ML20046A552
| ML20046A552 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 07/16/1993 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 93-182A, NUDOCS 9307290032 | |
| Download: ML20046A552 (18) | |
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VIRGINIA ELecTurc Axn Powna Comwxy )
Ricnnoxn,Vinom1A 20261 ;
July 16, 1993 ,
l United States Nuclear Regulatory Commission Serial No. 93-182A :'
Attention: Document Control Desk NA&F/GLD-CGL R0 Washington, D. C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF-4 o NPF ;
l Gentlemen:
i VIRGINIA ELECTRIC AND POWER COMPANY :
REPORT OF ECCS EVALUATION MODEL CHANGES ;
AND 30-DAY REPORT PER REQUIREMENTS OF 10CFR50.46 .
NORTH ANNA POWER STATION UNITS 1 AND_2 !
Pursuant to 10CFR50.46(a)(3)(ii), Virginia Electric and Power Company is providing information concerning changes to the ECCS Evaluation Models and their application in existing licensing analyses. Information is also provided which quantifies the effect of these changes upon reported results for North Anna Power Station and l demonstrates continued compliance with the acceptance criteria of 10CFR50.46.
Attachment 1 contains excerpted portions of the Westinghouse report describing the changes to the Westinghouse ECCS Evaluation Models that are applicable to North Anna and have been implemented during calendar year 1992. In addition to these.
generic changes, there were plant-specific changes associated with application of the evaluation models for the North Anna units. Attachment 2 provides a report describing these plant-specific evaluation model changes. As indicated in the Attachment 2 report, these changes have been concluded to be significant, based upon the criterion established in 10CFR50.46(a)(3)(i).-
Attachment 3 provides information regarding the effect of the ECCS Evaluation Model changes upon the reported LOCA results for the North Anna _ Power Station analyses of record. To summarize the information in Attachment 3, the calculated PCT for the small and large break LOCA analyses for North Anna are given below. As defined in 10CFR50.46(a)(3)(i), the change in large break LOCA PCT is a significant change, requiring a 30-day report to the NRC.
North Anna Units 1 and 2 - Small break: 1873 F North Anna Units 1 and 2 - Large break: 2019 F 270073 9 m 290032 93o716 '
'[h PDR ADOCK 05000338 j P PDR.
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. i Since none of the calculated temperatures exceed 2200 F, no further action is required if you have further questions or require additional information, please contact ca.
Very truly yours, ;
26p/w L
(#W. L. Stewart Senior Vice President - Nuclear ,
Attachments:
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- 1. Westinghouse Report of LOCA Evaluation Model Changes for 1992 - North Anna Units 1 and 2
- 2. Report of Changes in Application of Evaluation Model- North Anna Units 1 and 2 '
- 3. Effect of Evaluation Model Changes - North Anna !! nits 1 and 2 i
cc: U. S. Nuclear Regulatory Commission Region 11 101 Marietta Street, N. W. ;
Suite 2900 Atlanta, Georgia 30323 '
NRC Senior Resident inspector i North Anna Power Station !
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ATTACHMENT 1 ,
WESTINGHOUSE REPORT OF LOCA EVALUATION MODEL CHANGES i FOR 1992 NORTH ANNA UNITS 1 AND 2 i
LOCA' Evaluation Model Changes for 1992 1
Bessel Function Error
Background
An error was discovered in SUBkOUTINE BESSJO which led to calculation of incorrect values for the l zeroth order Bessel function of the first kind. This calculation is used in the algorithm designed to limit I heat transfer out of a quenching fuel rod to the theoretical conduction limit. This error existed only in one cycle of the NOTRUMP computer code (Cycle 23) and therefore only affects analyses performed with that version. Cycle 23 of NOTRUMP was in use from February of 1991 until the error was i corrected in February of 1992. 'Ihis error correction returned the NOTRUMP code to consistency with the applicable section of WCAP-10079-P-A and therefore is not a change to the Evaluation Model. ;
This was determined to be a Non-discretionary Change in accordance with Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451.
