ML18153A307
| ML18153A307 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 05/28/1998 |
| From: | Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 98-303, NUDOCS 9806020246 | |
| Download: ML18153A307 (19) | |
Text
- -**\
e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 May 28, 1998 United States Nuclear Regulatory Commission Serial No.98-303 Attention: Document Control Desk NL&OS/ETS: R1/
Washington, D.C. 20555 Docket Nos. 50-280/281 50-338/339 License Nos. DPR-32/37 NPF-4/7 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY AND NORTH ANNA POWER STATIONS UNITS 1 AND 2 REPORT OF EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL CHANGES PURSUANT TO THE REQUIREMENTS OF 10CFRS0.46 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 and Surry Power Stations, and demonstrates continued compliance with the acceptance criteria of 10CFR50.46.
Attachment 1 provides a report describing plant-specific evaluation model changes associated with the application of the large break and small break LOCA evaluation models for North Anna Unit 2.
Attachment 2 contains excerpted portions of Westinghouse reports describing the changes to the Westinghouse ECCS Evaluation Models which are applicable to North Anna and Surry and have been implemented during calendar year 1997.
Information regarding the effect of the ECCS Evaluation Model changes upon the reported LOCA analysis of record (AOR) results is provided for the North Anna and Surry Power Stations in Attachments 3 and 4, respectively. To summarize the information in Attachments 3 and 4,- the calculated peak cladding temperature (PCT) for the small and large break LOCA analyses for North Anna and Surry are given below. None of these results include significant changes, based on the criterion in 10CFR50.46(a)(3)(i).
North Anna Unit 1 - Small break: 1675°F Large break: 2068°F North Anna Unit 2 - Small break: 1676°F Large break: 2086°F Surry Units 1 and 2 - Small break: 1717°F Large break: 2113°F
'-\
., ,C\o
\ '- \J
\
'\
We have evaluated these issues and the associated changes in the applicable licensing basis* PCT results. These results demonstrate compliance with the requirements of 10CFR50.46(b). No further action is required to demonstrate compliance with 10CFR50.46 requirements.
10CFR50.46(a)(3)(ii) requires reporting ECCS analysis changes and .effects at least annually. Our reports for the previous calendar year have historically been transmitted by the end of the first quarter of the next year. This report for the calendar year 1997 was delayed to include the North Anna Unit 2 large break and small break LOCA evaluations due to the removal of the part-length control rod drive mechanisms discussed in Attachment 1.
If you have further questions or require additional information, *please contact us.
Very truly yours, James P. O'Hanlon Senior Vice President - Nuclear Commitments made in this letter:
- 1. None Attachments:
- 1) Report of Changes in Application of ECCS Evaluation Models - North Anna Unit 2
- 2) Westinghouse Report of ECCS Evaluation Model Changes - North Anna Units 1 and 2 and Surry Units 1 and 2
- 3) Effect of ECCS Evaluation Model Changes - North Anna Units 1 and 2
- 4) Effect of ECCS Evaluation Model Changes - Surry Units 1 and 2
e cc: U.S. Nuclear Regulatory Commission
- Region II Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. M. J. Morgan NRC Senior Resident Inspector North Anna Power Station Mr. R. Musser NRC Senior Resident Inspector Surry Power Station
e ATTACHMENT 1 REPORT OF CHANGES IN APPLICATION OF ECCS EVALUATION MODELS NORTH ANNA UNIT 2
Effect of Reactor Vessel Part-Length CRDM Removal Upon Large Break and Small Break LOCA Analysis North Anna Unit 2 1.0 Background This report provides a summary of changes in large break and small break LOCA analysis results from those last reported for North Anna Unit 2<1l. These changes are due to the removal of the part-length control rod drive mechanisms (CRDMs). It has been concluded that these changes are not significant, as defined in 10CFR50.46(a)(3)(i). The effects of the analysis have been implemented for North Anna Unit 2 via a station 10CFR50.59 evaluation<2l.
2.0 Description of Plant Changes This report presents the evaluation of large break and small break LOCA analyses performed to support North Anna Unit 2 operation following the part-length CROM removal. This modification was implemented during the spring 1998 Cycle 12-13 refueling outage.
Five part-length CROM housings were installed on the reactor vessel upper head as provided in the original plant design. The part-length control rods were not installed in the North Anna 2 vessel, but the lead screws of the part-length CRDMs remained in the vessel, latched in the fully withdrawn position. In the fully withdrawn position, the lead screw extends through a hole in the top of the upper guide tube structure in the reactor vessel upper head. The significant aspect of the modification (removal of part-length CROM lead screw) relating to the large break and small break LOCA analyses is the effect upon the flowpaths between the upper plenum and upper head.
