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Category:TEXT-SAFETY REPORT
MONTHYEARML20216G0111999-09-30030 September 1999 Year 2000 Readiness in U.S. Nuclear Power Plants ML20206N2191999-04-30030 April 1999 Operator Licensing Examination Standards for Power Reactors ML20205A5291999-03-31031 March 1999 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,October-December 1998.(White Book) ML20211K2851999-03-31031 March 1999 Standard Review Plan on Power Reactor Licensee Financial Qualifications and Decommissioning Funding Assurance ML20205A5991999-03-31031 March 1999 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,July-September 1998.(White Book) ML17313A7791999-02-0505 February 1999 Safety Evaluation Accepting Licensee Rev to Emergency Plan That Would Result in Two Less Radiation Protection Positions Immediatelu Available During Emergencies ML20203D0541999-01-31031 January 1999 Fire Barrier Penetration Seals in Nuclear Power Plants ML20155A9281998-10-31031 October 1998 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,April-June 1998.(White Book) ML20154C2081998-09-30030 September 1998 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,January-March 1998.(White Book) ML15261A4681998-09-0404 September 1998 Safety Evaluation Supporting Amends 232,232 & 231 to Licenses DPR-38,DPR-47 & DPR-55,respectively ML20203A1521998-07-31031 July 1998 Assessment of the Use of Potassium Iodide (Ki) as a Public Protective Action During Severe Reactor Accidents.Draft Report for Comment ML20153D3371998-07-31031 July 1998 Assessment of the Use of Potassium Iodide (Ki) as a Public Protective Action During Severe Reactor Accidents.Draft Report for Comment ML20236S9771998-06-30030 June 1998 Knowledge and Abilities Catalog for Nuclear Power Plant Operators.Pressurized Water Reactors ML20236S9681998-06-30030 June 1998 Evaluation of AP600 Containment THERMAL-HYDRAULIC Performance ML20236S9591998-06-30030 June 1998 Knowledge and Abilities Catalog for Nuclear Power Plant Operators.Boiling Water Reactors ML20217Q7971998-05-0404 May 1998 Safety Evaluation Supporting Amends 227 & 201 to Licenses DPR-53 & DPR-69,respectively ML20247E3951998-04-30030 April 1998 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,October-December 1997.(White Book) ML20217F3801998-03-31031 March 1998 Risk Assessment of Severe ACCIDENT-INDUCED Steam Generator Tube Rupture ML20202J3051997-11-30030 November 1997 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,July-September 1997.(White Book) ML20197B0431997-11-30030 November 1997 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,April-June 1997.(White Book) ML20211L2931997-09-30030 September 1997 Aging Management of Nuclear Power Plant Containments for License Renewal ML20210K7801997-08-31031 August 1997 Topical Report Review Status ML20149G9431997-07-31031 July 1997 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,January-March 1997.(White Book) ML20210R2131997-05-31031 May 1997 Final Safety Evaluation Report Related to the Certification of the System 80+ Design.Docket No. 52-002.(Asea Brown Boveri-Combustion Engineering) ML20140F0801997-05-31031 May 1997 Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design.Supplement No. 1.Docket No. 52-001.(General Electric Nuclear Energy) ML20140J4301997-05-31031 May 1997 Safety Evaluation Report Related to the Department of Energy'S Proposal for the Irradiation of Lead Test Assemblies Containing TRITIUM-PRODUCING Burnable Absorber Rods in Commercial LIGHT-WATER Reactors ML20141J9391997-04-30030 April 1997 Safety Evaluation Report Related to the Renewal of the Operating License for the Research Reactor at North Carolina State University ML20141C2411997-04-30030 April 1997 Circumferential Cracking of Steam Generator Tubes ML20141A5791997-04-30030 April 1997 Proposed Regulatory Guidance Related to Implementation of 10 CFR 50.59 (Changes, Tests, or Experiments).Draft Report for Comment ML20137A2191997-03-31031 March 1997 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,October-December 1996.