SBK-L-14049, Response to Request for Additional Information Regarding License Amendment Request 13-05, Fixed Incore Detector System Analysis Methodology (Redacted, Non-Proprietary Version)
| ML14078A059 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/12/2014 |
| From: | Walsh K NextEra Energy Seabrook |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14078A056 | List: |
| References | |
| SBK-L-14049 | |
| Download: ML14078A059 (36) | |
Text
Proprietary Information NE)F era' Withhold from public disclosure under 10 CFR 2.390 ENERGY ,,
SEABROOK March 12, 2014 10 CFR 2.390 10 CFR 50.90 SBK-L- 14049 Docket No. 50-443 1I. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Seabrook Station Response to Request for Additional Information Regarding License Amendment Request 13-05 Fixed Incore Detector System Analysis Methodology
References:
- 1. NextEra Energy Seabrook, LLC letter SBK-L- 13121, License Amendment Request 13-05, Fixed Incore Detector System Analysis Methodology, dated September 9, 2013 (ML13260A160)
- 2. NRC letter, Request for Additional Information for License Amendment Request 13-05, Fixed Incore Detector System Analysis Methodology (ML14034A381)
In Reference 1, NextEra Energy Seabrook, LLC (NextEra) submitted License Amendment Request (LAR) 13-05 to revise the Seabrook Station Technical Specifications (TS) in accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR).
The proposed change revises TS 6.8.1.6.b, Core Operating Limits Report, by adding AREVA Licensing Report ANP-3243P, "Seabrook Station Unit 1 Fixed Incore Detector System Analysis Supplement to YAEC-1 855PA," which supplements and modifies the previously approved methodology in YAEC-1 855PA, "Seabrook Station Unit I Fixed Incore Detector System Analysis," October, 1992. The proposed change also modifies the surveillance requirements associated with the heat flux hot channel factor and nuclear enthalpy rise hot channel factor to include revised uncertainty values when measurement is obtained using the fixed incore detector system (FIDS).
In Reference 2, the NRC submitted a request for additional information (RAI) regarding LAR 13-05 to obtain additional information to complete its review.
Attachment 2 of This Letter Contains Proprietary Information Withhold from public disclosure under 10 CFR 2.390 NextEra Energy Seabrook, LLC.
626 Lafayette Rd, Seabrook, NH 03874
Proprietary Information Withhold from public disclosure under 10 CFR 2.390 U.S. Nuclear Regulatory Commission SBK-L-14049/Page 2 of 3 to this letter provides the NextEra redacted, non-proprietary response to the RAI. provides the proprietary response. Information enclosed within brackets is proprietary to AREVA and is requested to be withheld from public disclosure. Responses to RAI questions 3 through 7 in Attachments 1 and 2 are based on AREVA Licensing Report ANP-3243Q1P, Seabrook Station Unit 1 Fixed Incore Detector System Analysis Supplement to YAEC-1855PA Topical Report - Request for Additional Information. Attachment 3 provides the AREVA request for withholding and affidavit.
This response does not alter the conclusion in Reference 1 that the change does not present a significant hazards consideration pursuant to 10 CFR 50.92.
No new commitments are made as a result of this letter.
Should you have any questions regarding this letter, please contact Mr. Michael Ossing, Licensing Manager, at (603) 773-7512.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on March 12-ifZ , 2014 Sincerely, Kevin T. Walsh Site Vice President NextEra Energy Seabrook, LLC Attachment 2 of This Letter Contains Proprietary Information Withhold from public disclosure under 10 CFR 2.390
Proprietary Information Withhold from public disclosure under 10 CFR 2.390 U.S. Nuclear Regulatory Commission SBK-L-14049/Page 3 of 3 Attachments cc: NRC Region I Administrator NRC Project Manager NRC Senior Resident Inspector Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 Mr. John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399 Attachment 2 of This Letter Contains Proprietary Information Withhold from public disclosure under 10 CFR 2.390
Attachment 1 to SBK-L-14049 Redacted, Non-Proprietary Version Response to Request for Additional Information Regarding License Amendment Request 13-05 Fixed Incore Detector System Analysis Methodology Page 1 of 33
RAI Question #1:
Describe the post modification testingprogram that is used to ensure that any manufacturing differencesfrom the originalequipment manufacturerprocess do not create deviationsfrom the originalplatinum incore detector specifications.
