ML993140059
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D.C. 20o55-0001
March 11, 1999
Mr. Glen A. Watford, Manager
General Electric Company
P.0. Box 780
Wilmington, NC 28402
SUBJECT:ACCEPTANCE FOR REFERENCING OF LICENSING TOPICAL REPORTS
NEDC-32601 P, METHODOLOGY AND UNCERTAINTIES FOR SAFETY LIMIT
MCPR EVALUATIONS; NEDC-32694P, POWER DISTRIBUTION
UNCERTAINTIES FOR SAFETY LIMIT MCPR EVALUATION; AND
AMENDMENT 25 TO NEDE-2401 1-P-A ON CYCLE-SPECIFIC SAFETY LIMIT
MCPR (TAC NOS. M97490, M99069 AND M97491)
Dear Mr. Watford:
The staff has reviewed the subject reports submitted by GE Nuclear Energy (GENE) by letters dated December 13, 1996, for NEDC-32601P; June 10, 1997, for NEDC-32694P; and
December 13, 1996, for Amendment 25 to NEDE-24011P. These submittals provide (1) the description of the procedures used to account for the reload-specific core design and operation in determining the cycle-specific safety limit minimum critical power ratio (SLMCPR) in NEDC 32601 P; (2) the power distribution uncertainty for the new GE 3D-MONICORE core surveillance
system in NEDC-32694P; and (3) the methodology and uncertainties required for the implementation of cycle-specific SLMCPR in Amendment 25 to NEDE-2401 1-P-A. The staff has found the subject reports to be acceptable for referencing in license applications to the extent specified and under the limitations stated in the GENE letter dated March 1, 1999, the enclosed
report, and the U. S. Nuclear Regulatory Commission (NRC) technical evaluation. The
evaluation defines the basis for acceptance of the report.
The staff will not repeat its review of the matters described in the GENE Topical Reports NEDC 32601P, NEDC-32694, and Amendment 25 to NEDE-2401 1-P-A and found acceptable when this letter request appears as a reference in license applications, except to ensure that the material presented applies to the specific plant involved. NRC acceptance applies only to the
matters described in the GENE Topical Reports NEDC-32601 P, NEDC-32694P, and
Amendment 25 to NEDE-2401 1-P-A. In accordance with procedures established in NUREG-0390, the NRC requests that GE publish accepted versions of the submittals, proprietary and non-proprietary, within 3 months of receipt of this letter. The accepted versions
shall incorporate this letter and the enclosed evaluation between the title page and the abstract
and an -A (designating accepted) following the report identification symbol.
'K>i
..............-
Mr. Glen A. Watford
If the NRC's criteria or regulations change so that its conclusions that the submittal is acceptable are invalidated, GE and/or the applicant referencing the submittal will be expected to revise and resubmit its respective documentation, or submit justification for the continued applicability of the submittal without revision of the respective documentation.
Sincerely,
erank
Generic Issues and Environmental Project Branch
Division of Reactor Program Management Office of Nuclear Reactor Regulation
Enclosures:
NEDC-32601P, NEDC-32694P, and Amendment 25 to NEDE-24011-P-A Evaluation
.... . .............
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D.C. 20555-0001
5 0 ENCLOSURE I
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO GENERAL ELECTRIC LICENSING TOPICAL REPORTS NEDC-32601 P, NEDC-32694P, AND AMENDMENT 15 to NEDE-2401 1-P-A
1. INTRODUCTION
By letters dated December 13, 1996, June 10, 1997, and December 13, 1996, from R. J. Reda (GE) to USNRC, General Electric Nuclear Energy (GENE) submitted licensing topical reports: NEDC-32601 P, 'Methodology and Uncertainties for Safety Limit MCPR Evaluation" (Reference 1); NEDC-32694P, "Power Distribution Uncertainties for Safety Limit MCPR Evaluation," (Reference 2); and Amendment 25 to NEDE-2401 1-P-A, Proposed Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR II) on Cycle-Specific Safety Limit MCPR," (Reference 3), respectively. The purpose of the submittal is (1) for NEDC32601P to update values of the CPR correlation uncertainties contained in NEDE-10958-P-A (GETAB, Reference 4) based on the most recent analysis of available data; (2) for NEDC32694P to update values of the power distribution uncertainties contained in NEDE-31152P, Revision 5 based on the most recent analysis of available data, and (3) for Amendment 25 to NEDE 24011-P-A to provide for cycle-specific Safety Limit Minimum Critical Power Ratios
(MCPRs).
The NRC staff was assisted in this review by its consultant, Brookhaven National Laboratory (BNL). The NRC staff's evaluation includes those three topical reports and the responses to staff's Request for Additional Information (RAI) dated January 8,1998 (GA W-98-002, MFN-004-98, Reference 5), January 9,1998 (GAW-98-003, MFN-005-98, Reference 6), January 28, 1998 (GAW-98-005, MFN-008-98, Reference 7), April 17, 1998 (GAW-98-009, Reference 8), and July 29, 1998 (GAW-98-012, MFN-01 7-98, Reference 9). The staff adopted the findings recommended in our consultant's Technical Evaluation Report (Enclosure 2).