The modified plant has five open holes in the upper guide tube structures and 48 through which pass the lead screws for the full length control rods. This change increases the total flow area between the upper plenum and upper head across the upper support plate by approximately 13%. The exchange of flow between the upper plenum and upper head during the large break and small break LOCAs is slightly altered from that of the existing plant design.
The change in flow area and resistance for the five part-length upper guide tube locations affect inputs which are modeled in the large and small break LOCA analysis. During LOCA events, there is total voiding of the upper head and there is steam flow, that passes from the core through the upper plenum and to the upper head through the guide tubes. Because of the complex dynamics of LOCA phenomena, it was expected that these physical changes could adversely affect the results. Limiting cases for large and small break LOCA accidents were reanalyzed to quantify the effects of the part-length CROM removal.
3.0 e
Evaluation of Model Changes 3.1 Small Break LOCA Analysis (North Anna Unit 2)
This discussion presents the results of a sensitivity analysis for the small break LOCA transient for North Anna Unit 2 operation with the part-length CROM modification.* Since the current small break analysis of record(3l has demonstrated that the upflow configuration bounds downflow for North Anna cores, the sensitivity analysis consisted of reanalyzing the limiting three-inch break case for the Unit 1 upflow design from Reference 3 with modifications to model the removal of the part-length CRDMs.
The key analysis input items from the existing analysis of record are listed below. These values are unchanged from those assumed in the analysis of record (3l_
Assumption of 7% uniform steam generator tube plugging Peak Heat Flux Hot Channel Factor, F(Q), of 2.32*
Peak value for Enthalpy Hot Channel Factor, F! of 1.65*
A minimum delivered HHSI flow rate calculated for LOCA analysis A full core of North Anna Improved Fuel (NAIF) with ZIRLO' cladding and Performance+ features (bounds operation with 17x17 Standard and NAIF mixed cores)
Upflow baffle/barrel design Safety Injection in all loops COSI Condensation Model
- These values bound the limits in the current North Anna Units 1 and 2 COLRs.
3.1.1 SBLOCA Method of Analysis As required by Appendix K of 10 CFR 50, certain conservative assumptions were made for the Small 8[eak LOCA-ECCS analysis. The assumptions pertain to the condition of the reactor and associated safety system equipment at the time that the LOCA is assumed to occur, and include such items as the core peaking factors, core decay heat and the performance of the Emergency Core Cooling System. Assumptions and initial operating conditions which reflect the requirements of Appendix K to 10 CFR 50 have been used in this analysis, and are the same as those used in Reference 3.
3.1.2 Results For this sensitivity analysis, only the limiting three-inch effective diameter cold leg break case was run. The sensitivity analysis resulted in an increase in the limiting peak clad temperature of 1°F. The maximum local cladding oxidation level and total core metal-water reaction were demonstrated to meet the limits of 10CFR50.46.
3.2 Large Break LOCA Analysis (North Anna Unit 2)
This discussion presents the results of a sensitivity analysis for the large break LOCA transient for North Anna Unit 2 operation with the part-length CROM modification. Since the current large break analysis of record<3l demonstrated that the downflow configuration bounds upflow, the sensitivity analysis consisted of reanalyzing the limiting double-ended cold-leg guillotine (DECLG) break with Cd=0.4 for the Unit 2 downflow design from Reference 3 with modifications to model the removal of the part-length CRDMs.
The key analysis input items from the existing analysis of record are listed below. These values are unchanged from those assumed in the analysis of record<3l_
Assumption of 7% uniform steam generator tube plugging Peak Heat Flux Hot Channel Factor, F(Q), of2.19*
Peak Enthalpy Hot Channel Factor, F:J, of 1.60*
Hot Assembly Relative Power Factor of 1.45*
Downflow baffle/barrel design North Anna Improved Fuel (NAIF) with ZIRLO' cladding and PERFORMANCE+ design features (bounds operation with 17x17 Standard and NAIF mixed cores)
- These values bound or equal the limits in the current North Anna Units 1 and 2 COLRs.
3.2.1 LBLOCA Method of Analysis As required by Appendix K of 10 CFR 50, certain conservative assumptions were made for the Large Break LOCA-ECCS analysis. The assumptions pertain to the condition of the reactor and associated safety system equipment at the time that the LOCA is assumed to occur, and include such items as the core peaking factors, core decay heat and the performance of the Emergency Core Cooling System. Assumptions and initial
operating conditions which reflect the requirements of Appendix K to 10 CFR 50 have been used in this analysis, and are the same as those used in Reference 3.
