ML18065A612
| ML18065A612 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/04/1996 |
| From: | Smedley R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML18065A613 | List: |
| References | |
| NUDOCS 9604110150 | |
| Download: ML18065A612 (19) | |
Text
consumers Power*
l'OWERINli
/llllCHlliAN'S l'ROliRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 April 4, 1996 U S Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
/
_/
. DOCKET 50-255 - LICENSE DPR PALISADES PLANT UPDATED REACTOR VESSEL FLUENCE VALUES This letter provides recent reevaluations of Palisades fluence data. New calculated and best estimate fluence values are presented and discussed. The new best estimate fluence values are derived utilizing a bias factor which is based on Palisades in-vessel and ex-vessel capsule measurements. The previously existing capsule measurements have been reevaluated and updated values are provided. Using the information provided in this letter, *a ~evised PTS screening criteria date, as defined in 10CFR50.61, has been calculated. The axial welds containing heat# W5214 remain the limiting vessel material and are now estimated to reach the screening date in the year 2011.
In past fluence evaluations, Consumers Power Company (CPCo) has estimated the future exposure rates at the peak fluence and axial weld locations using Cycle 9 fluence rates with an anticipated capacity factor of 75%. This submittal estimates future exposure rates at the peak fluence and axial weld locations using Cycle 11 fluence rates with an anticipated capacity factor of 85%.
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A CMS' ENER6YCOMPANY contains the Palisades evaluation of the updated reactor vessel fluence values and capsule fluence measurements. Attachment 2 contains Westinghouse report, WCAP14557, Revision 1, "Consumers Power Company Reactor Vessel Neutron Fluence Measurement Program for Palisades Nuclear Plant - Cycle 1 Through 11 ". contains AEA report, AEAT-0121, "Fluence Calculations for the Palisades Plant".
NRC staff concurrence with the appropriateness of this updated reactor vessel fluence methodology to project reactor vessel life is requested by November 1, 1996.
SUMMARY
OF COMMITMENTS This letter contains no new commitments and supersedes one existing commitment.
In CPCo letter dated February 23, 1994, we committed to use the fluence methodology described in our June 5, 1992 submittal. If the NRC staff concurs with the appropriateness of this new methodology, CPCo intends to use the methodology described in this letter for future fluence work.
Richard W Smedley Manager, Licensing CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector - Palisades Attachments
ATTACHMENT 1 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 PALISADES EVALUATION OF THE UPDATED REACTOR VESSEL FLUENCE VALUES AND CAPSULE FLUENCE MEASUREMENTS 15 Pages
1.0
SUMMARY
This evaluation transmits the results of our Engineering Analysis, Reference 1, and .
summarizes the information provided in the recent reevaluation of Palisades fluence data, Attachment 2. The new calculated and best estimate fluence values are presented and discussed. The best estimate fluence is derived utilizing a bias factor which is based on Palisades in-vessel and ex-vessel capsule measurements. The capsule measurements have been reevaluated and updated values are provided.
Using the information provided in this submittal, a PTS screening criteria date, as defined in 10CFRS0.61, can be calculated. The axial welds containing heat# W5214 remain the limiting vessel material and are estimated to reach the screening date in the year 2011. An independent fluence evaluation using a different methodology, , was conducted to verify the results of the reevaluation.
2.0 ANALYSIS
2.1 BACKGROUND
The Palisades reactor vessel fluence has been frequently evaluated since the late 1980's when the date that the reactor vessel would reach the PTS screening criteria
- was projected to be near the end of the license life. During this time, efforts have been made to both qualify a methodology and quantify the end results.
It has always been our position and the position of our fluence expert, Westinghouse, that inherent biases in the calculational methodology could be removed using information provided by the surveillance capsules. The ratio between the measured value at the capsules (both in-vessel and ex-vessel) and the calculated value at the capsules is used to derive this bias factor. The bias factor is then is employed to improve the accuracy of the calculated solution. This was not implemented in the past, however, because Westinghouse's general experience with their methodology suggested that the bias in their methods resulted in an under prediction of the measured values and not the over prediction that was shown at Palisades.