Affected Evaluation Models 1985 Small Break LOCA Evaluation Model l Estimated Effect Plant specific PCT effects were determined by reanalysis of the limiting break size transient with the corrected NOTRUMP version. These effects are included on the attached Margin Utilization Sheet.
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Auxiliary Feedwater Mow Table Error Backcround -
The Steam Generator Auxiliary Feedwater (AFW) flowrate is governed by the timing variable TIMESG(1). A minor logic error associated with this variable was discovered which led to a step change in the AFW flowrate once the transient time passed the value of TIMESG(7). Typically, this value is set equal to i1000 seconds and so this error would only affect very long transient calculations. In addition, the nature of the error is to allow the AFW flowrate to immediately revert to the full value of the Main Feedwater flowrate. This enormous step change has led to code aborts in the cases where it has occurred.
This logic was corrected as a Discretionary Change as described in Section 4.1.1 of WCAP-13451. This determination is based on the fact that SBLOCA transients are generally terminated before the logic error can have an effect coupled with the codes lack of capability to handle the step change if it does occur.
Therefore. it was reasoned that the logic could not affect LOCA results.
Affected Evaluation Models 1985 Small Break LOCA Evaluation Model Estimated Effect This error correction has no effect on any current or prior applications of the Evaluation Model.
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Steam Generator Secondary Side Modelling Enhancements
Background
A set of related changes which make steam generator secondary side modelling more convenient for the user were implemented into NOTRUMP. This model improvement involved several facets of feedwater flow modelling. First, the common donor boundary node for the standard Evaluation Model nodalization :
has been separated into two identical boundary nodes. Rese donor nodes are used to set the feedwater enthalpy. The common donor node configuration did not allow for loop specific enthalpy changeover times in cases where asymmetric AFW flowrates or purge volumes were being modeled for plant specific sensitivities. .
The second improvement is the additional capability to initiate main feedwater isolation on either loss cf offsite power co1ncident with reactor trip (Iow pressurizer pressure) or alternatively on safety injection 3 signal (low-low pressurizer pressure). The previous model allowed this function only on loss of offsite '
power coincident with reactor trip. The auxiliary feedwater pumps are still assumed to start after a loss of offsite power wi'h an appropriate delay time to model diesel generator start-up and buss loading times.
The final improvement is in the area of modelling the purging of high enthalpy main feedwater after auxiliary feedwater is calculated to start. This was previously modelled through an approximate time ',
delay necessary to purge the lines of the high enthalpy main feedwater before credit could be taken for the much lower enthalpy auxiliary feedwater reaching the steam generator secondary. This time delay was a function of the plant specific purge volume and the auxiliary feedwater flowrate, ne new modelling allows the user to input the purge volume directly. This then is used together with the code
- calculated integrated feedwater flow to determine the appropriate time at which the feedwater enthalpy can be assumed to change.
9 These improvements are considered to be a Discretionary Change as described in Section 4.1.1 of WCAP-13451. Since they involve only enhancements to the capabilities and useability of the Evaluation Model, and not changes to results calculated consistently with the previous model, these changes were implemented without prior review as discussed in Section 4.1.1 of WCAP-13451.
Affected Evaluation Models 1985 Small Break LOCA Evaluation Model Estimated Effects !
Because these enhancements only allow greater ease in modelling plant specific steam generator secondary side behavior over the previous model, it is estimated that no effect will be seen in Evaluation Model calculations.
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Basis Change'for Hot Assembly Rod Gap Pressure
Background
t In the past, the effective hot assembly average rod power assumed to calculate the gap pressure in the hot assembly average rod was based on the total power in the hot assembly spread over all available bundle positions. That is, the power was averaged over both active rods as well as thimble tubes, which generate no power. This led to an artificially low rod internal pressure in the hot assembly average rod due to the artificially low power. In the future, the hot assembly average rod power based only on active fuel rods in the assembly will be used to calculate the rod internal pressure. The power modeled in the hot assembly rod for the purposes of channel fluid heating is still the appropriate power averaged over both fuel rods and unpowered thimble tubes.