(White Book) ML20134L3601997-01-31031 January 1997 Standard Review Plan on Antitrust.Draft Report for Comment ML20134L3631997-01-31031 January 1997 Standard Review Plan on Power Reactor Licensee Financial Qualifications and Decommissioning Funding Assurance.Draft Report for Comment ML20138J2461997-01-31031 January 1997 Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,July-September 1996.(White Book) ML20135D5711997-01-31031 January 1997 Operator Licensing Examination Standards for Power Reactors ML20133E9161996-12-31031 December 1996 License Renewal Demonstration Program: NRC Observations and Lessons Learned ML20149L8261996-10-31031 October 1996 Reactor Pressure Vessel Status Report ML20135A4981996-10-31031 October 1996 Historical Data Summary of the Systematic Assessment of Licensee Performance ML20128P4381996-10-0909 October 1996 Safety Evaluation Accepting Review of Cracked Weld Operability Calculations & Staff Response to NRC Task Interference Agreement ML20107F5611996-04-17017 April 1996 Safety Evaluation Providing Guidance on Submitting plant- Specific Info W/Respect to IST Program Alternatives Request ML14183A6951995-09-18018 September 1995 Safety Evaluation Approving Relocation of Technical Support Ctr ML20236L5971994-12-29029 December 1994 SER in Response to 940314 TIA 94-012 Requesting NRR Staff to Determine Specific Mod to Keowee Emergency Power Supply Logic Must Be Reviewed by Staff Prior to Implementation of Mod ML20128Q0761994-11-0404 November 1994 Coordinating Group Evaluation,Conclusions & Recommendations ML20149H0671994-11-0404 November 1994 Safety Evaluation Supporting Amend 27 to Amended License R-103 ML20149G4281994-09-28028 September 1994 NRC Perspectives on Accident Mgt, Presented at 940928 Severe Accident Mgt Implementation Workshop in Alexandria, VA ML20149F7581994-08-25025 August 1994 Topical Rept Evaluation of WCAP-13864,Rev 1, Rod Control Sys Evaluation Program ML20149F4151994-08-0404 August 1994 Safety Evaluation Concluding That Unit 1 Can Be Safely Operated During Next Operating Cycle (Cycle 14) ML20149E8831994-08-0202 August 1994 Safety Evaluation Accepting Interim Relief Request IRR-03 Re Drywell Isolation Check Valves in Equipment Drain Lines & Reactor Equipment Closed Cooling Water Sys ML20248C5731994-07-19019 July 1994 SER Step 1 Review of Individual Plant Exam of External Fire Events for Millstone Unit 3 ML20059J4591994-01-25025 January 1994 Safety Evaluation Supporting Request for Relief from ASME Code Re Inservice Testing Requirements to Measure Vibration Amplitude Displacement ML20059H4991994-01-24024 January 1994 Safety Evaluation Accepting Revised Responses to IEB-80-04 Re MSLB Reanalysis 1999-09-30
[Table view]Some use of "" in your query was not closed by a matching "". Category:TOPICAL REPORT EVALUATION
MONTHYEARML20149F7581994-08-25025 August 1994 Topical Rept Evaluation of WCAP-13864,Rev 1, Rod Control Sys Evaluation Program ML20059L1061994-01-12012 January 1994 Draft Topical Rept Evaluation of B&Wog Rept 47-1223141-00, Integrated Plant Assessment Sys/Structure Screening.... Applicant for License Renewal That Refs B&Wog Sys Screening Methodology Will Be Required to Develop Own Procedures ML20059D1911993-12-30030 December 1993 Topical Rept Evaluation of RXE-91-005, Methodology for Reactor Core Response to Steamline Break Events ML20058P2181993-12-10010 December 1993 SER Accepting Siemens Nuclear Power Corp Submittal of Topical Rept EMF-92-081, Statistical Setpoint/Transient Methodology for W Type Reactors ML20058H9851993-11-26026 November 1993 Topical Rept Evaluation of WCAP-10216-P, Relaxation of Constant Axial Offset Control. Rept Acceptable ML20059H8481993-11-0202 November 1993 SER Accepting Proposed Changes in Rev 3 to OPPD-NA-8302-P, OPPD Nuclear Analysis,Reload Core Analysis Methodology, Neutronics Design Methods & Verification ML20134B4761993-10-30030 October 1993 Topical Rept Evaluation of Rev 3 to NP-2511-CCM Re VIPRE-01 Mod 2 for PWR & BWR Applications