3.2.2 Results For this analysis, only the limiting double-ended cold leg guillotine with Cd=0.4 case was run. The sensitivity analysis resulted in an increase in the limiting peak clad temperature of 18°F. The maximum local cladding oxidation level and total core metal-water reaction were demonstrated to meet the limits of 10CFR50.46.
4.0 Conclusions The large and small break LOCA analyses for North Anna Unit 2 have been modified to accommodate operation of North Anna Unit 2 with part-length CRDMs removed.
Sensitivity analyses were performed for the small break LOCA with the limiting 3 inch break and for the large break LOCA with the double-ended cold-leg guillotine (DECLG) break with Cd=0.4. The sensitivity cases resulted in an increase in the limiting peak clad temperature of 1°F for the small break LOCA and 18°F for the large break LOCA. The results of each case meet the applicable acceptance criteria as specified in 10CFR50.46.
Consequently, it is concluded that the North Anna ECCS continues to meet the acceptance criteria of 10CFR50.46 for operation following the part-length CROM modification.
5.0 References (1) Letter from J. P. O'Hanlon (Virginia Power) to USNRC, "Virginia Electric and Power Company, Surry and North Anna Power Stations Units 1 and 2, Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes Pursuant to the Requirements of 10CFR50.46," Serial No.97-174, March 27, 1997.
(2) North Anna Power Station 10CFR50.59 Safety Evaluation, 98-SE-MOD-11, Revision 1, "North Anna Power Station Unit 2 - Safety Evaluation for Removal of Part-Length Control Rod Drive Mechanisms," April 21, 1998.
(3) Letter from J. P. O'Hanlon (VEPCO) to Document Control Desk (USNRC),
'Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, 30-Day Report of ECCS Evaluation Model Changes Per Requirements of 10CFR50.46," Serial No.95-608, November 29, 1995.
1-~
e ATTACHMENT 2 WESTINGHOUSE REPORT OF ECCS EVALUATION MODEL CHANGES NORTH ANNA UNITS 1 AND 2 AND SURRY UNITS 1 AND 2 L
Westinghouse Report of ECCS Evaluation Model Changes SBLOCTA Clad Burst Strain Error
Background
An error has been discovered in the SBLOCTA code related to improper calculation of clad post-burst strain. Specifically, the error occurs because although the burst node is predicted to strain out upon the occurrence of burst, incorrect coding logic causes this burst strain to be neglected in subsequent timesteps. The main effect of this for small break transient calculations is that for high peak clad temperature cases (in excess of 1800°F) which are also limited by the rapid zirc-water reaction accompanying incipient clad burst, the smaller clad diameters reduce the burst temperature spike, and upon correction of the coding a net PCT penalty may result. This correction was determined to be a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect Plant specific evaluations demonstrate that the SBLOCTA Clad Burst Strain Error does not result in a PCT impact to the North Anna and Surry SBLOCA AORs since the current PCTs are below the threshold for the effect of the error.
Implicit Bubble Rise Derivative Error
Background
An error was discovered in the coding used for the implementation of implicit bubble rise for certain drift flux models. This error only affects advanced plant calculations which invoke these models. This error correction was determined to be a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451.
Affected Evaluation Model 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP-(AP600 Implementation)
Estimated Effect The nature of this error leads to no impact on the calculated PCT for all standard EM applications.
NOTRUMP Flow Regime Map Error
Background
An error was discovered in an exponent in a function used to define a boundary in the flow regime map used by NOTRUMP for some vertical flow situations. The error correction was determined to be a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451.
Affected Evaluation Model 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect Representative plant calculations have led to an estimated 0°F effect for this error.
lnterfacial Heat Transfer Courant Limit Error
Background
An error was discovered in the NOTRUMP logic dealing with the calculation of the material Courant timestep limit for interfacial heat and mass transfer. Due to this error, NOTRUMP allowed this limit to be violated, leading to the possibility of unrealistic fluctuations in the interfacial heat and mass transfer in a coolant leg with a small mixture or vapor region. The error correction was determined to be a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect Representative plant calculations have led to an estimated 0°F effect for this error.
e LOCBART Post-Burst Rod Internal Pressure *
Background
An error was discovered in a LOCBART code version related to improper calculation of post-burst rod internal pressure. After burst, the rod pressure should be set to the channel pressure, but due to an error in checking a flag in one portion of the code, the pressure continued to be calculated as if burst had not occurred. A review of the coding found that other than the pressure itself, the only other parameter incorrectly calculated was the thermal conductivity of the gas in the pellet-to-clad gap, which had a weak contribution to gap conductance. This was determined to be a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 1981 Westinghouse Large Break LOCA Evaluation Model with BART*
1981 Westinghouse Large Break LOCA Evaluation Model with BASH Estimated Effect Representative plant calculations have led to an estimated 0°F effect for this error.