For calculations in which rod burst is not predicted to occur, this change in basis will have a negligible :
effect. For calculations in which rod burst has been predicted, this change in basis will have the effect of increasing the tendency towards the limiting condition of coincident hot rod and hot assembly average l rod burst. This instance would lead to a change in PCT of less than 50*F.
This change in basis is being implemented as a result of information gathered during the course of an ongoing potential issue investigation as one of a closely related group of changes as described in Section 4.1.2 of WCAP-13451. The evaluation tool described in the following Section dealing with Limiting Time in Life in SBLOCA was made to conservatively bound the effect of this change and has been ,
applied to all affected plants to establish a Reasonable Assurance of Safe Operation as described in ~
Section 4.2.3 of WCAP-13451. Therefore, this Non-discretionary Change has been implemented (per Section 4.1 of WCAP-13451) as an Acceptable Change as described in Section 4.1.3 of WCAP-13451.
In addition, the ongoing evaluation of the issue is being performed in accordance with Section 4.2 of WCAP-13451 and so the effect of this change on the PCT is considered to be temporary pending final resolution of the issue.
Affected Evaluation Models 1975 SBLOCA Evaluation Model 1985 SBLOCA Evaluation Model j Estimated Effect Since the effect of this basis change is captured by the methodology used to assess the effect of the Limiting Time in Life basis change which follows, this change has no effect on previous analyses pending final resolution of the issue.
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Limiting Time in Life in SBLOCA Backcround It has historically been assumed that the limiting time in life for SBLOCA has coincided with the time of maximum fuel densification and, therefore, maximum fuel temperatures. It has recently been concluded that for some calculations performed under this assumption, a more limiting PCT will occur at some later time in life. His effect occurs ordy in cases where no rod burst has been predicted and the calculated PCT is greater than 1700'F.
This penalty arises due to both coolant channel flow blockage effects and from the heat deposited in the clad from the Zirc-water reaction which occurs on the clean interior surface of a newly burst rod. This rod burst later in life occurs due to the increase in rod internal pressure car" by the build up of fission gases in the fuel rod. While an evaluation tool has been developed to cor. .atively assess the impact of this issue, future analyses will include a plant specitic limiting time in lite determination.
- This change in basis is being implemented as a result of information gathered during the course of an ongoing potential issue investigation as one of a closely related group of changes as described in Section 4.1.2 of WCAP-13451. The evaluation tool described above was developed to conservatively bound the effect of this change and has been applied to all affected plants to establish a Reasonable Assurance of Safe Operation as described in Section 4.2.3 of WCAP-13451. Therefore, this Non-discretionary Change has been implf -ented (per Section 4.1 of WCAP-13451) as an Acceptable Change as described in Section 4.1.3 of WCAP-13451. In addition, the ongoing evaluation of the issue is being performed in accordance with Section 4.2 of WCAP-13451 and so the effect of this change on the PCT is considered to be temporary pending final resolution of the issue.
Affected Evaluation Models 1975 SBLOCA Evaluation Model 1985 SBLOCA Evaluation Model Estimated Effect Since the effect of this basis change is captured by the evaluation tool used to establish a RASO for an ,
ongoing potential issue investigation, this change has no effect on previous analyses pending final j resolution of the issue.
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Structural Metal IIcat Modeling,
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Rackcround A discrepancy was discovered during review of the finite element heat conduction model used in the i WREFLOOD-INTERIM code to calculate heat transfer from structural metal in the vessel during the reflood phase. It was noted that the material properties available in the code corresponded to those of stainlesssteel. While this is correct for the internal structures, it is inappropriate for the vessel wall !
which consists of carbon steel with a thin stainless internal clad. This was defined as a non-discretionary. l change per Section 4.1.2 of WCAP-13451, since there was thought to be potential for increased PCT with a more sophisticated composite model. The model was revised by replacing it with a more 'lexible one that allows detailed specification of structures.