ML20058M9851993-09-30030 September 1993 SE of Topical Rept, Transient Analysis Methodology for Wolf Creek Generating Station ML20056G4171993-08-18018 August 1993 Topical Rept Evaluation of Rev 4 to OPPD-NA-8303, Transient & Accident Methods & Verification. Proposed Changes in Rev 4 Acceptable Except for Use of Cents Computer Code for Transient Analyses ML20056E9661993-08-0606 August 1993 Sser Re Topical Rept HGN-112-NP, Generic Hydrogen Control Info for BWR/6 Mark III Containment Hydrogen Control ML20056E3811993-08-0505 August 1993 Safety Evaluation of RXE-89-002, Vipre-01 Core Thermal- Hydraulic Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications. Rept Is Acceptable for Ref in CPSES Core thermal-hydraulic Analyses ML20056E3961993-08-0505 August 1993 Safety Evaluation of RXE-90-006-P, Power Distribution Control Analysis & Overtemperature N-16 & Overpower N-16 Trip Setpoint Methodology. Methodology Acceptable ML20056E4681993-08-0505 August 1993 Supplemental Safety Evaluation for Topical Rept HGN-112-NP, Generic Hydrogen Control Info for BWR/6 Mark III Containments. Change Requests Consistent & Compatible W/ 10CF50.44 & Acceptable ML20056E2571993-08-0505 August 1993 Corrected Safety Evaluation for Topical Rept RXE-91-001, Transient Analysis Methods for Commanche Peak Steam Electric Station Licensing Applications. Corrections Made to Second Sentense of Second Full Paragraph on Page Two ML20056D9921993-07-29029 July 1993 Topical Rept Evaluation of OPPD-NA-8301,Rev 5, Reload Core Analysis Methodology Overview. Proposed Changes in Rev 5 Acceptable ML20057A2661993-07-14014 July 1993 Topical Safety Evaluation of CEN-387-P, Pressurizer Surge Line Flow Stratification Evaluation. C-E Owners Group Analysis May Be Used as Basis for Licensees to Update plant- Specific Code Stress Rept for Compliance W/Bulletin 88-011 ML20056E1261993-06-29029 June 1993 Safety Evaluation of CENPD-382-P, Methodology for Core Designs Containing Erbium Burnable Absorbers. Rept Acceptable for Reload Licensing Applications of Both CE CE 14x14 & 16x16 PWR Lattice Type Core Designs ML20057B5431993-06-26026 June 1993 Errata for Sser Re Topical Rept HGN-112-NP, Generic Hydrogen Control Info for BWR/6 Mark III Containments, for Use in Issuance of Final Approved Version of Topical Rept ML20128B8101993-01-19019 January 1993 Safety Evaluation Accepting Methodology Described in Topical Rept RXE-91-002 Reactivity Anomaly Events Methodology for Reload Licensing Analyses for CPSES ML20126E0381992-12-0909 December 1992 Safety Evaluation Accepting Topical Rept NEDC-31753P W/Ter Recommendations W/Listed Exceptions ML20056D9351991-01-11011 January 1991 Topical Rept Evaluation Accepting Proposed Methodology for Fuel Channel Bowing Anaylses & for Referencing in Reload Licensing Applications W/Listed Conditions ML20235Q7121989-02-22022 February 1989 Safety Evaluation Re Review of WCAP-10271,Suppl 2 & WCAP-10271,Suppl 2,Rev 1 on Evaluation of Surveillance Frequencies & out-of-svc Times for ESFAS ML20206L9611988-11-23023 November 1988 Topical Rept Evaluation of PECO-FMS-0004, Methods for Performing BWR Sys Transient Analysis. Rept Approved,But Limited to Util Competence to Use Retran Computer Code for Facility ML20205M0091988-10-25025 October 1988 Safety Evaluation of Topical Rept YAEC-1300P, RELAP5YA: Computer Program for LWR Sys Thermal-Hydraulic Analysis. Program Acceptable as Licensing Method for Small Break LOCA Analysis Under Conditions Stipulated ML20204G8371988-10-18018 October 1988 Safety Evaluation Accepting Topical Rept 151, Haddem Neck Plant Non-LOCA Transient Analysis, Except for Issue of Feedwater Event ML20155G7991988-10-12012 October 1988 Topical Rept Evaluation of TR-045, BWR-2 Transient Analysis Using Retran Code. Methods Described in Rept Acceptable for Reload Analysis When Listed Conditions Satisfied ML20155G3201988-09-26026 September 1988 Safety Evaluation of TS NEDC-30936P, BWR Owners Group TSs Improvement Methodology. GE Analyses Demonstrated Acceptability of General Methodology