At the time of our submittals dated June 5, 1992 and June 10, 1993, the NRC was suggesting that the Westinghouse standard bias (1.13) be applied to Palisades calculated fluence values. At this time, Westinghouse was suggesting that the Palisades derived bias (0.94) be applied to the calculated fluence values. CPCo and Westinghouse believe the main reason that the Palisades derived *bias was different from the Westinghouse standard bias for other Plants was that Palisades did not have a thermal shield. The lack of a thermal shield removes a significant amount of steel from between the vessel and the core. The derived steel cross sections (specifically iron) from ENDF/8-VI under predicted the forward scattering of the neutrons and caused the calculated flux to be under predicted when the neutrons had to penetrate 1
steel (ie. thermal shields) atthe other plants. This difference in iron cross-sections meant that at least one source of bias played a different role at plants with thermal shields, compared to plants without thermal shields (Palisades). At the time of the referenced submittals, the NRC had not accepted any calculation that applied a bias
- factor of less than 1.0. Palisades decided to report the conservative calculated fluence values until further NRC and industry evaluation of this issue was conducted. The primal"}' issues at the time were the lack of a commercially available cross-section library based on ENDF/B-VI and the acceptability of biasing calculations downward based on measurement data. The first issue was settled with the*release of BUGLE-93 in Reference 6. The second issue is addressed in the current draft version of DG 1025 dated March 14, 1994 in Reference 7.
Following the removal of the ex-vessel dosimetry program during the cycle 11/12 refueling outage, Westinghouse was contracted to do a complete reevaluation of Palisades reactor vessel fluence. This reevaluation included a recalculation of each cycle's fluence and all the dosimetry capsules using ENDF/B-VI cross-sections and fission spectrums.
In addition to the fluence reevaluation work performed by Westinghouse, CPCo also contracted with AEA to do an independent evaluation of our vessel fluence using MCBEND, Reference 8, which is a*Monte Carlo program for the solution of general radiation transp,art problems. This Monte Carlo program does not need cross-section sets that have been collapsed into groups as is necessary with DORT, Reference 9, nor is ,it necessary to approximate any geometrical structures whose actual geometry may be important to the problem solution. These two advantages made the use of MCBEND appropriate for verifying the DORT calculations.
The MCBEND data is currently best used to support conclusions about the vessel fluence based on DORT calculations. This MCBEND data is submitted to support the Westinghouse calculations. This strategy is based on two key points: first the PTS rule is based on compiled DOT/DORT calculations so the use of DORT is consistent, and second, the results of the two independent methods closely agree such that MC BEND verifies the results of DORT. MCB~ND is not- currently a NRC approved-methodology, and it is not the intent of this submittal to support approval of the methodology.
- 2.2 FLUENCE CALCULATIONS Several changes have been made with respect to the models used to calculate Palisades' fluence. First, the new calculations use the BUGLE-93, Reference 6, cross-section library and fission spectrum data, which are based on ENDF/B-VI. The source distributions used in the calculations were all provided by a new SIMULATE reactor physics code model. This change was made to allow the entire calculation to use a 2
consistent source database. The first seven cycles are now calculated individually, as opposed to combined cycles as was done in the past. The inlet temperatures used are consistent with the data submitted to the NRC in Section 1.1 of the Thermal Annealing Report, Reference 10. The water that makes up the bypass flow is now modeled at a temperature equal to the average temperature of exterior assemblies in that cycle. This is only a change for cycles that incorporated a low leakage core design, cycles 8 and beyond. In cycles 1 through 7 the core Tave and the exterior assembly Tave were essentially equal. All calculations are run using S16 order of angular quadrature. This is especially important for estimating the ex-vessel flux levels as shown in Reference 4.
The vessel inner radius (IR) was moved out by 1,il of an inch, Reference 11. Finally, some of the previous capsule evaluations used a vessel thickness of 8.5 inches. The current calculations use a more accurate nominal value of 8.79 +/- 0.213 inches, Reference 11. The change in IR and vessel thickness have not been incorporated into AEA's calculations because some of the data used to make these adjustments was not available when their calculations were run.
Table 2.2.1 provides the Westinghouse DORT calculation results and compares them to the values previously calculated using the old cross-sections and source data.
These calculations have been carried out using Westinghouse's approved methodology, Reference 12, and in accordance with the draft DG 1025, Reference 7.