Affected Evaluation Models 1981 ECCS Evaluation Model with BART 1981 ECCS Evaluation Model with BASH s l
Affected Codes WREFLOOD-INTERIM l
Estimated Effects The estimated effect of this correction is a 25*F PCT benefit.
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Spacer Grid IIeat Transfer Error in BART Backcround During investigations into anomolous wetting and dryout behavior demonstrated by the BART grid model a programming logic error was discovered in the grid heat transfer model. The error caused the solution to be performed twice for each timestep. The error was traced back to the original coding used in all of the BART and LOCBART codes. This was defined as a non-discretionary change per Section 4.1.2 of WCAP-13451. The error was corrected, and a complete reverification of the grid model was conducted and transmitted to the NRC (WCAP-10484, Addendum 1).
Affected Evaluation Models '
i 1981 ECCS Evaluation Model with BART 4 1981 ECCS Evaluation Model with BASH Affected Codes BART LOCBART Estimated Effects Calculations performed with the affected code have consistently demonstrated significantly better grid wetting and lower clad temperatures. A conservative estimate of zero degrees PCT penalty has been assigned for this issue. I 1
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POWER-SHAPE SENSITIVITY MODEL (PSSM)
Backcround Historically, chopped cosine power shapes have been assumed to produce limiting results in Westinghouse large break LOCA analyses. However, with the advent of more advanced models (BART and BASH) it was discovered that under certain circumstances, top skewed power shapes could potentially be more limiting. The PSSM was developed to allow the assessment of shape specific Peak Cladding Temperature (PCT) trends in large break LOCA. As described in WCAP-12909-P and further clarified in ET-NRC-91-3633 (currently under NRC review), the methodology was developed from a large database of large -
break LOCA analysis results which used a wide variety of full power power shapes in typical twejve foot core for Westinghouse supplied fuel.
This methodology change is considered to be a Non-discretionary Change as described in Section 4.1.2 of WCAP-13451 and has been implemented prior to final NRC review in accordance with Section 4.1.3 of WCAP-13451.
Affected Evaluation Models 1981 ECCS Evaluation Model with BART 1981 ECCS Evaluation Model with BASH Estimated Effect The implementation of this methodology reasonably assures that cycle specific power distributions will not lead to results more limiting than those of the analysis of record. Therefore, there is no PCT effect for this methodology.
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ATTACHMENT , 2
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REPORT OF CHANGES IN APPLICATION OF ECCS EVALUATION MODELS 1
NORTH ANNA UNITS 1 AND 2 Revised LBLOCA Analysis (North' Anna Units 1 and 2) l I
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1.0 Background
This report provides a summary of changes in LOCA analysis results from those last reported for North Anna Units 1 and 2 (1). These changes are described in Section 2.0 below. It has been concluded that these changes are significant, as defined in 10 CFR 50.46(a)(3)(i).
2.0 Evaluation Model Changes 2.1 Revised Large Break LOCA Analysis (North Anna Units 1 and 2)
Since our previous 10CFR50.46 report (1), a revised analysis of the large break LOCA transient has been performed for North Anna Units 1 and 2. This revised analysis has been implemented as the analysis of record via a station 10CFR50.59 evaluation, in conjunction with the provisions of North Anna Technical Specification 6.9.1.7 (relating to the Core Operating. Limits Report). This discussion summarizes the changes incorporated in this analysis. The analysis includes assumptions which have been made to reflect operation with extended steam generator tube plugging (SGTP) in addition to changes in other key analysis inputs. Results and limitations associated with this analysis are applicable to the operation of North Anna Units 1 and 2. The key changes in assumptions from the prior analysis are listed below.
- Assumption of 20% uniform steam generator tube plugging, including a reduced RCS total flowrate of 264400 gpm (supports operation with a peak SGTP of 20% in any SG).
- Hot assembly relative power factor of 1.40 (prior analysis assumed 1.362).
- Containment accumulator water temperature of 100oF (prior analysis assumed 86*F).
- Improved spacer grid heat transfer model (2).
Safety injection,1HHSl+1LHSI, spilling to 10 psig containment pressure (prior analysis assumed spilling at 0 psig backpressure).