for Considering TS Changes to ECCS Instrumentation Used in BWR Facilities ML20155B0501988-09-22022 September 1988 Topical Rept Evaluation of Suppl 1 to NEDC-30851P, Tech Spec Improvement Analysis for BWR Control Rod Block Instrumentation. Analyses Acceptable to Support Proposed Extensions to 3 Months ML20151K9921988-07-26026 July 1988 Topical Rept Evaluation of Nusco 140-1 Northeast Utils Thermal Hydraulic Model Qualification,Vol 1 (Retran). Rept May Be Generally Ref in Future Licensing Submittals.Further Justification by Util Required ML20150D9651988-03-21021 March 1988 Topical Rept Evaluation of Rev 0 to TR-040, Steady State & Quasi-Steady State Methods for Analyzing Accidents & Transients. Util Methods Acceptable for Performing Reload Assembly Mislocation Analysis W/Listed Exceptions ML20150D7671988-03-21021 March 1988 Topical Rept Evaluation of Rev 0 to TR-033, Methods for Generation of Core Genetics Data for RETRAN-02. Uncertainties in Input Parameters & Impact on Retran Results Should Be Determined for Qualification of Model ML20236D2621987-10-21021 October 1987 Topical Rept Evaluation of CEN-348(B)-P, Extended Statistical Combination of Uncertainties. Rept Acceptable ML20235D7041987-09-22022 September 1987 Safety Evaluation of Rev 0 to Topical Rept TR-021, Methods for Analysis of BWRs Steady State Physics. Rept,Methodology & Util Use of Methodology Acceptable ML20239A5461987-09-0909 September 1987 Safety Evaluation Supporting A-85-11, Retran Computer Code Reactor Sys Transient Analysis Model Qualification for Use in Performing plant-specific best-estimate Transient Analyses at Plant ML20215M3591987-05-0606 May 1987 Safety Evaluation Supporting Util Use of Suppl 1 to MSS-NA1-P, Qualification of Reactor Physics Methods for Application to PWRs of Middle South Utils Sys ML20212M7781987-02-17017 February 1987 Topical Rept Evaluation of WCAP-10325, Westinghouse LOCA Mass & Energy Release Model for Containment Design - Mar 1979 Version. Rept Acceptable for Ref in Licensing Actions ML20210N7331987-02-0404 February 1987 Safety Evaluation Supporting CEN-161(B)-P,Suppl 1-P, Improvements to Fuel Evaluation Model. Mods to Fission Gas Release & Fuel Thermal Expansion Models Acceptable ML20215B2331986-12-0404 December 1986 Corrected Page 1 to 861031 Topical Rept Evaluation of Rev 2 to STD-R-05-011, Mobile In-Container Dewatering & Solidification Sys (Mdss). Word Effective Inserted Before Words Pore Sizes in First Line of 4th Paragraph ML20214C5221986-11-14014 November 1986 Topical Rept Evaluation of Rev 0 to TR 020, Methods for Analysis of BWR Lattice Physics. Collision Probability Module Code Acceptable for BWR Fuel Lattice Calculations ML20213F6531986-11-10010 November 1986 Safety Evaluation of Rev 2 to Vol 3 of XN-NF-80-19(P), Exxon Nuclear Methodology for Bwrs,Thermex:Thermal Limits Methodology Summary Description. Rept Acceptable for Ref in Licensing Applications ML20207A8281986-11-0505 November 1986 Suppl 3 to Topical Rept Evaluation Re Submittal 2 to Rev 3 to CEN-152, C-E Emergency Procedure Guidelines. Rept Acceptable for Ref ML20215N6901986-11-0404 November 1986 Topical Rept Evaluation of BAW-10155, FOAM2 - Computer Program to Calculate Core Swell Level & Mass Flow Rate During Small-Break Loca. Rept Acceptable W/Listed Restrictions Re Ranges of Core Flow Rate & Pressure ML20215N3921986-10-31031 October 1986 Topical Rept Evaluation of STD-R-05-011, Mobile In-Container Dewatering & Solidification Sys (Mdss). Rept Acceptable for Ref in License Applications ML20211D6921986-10-16016 October 1986 Safety Evaluation of Nusco 140-2, Nusco Thermal Hydraulic Model Qualification,Vol II (Vipre). Rept Acceptable for Establishing Input Values & Selection of Correlation Options & Solution Techniques for Calculations ML20206S6511986-09-15015 September 1986 Topical Rept Evaluation of Addenda 3 to WCAP-8720, Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations/Application for BWR Fuel Analysis. Rept Acceptable for Ref in Licensing Applications ML20212N2001986-07-23023 July 1986 Topical Rept Evaluation of Rev 1 to XN-NF-85-67 (P), Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel. Rept Acceptable as Ref for Application to Jet Pump BWR Reload Cores,W/Listed Conditions 1994-08-25