Response to RAI Question #1:
The replacement In-Core Detector Assemblies (ICDAs) replicate or improve the physical characteristics of the original InCore Detector Assemblies (ICDA) including replication of the platinum Self-Powered Detector (SPD) design to minimize the potential for changes in the SPD's response to gamma and neutron fluxes.
These physical attributes of the SPD & ICDA are controlled by the specification, fabrication process controls and verified via the Factory Acceptance Testing and Seabrook oversight during fabrication.
Critical physical attributes of the Self-Powered Detector (SPD) include the following as listed below:
" Platinum emitter dimensions, surface area and purity,
- Alumina purity and density,
- Separation between the trailing end of the emitter & the leading end of the compensation lead wire,
" Sheath material composition and geometry,
- Insulation resistance,
Critical physical attributes of the In-Core Detector Assemblies (ICDA) include the following as listed below:
" Coiled flux window length,
- Flux window centerline position,
" Calibration Tube dimensions and material,
- Housing Tube dimensions and material.
A rigorous in-reactor post modification testing (PMT) program was employed at Seabrook Station to verify the detector functional attributes during operation of the initial replacement detectors over several cycles of operation. The PMT demonstrated equivalent operational results between the replacement and original detectors. The finalized fabrication process controls will assure conformance to design specifications and subsequent in-reactor performance of the replacement ICDAs. Subsequently, in-reactor routine performance monitoring of the original and replacement IDCAs will ensure their continued performance to functional design specification.
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Initial Demonstration PMT The PMT is functional verification of the replacement detectors in comparison to the original detectors and; as a system using the Incore Analysis Code. These are special test programs performed in conjunction with the installation of the initial replacement detectors as opposed to the continuing Routine Testing used in subsequent fuel cycles.
The Post Modification Testing of the replacement detectors evaluates the functional attributes of the installed replacement detectors individually as well as in concert with the other installed detectors operating as an integrated system. The individual detector tests include three functional evaluations to demonstrate replacement detector operation that is equivalent or superior to the original detectors.
The three functional evaluations involve trending the individual detector signals over time via the Fixed Incore Detector Data Acquisition System (FIDDAS) to assess detector signal dynamic response to include the following:
Individual SPD Functional Tests
- 1. The Zero Power Detector Signal Thermal Response,
- 2. The Low Power Detector Sensitivity & Linearity Evaluation and
- 3. Lastly the High Power Signal Stability & Reproducibility Determination The integrated system functional tests evaluate the detector incore power distribution analysis results generated with a full core compliment of detectors operating as a system. Here the detector signals are read by the FIDDAS and analyzed with SIMULATE-3 and Fixed Incore Detector Analysis Code (S3FINC). These analyses evaluate the measured to predicted power distribution results produced from signal snapshots of the full core compliment of detectors at various plant state points and include the following:
Integrated System Functional Tests
- 1. Symmetric Detector Signal Comparisons,
- 2. Relative Axial Power Distribution Evaluations and
- 3. Radial Power Distribution Evaluation.
The conclusion of the PMT validated the adequacy of the design specification and fabrication process controls to produce consistent ICDA in conformance to the design specification satisfying the functional performance requirements of the original platinum incore detectors. Factory acceptance testing and routine in reactor performance monitoring of the ICDAs provides additional assurance of the functional performance of the original and replacement detectors.
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RAI Question #2:
Describe what periodic validation will be done, ifany, to.
- a. Ensure that the uncertainty components listed in Table 3 (?bANP-3243P (page37) remain applicable to the Seabrook reactor over the remainderof the reactor's lIfetime.
- b. Ensure that the proposeddepletion correction flictor, wthich is empirically based of historicaldata, continues to be representativefor the new platinum incore detectors.
Response RAI Question #2a:
The Continued Validation for the five (5) Uncertainty Components listed in Table 3 of ANP-3243P is addressed below:
a) Signal Reproducibility, Ya b) Analytical Methods, ab c) Axial Signal Power Shape, u d) Total Detector Processing, Yd e) Integral Detector Processing, aT Signal Reproducibility, *a c Signal Reproducibility, aa only contributes a very small amount to the overall Fixed Incore Detector System (FIDS) analysis uncertainty. This small contribution indicates that close monitoring of subtle changes to the signal reproducibility is not warranted. Instead, monitoring for any significant signal degradation is sufficient.