All of the new calculations are done over a quarter of the core. Previous calculations only covered an octant of the core so comparisons to old calculations were made on an octant symmetric basis.
The largest changes in calculated flux levels occur in cycles 1, 2, and 7. These are all cycles that were previously analyzed in combination with other cycles. The use of individual cycle source information has removed this misrepresentation from the analysis of these cycles. In addition, it is important to note that the flux levels in cycles that were previously analyzed as a single cycle changed very little, with a few exceptions at isolated angles. Calculations done by Westinghouse using BUGLE-93 cross-sections for other plants do show a significant change, as do the cavity calculations of Attachment 2 when the change in vessel thickness is factored out. This shows that the new cross-sections do not have a major effect on Palisades vessel ID calculated fluence, which supports the position that the lack of a thermal shield is a key difference between Palisades fluence calculations and those done by Westinghouse for other thermal shield plants.
Table 2.2.2 gives a summary of AEA calculated flux values. This table provides AEA's flux calculations in a format that allows comparisons to be made with Westinghouse calculations. To prepare these values, averages had to be calculated from tables 25 through 31 of Attachment 3. These averages are also divided by the AEA bias factor of 0.91 since the tables in Attachment 3 are the best estimate values not th13 calculated 3
values. This division must be done to convert the best estimate values back into calculation values.
B~sed on the values presented in Tables 2.2.1 and 2.2.2, the calculated end of cycle 11 fluence at the key vessel locations can be determined. Table 2.2.3 provides these values for the previous Westinghouse calculations, the new Westinghouse calculations, and AEA's calculations.
The differences shown in Table 2.2.3 are consistent from location to location, except where quarter core to eighth core models are compared. These differences are largely related to the reanalysis of all the cycles individually. The changes in the model do not play a large part in the overall absolute change of the calculated IR fluence values. It should be noted, the calculated cavity fluence values would have changed by a much larger factor, due to the new ENDF/B-VI based cross-sections if the vessel model thickness had not been increased.
2.3 CAPSULE MEASUREMENTS Two items from the DORT calculations are used to determine the capsule-measurements. First, the neutron energy spectrum is used as an estimate of the actual spectrum the capsules encounter. Th.is estimate is then adjusted to get the best fit to the activation data from the capsule: The spedrum provided from the calculation does not have a large effect on the final results as long as it is close to the actual spectrum.
It is Westinghouse's experience that these spectrums do not vary significantly from one PWR calculation to the next. Second, the relative exposure rate from one cycle to the next is used to evaluate the capsule data. The only effect of the changes.in the calculated fluence values provided in Attachment 2 is the calc1,1lation of individual exposure rates for cycles 1 through 7. The other changes in the calculation provided in the report have little or no effect on the evaluation of measured capsule fluence values.
2.3.1 Results
- The capsule measurements for four in-vessel capsules (A240, W290, W290-9, and W110) and ten ex-vessel dosimetry capsules have been reported to the NRC in past fluence submittals, References 4, 5, 13 and 14. All of these capsules have been reevaluated by Westinghouse and presented in Attachment 2 along with seven more ex-vessel capsules. In AEA's report, Attachment 3, three of the in-vessel fluence capsules and six of the ex-vessel capsules have been evaluated.
The measured fluence values determined for these capsules have changed from those submitted in the past for reasons specific to their location. The in-vessel capsule evaluations of past submittals assumed that photo fissions within the U238 and Np239 dosimeters did not occur in significant proportions. This approach was consistent with 4
how Westinghouse treated other plants. However, since no thermal shield exists at Palisades, the gamma flux at the in-vessel capsules is significantly higher than at a plant with a thermal shield. When the effect of photo fission was removed from these dosimeters, the actual reaction rate due to neutrons was reduced, and, in turn, the total fluence measurement dropped. In addition during the reanalysis of the data, cycle specific cases were run for all of the first 11 cycles. This allowed Westinghouse to differentiate between the relative exposure rates of cycles 1 through 7. In past reports these cycles were analyzed with two calculations, one for cycles 1 through 5, and one for cycles 6 and 7. The more detailed flux data gathered from the explicit calculations done for each cycle resulted in further reductions of the measured values for the in-vessel capsules.