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- ENCLOSURE SAFETY EVALUATION REPORT'ON THE WESTINGHOUSE CONDOR CODE FOR THERMAL-HYCRAULIC ANALYSES OF BOILING WATER REACTORS (TACS 48567)
Report Number: WCAP-10107(Proprietary)
Report
Title:
CONDOR-A Thermal-Hydraulic Code for Boiling Water Reactors Report Date: June 1982
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Originating Organization: Westinghouse Reviewed By: Core Performance Branch / Pacific Northwest Laboratory
1.0 INTRODUCTION
In WCAP-10107 (Refs. I and 2), Westinghouse presented the CONDOR computer code which will be used to calculate the axial distribution of coolant flow, enthalpy, pressure, void fraction and other associated parameters in a BWR core under steady-state conditions.
The CONDOR code solves the steady-state conservation equations of mass, momentum, and energy in one dimensional channels. Since it only' solves the steady-state equations, it can not calculate the thermal-hydraulic conditions during transients. The code takes the total core flow, reactor power, inlet enthalpy, system pressure, axial and radial power distributions, core geometries and component pressure loss coefficients as inputs for the boundary conditions to the numerical solution, and then iterates on the flow distribution among the channels until a converged solution is obtained. The constitutive relations for the void fractions, friction losses, two-phase pressure drop and bypass flow are also discussed in the report (WCAP-10107).
$k c
k l
I
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.s 2.0 STAFF REVIEW AND EVALUATION .
The objective of this review is to evaluate the validity and applicability of the CONDOR code for use in thermal-hydraulic calculations for licensing applications. The review includes the evaluation of the hydraulic models, numerical schemes, code verification and assessment of the limitations arising from modeling assumptions.
_. The results of our review are discussed in the following sections.
2.1 HYDRAULIC MODELS 2.1.1 Conservation Equations The mass continuity used in CONDOR is satisfied by having the sum of the flows from all core channels equal to the total flow to the core. No cross flow between heated channels in the core is assumed. The energy balance is achieved by requiring that the energy released to a given channel axial node divided by the flow rate is equal to the enthalpy rise for that channel node.
The heat transfer across the fuel channel wall from the active flow to the bypass flow is accounted for in the energy balance. We find that the approaches used for mass and energy conservation in the code'are conventional for modeling a BWR core and are acceptable. However, during the course of the review we requested that a sensitivity study be performed to establish the accuracy in the calculation of the total pressure drop and the enthalpy rise with respect to the axial node size.
In response, Westinghouse submitted the results of a sensitivity study (Ref. 8) indicating that the changes in the calculated total pressure and minimum critical power ratio are small (less than 0.26% and 0.9%, respectively) when the number of axial nodes is reduced from 24 nodes to 12 nodes. Westinghouse indicates that since the CONDOR code used 24 nodes for code verification, this same nodal scheme will be used for licensing calculations. Westinghouse also agrees that 2
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if a different number of nodes is used, additional calculations will be performed to identify the uncertainties associated with the change of the number of nodes. We find that the Westinghouse responses adequately address the concern related to the sensitivity of number of nodes on the CONDOR calculations and are acceptable.
