ML17158B632
| ML17158B632 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 05/29/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17158B631 | List: |
| References | |
| NUDOCS 9605310047 | |
| Download: ML17158B632 (9) | |
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'NITED STATES NUCLEAR REGULATORY COMMlSSlON
'ASHINGTON, D.C. 2055&4001 V
UATION BY THE OFFICE OF NUCLEAR REACTOR REGU TION E AT NG T TOPICAL REPORT PL-NF-90-001 SUPP H NT 2 8
NCHMARK OF CASHO-3G ANF-P NNSYLVANIA POWER AND LIGHT COMPAN SUS U HANNA STEAM ELECTRIC STATIO OPERATING LICENSE NUMBERS NPF-14 AND NP -22 DOCKET NOS.
50-387 AND 50-388
1.0 BACKGROUND
Pennsylvania Power and Light Company (PP&L), the licensee for Susquehanna Steam Electric Station (SSES),
by letter dated August 1,
- 1995, submitted the Topical Report PL-NF-90-001, Supplement 2, "Application of Reactor Analysis Methods for BWR Design and Analysis".
The report describes two changes to PP&L's current licensing methods to support advanced fuel designs at the Susquehanna Station.
The first change is to replace the CPH-2 lattice physics code with the CASMO-3G code.
The second change is to replace the XN-3 critical power correlation with the ANF-B correlation.
The POWERPLEX-II core monitoring system will replace the current core monitoring system in parallel with these changes.
The staff requested additional information on the topical report by letter dated December 5,
1995.
By letter'ated January 22,
- 1996, the licensee submitted a response to the staff's request.
The licensee intends to use the proposed methodology and CASMO-3G for a number of fuel designs, including Siemens 9x9-2 and ATRIUM des'igns, ABB's SVEA
- designs, and GE 11, 12, and 13 fuel.
The licensee has stated that prior to using this software for licensing applications for fuel designs not previously used at Susquehanna, PP&L will perform validation of the CASMO-3G models.
The validation process typically involves comparison to vendor provided data or modeling techniques, measured data from another reactor or lead use assembly (LUA) data.
Validation of the ANF-B correlation for new fuel designs is performed by Siemens Power Corporation (SPC) in accordance with approved fuel design criteria.
- 2. 0 EVALUATION PP&L has proposed two changes to current licensing analysis methods to support SPC advanced fuel designs.
The first change is to replace the CPH-2 lattice physics code with the CASHO-3G code, which will be used to support off-line reload design analyses including calculation of neutronics data for SIMULATE-E and generation of kinetics data for RETRAN.
The second change is to replace the XN-3 critical power correlation with the ANF-B correlation, for use with SPC fuel designs.
The POWERPLEX-II core monitoring system will replace the current monitoring system in parallel with these changes.
POWERPLEX-II, developed by SPC, uses CASMO-3G to generate cycle-specific neutronics input and uses the HICROBURN-B nodal simulator code for calculation of power distributions and thermal limits through use of the ANF-B critical power correlation.
9605310047 960529 PDR ADQCK 05000387 P
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The use of CASHO-3G has been approved for other utilities in staff safety evaluations (SEs) dated March 22,
- 1993, March 21, 1990 and March 15, 1990.
Previous topical report submittal s included benchmarking of CASMO-3G/HICROBURN for use at the Duad Cities, Dresden and LaSalle plants', validafion of CASMO-3G for the Yankee
- Rowe, Vermont Yankee, and Maine Yankee plants and benchyarking of MICBURN-3/CASMO-3/TABLES-3/SIMULATE-3for the Vermont Yankee Plant Use of the ANF-B Critical Power correlation has been approved in the staff SE for ANF-1125, "ANFB Critical Power Correlation,"
dated March 8, 1990.
- 2. 1 Benchmarking of the CASMO-3G Code Benchmarking of CASMO-3G included development of nodal simulation models and transient analysis models.