2 EVALUATION
This review includes three topical reports involving the Safety Limit Minimum Critical Power Ratio (SLMC PR) methodology and input uncertainties described in NEDC-32601 P, the methodology for constructing the bounding statepoint power distribution described in NEDC-32694P, and the overall procedures for determining the cycle-specific SLMCPR described in Amendment 25 to GESTAR II. The details of the evaluation are provided in Enclosure 2.
2.1 Methodology and Uncertainties for Safety Limit MCPR Evaluation (NEDC-32601 P)
The topical report, NEDC-32601 P, provides an update to the Safety Limit MCPR methodology and inputs to be used in the evaluation of the Safety Limit MCPR for BWRs (GETAB, Reference 4) including plant surveillance measurement uncertainties and local R-Factor uncertainties. The plant surveillance component uncertainties include the reactor pressure, feedwater temperature and flow, core inlet temperature and flow, and channel flow area and friction factors. The plant surveillance uncertainty revisions are based on current BWR practice, and are generally evaluated using the error methodology (Reference 10). The
2
R-Factor provides the critical power dependence on the local pin power distribution (References 4 and 11) in the GEXL correlation. The R-factor uncertainty analysis includes an allowance for power peaking modeling uncertainty, manufacturing uncertainty and channel
bow uncertainty.
Based on the review of the NEDC-32601 P topical report and the responses to the staff's request for additional information (RAI) (References 5, 8, and 9), we find the SLMCPR methodology and associated uncertainties to be acceptable, however, actions should be taken
as follows:
(1) The TGBLA fuel rod power calculational uncertainty should be verified when applied to fuel designs not included in the benchmark comparisions of Table 3.1 of Reference 1, since changes in fuel design can have a significant effect on calculation accuracy.
(2) The effect of the correlation of rod power calculation uncertainties should be reevaluated to insure the accuracy of the R-Factor uncertainty when the methodology is applied to a
new fuel lattice.
(3) In view of the importance of MIP criterion and its potential sensitivity to changes in fuel bundle designs, core loading and operating strategies, the MIP criterion should be reviewed periodically as part of the procedural review process to insure that the specific value recommended in NEDC-32601p is applicable to future designs and operating strategies.
2.2 Power Distribution Uncertainties for Safety Limit MCPR Evaluations (NEDC-32694P)
The power distribution uncertainty topical report NEDC-32694P provides a description of the 3D-MONICORE core surveillance system and the determination of the associated bundle
power uncertainty for use in SLMCPR calculation. The 3D-MONCORE system uses three dimensional coarse-mesh diffusion theory methods, together with models for interfacing with the incore TIP and LPRM instrumentation, to determine the detailed core statepoint. The uncertainty in the 3D-MON ICORE prediction of bundle power was determined by comparisions of measured and calculated TIP integrals and gamma scanned bundle powers.
Based on the review of Reference 2 and the responses to the staff's RAI (References 6 and 8) we have found that the 3D-MONICORE power distribution uncertainties are acceptable for determining the SLMCPR, MAPLHGR and LHGR core limits, however, the 3D-MONICORE bundle power calculational uncertainty should be verified when applied to fuel and core designs not included in the benchmark comparisons of Tables 3.1 and 3.2 of Reference 2.
2.3 Amendment 25 to GE Licensinq Topical Report NEDE-2401 1-P-A (GESTAR II) on Cycle Specific Safety Limit MCPR
Amendment 25 to GESTAR I provides the methodology and uncertainties required for the implementation of cycle-specific Safety Limit MCPRs that replace the former generic, bounding SLMCPR. General procedures are given describing the analysis to be used in determining the cycle-specific SLMCPR. These procedures require that the analysis be performed for the specific fuel bundle design and core loading used in the cycle reload design.
Based on the review of References 3 and 7, we have found that the proposed methodology to be acceptable for performing cycle-specific SLMCPR analyses.
3
3 CONCLUSION
Based on our review of Topical Reports NEDC-32601 P, NEDC-32694P, and Amendment 25 to NEDE-2401 1-P-A (GESTAR II), the staff concludes that the input plant system uncertainties, the power distribution uncertainties associated with the application of 3D-MON ICORE, and the proposed cycle-specific determination of the SLMCPR are acceptable. In letter dated PM TO SUPPLY, GENE has stated that they will take the following actions whenever a new fuel
design is introduced.
(1) The TGBLA fuel rod power calculational uncertainty should be verified when applied to fuel designs not included in the benchmark comparisions of Table 3.1 of NEDC-32601 P, since changes in fuel design can have a significant effect on calculation accuracy.
(2) The effect of the correlation of rod power calculation uncertainties should be reevaluated
to insure the accuracy of R-Factor uncertainty when the methodology is applied to a new
fuel lattice.
(3) In view of the importance of MIP criterion and its potential sensitivity to changes in fuel bundle designs, core loading and operating strategies, the MIP criterion should be reviewed periodically as part of the procedural review process to insure that the specific value recommended in NEDC-32601 P is applicable to future designs and operating
strategies.
(4) The 3D-MON ICORE bundle power calculational uncertainty should be verified when
applied to fuel and core designs not included in the benchmark comparisions in
Tables 3.1 and 3.2 of NEDC-32694P.