Signal Reproducibility, Ga is dependent on physical attributes of the detector and the accuracy and stability of the Fixed Incore Detector Data Acquisition System (FIDDAS). Stability of the FIDDAS and the individual Self-Powered Detectors (SPD) is dependent on the circuit continuity of the detectors leads, signal connectors and electro-magnetic noise immunity. The FIDDAS automatically monitors each detector signal for stark or abrupt changes in circuit continuity that may result in a loss of detector compensation or emitter signal whereby the FIDDAS tags the signal unreliable. The FIDDAS instrumentation is also monitored by the Main Plant Computer System to identify detector FIDDAS failure or malfunctions and generates alarms in the Control Room. This automatic periodic validation function is performed at the one minute cycle time of the FIDDAS.
Analytical Methods, rb The physics analysis method uncertainty, 0 b , has not changed. The methods used in the FIDS analysis system (CASMO-3 and SIMULATE-3) have not changed since the licensing of the system.
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Axial Signal Power Shape, a, Axial power shape uncertainty, Tc , was determined by comparing predicted and measured axial power shapes. Data from the SIMULATE-3 code and movable fission chamber measurements were used to determine this component of uncertainty. Since SIMULATE-3 has not been modified, no change in this uncertainty component is expected.
2D Integral Detector Processing, a AND 3D Total Detector Processing, (d Development of the 2D & 3D Detector Processing Uncertainty Components was initially based on the limited data from Cycles 1 and 2 for the initial FIDS license granted via LAR 92-14.
These same Uncertainty Components have now been regenerated within ANP-3242P based on the analysis of data from Cycles 1 through 15 to accommodate many variations in core design and cycle operation. This reanalysis demonstrates very good agreement with the original analysis presented within YAEC-1855PA. Going forward the 2D and 3D Detector Processing RMS values will continue to be evaluated by procedure for each use of the S3FINC code to satisfy Technical Specification surveillance requirements.
Routine use of the FIDS for surveillance purposes requires the procedural application of review criteria values to the following:
Review Criteria Included the Incore Power Distribution Analysis Procedure
- 2D RMS of Symmetric Signal Ratios ( 1% to 1.5%)
- 3D RMS of Symmetric Signal Ratios ( 1% to 2%)
- RMS % Radial Signal Difference for the Average Plane (3% BOL & 1.75% EOL)
- RMS % Radial Signal Difference for the All Detectors (3% BOL & 2.5% EOL)
- Radial Power Distribution Batch Average Relative Error (< or = 2%)
- Radial Power Distribution RMS % for the Average Plane (2.5% BOL & 1.5% EOL)
These procedural RMS review criteria are more restrictive than the 95/95 detector processing uncertainty component (m-p) in ANP-3243P. Evaluation or Review Criteria are not operability criteria for the FIDS analysis system. Satisfaction of these criteria assure that the detector processing uncertainty component will remain bounding. Exceeding these criteria will require an evaluation to determine if the detector processing uncertainty component is affected or whether it is due to real core power distribution phenomena that is being accurately captured by the FIDS analysis system.
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Response to RAI Question #2b Detector depletion is a small effect. See the response to Question 5 for a discussion of the small effect of detector depletion. As a result, deviations in the depletion correction factor (DPC) would not be expected to significantly impact the overall uncertainty of the FIDS analysis system. Nevertheless, routine monitoring of detector performance and system performance during individual flux map analyses discussed above would be expected to capture any significant deviations in DPC and Core Summary Reports that review performance over the cycle would be expected to capture any trends caused by smaller deviations over several cycles.
As indicated in the response to question 2a, above, the S3FINC code calculates signal differences in symmetric detectors as well as differences between measured and predicted signals. The S3FINC code also calculates RMS values for detectors differences for the Average Plane and RMS values for all detectors within each FINC analysis. These symmetric and measured to predicted signal differences as well as the RMS values are evaluated by procedure for significant detector signal anomalies on a monthly basis. These existing monitoring practices continue today and will continue in the future to ensure that the DPC factor remains valid.
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RAI Question #3:
In Figure 5-2 of 32-9161509-000, Seabrook Cycle 14 2-D/3-D incore detector root mean square (RMS) errors,the trend of RMS errors versus core average burnup is significantlyhigher at beginning of life, trends downward, and stabilizes after -6 GWd/MT in core average burnup.
[I Page 7 of 33
RAI Question #3a:
Explain this systematic behavior of 2-D1 3-D incore detector RM/S errors versus core average burnup.