The CONDOR axial momentum equation (Equation 3.1) expresses the total axial pressure drop in terms of the pressure drops in elevation, acceleration,
. friction, and local form loss. We find that axial area variations are not considered in the acceleration term. However, this will not result in any deviation in channel pressure drcp and flow calculation because of the absence of flow exchange between channels in a BWR core.
In summary, the steady-state conservation equations of mass, momentum, and energy given in the report (WCAP-10107) are fonnulated correctly for the intended use of the code and are acceptable.
2.1.2 Bypass Flow Models The bypass flow is not solved through the rigorous differential momentum equations. Instead, an empirical correlation (Equation 3.8) establishing the bypass flow in terms of the pressure differences across the leakage paths is used as follows:
C W= C1+ cap2 1/2 + C3a P 4, This approach is acceptable for the steady-state calculations of CONDOR when the dominating driving force for the bypass flow is the pressure differences across the leakage paths. However, the accuracy of using this approach depends on the selection of the constants C , C , C , and C . These constants 1 2 3 4 have to be determined through comparison with experimental data for the prototypical assemblies with bypass geometries. Therefore, we require that the determination of these constants be made through comparison with appropriate 3
m _ _ _ _
test data on a plant-specific basis. Factors affecting these values, such as the crud buildup, geometry changes due to irradiation, and mixed core due to fuel designs provided by different vendors, have to be considered.
With the coefficients Cy to C carefully selected and justified through 4
comparison with appropriate test data, we conclude that the bypass flow model used by the COND0R code is acceptable.
2.1.3 Yoid Models The void model in both subcooled and bulk boiling regions is based on a modifi-cation of Zuber's drift flux model. The formulation is the Zuber model (Ref. 3) with coefficients adjusted to give agreement with test data. In the subcooled
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region, the voids are predicted using the Zuber model in conjunction with the effective flow quality for the subcooled region obtained from the Levy correlation (Ref. 4). These models give acceptable results for void fractions for both sub-cooled and bulk boiling. This was verified by the comparisons of the void models with five FRIGG data sets discussed in the topical report.
We conclude that the void models are acceptable.
2.1.4 Pressure Drop Correlations The single ~ phase friction factor, f, is based on the Blasius equation. The multiplier for the two-phase pressure drop is based on the modified Baroczy (Ref. 5) and Chisholm (Ref. 6) correlations. Data comparison by Westinghouse with FRIGG experiments showed that these correlations are satisfactory.
The single and two-phase form losses are based on a multiplier, k, to the velocity head and a homogeneous two-phase multiplier to k. These models are adequate and acceptable. Westinghouse indicates that in the actual coding of CONDOR, another factor is multiplied to k to account for the effects of crud buildup at the grid spacers on pressure drop. As a result of our review we 4
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require that the selection of the form loss coefficients with the effects of the crud buildup, geometry change due to irradiation, different fuel designs by different vendors, be considered on a plant-specific basis.
In sumary, the pressure drop correlations in the CONDOR code are acceptable provided that the multiplier accounting for the effects of crud buildup and geometry change on the pressure drop are considered on a plant-specific basis.
2.2 NUMERICAL SOLUTION TECHNIQUES In the CONDOR solution, the core and bypass regions are divided into a given number of coolant channels with given inlet conditions. The power to each channel is given. Each channel is axially subdivided into a number of nodes.
The code then iterates on the flow distributions among the channels until a converged solution is obtained.
We conclude that the solution scheme is a common scheme for BWR core flow calculations and is, therefore, acceptable.
2.3 CODE VERIFICATION CONDOR predictions were compared with two analytical solutions. One was for the case of homogeneous equilibrium two-phase flow in a vertical heated tube with uniform heat flux. The other is for the case with a cosine-shaped heat flux. To do this, the void fraction expression and two-phase friction multiplier expression used in the analytical model were inserted into the CONDOR code.
There is a small difference between the analytical solution and the COND0R evaluations. This difference results from the different calculational methods:
CONDOR evaluates the local steam / water properties using the local coolant pressure of the calculational nodes whereas the analytic solutions are based on evaluation of these properties at a given constant system pressure. The effect of this difference in calculation is insignificant 1y small at typical BWR steady-state operational conditions, because the pressure variations along a channel are usually very small when compared to a system pressure of 1000 psia.