4 REFERENCES
1. GE Letter RJR-96-139 MFN-185-96 dated December 13, 1996 from R. J. Reda to USNRC transmitting a topical report, NEDC-32601 P, "Methodology and Uncertainties for Safety Limit MCPR Evaluations," December 1996.
2. GE Letter RJR-97-074 MFN-022-97 dated June 10, 1997 form R. J. Reda to USNRC
transmitting a topical report, NEDC-32694P, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations," January 1997.
3. GE Letter RJR-96-133 MFN-179-96 from R. J. Reda to USNRC, "Proposed Amendment 25 to GE Licensing Topical Report NEDE-2401 1-P-A (GESTAR II) on
Cycle Specific Safety Limit MCPR," December 13,1996.
4. General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDE-10958-PA, January 1977.
5. GE Letter GAW-98-002 MFN-004-98 from Glen A. Watford to USNRC, Responses to Request for Additional Information for GE Topical Report NEDC-32601 P, Methodology and Uncertainties for Safety Limit MCPR Evaluations, January 8,1998.
6. GE Letter GAW-98-003 MFN-005-98 from Glen A. Watford to USNRC, Responses to
Request for Additional Information for GE Topical Report NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evaluations, January 9,1998.
4
7. GE Letter GAW-98-005 MFN-008-98 from G. A. Waterford to USNRC, Responses to Request for Additional Information for Amendment 25 to GE Topical Report NEDE-24011-P-A (GESTAR 11) on Cycle-Specific Safety Limit MCPR (TAC No. M97491), January 28, 1998.
8. GE Letter GAW-98-009 MFN-014-98 from Glen A. Watford to USNRC, Responses to NRC Request for Additional Information associated with SLMCPR Methodology and Uncertainty Topical Reports NEDC-32601P and NEDC-32694P, April 17, 1998.
8. GE Letter GAW-98-012 MFN-017-98 from Glen A. Watford to USNRC, Additional Information Associated with SLMCPR Methodology and Uncertainty Topical Report NEDC-32601P, July 29, 1998.
10. "Recommended Practice - Setpoint Methodologies," Part II, ISA-RP 67.04, Instrument
Society of America, September 1994.
11. NEDC-32505, Revision 1, "R-Factor Calculation Method for GEII, GEI2, and GEI3
Fuel," June 1997.
ENCLOSURE 2
TECHNICAL EVALUATION REPORT
1) Power Distribution Uncertainties for Safety Limit MCPR Evaluations
2)
3)
Report Numbers:
Methodology and Uncertainties for Safety Limit MCPR
Amendment 25 to GE Licensing Topical Report NEDE-2401 I-P-A
(GESTAR) on Cycle-Specific Safety Limit MCPR
1) NEDC-32694P
2) NEDC-32601P
3) NEDE-24011-P-A
Report Dates: 1)
2)
January 1997
December 1996
3) December 1996
Originating Organization: General Electric Company
1.0 INTRODUCTION
In Reference-i, the General Electric Company (GE) has submitted the proposed GESTAR
modifications for including the cycle-specific Safety Limit MCPR (SLMCPR), replacing the generic
bounding SLMCPR methodology included in GESTAR, for NRC review and approval. These
modifications provide the licensing methods to be used in determining the cycle-specific SLMCPR
for each plant reload. In support of these modifications, GE has submitted the two additional
licensing topical reports: (1) NEDC-32601P (Reference-2), "Methodology and Uncertainties for
Safety Limit MCPR," and (2) NEDC-32694P (Reference-3), "Power Distribution Uncertainties for
Safety Limit MCPR Evaluations." The NEDC-32601P Topical Report describes the procedures used
to account for the reload-specific core design and operation in determining the cycle-specific
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Report Titles:
SLMCPR. In this topical report, the values of the plant monitoring uncertainties and local R-Factor
uncertainty used in the SLMCPR determination are also reviewed and updated to reflect current
recommended practices.
The NEDC-32694P Topical Report provides the power distribution uncertainty for the new GE 3D
MONICORE core surveillance system. The 3D-MONICORE power distribution uncertainties are
determined based on an uncertainty propagation analysis and on comparisons with benchmark
measurements. The resulting 3D-MONICORE uncertainties are used in the determination of the
SLMCPR for the plants employing the 3D-MONJCORE system.
The review of the GE core monitoring and SLMCPR analysis was included in the NRC vendor
inspections (Nos. 99900003/95-01 and 99900003/96-01) at the General Electric Nuclear Energy
Facility in Wilmington, NC during the weeks of August 14 through September 1, 1995 and May 6
through May 10, 1996. Several important concerns were identified during these reviews including:
(1) the level of conservatism in the operating state assumed in the cycle-specific determination of the
SLMCPR and (2) the effect of the 3D-MONICORE uncertainties on the SLMCPR uncertainty
analysis. These concerns are addressed in the safety limit methodology and uncertainty analysis
Topical Report NEDC-32694P and the power distribution uncertainty Topical Report NEDC32694P,
respectively.
The purpose of this review was to evaluate these methodology modifications and updates to insure
that the changes in the monitoring uncertainties are acceptable and that adequate margin is included
in the determination of the SLMCPR. The methodology changes are summarized in Section 2, and
the evaluation of the important technical issues raised during this review is presented in Section 3.