The A240 capsule has a significant change in its measured exposure. This new value is based on an explicit modeling of the accelerated capsule position. Battelle Columbus Laboratories had performed this calculation originally in Reference 14 and the value was later updated with an approximation provided by Westinghouse in Reference 4. The Westinghouse approximation used the cycle 1 through 5 DOT fluence calculation as a basis, which significantly over predicted the flux level present in the vessel during the first two cycles, see Table 2.2.1. The new evaluation shows that the use of this approximation was not appropriate. The new more explicit modeling, and reanalysis of the measurement data made in Attachment 2, provides support for the value calculated by Battelle Columbus Laboratories and actually reduces its fluence.
The measured values of the ex-vessel capsules have changed for a different reason.
These values are based on the measured disintegration rates of the dosimeter
- materials, but the U238 is now only evaluated using it Cs 137 daughter product. This is done in accordance with Westinghouse's standard practice.
Table 2.3.1 provides the capsule measurements, along with the values *previously reported for these capsules, and the measurements provided by AEA where available.
The ex-vessel capsules are all core. centerline capsules, except those noted as "low,(which meant they resided at an elevation equal to the bottom of the active core).
The previous capsule flux for A240 is provided from the Battelle Columbus Laboratories report, Reference 14.
2.3.2 Interpretation The changes from Westinghouse's old values to the new values all fall within two standard deviations (+/-20) of the old reported values, and in general most fall near or within +/-1 a. (The reasons for these changes have been described in Attachment ~-) *In general, the differences are a result of three changes from the old calculations:
5
a) the photo fission correction of the uranium and neptunium dosimeter materials was ignored in the past calculations, b) the method of interpreting the results from the uranium capsules has been changed to be consistent with Westinghouse's current practices, and c) the explicit calculation of cycles 1 through 7 changed the C/s of the FERRET calculation which impacted the final results of the A240, W290, and W110 in-vessel capsules.
The majority of the previously evaluated capsules were also calculated by AEA using MCBEND. AEA's methodology is different from the methods used by Westinghouse and provides a valuable alternative calculation for the verification of Westinghouse's values. The two methods arrived at very similar results and were within or, for one capsule, just outside +/-1 o of each other.
Using the revised fluence values for the surveillance capsules and the measured shifts reported in Section 1.1 of the Thermal Annealing Report, the chemistry factor (CF) increases slightly for each of the surveillance materials from that reported on June 21, 1994, with the W-110 surveillance capsule results. The measured 8RTNoT agrees more closely with the revised projected 8RTNOT which was determined using the revised capsule fluences in Equation 3 of 10 CFR 50.61. Measured 8RTNOT continues tq be adequately represented by projected 8RTNOT*
In summary, the values as calculated by Westinghouse in Attachment 2 and shown in Table 2.3.1, are being submitted as the new capsule flux values. The new fluence values for the three surveillance program capsules are shown in Table 2.3.2.
2.4 BEST ESTIMATE FLUENCE Table 7 .1-1 of Attachment 2 provides a comparison of the values of the capsule measurements and DORT capsule calculation~. This table shows a consistent trend in which the calculations over estimate the flux as measured in the capsules. This is the same trend that existed in information provided in previous submittals. As suggested in
. Reference 7, and as applied in Attachment 2, the capsule measurements are used.to calculate an adjustment factor or bias factor for the calculated flux from DORT. This is the same method that Westinghouse has always used in the determination of the best estimate fluence at the clad-base metal interface. However, prior to the implementation of the BUGLE-93, the light water shielding cross-sections derived from ENDF/8-VI, the bias factor for the other plants for which Westinghouse had evaluated fluence data was always greater than .1.0. In general this factor was approximately 1.13, Reference 4.
Using old cross-sections and fluence calculations, the Palisades bias factor was 0.94, Reference 5.
6
As detailed in Section 2.1, calculated fluence values without the application of any bias were used to support the SER issued on April 12, 1995, Reference 15. The Technical Evaluation, Reference 16, that preceded this SER, however, tended to support the Palisades position that the DORT calculations used as the basis of the fluence portion of the SER were conservative. This evaluation did not use capsule measurements to adjust the calculations, but instead looked at the inputs used and their possible biases, and stated that the calculations from DORT were approximately 10% conservative.