- Assumed fuel temperature and rod internal pressure associated with core average burnup of 500 MWD /MTU, including conservatism of 65 psi rod pressure.
These changes are discussed further in the following paragraphs.
The analysis assumes that 20% of the tubes in each steam generator are plugged.
Since large break LOCA results are sensitive to SGTP, this assumption is necessary to demonstrate continued compliance with the 10CFR50.46 ECCS acceptance criteria for operation with extended SGTP. In conjunction with extended SGTP, a reduced RCS total flowrate of 264400 gpm has been assumed. This value bounds the expected flow associated with 20% SGTP.
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A sensitivity study was performed as part of the revised analysis to quantify the effects of small variations in the assumed hot assembly relative power. The analysis of record assumed a hot assembly relative power of 1.40, which is the maximum value analyzed in the study.
The assumed initia! accumulator water temperature is increased from 86 F (used in the previous analysis) to 100 F. The water temperature in each accumulator is assumed to equal the temperature in the surrounding containment compartment during full power operation. At North Anna, the accumulators are located on the containment floor. A review of temperature detector data at this location has confirmed :
that a temperature of 100 F is a representative conservative nominal value for use in -
the large break LOCA analysis. This assumption has been changed and validated in accordance with Westinghouse guidance which implements the recommendations of Reference (3).
The North Anna analysis uses a corrected version of the LOCBART code, which is part of the BASH Evaluation Model (EM). Westinghouse has corrected and improved the spacer grid heat transfer model used in the BART and BASH ECCS Evaluation Models (2). Since this model change is primarily a correction to the EM, it has been ,
implemented in all versions of the BART and BASH ems without prior NRC review.
This process for addressing model changes is documented in WCAP-13451 (4).
The safety injection is assumed to spill to 10 psig containment pressure instead of the value of 0 psig used in the previous analysis. This assumption has been validated by confirming that containment pressure (conservatively calculated using the Westinghouse evaluation model) does not decrease to less than 10 psig during the transient prior to the time of peak clad temperature.
An additional uncertainty of 65 psi is added to the design values for fuel rod internal pressure assumed as initial conditions for the accident. This allowance has been '
established by Westinghouse to bound the expected magnitude of a potential issue concerning the calculational methodology for this design input parameter. Adding the uncertainty is conservative for large break LOCA analysis, since it causes the predicted time of cladding burst to occur earlier and with more severe channel blockage. .
Employing these assumptions in the current version of the 1981 ECCS Evaluation Model with BASH, it has been demonstrated that operation at the rated thermal power ,
of 2893 MWt with SGTP up to 20% in any SG will comply with all of the acceptance criteria specified in 10 CFR 50.46. Attachment 3 provides the PCT result for the revised analysis of record, in conjunction with appropriate margin assessments which address BASH evaluation modelissues.
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l 3.0 References (1) Letter from W. L. Stewart (Va. Electric & Power Co.) to NRC, "Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2 - Report of ECCS Evaluation Model Changes Pursuant to Requirements of 10 CFR 50.46," Serial No.92-560, August 31,1992.
(2) Letter from Nick Liparulo (Westinghouse-Manager, Nuclear Safety & Regulatory l Activities) to USNRC, " Notification of Changes to the Westinghouse Large Break LOCA ECCS Evaluation Model," ET-NRC-92-3787, December 22,1992 transmits WCAP-10484, Addendum 1, " Spacer Grid Heat Transfer Effects During Reflood."
(3) Letter from Nicholas J. Liparulo (Westinghouse-Manager, Nuclear Safety &
Regulatory Activities) to USNRC,"Results of Technical Evaluation of Containment Initial Temperature Assumptions for Large Break Loss of Coolant Accident Analysis," ET-NRC-92-3699, June 1,1992. ,
(4) Letter from N. J. Liparulo (Westinghouse-Manager, Nuclear Safety & Regulatory Activities) to USNRC, " Westinghouse Methodology for Implementation of 10CFR50.46 Reporting," ET-NRC-92-3755, October 30, 1992 transmits WCAP-13451, " Westinghouse Reporting Methodology for Implementation of 10 CFR 50.46 Reporting."