. Core-Average LOCTA Fuel Rod Initialization
Background
An error was discovered in the BASH code related to improper calculation of core-average LOCTA fuel rod initialization. After the fuel rod initialization converges to the appropriate temperature, the density correction should be applied during the transient as during the initialization, to adjust the transient pellet radial node power density. A review of the coding found that this density correction was being applied incorrectly during the steady-state convergence. This was determined to be a No,n-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model 1981 Westinghouse Large Break LOCA Evaluation Model with BASH Estimated Effect Representative plant calculations have led to an estimated 0°F effect for this error.
ATTACHMENT 3 EFFECT OF ECCS EVALUATION MODEL CHANGES NORTH ANNA UNITS 1 AND 2 L
e Effect of ECCS Evaluation Model Changes - North Anna Unit 1 The information provided herein is applicable to North Anna Power Station, Unit 1. It is based upon reports from Westinghouse Electric Corporation for issues involving the ECCS evaluation models and plant-specific application of the models in the existing analyses. Peak cladding temperature (PCT) values and margin allocations represent issues for which permanent resolutions have been implemented. The assessments for small break and large break LOCA are presented in Sections A and B, respectively.
Section A - Small Break LOCA Margin Utilization - North Anna Unit 1 A PCT for Analysis of Record 1704°F (1)
B. Prior PCT Assessments Allocated to AOR -29°F
- 1. NOTRUMP Specific Enthalpy Error +20°F (2)
- 2. SALIBRARY Double Precision Errors -15°F (2)
- 3. Fuel Rod Initialization Error +10°F (3)
- 4. Loop Seal Elevation Error -44°F (3)
SBLOCA Augmented PCT for AOR 1675°F C. PCT Assessments for 10CFR50.46(a)(3)(i) Accumulation 0°F SBLOCA Licensing Basis PCT (AOR PCT+ PCT Ass~ssments) ~,,. 1675°F Section B - Large Break LOCA Margin Utilization - North Anna Unit 1 A PCT for Analysis of Record 2013°F (1)
B. Prior PCT Assessments Allocated to AOR 40°F
- 1. LBLOCA/Seismic SG Tube Collapse +30°F (1)
- 2. BASH Accumulator Empty Flag +10°F (1)
LBLOCA Augmented PCT for AOR 2053°F C. PCT Assessments for 10CFR50.46(a)(3)(i) Accumulation {1} 15°F
- 1. Translation of Fluid Conditions from SATAN to LOCTA +15°F (4)
LBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 2068°F Notes { } and References ( ) on the final page of this attachment.
Effect of ECCS Evaluation Model Changes - North Anna Unit 2 The information provided herein is applicable to North Anna Power Station, Unit 2. It is based upon reports from Westinghouse Electric Corporation for issues involving the ECCS evaluation models and plant-specific application of, the models in the existing analyses. Peak cladding temperature (PCT) values and margin allocations represent issues for which permanent resolutions have been implemented. The assessments for small break and large break LOCA are presented in Sections A and B, respectively.
Section A - Small Break LOCA Margin Utilization - North Anna Unit 2 A PCT for Analysis of Record 1704°F (1)
- 1. NOTRUMP Specific Enthalpy Error +20°F (2)
- 2. SALIBRARY Double Precision Errors -15°F (2)
- 3. Fuel Rod Initialization Error +10°F (3)
- 4. Loop Seal Elevation Error -44°F (3)
SBLOCA Augmented PCT for AOR 1675°F
)
C. PCT Assessments for 10CFR50.46(a)(3)(i) Accumulation {1} 1°F
- 1. Removal of Part-Length CRDMs {2} {3} +1°F SBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 1676°F Section B - Large Break LOCA Margin Utilization - North Anna Unit 2 A PCT for Analysis of Record 2013°F (1)
- 1. LBLOCA/Seismic SG Tube Collapse +30°F (1)
- 2. BASH Accumulator Empty Flag' +10°F (1)
LBLOCA Augmented PCT for AOR 2053°F C. PCT Assessments for 10CFR50.46(a)(3)(i) Accumulation {1} 33°F
- 1. Translation of Fluid Conditions from SATAN to LOCTA +15°F (4)
- 2. Removal of Part-Length CRDMs {2} {3} +18°F LBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 2086°F Notes { } and References ( ) on the following page.