This would have led to a bias factor of 0.90.
- Based on the results of the Technical Evaluation and the plant's need to recover PTS margin, additional steps were taken to provide more justification for the use of best estimate fluence values at Palisades. Two different efforts were started.
Westinghouse was asked to do a complete reanalysis of the Palisades fluence using the new BUGLE-93 cross-sections and more consistent and complete neutron source data. This new source data allowed calculation of cycles 1 through 7 td be run as individual cases. Separately, AEA was asked to take an independent look at Palisades reactor vessel fluence using a different methodology and code that did not have some of the limitations of DORT solutions. This independent look was intended to in'.)ure that DORT calculations were reasonable and not concealing some bias which may be.
significant and exclusive to Palisades.
These two evaluations are now complete and the results support a plant specific bias of less than 1.0. In addition, since Westinghouse has converted their fluence methodology to use ENDF/B-VI, other plants have submitted capsule reports to the NRC that show bias factors of less than 1.0. This information supports the use of a plant specific bias at Palisades as the correct method of calculating the plant's best estimate fluence as required by the Reg Guide 1.99.
- The bia~ factor for Palisades is calculated by taking the individual bias factors from each of the capsules analyzed at the core centerline and averaging them. This is done in Table 7.1-1 of Attachment 2. This averaging provides a bias factor for the flux above 1.0 MeV. of 0.831 +/- 0.067. All but two of the individual bias factors fall within +/-1 o of the average, and the two that do not, fall within +/-20. AEA calculated a different bias factor (0.91) using the three in-vessel capsules that they evaluated. Table 2.4.1 provides the best estimate fluence at the end of cycle 11 as calculated in Attachment 2 and Attachment 3.
The uncertainty on Westinghouse's best estimate is 14.5%. The uncertainty on AEA's best estimate is 13.0%. The two estimates agree to within 1o. AEA's fluence values at the inner radius of the vessel is calculated by averaging the fluence in the vessel clad.
Additionally, AEA did not shift the vessel IR of their calculation by Ye of an inch to be consistent with the more accurate value used by Westinghouse. Both of these factors over predict the fluence at the clad base metal interface by a small amount so the 7
agreement between the two calculations is even better than what is shown in Table 2.4.1.
2.5 PTS SCREENING CRITERIA DATE As shown in Reference 4 and Reference 10, the limiting fluence for both the base metal material and the axial weld material can be calculated using the following formula;
./ (RTprs - I - M)
- 0. 28 - '{ 0. 0784 - 0. 4 log CF Equation 1 f=lO 0* 2 Where; RTprs = 270°F for Axial Welds and Shell Plates 300°F for Circ. Welds
= O °F for the Limiting Shell Plate (D-3803-3 heat# C-1279-1)
-56 °F for Axial and Circ. Welds M = 34 °F for the Shell Plate 66 °F for the Axial and Circ. Welds CF = 158 for the Limiting Shell Plate (D-3803-3 heat# C-1279-1)
.229 for the Intermediate/Lower Circ. Weld (heat #27204) 232 for the Limiting Axial Welds (heat# W5214).
f = Fluence x 1019 n/cm 2 Using Equation 1 and the values provided above, the following PTS screening criteria limiting fluence values can be calculated; Limiting Shell Plate Fluence = 8.58 E+19 n/cm 2 Limiting Circ Weld Fluence = 2. 71 E+19 n/cm 2 Limiting Axial Weld Fluence = 1.55 E+19 n/cm 2 The Shell Plate material and the Circumferential Weld materials both are exposed to the peak vessel fluence at the 75° azimuth. This means that the Circumferential Weld material will reach its screening criteria prior to the Shell Plate material. The axial weld position has a lower fluence rate. The calculations below provide the date at which the 8
screening criteria will be met for both the Axial and Circumferential Weld materials. The fluence rates are based on cycle 11 fluxes and an 85% capacity factor.
