Review of Combustion Engr Topical Rept#CENPD-169-P COLSS- Assessment of the Accuracy of PWR Oper Limits as Determined by the Core Oper Limit Supervisory Sys. Finds Rept Is Acceptable Reference for MethodologyML20147E363 |
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12/13/1978 |
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Office of Nuclear Reactor Regulation |
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ML20147E348 |
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NUDOCS 7812210051 |
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Category:TEXT-SAFETY REPORT
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[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. 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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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.. . OEC 131973 REVIEW 0F TOPICAL REPORT CENPD-169-P Report
Title:
Assessment of the Accuracy of PWR Operating Limits as Determined by the Core Operating Limit Supervisory System (COLSS)
Report Number: CENPD-169-P (Proprietary)
Report Date: July 1975 Originating Organization: Combustion Engineering, Inc.
l Reviewed by: Core Performance Branch, DSS Summary of Topical Report The primary objective of CENPD-169-P is to present the methodology fnr ascertaining the uncertainty in the core power distribution determined by the Core Operating Limit Supervisory System (COLSS) synthesis pro-cedure. In addition, the report serves as a reference for the basic concepts employed to transform the signals from five-level incore detectors into forty-node representations of the radially averaged core axial power shape (CAPS). Also included is a description of the construction of the hot pin and hot channel power distributions using precalculated planar Radial Peaking Factors (RPF) and the CAPS.
The subject report should only be used as a generic reference for the synthesis and uncertrinty methodology, since implementation of the methodology and the precise values of the uncertainty factors are plant specific.
Summary of Regulatory Evaluation The following sections are intended to clarify the essential points pertiment
, to the evaluation of the CENPD-169-P methodology.
"I812210 @
_2 OEC 131978 Applicability of the Report The COLSS methodology described in the report will generically encompass CE plants of the Arkansas Nuclear One-Unit 2 and future designs. However, the precise value of the appropriate uncertainty factors depend on basic-ally three design parameters: the number and burnup of the fuel assemblies the Control Element Assembly '(CEA) programming, and the number of operable -
incore detectors. Thus the Linear Power Density (LPD) uncertainty will depend on the plant design.
Core Power Distribution The core axial power shape (CAP) is calculated from the integral responses of 5 levels of incore detector segments. A simple 5 element Fourier fit to the power distribution is employed. Sets of integral detector signal equations are solved for the combining coefficients and the core axial power distribution is constructed for 40 axial nodes. For each reactor empicying a COLSS, these coefficients are calculated with a reactor simulator model and are experimentally verified during startup testing.
The detailed radial power distribution is not measured in COLSS. Radial peaking factors calculated with a reactor simulator model (the simulator l model is a three dimensional nodal core physics model for the calculation l
of the core power distribution) are employed to construct the hot pin and hot channel power distribution. The radial peaking factors are functions
' of CEA configuration and are stored in the COLSS computer. Tests will be
. , DEC 13 'D performed during the reactor start-up program to verify that the pre-calculated values of radial peaking factors are appropriate for the as-built reactor. Updating of COLSS power distribution constants will be based on the results of the measurement of core power distributions by the INCA (ref. 1) program.
Uncertainty Analysis _
The technique described for determining the uncertainty in the COLSS calculated Fg combines the CAPS synthesis error with the INCA (ref.1) verified planar RPF uncertainty. To determine the CAPS uncertainty, the Fg 's for more than 4000 core configurations, including steady-state, transients, and non-normal CEA configurations, calculated by the reactor simulator and the COLSS procedures were compared. In these computations, the COLSS procedures employed the planar RPFs obtained from the simulator for each core configuration.
Thus, the differences in Fg reflect only CAPS synthesis errors. The uncertainty in the RPF's attributable to the accuracy of the INCA (refs. I and 2) system used to verify the RPF's is combined with the CAPS uncertainty to yield the total uncertainty value.
The increase in the uncertainty in the COLSS synthesized Fg as a function of inoperable incore detectors was evaluated. The calculations show that the error is small, even with a large number of inoperable detectors. This is because the axial distribution is obtained from the average of all operable detectors at a given axial level and therefore even a large number of inoperable detectors does not greatly affect the average values.
Orc 1 a 570 Regulatory Position Based on our review of the areas described above, we conclude that the subject report is an acceptable reference for the general methodology of COLSS to determine core power distributions from incore detectors.
In addition, the procedure described for calculating the overall uncertainty in Fg including uncertainties in the COLSS synthesis pro- _
cedure is satisfactory. The values for uncertainties used in references 1 and 2 are intentionally excluded from this review, since review of those documents has not been completed.
REFERENCES
- 1. CENPD-145 INCA - Method cf Analyzing In-Core Detector Data, April 1975,
- 2. CENPD-153 Evaluation of Uncertainty in Fg Measured by Self-Powered Fixed In-Core Detector Systems, August 1974.
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