ML20212R259
| ML20212R259 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 04/16/1987 |
| From: | Wilson R GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 2659C, 5000-87-1228, NUDOCS 8704270027 | |
| Download: ML20212R259 (41) | |
Text
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GPU Nuclear NMCIMr 100 Interpace Parkway Parsippany, New Jersey 07054 201 263-6500 TELEX 136-482 Writer's Direct Dial Number:
April 16, 1987 5000-87-1228 U.S. Nuclear Regulatory Comission Attention: Document Control Desk Washington, D.C. 20555 Gentlemen:
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 Reload Topical Report 021 Pursuant to your inquiry of January 8, 1987, please find attached GPU Nuclear's resoonse to the request for additional information concerning Topical Report 021, entitled " Methods for the Analysis of Boiling Water Reactors Steady State Physics". As discussed previously, the response to question 12 will be provided at a later date.
If you have any questions, please contact M. W. Laggart at (201) 263-6205.
S cer ly, f h s....
ko[2 DO 5 00 19 ice P e ent p
PDR Technical Functions RFW/JDL/pa(2071 )
9 Att.
cc: Dr. Thomas E. Murley, Administrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA.
19406 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, N.J.
08731 Mr. Jack N. Donohew, Jr.
h U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue
'lg Phillips Building, Mail Stop 314 1
Bethesda, Maryland 20014
'GPU Nuclear is a part of the General Pubhc Utilities System
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ATTACMENT OYSTER CREEK NUCLEAR GENERATING STATION RELOAD TOPICAL REPORT 021 REQUEST FOR ADDITIONAL INFORMATION i
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G QUESTION #1:
What specific quantitles will be calculated by the GPUN code for the reload core analysis? Have the uncertainties in all of these quantities been provided in TR-021 and, if not, what are they? How are these uncertainties incorporated in the design analysis (fuel limits, thermal margin, LOCA limits, etc.)?
RESPONSE
The GPUN nodal code is used to calculate the following information:
3-D power, void and exposure distribution, Core K-effective, and for each fuel type peak KW/FT, margins to LOCA limits and minimum critical power ratio.
The code provides a Haling calculation for the EOC exposure distribution to be used in the safety analysis. The code also has a cold model for calculating a K-effective for the cold shutdown margin determination.
The uncertainties associated with the above parameters are assessed in the power distribution and the hot and cold reactivity.
TR-021 reports a TIP 1 41% and a hot critical k-effective of 0.98625 nodal uncertainty of 7.65 1
1 00177 for Oyster Creek Cycles 8 and 9.
A cold k-effective of 1.00193 0
1 00293 is based on Cycles 8, 9 and 10.
The Haling exposure distribution 0
becomes a bounding distribution which must be met within a specified tolerance to ensure the validity of the safety analysis.
As part of the design analysis, the core is stepwise depleted in 0.5 GWD/MT intervals to EOC. Control Rod patterns are set at each interval to remain 2659C
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G, within thermal limits and meet the target k-effective.
The target K-effective is met with a il sigma band. Uncertainties in power distribution are not considered in this aspect of the design analysis.
Integrated Exposure Ratio (INER) calculations are performed.using the stepwise depletion to see if projected control rod patterns and operation will be within the bounds of the Haling exposure distribution. At each 0.5 GHD/MT interval shutdown margin is calculated.
The required design cold shutdown margin of 1.0% AK must be met using the cold critical k-effective with a i sigma uncertainty.
It is during actual operation that the uncertainty in power distribution is considered with regard to fuel limits, thermal margins, LOCA limits, etc.
The Power Shape Monitoring System (PSMS) is used to monitor the fuel thermal limits. The PSMS uses the same nodal code (NODE-B) used in the reload design analysis, but in an online manner. Core heat balance sensor data and control rod patterns are automatically input to the code at predetermined intervals or on demand to obtain a ' ore power distribution and calculation of thermal limits.
The online PSMS uses full core measured TIP data to determine model performance and uses LPRM data to provide a continuous performance evaluation.
The online code is required to maintain a 7.5% nodal power uncertainty which corresponds to about an 8.0% nodal TIP uncertainty.
The 7.5% nodal power uncertainty is used to be within the power uncertainty used by the' fuel vendor to establish the MCPR safety limit.
If the base model does not meet the 7.5%
performance requirement, model corrections can be made on the basis of plant measured TIP or LPRM data. The use of LPRM feedback to improve model performance by correcting the nodal code axial power shape to match the LPRMs has been sufficient to maintain the required uncertainty.
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Since the online core monitoring code and offline reload analysis code are the same, the performance of the unadjusted online code will reflect the performance of the offline code.