Response to RAI Question #3a:
The 2D and 3D RMS behavior exhibited in cycle 14 is typical for Seabrook. Examples for cycles 3, 13, and 14 are shown in Figure 1 through Figure 3.
The 2D and 3D RMS values are based on the differences between the measured and predicted detectors signals. The behavior is attributed to one or more of several phenomena that are captured by the measured detector signals but are not captured or not precisely captured by the predicted detector signals. The possible phenomena are:
" Manufacturing variations in the fuel pellets, fuel enrichment, and burnable absorber loading.
These variations will burn down over time. Manufacturing variations are not captured in the predicted detector signals.
- Hydraulically driven tilts in the core power distribution. Figure 4 through Figure 6 show the measured incore tilt versus cycle exposure for cycles 3, 13, and 14. The power distribution tilts tend to burn down around the same time that the 2D and 3D RMS trends stabilize.
Figure 7 shows the 2D RMS and Figure 8 shows the 3D RMS plotted against the measured incore tilt, showing a correlation between the measured incore tilt and the RMS values. Core power distribution tilts are not captured by the predicted detector signals.
- High sensitivity to boron loading variations in burnable absorber modeling. The predicted detector signals may not precisely capture the effects of burnable absorber and do not capture the effects of manufacturing variations in burnable absorber loading. Seabrook has used Integral Fuel Burnable Absorber (IFBA) rods since cycle 2. The IFBA loading for cycles 3, 13, and 14 are shown in Table 1. The IFBA burns out at about the same time that the 2D and 3D RMS trends stabilize. The reactivity effect of IFBA depletion can be seen as the peak in the boron concentration versus cycle exposure, shown in Figure 9 through Figure 11 for cycles 3, 13, and 14.
All of these phenomena are captured by the measured detector signals and burn out at about the same time in core life that the 2D and 3D RMS stabilize. The phenomena are not captured or not perfectly captured by the predicted detector signals and thus lead to the downward trend in the RMS at the beginning of cycle as shown in Figure 1 through Figure 3, below.
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Table I Burnable Absorber Loading for Cycle 3, 13, and 14 Cycle Number of IFBA IFBA loading Pins (mg/in) 3 9280 1.57 13 6816 2.355 14 8992 2.355 Page 9 of 33
Figure 1 Cycle 3 Detector Measured to Predicted RMS versus Cycle Exposure Page 10 of 33
Figure 2 Cycle 13 Detector Measured to Predicted RMS versus Cycle Exposure Page 11 of 33
Figure 3 Cycle 14 Detector Measured to Predicted RMS versus Cycle Exposure Page 12 of 33
Figure 4 Cycle 3 Incore Measured Tilt versus Cycle Exposure 1.020-1.018-1.016-1.014 1.012 V 1-010 to 8
1.00*
S1.004 1.006 1.004 1.002 0 2 4 6 8 10 12 14 16 18 20 22 Cyde Exposure (GWD/MTJ)
Page 13 of 33
Figure 5 Cycle 13 Incore Measured Tilt versus Cycle Exposure 1.020 1.018 1.016 1.014 NOI E 1.012 1.010 S1.008 1.006 1.004 1.002 1.000 _ i_ _ i_ _ i_ i + i i j i 0 2 4 6 8 10 12 14 16 18 20 22 Cydle Exposure (GWD/MTU)
Page 14 of 33
Figure 6 Cycle 14 Incore Measured Tilt versus Cycle Exposure 1.020 1.018 1.016 1.014 1.012
=.
j 1.010 8
A1.008 1.006 1.004 1.002 1.000 0 2 4 6 8 10 12 14 16 18 20 22 Cyde Exposure (GWD/MTU)
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Figure 7 2D RMS versus Measured Incore Tilt for Cycles 3, 13, and 14 Page 16 of 33
Figure 8 3D RMS versus Measured Incore Tilt for Cycles 3, 13, and 14 Page 17 of 33
Figure 9 Cycle 3 RCS Boron Concentration versus Cycle Exposure 1200 10 8
4001_241 1_
200 ______
0 2 4 6 8 10 12 14 16 18 Cyde Expmare IGWD/MTU)
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Figure 10 Cycle 13 RCS Boron Concentration versus Cycle Exposure 2000 1800 1600 1400 _________ _
-91200
- 1000 1am800_o__
600_ _
400 _\_I 200 -_____
0 2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 22000 Cyde Exposure (MWD/MTU)
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Figure 11 Cycle 14 RCS Boron Concentration versus Cycle Exposure 2000 ism 1600 1400 i
.* 1200 a
.2 51000 8w 600 400 200 0
0 2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 Cyde Exp-ure IMWD/MPIU)
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RAI Question #3b:
Demonstrate whether this bias is seen versus the instrumented assembly's assembly average burnup.