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ATTACHMENT 3 EFFECT OF WESTINGHOUSE ECCS EVALUATION MODEL MODIFICATIONS NORTH ANNA UNITS 1 AND 2 7
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Effect of Westinghouse ECCS Evaluation Model Modifications -
1 North Anna Units 1 and 2 l
The information provided herein is applicable to North Anna Power Station, Units 1 and 2. It is based upon reports from Westinghouse Electric Corporation for issues i involving the ECCS evaluation models and plant-specific application of the models in !
the existing analyses. Peak cladding temperature (PCT) values and margin l allocations represent issues for which permanent resolutions have been implemented. I Section A presents the detailed assessment for small break LOCA. The large break l LOCA details are given in Section B. j Section A - Small Break LOCA Margin Utilization - North Anna Units 1 and 2 A. PCT for Analysis of Record (AOR) 1873 F (1)
B. Prior Evaluation Model PCT Assessments 1.1990 -Total Permanent Assessment {1} (2) 0F 2.1991 -Total Permanent Assessment {2} (2) 0F 3.1992 - Total Permanent Assessment {2} (3) 0F i C. Current Evaluation Model PCT Assessments {3} l
- 1. NOTRUMP Bessel Function Error {2} OoF ,
1 SBLOCA Licensing Basis PCT 1873 F (AOR PCT + PCT Assessments)
Section B - Large Break LOCA Margin Utilization - North Anna Units 1 and 2 A. PCT for Analysis of Record (AOR) 2044 F (4) 1 B. Prior Evaluation Model PCT Assessments 1.1990 - Total Permanent Assessment {1} (2) OoF 2.1991 -Total Permanent Assessment {1} {2} 0F 3.1992 - Total Permanent Assessment {1} (3) 0F C. Current Evaluation Model PCT Assessments {3}
- 1. Structural Metal Heat Modeling - 25 F ,
- 2. Spacer Grid Heat Transfer Error in BART {2} 0F-
- 3. Power Shape Sensitivity Model (PSSM) {4} 0F LBLOCA Licensing Basis PCT 2019 F (AOR PCT + PCT Assessments)
Notes { } and References ( ) on the following page.
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'Effect of Errors / Changes in ' Application of ECCS Evaluation Models -
North Anna Units 1 and 2 Notes ]
{1} No issues which were applicable to the evaluation model used in the AOR were permanently resolved for this reporting period.
{2} The AOR was performed with an evaluatio'n model version that includes corrections and/or input changes to address the applicable issues. ]
{3} Refer to the 1992 Westinghouse Report of LOCA Evaluation Model Changes ;
provided in Attachment 1. ;
{4} Final resolution of this issue involves no permanent PCT assessment. Virginia Power applies the Westinghouse methodology documented in Reference (5) to reload cores. ,
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References _j (1) " North Anna Power Station Units 1 and 2 - Implementation of Extended SGTP 1 Small Break LOCA Analysis," 10CFR50.59 Safety Evaluation 92-SE-OT-005, 1 January 21,1992.
(2) Letter from W. L. Stewart (Va. Electric & Power Co.) to NRC, "Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and Report of ECCS q Evaluation Model Changes Per Requirements of 10CFR50.46,"_ Serial No. J 91-428, August 23,1991.
(3) Letter from W. L. Stewart (Va. Electric & Power Co.) to NRC, " North Anna Power Station Units 1 and 2, North Anna Power Station Units 1 and 2- Report of ECCS Evaluation Model Changes Pursuant to Requirements of 10CFR50.46," Serial No.'92-560, August 31,1992. j
. 1 (4) " North Anna Power Station Units 1 and 2 - Large Break LOCA Analysis for 20%
SGTP," 10CFR50.59 Safety Evaluation 93-SE-OT-048, June 17,1993.
(5) " Westinghouse ECCS Evaluation Model: Revised Large Break'LOCA Power Distribution Methodology," WCAP-12909-P, June 1991.
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