If there is a decline in base (unadjusted) model performance from one cycle to the next or within a cycle the model is adjusted to assure that uncertainties will be maintained within required accuracy.
The core is constrained to operate within the technical specification for thermal limits and the INER calculation will be performed using the online model to maintain operation such that the target Haling exposure distribution is met. As described in the response to the following questions, the decline in the base model performance from one cycle to the next can be corrected by using the PSMS to provide an improved model with reduced uncertainty for the next reload analysis.
QUESTION #2:
How sensitive are the a's and g's (discussed in Page 15) to exposure, rod pattern, temperature, core loading, etc? Describe the procedure used in the selection of albedos and leakage constants.
If k= multiplier is applied, how is it determined and how frequently is it adjusted? What are the criteria for adopting new normalization parameters?
RESPONSE
An evaluation of the albedos and leakage constants (normalization parameters) is performed by using the optimization routine in the PSMS.
This routine iterates on the model normalization parameters to obtain the best agreement between PSMS predicted and plant measured TIP data on a statepoint by statepoint basis.
The procedure for the evaluation of the normalization 2659C
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=0 parameters of a given cycle begins by redepleting from the beginning of the 4
Cycle out to the first high powered (> 701.) steady-state (equilibrium xenon) statepoint where a full set of measured TIP data is available.
This depletion process is done using the normalization parameters from Cycle N-1.
A parameter optimization calculation is then performed.
The parameter optimization routine adjusts the normalization parameters to minimize the errors in TIP nodal. uncertainty.
Tha resulting set of optimized parameters are averaged with the values used in the depletion calculation as follows:
i l
j Let a be an optimal value of a normalization parameter for statepoint I
k.
Then, the value of the normalization. parameter.to be used for the burnup calculations (b.)between statepoint k and the next statepoint k +
1, is given by:
b - 1/2 (a.) + 1/2 (b.. )
where b=_i is the value that was used for the burnup calculations between statepoints k-1 and k.
This simple filtering technique dampens random changes in the parameters and resulting in a high predictive accuracy.
This newly created averaged set of parameters is then used in the cycle depletion from the current statepoint out to the next statepoint.
This process is repeated until end of cycle. A set of optimized normalization-parameters for each statepoint is compiled. An arithmetic average for each parameter is calculated and an averaged set of optimized parameters is obtained.
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This procedure was used by EPRI to develop a set of normalization parameters for Cycle 8 which are given in Table 2-1 (Ref. 3 in TR-021).
There is no significant variation in the parameters until the last two statepoints.
The reason for the variation in these statepoints is that the full set of TIPS were taken following a reactor startup and the core was not at equilibrium xenon. The GPUN procedure for parameter optimization requires equilibrium xenon.
Variation'in the normalization parameters due to exposure, rod pattern and core loading does not appear to be significant.
The Cycle 9 core depletion used the same set of normalization parameters as Cycle 8 and model performance for Cycle 9 was good.
However, no new fuel designs were introduced into the Cycle 9 core.
From the Cycle 10 data (see response to Question #3), it was concluded that the normalization parameters required reoptimization for new fuel designs with natural U ends, and/or large changes in enrichment or gadolinium loading.
The criteria for adopting a new set of normalization parameters is tied to model performance. GPUN requires a 7.5% nodal power uncertainty which is equivalent to about an 8.0% TIP nodal uncertainty for the online nodal model.
If the online nodal model performance (and hence the offline model performance) cannot meet the criteria, the LPRM feedback mechanism is used to correct the online model.
If the model performance without the LPRM feedback continues to be poor over the course of the cycle a new set of normalization parameters would be adopted for use in the reload design analysis and for use in the following cycle.
The new parameters could also be used in the current cycle in conjunction with the LPRM feedback.
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1 The GPUN nodal code does not use a K= multiplier, however the bias in K-effective is considered in the design analysis.
QUESTION #3:
What normalizations of the GPUN code, model, input data, etc., to measured data have been performed, and are any of these normalizations based on the verification data of Chapter 4? If so, what increase in the GPUN code uncertainties will occur in comparisons where the normalization is not made (i.e., Cycle 10 and later cycles)?
. RESPONSE:
As stated in response to Question #2, the normalization of the GPUN code was done using Cycle 8 data.
No additional normalization was performed for the Cycle 9 data. The Cycle 9 performance showed no increase in code uncertainties.
In going to Cycle 10 there was noticeable decline in nodal code performance (Cycle 10 data is supplied in response to Question #14).