Date =Date(EOC11) +(Limiting Fluence - Accumulated Fluence (EOC11))/Fluence Rate Circ. Weld =1995.65 + (2. 71 E+19 - 1.32E+19)/(1.74E+10 x 365.25 x 24 x 3600 x 0.85)
= 2025.4 Axial Weld =1995.65 + (1.55E+19- 9.82E+18)/(1.30E+10 x 365.25 x 24 x 3600 x 0.85)
. = 2011.9 These dates provide significant margin beyond the end of license life in March of 2007.
It is possible that the bias factor calculated by Westinghouse will change as additional
- capsules are evaluated. If the bias value were to increase by 1o, the plant could still reach the end of license life without exceeding the screening criteria, however, the tight agreement of current capsule biases suggests that the bias factor will not change much.
3.0 CONCLUSION
Several significant items have been accomplished since the last CPCo submittal of fluence calculations. Westinghouse has reanalyzed Palisades fluence data using the BUGLE-93 cross-section library based on ENDF/8-VI .and shown that the lack of a thermal shield is likely the source of the difference in bias factors between Palisades and the other plants that they provide fluence calculations for. The reanalysis of the fluence data involved many small model improvements and the evaluation of the first seven cycles as individual calculations instead of two grouped calculations.. In addition, Westinghouse removed an inappropriate assumption that the photo fission effect in the in-vessel capsule evaluation was small, and the new calculations no longer neglect this effect. AEA was contracted to do an independent review of Palisades fluence data using a Monte Carlo code. This evaluation supports the calculations performed by Westinghouse. Finally, Westinghouse has provided calculations for other plants that show bias factors of less than 1.0.
All these factors support the use of a bias factor of less than 1.0 at Palisades. It is CPCo's position that:
a) the updated capsule measurements provide the !TIOSt accurate values for the capsules fast neutron exposure, and b) the best estimate fluence at the reactor vessel ID calculated in section 2.4 be used to calculate the date that the Palisades reactor vessel will reach the PTS screening criteria as defined in 1OCRFS0.61.
Using this data provides a PTS screening criteria date in the year 2011.
9
4.0 REFERENCES
- 1) Consumers Power Company, EA-PTS-96-01, "Evaluation of Westinghouse and AEA Fluence Reports", March 1996.
- 2) Anderson, S.L and Perock, J. D.,"Consumers Power Company Reactor Vessel Neutron Fluence Measurement Program for Palisades Nuclear Plant - Cycle 1 through 11", WCAP-14557, Revision 1, Pittsburgh, Pa., March 1996.
- 3) Avery A. F., Wright, W. V., "Fluence Calculations for the Palisades Plant",
AEAT-0121, Pittsburgh, Pa., April 1996.
- 4) Consumers Power Company, "10CFR50.61 Pressurized Thermal Shock, Revised Projected Values of RTPTs for Reactor Beltline Materials", Docket 50-255, June 5, 1992.
- 5) Consumers Power Company, "1 OCFR50.61 Pressurized Thermal Shock, Reactor Vessel Neutron Fluence, Additional Information", Docket 50-255, June 10, 1993.
- 6) Oak Ridge National Laboratory, RSIC Data Library Collection, DLC-175, "BUGLE-93 Production and Testing of the VITAMIN-86 Fine ~roup and the BUGLE-93 Broad-Group Neutron/Photon Cross-Section Libraries Derived From ENDF/B-VI Nuclear Data".
- 7) U. S. Nuclear Regulatory Commission, "Calculational and Dosimetry Methods for
- Determining Pressure Vessel Neutron Fluence", Draft Regulatory Guide, DG 1025, March 14, 1994.
- 8) AEA Technology, "MCBEND -A Monte Carlo Program for General Radiation Transport Solutions", User Guide for Version 98, September 1995.
- 9) Oak Ridge National Laboratory, RSIC Computer Code Collection, DLC-175, "TORT-DORT Two- and Three-Dimensional Discrete Ordinates Transport",
Version 2.8.14.
- 10) Consumers Power Company, "Palisades Plant - Preliminary Thermal Annealing Report, Thermal Annealing Operation Plan, Section 1.1, General Considerations, and Section 1.2, Description of the Reactor Vessel", Docket 50-255, December 12, 1995.
- 11) Snuggerud, RD., Consumers Power Company letter to John Perock, Westinghouse Energy Systems, March 8, 1996.