Response to RAI Question #3b:
The difference between predicted and measured signals that was plotted in Figure A-12 of ANP-3243P is plotted against assembly average exposure in Figure 12. The figure shows a trend against assembly average exposure. The overall trend with assembly average exposure is expected since the differences must sum to zero in any fixed incore detector core map and each core map is comprised of detectors in fuel assemblies over a large range of exposures. The steeper trend and greater variability at the beginning, from 0 to approximately 7 GWD/MTU, is expected from the beginning of cycle or and early fuel exposure phenomena discussed in the response to RAI 3.a. The beginning of cycle phenomena would also affect the second and third burn fuel but that is obscured because there is no clear demarcation of beginning of cycle for the older assemblies.
As discussed in the response to RAI 3.a, the beginning of cycle and early fuel exposure phenomena leading to this trend are captured by the measured signal such that they do not contribute to the measurement uncertainty. On the other hand, these phenomena are not captured or not perfectly captured by the predicted signal thus leading to this trend. Nonetheless, the trend is included in the 2D and 3D RMS values that are used in the fixed incore detector uncertainty analysis described in Section 6.1 of ANP-3243P.
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Figure 12 Difference Between Predicted and Measured Signals versus Assembly Average Exposure Page 22 of 33
RAI Question #3c:
Explain why it is acceptable to apply a single RMS error valuefor an entire cycle in the calculation of Fdh andFq uncertainty when the average RMS is non-conservativefor early in cycle (e.g., use of 3-D average RMS value of[ ] in Equation 6 and 2-D average RMS value of [ ] in Equation 7 are not representative of actual RMS values calculationfor Seabrook in the first - 5 GwdiMT of cycle exposure).
Response to RAI Question #3c:
This response is in two parts. The first part demonstrates that the uncertainty analysis provides a conservative value. The second part demonstrates that the fixed incore detector system can accurately construct a set of measured surveillance parameters (FQ and FAH) regardless of any disparity between the measured and predicted detector signals that are discussed in the responses to RAI 3.a and RAI 3.b.
3c.1 Uncertainty Analysis The uncertainty analysis for FQ, as described in Section 6.1 of ANP-3243P, includes the following terms:
[
- ]
In addition, outside the scope of the methodology described in YAEC-1 855PA and ANP-3243P, FQ is increased by an uncertainty term to account for manufacturing tolerances, as required by the Technical Specifications.
The uncertainty analysis for FAH, as described in Section 6.1 of ANP-3243P, includes the following terms:
[
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]
The 2D and 3D RMS values are based on the differences between the measured and predicted detector signals. The 3D RMS is based on the differences between individual detector signals and the 2D RMS is based on the difference between string averaged detector signals.
Since the values of 2D and 3D RMS are based on the differences between measured and predicted detector signals, [
] and the manufacturing tolerance is also explicitly applied to the FQ surveillance parameter as required by the Technical Specification.
Since, as discussed earlier, the measured detector signals captures the actual core phenomena, the algorithms described in YAEC- 1855PA apply the differences between the measured and predicted detector signals to adjust the predicted power distribution to create a measured power distribution that matches the measured detector signals and represents the actual reactor power distribution.
Therefore, the uncertainty values based on the 2D and 3D RMS of the differences between measured and predicted detector signals must conservatively bound the uncertainty of the measured power distribution.
3c.2 Surveillance Parameters For the new methodology proposed in ANP-3273P, the ability of the fixed incore detector system to construct an accurate measured power distribution is demonstrated by comparing measured FQ and FAH results to the moveable system results for Cycle 3.
Figure 13 shows the fixed incore measured values of FQ and the moveable incore measured values of FQ. The values in this figure have uncertainty factors applied per Technical Specification requirements. Figure 14 shows predicted FQ values and the fixed incore measured values of FQ without the uncertainty factor.