The output of the CASHO-3G based SIMULATE-E model was compared to operating data from the two SSES boiling water reactors.
These analyses consisted of comparisons to both hot and cold core critical k,<< data and traversing incore probe (TIP) neutron flux measurements.
The data used for the benchmarking analysis includes Susquehanna 1 cycles 1
through 7, Susquehanna 2 cycles 1 through 6 and Peach Bottom 2 cycles 1 and 2; The Susquehanna data was used for TIP and core reactivity comparisons.
The Peach Bottom data was used to support the transient analysis benchmarks of the three turbine trip tests and for hot critical k,<< data.
2.1.1 Hot Benchmark Data PPIlL provided results of CASHO-3G/SIMULATE-E based hot k,> calculations and also provided the corresponding CPH-2/SIMULATE-E calculafsons versus core average exposure.
For the hot critical data, operating core conditions are said to be steady state if the core power, core flow, core inlet sub-cooling, core pressure, and control rod pattern have not changed for at least 3 days to allow xenon to equilibrate.
The data base includes data ranging from initial cores which were composed entirely of GE Bx8 fuel to recent core designs which are composed entirely of SPC 9x9-2 fuel.
The total number of data points for the hot k,<< benchmarking from both Susquehanna units was 1189.
The mean and standard deviation of the hot k << for each reactor cycle and for entire database were also provided.
Tfie mean k, for the entire database is 0.99999, with a standard deviation of O.tl017.
These results are typical of previous benchmarking analyses and are within the range of values previously accepted by the staff.
Hot core k,<< CASMO-3G/SIMULATE-E calculations were also compared to data from Peach Bottom 2 cycles l,and 2.
The Peach Bottom 2 cycle 1 core was entirely composed of GE 7x7 fuel, while cycle 2 contained a mix of GE 7x7 and GE 8x8 fuel.
The total number of data points was 24.
The mean k,<< for the database is 0.99590, with a standard deviation of 0.00325.
- 2. 1.2 Cold Benchmark Data CASMO-3G/SIMULATE-E cal'culations for cold k,<< data were provided by the licensee.
Cold critical conditions are considered steady state if the reactor has been shutdown a sufficient amount of time to allow the core to be
t considered xenon-free and the reactor coolant temperature is less than 212 'F to eliminate the possibility of voiding.
The cold k,<< results have been adjusted for the measured moderator temperature and reactor period.
The data included 73 points with varied core average exposures.
The average cold k,<<
for the 73 points was 0.99840, with a standard deviation of 0.00185.
These values are typical of previous benchmarking analyses.
The standard deviation between hot and cold k,<< values was also calculated and is comparable to the CPH-2 value.
Therefore, the licensee will continue to use CPH-2 uncertainties with CASHO when modeling cold reactivity events.
- 2. 1.3 Peach Bottom Turbine Trip Test Benchmark Data taken from three turbine trip tests at the Peach Bottom plant was compared to a CASMO-3G/SIMULATE-E model of the Peach Bottom plant.
A CASMO-3G/SIMULATE-E model for Peach Bottom was developed to prepare the data for RETRAN to determine whether the transient analysis models were adversely impacted by the switch to CASHO-3G.
RETRAN produced results on reactor power for the turbine trip tests which was compared to measured data from Peach Bottom.
The comparisons of measured and CASMO-3G based RETRAN calculated average power range monitor (APRH) core power for the three turbine trip tests were provided, as well as CPH-2 based results.
The table below compares the peak power during the transient (normalized reactor power) of the actual data to the two code models for the three tests.
The results for TT2 and TT3 exhibit an acceptable comparison to the measured response.
For TT1, the result generated by CASHO-3G/RETRAN bounds by the measured value and the value obtained by CPH-2/RETRAN.
These results are acceptable to the staff.