10
- 12) Grimes, Christopher I., NRC Letter Dated October 16, 1995, "Safety Evaluation of Topical Report - Methodology Used to Develop Cold Overpressure Mitigating System and RCS Heatup and Cooldown Limit Curves", WCAP 14040, Revision 1, Westinghouse Electric Corporation, TAC No. M917 49.
- 13) Consumers Power Company, "1 OCFR50.61 Pressurized Thermal Shock, Reactor Vessel Material Surveillance Capsule Test Report", Docket 50-255, June 21, 1994.
- 14) Consumers Power Company, "Palisades Plant - Technical Specification Change Related to Heatup and Cooldown Curves", Docket 50-255, July 2, 1979.
- 15) Adensam, Elinor G., NRC Letter Dated April 12, 1995, "Palisades Plant -
Pressurized Thermal Shock Safety Evaluation", TAC No. M83227.
- 16) Hsia, Anthony J., NRC Letter Dated September 2, 1994, "Palisades Plant -
Transmittal of Technical Evaluation Report", TAC No. M83227.
11
Cycle .... g 1 2 3 4 5 6 7 8 10 11 Location!'
New Palisades Rea~torVessel ID Neutron Flux (E > 1 Mev.) n/(cm 2-sec.) x10 10 oo 3.694 3.358 4.461 4.511 4.211 4.560 4.355 1.940 2.014 1.471 1.348 15° 4.868 4.298 5.866 5.946 5.736 5.904 5.628 4.641 3.018 2.260 1.891 30° 3.579 3.299 4.351 4.704 4.516 4.544 4.352 2.222 1.961 1.886 1.506 60° 3.700 3.413 4.348 4.704 4.516 4.544 4.352 2.220 1.961 1.890 1.556 75° 4.899 4.324 5.855 5.942 5.731 5.899 5.624 4.636 3.016 2.261 2.089 goo 3.693 3.357 4.466 4.510 4.209 4.560 4.354 1.940 2.015 1.470 1.388 Previous Palisades Reactor Vessel ID Neutron Flux (E > 1 Mev.) n/(cm 2-sec.) x10 10 oo 4_5g 4_5g 4_5g 4_5g 4_5g 4.87 4.87 2.16 2.08 1.51 1.31 15° 6.03 6.03 6.03 6.03 6.03 6.25 6.25 4_9g 3.06 2.40 2.07 30° 4.70 4.70 4.70 4.70 4.70 4_7g 4_7g 2.34 2.00 1_g4 1_5g Percent Change From Previous to New oo -1g_5 -26.8 -2.8 -1.7 -8.3 -6.4 -10.6 -10.2 -3.2 -2.6 2.g 15° -1g_3 -28.7 -2.7 -1.4 -4.g -5.5 -10.0 -5.1 -1.4 -5.8 -8.6 30° -23.g -2g_9 -7.4 0.1 -3.g -5.1 _g_1 -5.0 -2.0 -2.8 -5.3 60°. -21.3 -27.4 -7.5 0.1 -3.g -5.1 _g_ 1 -5.1 -2.0 -2.6 -2.1 75° -18.8 -28.3 -2.g -1.5 -5.0 -5.6 . -10.*0 -5.2 -1.4 -5.8 o.g goo -1g_5 -26.g -2.7 -1.7 -8.3 -6.4 -10.6 -10.2 -3.1 -2.6- 6.0 Table 2.2.1 Westinghouse's Calculated Flux Values Old and New 12
Cycle-.
1 2 3 4 5 6 7 8 9 10 11 Location!
