% of design RCS flow for 4 pumps; appropriately lower for less 10
than 4-pump operation;
- 4. minimum coolant temperature; and
- 5. generic maximum allowable peaking (MAP) limit curves.
Having developed the P-T curves,
- DPC, as part of its reload analysis, performs a two-pump coastdown transient to determine the flux/flow trip setpoint.
This trip provides DNB protection during a loss of one or more reactor coolant pumps.
For this 2-pump coastdown analyses, the input to the fuel rod heat conduction model in VIPRE were determined by sensitivity studies evaluating impact of pellet/clad gap, gas composition and pellet radial power profile to the DNBR.
Results led to a conservative set of eight fuel parameters for the conduction model input.
The methodology described in the report is acceptable.
3.0 CONCLUSION
S We find that the subject topical report, together with DPC responses, contains sufficient information to satisfy the VIPRE-01 SER requirement that 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 reload stead-state and pump coastdown analyses.
We further find that the manner in which the code is to be used for such analyses, selection of nodalization, models, and correlations provides, except as limited below, adequate assurances of conservative results and is therefore acceptable.
The following items are limitations regarding application of DPC-NE-2003:
- 1. An MDNBR limit of less than 1.18 with the BWC CHF correlation, as described in Section 5.13 of DPC-NE-2003, requires further justification based on broader CHF database for detailed review.
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- 2.
Studies presented in this report are performed using design data for Mark-BZ fuel assemblies, which are currently used in Oconee.
Although the approach described in this report is acceptable, for future analysis of reloads which incorporate other fuel, DPC should assure that the VIPRE-01 computer code be used within the range of applicability.
- 3. The scope of this review and the applicability of findings are limited to DPC's use of VIPRE-01 for core reload steady-state andsa two-pump coastdown transient analyses.
4.0 REFERENCES
- 1. "Core Thermal-Hydraulic Methodology Using VIPRE-01,"
DPC-NE-2003, August 1988.
- 2. "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores, EPRI NP-2511-CCM Revision 2, EPRI, July 1985.
- 3. Letter from C.E.
Rossi (NRC) to J.A. Blaisdell (UGRA),
(Transmittal of VIPRE-01 Safety Evaluation Report), May 1, 1986.
- 4. Letter from H.B.
Tucker (DPC) to USNRC, "Response to the Request for Additional Information," May 3, 1989.
- 5. Letter from H.B. Tucker (DPC) to USNRC, "Response to Questions Regarding Differences Between Duke Topical Reports DPC-NE-2003 and DPC-NE-3000,"
June 19, 1989.
- 6. "Duke Power Company Oconee Nuclear Station Reload Design Methodology,"
DPC-NE-1001A, Rev. 4, April 1981.
- 7. "BWC Correlation of Critical Heat Flux," BAW-10143P-A, April 1985.
- 8. "LYNX2-Subchannel Thermal-Hydraulic Analysis Program,"
BAW-10130A, October, 1976.
- 9. "Duke Power Company Oconee Nuclear Station Reload Design Methodology II," DPC-NE-1002A, October 1985.
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