Table 1:
Normalized Peak Reactor Power for Peach Bottom Turbine Tri Tests Measured Data RETRAN CASHO RETRAN(CPM) 4.8 6.6 6.0 TT2 4.5 4.4 4.4 TT3 4.9 4.8 4.9 2.1.4 TIP Comparison Data PP8L compared CASMO-3G/SIMULATE-E model calculations with measured TIP data to determine whether the model is capable of predicting core power distribution.
The TIP detectors are neutron fission detectors located in the local power range monitor (LPRH) instrument.tubes.
SIMULATE-E calculates the TIP response for each six-inch segment by calculating and summing the contributions from each of the four surrounding assemblies.
Heasured TIP data was taken from SSES Unit 1 cycles 1 through 7 and Unit 2 cycles 1 through 6 for a total of 226 TIP comparison data points.
Nodal and radial TIP comparisons have been made to the SSES measured TIP data.
The individual data points (percent variation from measured data) are provided in separate tables by unit and cycle.
The corresponding CPH-2/SIMULATE-E comparisons are also provided for each case.
Overall results are also provided.
The overall average nodal
-C
results from the CASHO-3G based model is 6.9X, compared with 6.3X for the CPH-2 based model.
The axially integrated (radial) overall average is 2.93X for the CASHO-3G based
- model, compared with 2.9X for the CPH-2 model.
The TIP measurements provide a measure of the uncertainty in the SIMULATE-E calculated radial and nodal power distribution.
Because root mean squared (RHS) values are essentially unchanged, the use of CASNO will not significantly effect SIMULATE-E modeling uncertainties.
These results are similar to results obtained for other plants in the benchmarking of CASHO.
If, during operation, the nodal and radial standard deviations approach the 10X or 6X values, respectively, they are beyond what has been historically seen and should be investigated.
Nearly all of the individual TIP measurement points are within the criteria.
2.2 Honitoring System Methods The POWERPLEX-II core monitoring system will be installed at SSES as a
replacement for the current POWERPLEX monitoring system.
POWERPLEX-II employs the MICROBURN-B 3-D simulation code supplied by SPC.
The neutronics input to MICROBURN-8 is generated by CASMO-3G for each cycle.
The staff has previously approved the use of CASMO-3G/HICROBURN-B in reload design apd licensing applications in the safety evaluation dated August 13, 1990 The licensee states that the methods employed in MICROBURN-B in the POWERPLEX-II system and the version used for benchmarking analysis by SPC differ in two ways.
One difference is that PP8L is using CASNO-3G version 4.7 and SPC is using CASHO-3G version 4. 1 for benchmarking.
Both versions use the same neutronics methods and cross section libraries, but version 4.7 can model a wider range of fuel assembly designs and additional isotopic depletion chains.
The other difference is due to PP&L's use of ESCORE to calculate the fuel and cladding temperatures for the input deck.
The licensee states that the net effect of both of these differences is approximately 0.004 delta k for exposures up to 40 GWD/NTU.
The changes are acceptable due to the relatively minor effect on the calculation of k,ff.
2.3 Use of the ANF-B Correlation The ANF-B correlation was developed to provide a generic tool for evaluation of the DNB heat flux for all ANF BWR fuel designs and has been incorporated into SINULATE-E.
[he staff has previously approved of the ANF-B correlation on a generic basis The CPR related events that utilize steady-state core physics methods analyzed as part of a reload analysis include Rod Withdrawal Error, Nislocated
- Bundle, Loss of Feedwater Heating and Rotated Bundle.
The licensee has stated that evaluation of the rod withdrawal error will continue to be performed in accordance with previously approved methodology, except that the ANF-B correlation'ill be used to calculate delta-CPR instead of XN-3.
The mislocated bundle analysis is performed on a cycle-s~ecific basis.
A bounding analysis for the mislocated bundle developed by SPC is currently under review by the staff.
The licensee acknowledged this separate review effort and intends to implement the SPC bounding analysis if the analysis is approved by the staff.