The Technical Position is given in Section 4.
2.0 SUMMARY OF THE REVISED SLMCPR METHODOLOGY
2.1 Power Distribution Uncertainties for Safety Limit MCPR Evaluations (NEDC-32694P)
The power distribution uncertainty Topical Report NEDC-32694P provides (1) a description of the
3D-MONICORE core surveillance system and (2) the determination ofthe associated bundle power
uncertainty for use in SLMCPR calculations. The 3D-MONICORE system uses three-dimensional
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coarse-mesh diffusion theory methods, together with models for interfacing with the incore TIP and
LPRM instrumentation, to determine the detailed core statepoint. The physics methods used in 3D
MONICORE are identical to those used in BWR fuel design calculations and core management
evaluations. 3D-MONICORE solves a modified diffusion theory equation in order to allow the local
normalization of the power distribution to the TIP and LPRM incore measurements. However, prior
to this normalization, the TIP/LPRM measurements are compared to the instrument responses
predicted by 3D-MONICORE. If these comparisons indicate that certain measurements are suspect,
this data is rejected and the normalization is performed with the remaining reliable TIP/LPRM
measurements.
The uncertainty in the 3D-MONICORE prediction of bundle power was determined by comparisons
of measured and calculated TIP integrals and gamma scanned bundle powers. These comparisons
included a wide range of fuel enrichments, poison loadings and operating conditions. The increased
uncertainty between TIP measurements was determined by comparing LPRM-updated TIPs and TIP
measurements taken immediately following the LPRM update. The uncertainty analysis also accounts
for TIP and LPRM failures (i.e., measurement rejection). The NEDC-32694P uncertainty analysis
indicates that the 3D-MONICORE power distribution uncertainty is less than the value presently
used in the GETAB SLMCPR determination.
2.2 Methodology and Uncertainties for Safety Limit MCPR (NEDC-32601P)
The NEDC-32601P Topical Report documents the latest updates to the GETAB (Reference-4)
(1) plant surveillance measurement uncertainties, (2) local R-Factor uncertainties and (3) SLMCPR
methodology. The plant surveillance component uncertainties include the reactor pressure, feedwater
temperature and flow, core inlet temperature and flow, and channel flow area and friction factors.
The plant surveillance uncertainty revisions are based on current BWR practice, and are generally
evaluated using the error methodology of Reference-5. The uncertainty analysis accounts for the
overall instrument channel accuracy, drift, calibration, process uncertainty, and plant environmental
effects. In most cases, a simple sum-of-the-squares combination of the contributing uncertainties is
employed, however, the uncertainty in the inlet subcooling (i.e., core inlet temperature) is determined
using the process computer heat balance to propagate the uncertainties.
In most cases, the reevaluation of the plant surveillance uncertainties concluded that the presently
accepted GETAB uncertainty values are conservative. A detailed analysis is provided to support the
revised values in the cases where the reevaluation results in a reduction in the component uncertainty.
In the GEXL correlation, the R-Factor provides the critical power dependence on the local pin power
distribution (References 4 and 6). The R-Factor uncertainty analysis includes an allowance for
power peaking modeling uncertainty, manufacturing uncertainty and channel bow uncertainty. The
TGBLA (Reference-7) power peaking modeling uncertainty is determined by comparisons of
TGBLA with MCNP (Reference-8) and quarter-core benchmark calculations for a range of BWR
fuel bundle and core designs. The power peaking uncertainty determined by this analysis was
confirmed with gamma scan measurements taken following Cycle-8 of the Duane Arnold Plant
(Reference-9).
The uncertainty in the power peaking resulting from channel bow is determined using the procedures
of Reference- 10, and the uncertainty introduced by the manufacturing process is baseci on estimated
fuel enrichment and density measurements. The R-Factor uncertainty is determined by propagating
the resulting local power peaking uncertainty using the R-Factor dependence on peaking factor.
The revised methodology includes updates to the calculation process used to determine the
SLMCPR. The operating core statepoint is determined using the PANACEA (Reference-7)
3D-simulator program. The statepoint information used in the SLMCPR calculation includes the
channel flows, bundle powers, local void fraction and the TIP detector responses. In addition, the
bundle and exposure dependent R-Factors are obtained from the PANACEA statepoint data and used
to determine the critical power. The SLMCPR is determined by randomizing the statepoint
surveillance input and correlation data to determine the MCPR margin required to insure that 99.9 %
of the rods avoid boiling transition.
The SLMCPR is sensitive to the assumed statepoint radial power distribution. In the cycle-specific
methodology, the power distribution is selected to provide a reasonable bound on the number of rods
expected to experience boiling transition. This selection is made subject to the condition that the
core is critical and within thermal limits. For current BWR reload designs the limiting radial power
distribution includes a centrally located high powered region which is either circular or annular in
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...... ..... .
shape. Control rod patterns which provide these limiting power distributions are described and
recommended. In order to quantify the severity of power distributions with respect to the number
of rods in boiling transition a core weighting parameter is defined. The frequency distribution of this
parameter is used to compare and select the limiting power distribution.
The determination of the SLMCPR using the revised methodology and input uncertainties is
compared to the presently accepted GETAB methodology for several plants. For the cases evaluated,
the effect of the changes in methodology and uncertainties is small -. 01 ASLMCPR.