Effect of ECCS Evaluation Model Changes - North Anna Notes:
{1} The accumulation of changes (sum of absolute magnitudes) is less than 50°F and is not significant, as defined in 10CFR50.46(a)(3)(i).
{2} The current report is the initial quantification of effects for this issue.
{3} Refer to the Report of Changes in Application of ECCS Evaluation Model Changes, North Anna Unit 2, provided in Attachment 1. It has been determined that a +1 °F adjustment is applicable to the small break LOCA and a +18°F adjustment is applicable to the large break LOCA for North Anna Unit 2.
References:
(1) Letter from J. P. O'Hanlon (VEPCO) to Document Control Desk (USNRC), "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, 30-Day Report of ECCS Evaluation Model Changes Per Requirements of 10CFR50.46,"
Serial No.95-608, November 29, 1995.
(2) Letter from J. P. O'Hanlon (Va. Electric & Power Co.) to USNRC, 'Virginia Electric and Power Company, North Anna and Surry Power Station Units 1 and 2, Report of ECCS Evaluation Model Changes and 30-Day Report of ECCS Evaluation Model Changes Per Requirements of 10CFR50.46," Serial No.96-111, March 14, 1996.
(3) Letter from J. P. O'Hanlon (Va. Electric & Power Co.) to USNRC, "Virginia Electric and Power Company, North Anna Power and Surry Power Station Units 1 and 2, Report of ECCS Evaluation Model Changes and 30-Day Report of ECCS Evaluation Model Changes Per Requirements of 10CFR50.46," Serial No.96-390, August 1, 1996.
(4) Letter from J. P. O'Hanlon (Va. Electric & Power Co.) to USNRC, "Virginia Electric and Power Company, Surry and North Anna Power Stations Units 1 and 2, Report of Emergency Core Cooling System (ECCS) Evaluation Changes Pursuant to the Requirements of 10CFR50.46,II Serial No.97-174., March 27, 1997.
e ATTACHMENT 4 EFFECT OF ECCS EVALUATION MODEL CHANGES SURRY UNITS 1 AND 2
e e Effect of Westinghouse ECCS Evaluation Model Changes - Surry The information provided herein is applicable to Surry Power Station, Units 1 and 2. It is based upon reports from Westinghouse Electric Corporation for issues involving the ECCS evaluation models and plant-specific application of the models in the existing analyses. Peak cladding temperature (PCT) values and margin allocations represent issues for which permanent resolutions have been implemented. The assessments for small break and large break LOCA are presented in Sections A and B, respectively.
Section A - Small Break LOCA Margin Utilization - Surry Units 1 and 2 A. PCT for Analysis of Record (AOR) 1717°F (1)
B. Prior PCT Assessments Allocated to AOR SBLOCA Augmented PCT for AOR 1717°F C. PCT Assessments for 10CFR50.46(a)(3)(i) Accumulation {1}
SBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 1717°F Section B - Large Break LOCA Margin Utilization - Surry Units 1 and 2 A. PCT for Analysis of Record (AOR) 2120°F (2)
B. Prior PCT Assessments Allocated to AOR -16°F
- 1. ZIRLO' Cladding -16°F LBLOCA Augmented PCT for AOR 2104°F C. PCT Assessments for 10CFR50.46(a)(3)(i) Accumulation {1} 21°F
- 1. Vessel & SG Calculation Errors in LUCIFER -6°F (2)
- 2. LBLOCA Rod Internal Pressure Issues 0°F (2)
- 3. Translation of Fluid Conditions from SATAN to LOCTA +15°F (3)
LBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 2113°F Notes { } and References ( ) on the following page.
Effect of Westinghouse ECCS Evaluation Model Changes - Surry Notes:
{1} The accumulation of changes (sum of absolute magnitudes) is less than 50°F and is not significant, as defined in 10CFR50.46(a)(3)(i).
References:
(1) Letter from W. L. Stewart 0fa. Electric & Power Co.) to NRC, "Surry Power Station Units 1 and 2 - Proposed Technical Specifications Changes F! Increase/Statistical DNBR Methodology," Serial No.91-374, July 8, 1991.
(2) Letter from W. L. Stewart 0/EPCO) to Document Control Desk (USNRC), "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, 30-Day Report of ECCS Evaluation Model Changes Per Requirements of 10CFR50.46," Serial No.94-254, April 27, 1994.
(3) Letter from J. P. O'Hanlon 0fa. Electric & Power Co.) to USNRC, 'Virginia Electric and Power.Company, Surry and North Anna Power Stations Units 1 and 2, Report of Emergency Core Cooling System (ECCS) Evaluation Changes Pursuant to the Requirements of 10CFR50.46," Serial No.97-174, March 27, 1997.