Figure 15 shows the fixed incore measured values of FAH, the moveable incore measured values of FAR, and the predicted values of FAR. These values are shown without any uncertainty because the uncertainty factor is applied to the limit as specified in the core operating limits report.
Figure 14 and Figure 15 provide an indication of how much the power distribution was adjusted by the algorithms described in YAEC-1855PA. Figure 13 and Figure 15 show good agreement between the fixed incore system and the moveable system for both FQ and FAH, demonstrating that the measured surveillance parameters produced by the fixed incore system are a good representation of the true reactor conditions.
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3c.3 Conclusion The calculated uncertainty values based on the 15 cycle average of 2D and 3D RMS may be applied to the entire cycle, despite the typically higher values at BOC, for two reasons: 1) The fixed incore detector system uncertainty based on the values of 2D and 3D RMS of the difference between measured and predicted detector signals is conservatively bounding to the true measurement uncertainty, and 2) The higher observed values of 2D and 3D RMS typically seen at beginning of cycle have a negligible impact on the ability of the fixed incore detector system to determine accurate measured peak values of FQ and FAH.
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Figure 13 Cycle 3 Fixed Incore Measured FQ Compared to Moveable System Measured FQ 2.4 2.2 2
is 1.4 1.2 1
0 2000 400(0 6000 8000 100D0 12000 14000 16000 18000 200D0 22000 Cyde Exposure (MWDI/MT)
-- Moveable w/ Uncertainty -- Fixed Incore w/ Uncertainty Page 26 of 33
Figure 14 Cycle 3 Fixed Incore Measured FQ Compared to Predicted FQ 2.4 2.2 4 +/- 4 + 4 2
1.8
-j 1.4 1.2 1
0 2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 22000 Cvde Exposure (MWD/MTU)
-"-1 Predicted -"-Fixed Incore w/o Uncertainty Page 27 of 33
Figure 15 Cycle3 Fixed Incore Measured FAH Compared to Moveable System Measured FAH 1.6 1.5 1.4 S1.3 1-2 1.1 1
0 2000 40W0 6000 80(0 100D0 12000 140DO 16000 18000 20000 22000 Cyde Exposure (MWD/MTU)
-- Moveable --U Predicted -o- Fixed incore Page 28 of 33
RAI Question #4:
Explain andjustify the removal of the 39 statepoints in the uncertainty analysis. What criteriawere used to exclude statepoints? How are the statepoints covered by the current thermal limits surveillance methodology?
Response to RAI Question #4:
A few (three to five) state points at the beginning of cycles 5 through 15 were excluded from the calculation of the average values of 2D and 3D RMS. The excluded state points are from the non-equilibrium initial power escalation to full power.
These state points are excluded because the SIMULATE-3 cases were run with equilibrium xenon and samarium, thus making the comparison invalid. The fixed incore system is intended to be a replacement for the moveable incore system, which is not used in a transient situation. It should be noted such non-equilibrium state points were automatically excluded from the results reported in YAEC-1855PA and YAEC-1931.
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RAI Question #5:
Explain how the depletion correctionfactor (DPC)and the uncertainty on DPC are directly or indirectly accountedfor in the uncertaintyanalysis.
Response to RAI Question #5:
[
] As shown in Figure A-il of ANP-3243P it increases over time. As the factor grows over time the magnitude of the uncertainty on the factor also grows over time. The DPC uncertainty (one sigma) is projected to be [ ] at a future time where the detectors have twice the operating time they have seen at the end of cycle 15. Including this conservative estimate of the DPC uncertainty in Equations 4 and 5 in Section 6.1 of ANP-3243P has a negligible impact on the uncertainty factors for FAH and FQ and therefore may be ignored.
RAI Question #6:
Explain the difference in the calculationof the neutron conversionfactor (NCF)and/or (how the) applicationof the NCFdiffers from the previously approvedmethod Response to RAI Question #6:
For the methodology approved in YAEC-1855PA, the neutron portion of the detector signal was assumed to be some fraction of the total detector signal. This fraction was distributed amongst the detectors by the ratio of thermal neutron flux at the detector to the average thermal neutron flux over all detectors. The fraction was found to be 25% by a parametric study documented in Appendix B of YAEC-l1855PA, where the fixed incore detector measured results with fractions ranging from 20% to 30% were compared to results from the moveable incore system over cycle 1 and most of cycle 2. A value of 25% was found to provide the best match.