2.3 GESTAR II Amendment 25 on Cycle-Specific Safety Limit MCPR (NEDE-24011-P-A)
Amendment 25 to GESTAR II provides the methodology and uncertainties required for the
implementation of the cycle-specific Safety Limit MCPR. A set of general procedures are given describing the analysis to be used in determining the cycle-specific SLMCPR. These procedures require that the analysis be performed for the specific fuel bundle design and core loading used in
the cycle reload design. The core radial power distribution must represent a reasonable bound on
the number of fuel bundles at or near thermal limits, and the fuel assembly local power distribution
must be based on the actual bundle design. The cycle-specific analysis is performed at multiple
exposure points throughout the cycle, and either the most limiting or an exposure-dependent
SLMCPR is used in determining the Operating Limit MCPR (OLMCPR). The cycle-specific procedures require that the SLMCPR be recalculated or' reconfirmed for each plant operating cycle.
In the reload process, the final core loading plan is evaluated relative to the reference design criteria
including the OLMCPR. If the cycle-specific determination results in an increased SLMCPR, the
final core loading plan may fail to satisfy the specified acceptance criteria. In this case, calculations
of the sensitivity of the OLMCPR to changes in the SLMCPR are used to determine the acceptability
of the calculated cycle-specific SLMCPR.
While Amendment 25 provides the overall procedures for determining the cycle-specific SLMCPR, the detailed SLMCPR methodology and input uncertainties are described in NEDC-32601P and the
methodology for constructing the bounding statepoint power distribution is described in NEDC 32694P.
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3.0 SUMMARY OF THE TECHNICAL EVALUATION
The GE Topical Reports NEDC-32694P, NEDC-32601P and Amendment 25 to NEDE-24011-P-1
(GESTAR II) provide the basis for the cycle-specific determination of the SLMCPR, input plant
system uncertainties and the power distribution uncertainties associated with the application of
3D-MONICORE. The review of the GE methodology focused on: (1) the assumptions made in the
cycle-specific SLMCPR methodology and the changes relative to the presently approved generic
SLMCPR approach and (2) the basis for the changes in the SLMCPR uncertainty values. As a
result of the review of the methodology, several important technical issues were identified which
required additional information and clarification from GE. This information was requested in
References-11 and 12 and was provided in the GE responses included in References 13-16. This
evaluation is based on the material presented in the topical reports (References 1-3) and in
References 13-19. The evaluation of the major issues raised during this review are summarized in the
following.
3.1 Power Distribution Uncertainties for Safety Limit MCPR Evaluations (NEDC-32694P)
The 3D-MONICORE system is used to perform the steady-state on-line core performance
evaluation. The 3D-MONLCORE models are based on accepted BWR calculational methods. The
neutronics model is essentially the same as that described in Reference-7 and the thermal-hydraulics
model is the same as presently used in the P-I Process Computer Analysis (Reference- 13, Response
11.4).
The 3D-MONICORE power distribution uncertainties are required for determining the SLMCPR,
LHGR and MAPLHGR limits. The (axially integrated) bundle power uncertainty is required for the
SLMCPR and the nodal power uncertainty is required for detennining MAPLHGR and LHGR. The
radial bundle power uncertainty is considered to be a statistical combination of: (1) the uncertainty
in the four-bundle power associated with the TIP location and (2) the uncertainty in the allocation
of the four-bundle power to the individual bundles. The four-bundle power uncertainty is
determined by a comparison of the predicted and measured TIP responses, and the uncertainty in the
power allocation is determined by comparisons of calculated and measured (gamma-scanned) bundle
powers.
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While the calculated bundle powers were determined with the "core tracking" system, rather than
with 3D-MONICORE, GE has indicated (in Reference- 13, Responses 1.2 and 1.6) that the difference
in these codes has no effect on the uncertainty estimates. The TIP comparisons include cores with
both part length fuel rods and axially zoned gadolinium, and all current fuel product lines except for
GE13. However, in view of the similarity of the GE13, GEl1, and GE12 fuel designs, this is
considered acceptable (Reference- 13, Response-II.4). In addition, GE has indicated that the core
follow calculations employed the same methods to process and accumulate the void-history and fuel
exposure as used in the on-line core surveillance (Reference-I 3, Response-tI.8). However, it is
concluded that since changes in the fuel and core design can have a significant effect on the
calculation accuracy, the 3D-MONICORE bundle power calculational uncertainty should be verified
when applied to fuel and core designs not included in the benchmark comparisons of Tables-3.1 and
3.2 of NEDC-32694P.
The review of the calculation-to-measurement (C/M) comparisons indicated an increased uncertainty
at end-of-cycle. However, the cycle-average four-bundle power uncertainty is considered acceptable
since the uncertainty estimate does not take credit for the uncertainty increase due to TIP
measurement uncertainty. The nodal power uncertainty is determined by a statistical combination
of the 3D-MONICORE bundle power uncertainty and an accepted TIP axial power uncertainty. The
TIP uncertainty is measured once per cycle to ensure that it satisfies the specified acceptance criteria.