For the Loss of Feedwater Heating event, the licensee intends to use the result of previously approved SPC generic analysis
Therefore, incorporation of the ANF-B correlation is acceptable.
The staff will review changes to PP&L's rotated bundle analysis methodology separately and is not approved herein.
- However, PP&L should continue to use the approved rotated bundle methodology and can incorporate the ANF-B critical power correlation into the rotated bundle methodology.
This issue is currently under generic review by the staff.
- 3. 0 CONCLUSION The licensee has proposed two changes to the current licensing methodology to support advanced fuel designs at the SSES.
The first change is to replace the CPM-2 lattice physics code with the CASMO-3G code.
The second change is to replace the XN-3 critical power correlation with the ANF-B correlation.
The POWERPLEX-II core monitoring system will replace the current core monitoring system in parallel with these changes.
Benchmarking of CASMO-3G for use at SSES included hot and cold k,ff calculations, comparison to Peach Bottom turbine trip data, and comparison to TIP values.
The benchmarking of the codes by the licensee relative to measurements from operating reactors and experimental configurations resulted in agreement typical of that observed with accepted methods.
The ll89 hot k << and 73 cold k,<< data points provide a sufficient database for comparison.
Tfie nodal and integral RMS values for 226 TIP observations were also provided.
The RMS values were comparable to CPM-2 results and are typical of the range of RMS values previously approved by the staff.
The results of the Peach Bottom turbine trip analyses were comparable to both actual data and CPM-2 benchmarking data.
Therefore, use of CASMO-3G is acceptable.
The ANF-B critical power correlation has also been previously approved by the staff on a generic basis.
Use of the ANF-B correlation and the methodology used for analysis of reload specific events as described in Supplement 2 is acceptable.
The staff will review changes to PP&L's rotated bundle analysis separately.
This issue is currently under generic review by the staff.
The licensee must also consider the limitations placed upon the approval of the CASMO-3G code and the ANF-B critical power correlation, as stated in references 4 and 5 and ensure that these limitations are not exceeded during use of CASMO-3G and the ANF-B correlation.
Principal Contributors:
G. Golub E. Kendrick Date:
May 29 1996
~REFERE CES Letter'rom C.
P. Patel, NRC, to T. J.
- Kovach, Commonwealth Edison
- Company, "Commonwealth Edison Company Topical Report NFSR-0091,
'Benchmark of CASHO/MICROBURN BWR Nuclear Design Methods'," dated March 22, 1993.
2.
3.
4, 5.
6.
7.
Letter from A. C. Thadani, NRC, to G. Papanic, YAEC, "Acceptance for Referencing of Topical Report YAEC-1363,
'CASHO-3G Validation'," dated March 21, 1990.
Letter from A. C. Thadani, NRC, to R.
W. Capstick, Vermont Yankee Nuclear Power Corporation, "Acceptance for Referencing of Topical Report YAEC-1683,
'HICBURN-3/CASMO-3/TABLES-3/SIMULATE-3 Benchmarking of Vermont Yankee Cycles 9 through 13'," dated March 15, 1990.
Safety Evaluation for the Topical Report ANF-1125P and Supplement 1,
"ANFB Critical Power Correlation," dated March 8, 1990.
Letter from A. C. Thadani, NRC, to R. A. Copeland, ANF, "Acceptance for Referencing of Topical Report XN-NF-80-19(P), Vol. 1, Sup. 3, "Advanced Nuclear Fuels methodology for Boiling Water Reactors; Benchmark Results for the CASHO-3G/HICROBURN-B Calculation Methodology," dated August 13, 1990.
"Bounding Fuel Assembly Hislocation Accident Analysis for Boiling Water Reactors",
EHF-93-205, Siemens Power Corporation, February 1994 (P).
"Loss of Feedwater Heating Transient in Boiling Water Reactors",
ANF-1358(P)(A),
Rev.
1, September 1992.