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The good comparison between the analytical solutions and the CONDOR predictions shows that the numerical solution in the code is correct in predicting the total and component pressure drop for a given inlet flow rate.
In addition to the comparison with analytical solutions, CONDOR predictions were compared with the P1 process computer output. The results showed that the CONDOR predictions matched the P1 computer output closely. However, the P1 results were not measured data; instead, they were based on a numerical cal-
._ culation with given boundary conditions such as core power, total flow, system pressure, and inlet enthalpy similar to CONDOR. Therefore, they are not a truly independent source for comparison purposes. However, the benchmark of CONDOR with P1 results may be combined with the independent check of the void model, the pressure drop calculations, and the analytical solution, as discussed in the CONDOR topical report, to verify that the code is able to calculate the flow and enthalpy distributions satisfactorily.
In summary, we conclude that the verification calculations showed that different models in CONDOR performed correctly, the void and pressure drop correlations compared favorably with experimental data and comparisons of CONDOR predictions with analytical solutions verified the numerical and programming algorithms.
3.0 CONCLUSION
S Based on the review which is described above we conclude that topical report WCAP-10107 is acceptable for referencing in licensing actions by Westinghouse with respect to the steady-state thermal-hydraulic performance of a BWR. The following restrictions apply to the use of the CONDOR code:
(1) The CONDOR code is claimed, in the topical report, to be able to perform BWR loop calculations. However, no description of loop modeling is given.
Therefore, the current version of CONDOR is restricted to the calculation of core flow and enthalpy distribution.
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(2) CONDOR does not have a verified CHF correlation in the code at this time. Any correlation to be incorporated in the code for MCPR licensing analysis has to be reviewed and approved by the NRC separately.
(3) Since the core bypass flow calculation is based on a simplified correlation with AP as the independent parameter, the correlation coefficients should be determined by comparing with the test data on a plant-specific basis. Factors affecting the coefficients, such as mixed core with fuels from different vendors, the crud buildup, and
-irradiation effects have to be considered.
(4) The CONDOR code uses 24 nodes to represent a flow channel for code verification, this number of nodes should be used for licensing cal-culations. If any reduced number of nodes is used, additional calculations should be performed to identify uncertainties associated with the reduced number of nodes.
(5) Selection of the loss coefficients with the effects of the crud build-up, geometry change due to irradation, different fuel designs by ,
different vendors should be considered on a plant-specific basis.
(6) During the course of our review, we raised questions regarding the proposed models and data. Westinghouserespondedtotbesequestions in Reference 7. These questions and answers should be included (Ref. 2) in the final topical report on CONDOR submitted by Westinghouse.
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4.0 REFERENCES
- 1. C. A. Olson, " CONDOR-A Thermal-Hydraulic Performance Code for Boiling Water Reactors", WCAP-10107, June 1982.
- 2. C. A. Olson, "COND0R-A Thermal-Hydraulic Performance code for Boiling ,
Water Reactors", WCAP-10107 (Rev.1), December 1983.
- 3. N. Zuber, et al., " Steady-State and Transient Void Fraction in Two-Phase
_ Flow Systems", Final Report, Vol. 1, GEAP-5417, January 1967.
- 4. S. Levy, " Forced Convection Subcooled Boiling Prediction of Vapor Volumetric Fraction," GEAP-5157, April 1966.
- 5. C. Baroczy, "A Systematic Correlation for Two-Phase Pressure Drop,"
Heat Transfer Conference (Los Angeles), Chemical Engineering Program Symp. Series No. 64, Vol. 62, 1966.
- 6. D. Chisholm, " Pressure Gradients Due to Friction During the Flow of Evaporating Two-Phase Mixtures in Smooth Tubes and Channels",
Intl. J. Heat & Mass Transfer, Vol.16, pp. 347-358,1973.
- 7. LetterwithAttachmentfromE. Rahe (Westinghouse) toc. Thomas (NRC),
dated December 1983.
- 8. Letter from E. Rahe (Westinghouse) to C. Thomas (NRC), dated July 23, 1984.
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