The 3D-MONICORE system allows rejection of the TIP measurement data based on a specified
acceptance criteria. During the review it was noted that the 3D-MONICORE acceptance criteria
will, under certain conditions, reject good TIP measurement data. However, in Responses-I.7 and
1.10 (Reference- 13), GE has indicated that the rejection of TIP data is very rare. In addition, in most
cases TIP rejection is due to poor agreement between measured and calculated data and, when the
acceptance criteria results in rejection of measurements which are in good agreement with the
calculations, the effect on the core power distribution uncertainty is negligible.
The uncertainty methodology determines the effect of TIP and LPRM rejection and the LPRMupdate
of the power distribution using comparisons of calculations and measurements. In these comparisons
the recommended value for the rejection criteria parameter is used. It is noted that after ten years of
operation, no correlation has been observed between the rejected TIP locations and the
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core locations that are difficult to calculate, such as the peripheral core locations, part-length fuel
bundles and partially controlled fuel bundles. It is concluded that the TIP rejections are generally
a result of erroneous measurement data rather than miscalculation. It is also noted that the TIP
rejection only affects the 3D-MONICORE system and the other BWR surveillance systems use the
measured TIP/LPRM data.
The process computer monitors kw/ft and LHGR as well as the SLMCPR. The uncertainty analysis for the 3D-MONICORE LHGR evaluation is provided in Response-II.5 (Reference- 14) and accounts
for the effect of both the TIP and LPRM update uncertainties on the nodal power calculation.
Based on the review of the NEDC-32694P topical report and supporting documentation provided in References 13 and 14, it is concluded that the 3D-MONICORE power distribution uncertainties
are acceptable for determining the SLMCPR, MAPLLIGR and LHGR core limits subject to the
condition identified above (in the third paragraph of this section).
3.2 Methodology and Uncertainties for Safety Limit MCPR (NEDC-32601P)
3.2.1 Process Computer Uncertainties
The reevaluation of the process computer uncertainties provided in the NEDC-32601 P topical report
were reviewed in detail. The topical report provides a description of both the instrumentation and
modeling uncertainties that are required for the SLMCPR analysis. The evaluation of the core inlet
subcooling uncertainty employs the heat balance used by the process computer to relate the inlet
subcooling to the available instrumentation signals. Using this relation, the inlet subcooling variance
is determined by the individual component variances (e.g., feedwater flow and temperature, core
flow, steam carry under fraction). While the coefficients that weight the individual uncertainty
components depend on the reactor statepoint, the analysis neglects this dependence and assumes
constant weighting coefficients. In Response-I. 1 (Reference-15), GE has shown using a Monte Carlo
procedure that these constant weighting coefficients are conservative.
The calculation of the bundle critical power is sensitive to the channel flow area and friction factors.
The two-phase friction factor is based on measurements made at the full scale ATLAS test facility covering a range of power and flow The uncertainty in the two-phase friction factor is based on the
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comparisons with test data. The uncertainty in the single-phase friction factor is determined by
comparison of the calculations to total pressure drop measurements made at the ATLAS facility. In
Response-II.6 (Reference-15), it is noted that, since the total pressure drop measurement includes
both the single-phase and two-phase losses, the inferred single-phase loss coefficient is conservative.
The channel flow area is subjected to random variations due to non-uniform channel bulge and
crud/corrosion buildup which result in channel-to-channel variations in flow. The SLMCPR
uncertainty analysis accounts for the effect of these variations by increasing the uncertainty in the
channel-to-channel friction factor multiplier (Response-I.2, Reference- 15).
3.2.2 R-Factor Uncertainties
The fuel rod power calculational uncertainty determines the GEXL R-Factor uncertainty and is
separated into three components; modeling, manufacturing and bowing. The modeling uncertainty
is determined by comparison of the TGBLA calculation to MCNP benchmark lattice calculations.
The Table-3.1 TGBLAIMCNP rod power comparisons include all GE fuel designs which are
currently loaded in operating BWRs (Response-lI.1, Reference-iS). A range of gadolinium rods is
included in the comparisons in order to simulate the effects of depleted fuel rods (Response-II.2,
Reference- 15). The fuel rod power peaking uncertainty is determined by weighting the variance for
each fuel design by the number of rods in the lattice (Response-II.3, Reference-iS). However it is
concluded that since changes in the fuel lattice design can have a significant effect on the calculation
accuracy, the TGBLA fuel rod power calculational uncertainty should be verified when applied to
fuel designs not included in the benchmark comparisons of Table-3.1 of NEDC-32601P.
In addition to the TGBLA!MCNP comparisons, GE has evaluated the effect of void fraction
uncertainty on the fuel rod power calculation (Response-II.10, Reference-15). Estimates of the
lattice-average void fraction uncertainty were determined by comparison with measurement. The
local void fraction uncertainty was determined by comparison with detailed subchannel calculations.
The effect of the lattice-average and local void fraction uncertainties on the fuel rod power
calculation was determined by sensitivity calculations and found to be negligible.