For the methodology presented in ANP-3243P, the neutron portion of the signal of each detector is calculated from the predicted neutron reaction rate of Pt- 195 in the detector times a Neutron Conversion Factor (NCF). The values for NCF and the Depletion Correction Factor (DPC) were determined parametrically to provide the best match between measured and predicted detector signals over 15 cycles of operations, as documented in Appendix A of ANP-3243P. Using the NCF, the neutron portion of the signal is still approximately 25% of the total signal.
In summary, in YAEC-1 855PA the neutron portion of the signal (25% of the total detector signal) was distributed by the core average weighted thermal flux at each detector location maintaining a core average neutron component of the signals at a value of 25%. In ANP-3243P the neutron portion of the signal is distributed by the Pt-195 neutron reaction rate in each detector so that the core average neutron component of the signal is not forced to a value of 25% and may vary. The individual detector Pt-195 neutron reaction rate is dependent not only on the thermal neutron flux, but also the fast neutron flux. Using the total neutron reaction rate better captures the effect of a broader range of core design and Page 30 of 33
operating conditions that make it more suitable for predicting the neutron portion of the signal for each detector.
RAI Question #7:
Explain the applicationof the DCP and NCF to overall system used in process of surveillances of Fdh, Fq, and axialflux difference. Provide a block diagram of the hardware,software, andplantprocedures that used these surveillancesas discussed during the January2014 audit.
Response to RAI Question #7:
A diagram of the data flowing through major modules of the Reactor Analysis Workstation (RAW) is shown in Figure 16.
The process of setting up the RAW system for a new cycle is performed under an AREVA administrative procedure in conjunction with a Seabrook station procedure specifically for updating the RAW system. The RAW system is used for surveillance purposes under a Seabrook station procedure for incore power distribution analysis.
An explanation of each module is provided below:
" CASMO-3: This program creates all the fuel dependent information that SIMULATE-3 needs to model the reactor core. Fuel cross sections, assembly pin power distributions, detector gamma signals and detector Pt-195 neutron reaction rates are functionalized against fuel exposure, temperature, boron concentration, control rod presence, exposure history, etc.
- SIMULATE-3: This program uses the fuel cross section tables from CASMO-3, a core loading map, and a fuel exposure distribution from a previous SIMULATE-3 case to determine a reactor core power distribution. For a core follow case (update to current conditions) the primary output is the up-to-date fuel exposure distribution.
For a SIMULATE-3 that provides input to FINC, there are several outputs: a predicted 3D nodal fuel exposure distribution, a predicted 3D nodal power distribution, a predicted 3D nodal distribution of FQ, form factors that allow the FQ distribution to be expanded in the axial direction to a finer mesh, predicted detector gamma signals, and predicted Pt-195 reaction rates for the detectors.
" Fixed Incore Detector Data Acquisition System (FIDDAS): This system records the measured fixed incore detector signals once per minute.
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FINC: This program combines the measured detector signals with the SIMULATE-3 predicted data to obtain a measured power distribution with the measured peak values of FAH and FQ. In broad outline:
o The sensitivity factor, Gamma Correction Factor (GCF) and Depletion Correction Factor (DPC) are applied to the measured signals. The GCF is unity for all original detectors and dependent on detector batch for replacement detectors. The DPC is based on up-to-date detector exposure using 3D fuel exposure data from SIMULATE-3.
0o1 1 o As described in YAEC-1855PA section 4.2, a sophisticated algorithm is used to replace detectors that have been flagged as failed by a user input file. Note that the user input file is the only mechanism to flag a detector as failed.
o The full-core, measured nodal power distribution is determined from the predicted power distribution times the ratio of the measured detector signal to the predicted detector signal. The actual algorithm is described in YAEC-1855PA, Section 4.4. The full-core nodal values of FQ are determined with the same algorithm.
o The nodal measured values of FQ are expanded to 61 locations in the axial direction by use of the SIMULATE-3 form factors as described in YAEC-1855PA. This allows the selection of an FQ value that is consistent with the moveable incore system and consistent with the Seabrook Technical Specifications.
o The axially expanded FQ data is integrated to find FAH.
o The measured nodal power distribution is used to calculate the incore axial offset and the incore tilt.
- The values of FQ and FAH are used for required Technical Specification surveillance.
The value of incore axial offset is used to calibrate the excore detector AFD function.
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Figure 16 Reactor Analysis Workstation Data Flow Diagram Page 33 of 33