New Palisades Reactor Vessel ID Neutron Flux (E > 1 Mev.) n/(cm 2-sec.) x10 10 oo 3.75 3.75 3.75 3.75 4.01 4.40 4.40 2.16 2.04 1.53 1.42 15° 4_g8 4.g8 4.g8 4.g8 5.42 5.40 5.40 4.4g 2.g1 2.1g 1.82 30° 3.66 3.66 3.66 3.66 4.27 4.27 4.27 2.1g 1_g7 1.86 1.46 60° 3_g5 3.g5 3_g5 3.g5 "4.43 4.33 4.33 2.18 1.g7 1.85 1.53 75° 4.g3 4.g3 4.g3 4.g3 5.33 5.47 5.47 4.55 2.g1 2.20 2.07 goo 3.75 3.75 3.75 3.75 4.01 4.40 4.40 2.16 2.04 1.53 1.42 Table 2.2.2 AEA's Calculated Flux Values Location Previous New AEA % difference % Difference Westinghouse Westinghouse New to Previous ,AEAto New Palisades Reactor Vessel ID Neutron Fluence (E > 1 Mev.) (n/cm 2 ) x10 19 oo 1.27 1.14 1.12 -10.2 -1.8 15° 1.77 1.5g 1.52 -10,2 -4.4 30° 1.31 1.17 1.11 -10.7 -5.1 60° 1.31 1.18 1.16 _g.g -1.7 75° 1.77 1.5g 1.52 -10.2 -4.4 goo 1.27 1.14 1.12 -10.2 -1.8 Table 2.2.3 End of Cycle 11 Accumulated Fluence Values 13
Current Previous AEA Capsule Location Capsule 1o Capsule 1o Percent Capsule Flux Flux Flux Difference n/(cm2 -sec.)
n/(cm2 - n/(cm2 -
sec.) sec.)
Internal Capsules
- A240 (30°) 5.36E+11 11% 6.15E+11 NA -12.8 NA W290 (20°) 5.63E+10 9% 6.71E+10 9%* -16.1 5.92E+10 W290-9 (20°) 3.12E+10 7% 3.52E+10 7% -11.4 3.27E+10 W110 (20°) 5.06E+10 11% 5.67E+10 11% -10.8 5.26E+10 Cavity Capsules Cycle 8 (16°) 1.34E+9 7% 1.46E+9 8% -8.2 1.33E+9 Cycle 8 (26°) 9.97E+8 7% 1.10E+9 8% . -9.4 9.26E+8 Cycle 8 (26°) Low 3.45E+8 8% 3.86E+8 8% -10.6 NA Cycle 8 (39°) 6.94E+8 7% 7.62E+8 8% -8.9 6.76E+8 Cycle 8/9 (6°) 9.57E+8 7% 9.81E+8 8% -2.4 NA Cycle 8/9 (6°) Low 2.13E+8 8% 2.18E+8 8% -2.3 NA Cycle 9 (16°) 8.56E+8 8% 8.92E+8 8% -4.0 8.82E+8 Cycle 9 (26°) 7.83E+8 8% 7.77E+8 8% 0.8 7.93E+8 Cycle 9 (26°) Low 2.48E+8 8% 2.64E+8 8% -6.1 NA Cycle 9 (39°) 4.87E+8 8% 5.26E+8 9% -7.4 4.66E+8 Cycle 10/11 (6°) 6.43E+8* 8%
Cycle 10/11 (16°) 6.51E+8 8%
Cycle 10/11 (16°) Low 1.79E+8 11%
Cycle 10/11 (26°) 6.05E+8 8%
Cycle 10/11 (36°) 4.89E+8 8%
Cycle 10/11 (39°) 4.64E+8 8%
Cycle 10/11 (24°) 5.46E+8 8%
Table 2.3.1 Changes in Capsule Measurements 14
Capsule Identification Capsule Measured Fluence E > 1 Mev.
A240 3.84 x 1019 n/cm 2 W290 9.24 x 1018 n/cm 2 W110 1.59 x 1019 n/cm 2 Table 2.3.2 Measured Fluence of Surveillance Program Capsules oo 15° 30° 60° 75° goo Westinghouse 9.46E+18 1.32E+19 9.73E+18 9.82E+18 1.32E+19 9.47E+18 AEA 1.01E+19 1.38E+19 1.01E+19 1.05E+19 1.39E+19 1.02E+19 Table 2.4.1 Best Es_timate Fluence at the Reactor Vessel ID EOC 11 E >1.0 MeV. n/cm2 15
ATTACHMENT 2 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 CONSUMERS POWER COMPANY REACTOR VESSEL NEUTRON FLUENCE MEASUREMENT PROGRAM FOR PALISADES NUCLEAR PLANT -
CYCLE 1 THROUGH 11 WCAP-14557, Revision 1 MARCH 1966 267 Pages