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The fuel rod manufacturing uncertainty includes the effects of fuel enrichment, density and rod
position uncertainty. The uncertainty in fuel enrichment and density was determined from
measurements on a large number of fuel rods performed as part of manufacturing studies. The fuel
rod position uncertainty was determined from a series of rod spacing measurements performed on
a high bumup fuel bundle. In Responses 11.4, 11.5, and 11.9 of References 14 and 15, GE has shown
that the effects of these uncertainties are conservatively included in the R-Factor analysis. In
Response-II.10 (Reference- 14), the effect of local fuel bundle exposure uncertainty on rod power is
shown to be negligible. It is important to note that the power peaking uncertainty is determined
using a components of uncertainty approach and then independently confirmed by a comparison with
gamma scan measurements.
In the approved GETAB methodology of Reference-4, the power peaking calculation errors in
neighboring fuel rods are assumed to be correlated so that each of the fuel rods has exactly the same
calculational error. In the proposed methodology, the modeling errors in neighboring fuel rods are
assumed to be uncorrelated. As a result, the uncertainty in the R-Factor is reduced significantly in
the proposed methodology. In Response-lI. 13 of Reference 14 and in References 17-19, GE has
evaluated this effect for the 8x8, 9x9 and 1WxiO lattices and has indicated that the R-Factor
uncertainty will be increased (relative to the presently approved value of Reference-4) to account for
the correlation of rod power uncertainties. However, in References-18 and 19 (Table-i), it is noted
that the effect of the rod-to-rod correlation has a significant dependence on the fuel lattice (e.g., 9x9
versus lWx 0). Therefore, in order to insure the adequacy of the R-Factor uncertainty, the effect of
the correlation of rod power calculation uncertainties should be reevaluated when the NEDC-32601P
methodology is applied to a new fuel lattice.
3.2.3 SLMCPR Evaluation Methodology
The SLMCPR is sensitive to the "flatness" of the bundle power distribution of the initial reactor
statepoint. GE has defined a MCPR Importance Parameter (MIP) to allow a quantitative assessment
of the flatness of the power distribution and identify limiting statepoints for SLMCPR analysis. In
Response-1II.2 (Reference- 15), the expression for determining MIP is derived and shown to provide
a quantification of the effect of the bundle power distribution on the SLMCPR. In Reference- 15,
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GE provides the specific MIP criterion (Response-III.5) and the thermal limits and reactivity
constraints (Response-III.6) for selecting the bundle power distribution to be used in the SLMCPR
analysis.
The determination of the selected MIP criterion is based on an extensive evaluation of operating
reactor statepoints. In view of the importance of this MIP criterion and its potential sensitivity to
changes in fuel bundle designs, core loadings and operating strategies, there is a need to insure that
the specific value recommended in NEDC-32601P is applicable to future designs and operating
strategies. In response to this concern, GE has indicated that the MIP criterion will be reviewed
periodically as part of the procedural review process (Response 111.6, Reference-15).
In the presently approved GETAB methodology (Reference-4), the bundle power calculation error
is assigned to the four bundles surrounding the TIP in a correlated manner so that each of the four
bundles is perturbed simultaneously by the same amount. In the proposed methodology, the
calculational error is assumed to be uncorrelated and the individual bundle powers are varied
independently in the Monte Carlo uncertainty propagation. The increased variability in the proposed
methodology results in a (non conservative) reduction in the SLMCPR. In Response-II.,1 of
Reference-14, GE has revised the NEDC-32601P methodology to allow for the correlation of the
bundle power calculation modeling errors.
Based on the review of the NEDC-32601P topical report and supporting documentation provided
in References 14 and 15, we find the SLMCPR methodology and associated uncertainties to be
acceptable subject to the conditions identified in Sections-3.2.2 and 3.2.3.
3.3 GESTAR II Amendment 25 on Cycle-Specific Safety Limit MCPR (NEDE-24011-P-A)
Amendment 25 to GESTAR II provides the modifications required for performing the cycle-specific
SLMCPR analysis. In the cycle-specific analysis, a search is performed to determine the initial
reactor statepoint for use in the Monte Carlo statistical analysis. The purpose of this search is to
determine a reactor statepoint that satisfies both (1) the operations criteria required for operating
statepoints and (2) the MIP flatness criterion to insure that the statepoint provides a bounding
SLMCPR. In the information provided in support of Amendment 25 (Reference-i, Corrective
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Action-4, Item-3), it is noted that this search may be terminated before all criteria are satisfied.
However, in Responses 2 and 3 (Reference- 16), GE has indicated that if the MIP criterion is not
initially satisfied the search will be expanded, by relaxing the operations criteria, to insure that the
MIP criterion is satisfied.
In the presently approved GETAB methodology, the limiting power shape is assumed to include a
centrally located annular ring of high-powered fuel bundles. While the proposed cycle-specific
methodology does not require the power distribution to include this high-powered annular zone, it
is indicated in Response-4 (Reference- 16) that the control rods are selected so that this power shape
is not precluded from the search for the bounding statepoint.
Based on the review of Amendment 25 and the supporting information provided in Reference-16,
we find the proposed methodology to be acceptable for performing cycle-specific SLMCPR analyses.
4.0 TECHNICAL POSITION
The Topical Reports NEDC-32694P, NEDC-32601P and Amendment 25 to NEDE-24011-P-A
(GESTAR 1I) and supporting documentation provided in References 13-16 have been reviewed in
detail. Based on this review, it is concluded that the proposed cycle-specific determination of the
SLMCPR, the input plant system uncertainties, and the power distribution uncertainties associated
with the application of 3D-MONICORE are acceptable subject to the conditions stated in Section
3 of this evaluation and summarized in the following.
1) Since changes in the fuel and core design can have a significant effect on the calculation
accuracy, the 3D-MONICORE bundle power calculational uncertainty should be verified when
applied to fuel and core designs not included in the benchmark comparisons of Tables-3.1 and
3.2 of NEDC-32694P (Section 3.1).
2) Since changes in fuel design can have a significant effect on the calculation accuracy, the
TGBLA fuel rod power calculational uncertainty should be verified when applied to fuel designs
not included in the benchmark comparisons of Table-3.1 of NEDC-32601P (Section-3.2.2).
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3) In order to insure the adequacy of the R-Factor uncertainty, the effect of the correlation of rod
power calculation uncertainties should be reevaluated when the NEDC-32601P methodology is
applied to a new fuel lattice (Section-3.2.2).
4) In view of the importance of the MIP criterion and its potential sensitivity to changes in fuel
bundle designs, core loadings and operating strategies, the MIP criterion should be reviewed
periodically as part of the procedural review process to insure that the specific value
recommended in NEDC-32601P is applicable to future designs and operating strategies
(Section-3.2.3).
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5.0 REFERENCES
1. "Proposed Amendment 25 to GE Licensing Topical Report NEDE-240 11-P-A (GES TAR II)
on Cycle-Specific Safety Limit MCPR," RJR-96-133, Letter R. J. Reda (GE) to U.S. NRC,
dated December 13, 1996.
2. "GE Licensing Topical Report, Methodology and Uncertainties for Safety Limit MCPR
Evaluations," RJR-96-139, Letter, R. J. Reda (GE) to U.S. NRC, dated December 13, 1996.
3. "GE Licensing Topical Report, Power Distribution Uncertainties for Safety Limit MCPR
Evaluations," RJR-97-074, Letter, R. J. Reda (GE) to U.S. NRC, dated June 10, 1997.
4. "General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design
Application," NEDO- 10958-A, January 1977.
5. "Recommended Practice - Setpoint Methodologies," Part II, ISA-PP 67.04, Instrument Society
of America, September 1994.
6. "R-Factor Calculation Method for GEl 1, GE12, and GE13 Fuel," NEDC-32505P, November
1995.
7. "Steady State Nuclear Methods," NEDE-30130-P-A, April 1985.
8. "MCNP A General Monte Carlo N-Particle Transport Code, Version 4a," LA-12625,
J. F. Breismeister, Ed., Los Alamos National Laboratory (1993).
9. "Gamma Scan Measurements of the Lead Test Assembly at the Duane Arnold Energy Center
Following Cycle-8," NEDC-3 1569-P. April 1988.
10. "Fuel Channel Bow Assessment," GENE Report MFN086-89, Letter, J. S. Charnley (GE) to
R. C. Jones (NRC), dated November 15, 1989.
11. "Request for Additional Information for GE Topical Reports NEDC-32601 P and NEDC 32694P," Letter, J. H. Wilson (NRC) to R. J. Reda (GE), dated August 20, 1998.
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12. "Request for Additional Information for Amendment 25 to GE Topical Report NEDE-24011
P-A (GESTAR II) on Cycle-Specific Safety Limit MCPR (TAC No. M97491)," Letter,
J. H. Wilson (NRC) to R. J. Reda (GE), dated October 21, 1998.
13. "Responses to Request for Additional Information for GE Topical Report NEDC-32694P,"
"Power Distribution Uncertainties for Safety Limit MCPR Evaluations," GAW-98-003, Letter,
G. A. Watford (GE) to U.S. NRC, dated January 9, 1998.
14. "Responses to NRC Request for Additional Information Associated with SLMCPR
Methodology and Uncertainty Topical Reports NEDC-32601P and NEDC-32694P," GAW 98
009, Letter, G. A. Watford (GE) to U.S. NRC, dated April 17, 1998.
15. "Responses to Request for Additional Information for GE Topical Report NEDC-32601P,
Methodology and Uncertainties for Safety Limit MCPR Evaluations," GAW-98-002, Letter, G.
A. Wafford (GE) to U.S. NRC, dated January 8, 1998.
16. "Responses to NRC Request for Additional Information for Amendment 25 to GE Topical
Report NEDE-2401 1-P-A (GESTAR II) on Cycle-Specific Safety Limit MCPR (TAC No.
M97491)," GAW-98-005, Letter, G. A. Watford (GE) to U.S. NRC, dated January 28, 1998.
17. "Additional Information Associated with SLMCPR Methodology and Uncertainty Topical
Report NEDC-32601P," GAW-98-012, Letter, G. A. Watford (GE) to U.S. NRC, dated June
12, 1998.
18. "Additional Information Associated with SLMCPR Methodology and Uncertainty Topical
Report NEDC-32601P," GAW-98-012, Letter, G. A. Watford (GE) to U.S. NRC, dated July 29,
1998.
19. "Additional Information Associated with SLMCPR Methodology and Uncertainty Topical
Report NEDC-32601P," GAW-98-017, Letter, G. A. Watford (GE) to U.S. NRC, dated
September 9, 1998.
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