TIP nodal uncertainty increased to around 10% averaged over the cycle which was due to some of the factors highlighted in response to Question #2. Two new GE fuel designs were introduced (see response to Question #4) and. Cycle 10 operation followed a prolonged coastdown (to 40% power) in Cycle 9.
During Cycle 9 operation the LPRM feedback-mechanism was not used. Model uncertainty increased during the coastdown (14% TIP uncertainty at EOC) and was carried over to Cycle 10 with the errors introduced to the exposure and exposure weighted void (E and V) arrays. As a result of the cycle 10 performance, LPRM feedback will be used to correct model performance and maintain a correct E and V array.
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Cycle 10 performance is considered the largest increase (about 4%) in uncertainty that would occur in going from one cycle to'the next.
Increased
,:ncertainties that existed in the E and V due to poor model performance will be eliminated, by using LPRM feedback, leaving only the impact of changes in fuel design. Optimized normalization parameters, developed during the cycle if poor model performance necessitated it, would be applied to the next cycle. This would ensure that the best available model would be used in the reload design.
The optimized parameters based on Cycle 10 operation were used
~
for Cycle 11.
00ESTION #4:
Describe the fuel loadings of the Cycle 8 and Cycle 9 cores which are included in the verification process of the GPUN code.
Provide information on fuel types, U s - enrichment, gadolinia, etc.
2
RESPONSE
The fuel loadings for Cycles 8 and 9 are shown in Figures 4-1 and 4-2.
Since information is also being supplied for Cycle 10, the Cycle 10 fuel loading is included in Figure 4-3.
The fuel designs included are:
4 ENC III-E 7X7 fuel design 2.63 U-235 enriched, 1% gd ENC III-F 7X7 fuel design 2.63 U-235 enriched, 1% gd ENC V 8X8 fuel design 2.65 U-235 enriched, 1% gd ENC V-B 8X8 fuel design 2.5 U-235 enriched, 1% gd GE P8X8R fuel design 2.54 U-235 enriched lattice with Nat U ends, 2% gd GE P8X8R fuel design 2.85 U-235 enriched lattice with Nat U ends,'3% gd 8@blC.
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QUESTION #5:
What is the range in temperatures and reactor periods during the performance of cold critical experiments? Describe the procedures for correcting temperature and reactor period.
RESPONSE
The local criticals are performed with the reactor head off the vessel.
The coolant temperature ranged between 90*F to 97*F during Cycles 8, 9 and 10.
The spread criticals, or insequence criticals during reactor startup, have the reactor head on the vessel and require a minimum temperature of around 190*F.
The reactor periods during the local criticals ranged from 28 to 71 seconds for the positive periods and 113 to 341 seconds for negative - -iods.
The periods are calculated from measurements on the IRMs.
For tne spread criticals the periods were around 100 seconds.
This is based on estimates from the period meier.
For the local criticals, the fuel constants for the GPUN code are calculated at 68'F.
The temperature correction is calculated as follows:
K.. (1 + AK
+ AK )
Ko re Knoo Knoo where
- Keor,
- Corrected core average K effective K..
- Code calculated K effective using 68'F fuel _ constants.
AK Knoo
- Temperature correction for moderator AK K?oo
- Temperature correction for control rod worth 2659C j
d The temperature correction for moderator and control rod worth use a volume weighted coefficient calculated from the lattice physics code.
The code bias is the difference between the corrected K-eff and the measured K-effective of the reactor BIAS = Koor, - (1.00 + RHO) where RHO is the reactivity calculated from the reactor period measurement.
The same procedure is used for the spread criticals exc.pt that the fuel constants used are calculated at 200*F.
QUESTION #6:
Provide the rod configurations for the local and spread criticals in Tables 4.1 - 4.3 for Cycles 8, 9 and 10.
Have any additional criticals for Cycle 10 and later cycles been analyzed? If so, please provide the results.
RESPONSE
The rod configurations for the criticals in Cycles 8, 9, and 10 are provided in Figures 6-1, 6-2 and 6-3.
The rod configuration for the Cycle 11 local critical is also provided in Figure 6-4.
Unlike the previous cycles only a single critical was performed for Cycle 11.
The calculated K-effective with temperature correction for Cycle 11 is 1.00149 and the measured K-effective is 1.00056 for a code bias of 0.00093.
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QUESTION #7:
Have any spread criticals been analyzed with the GPUN code? If so, please indicate control rod configuration and show calculated results and corrections, if any.
RESPONSE
Spread criticals have been analyzed at four different points during Cycle 9.
The criticals were at BOC and following shutdown periods where the reactor coolant was near 200*F and the core was xenon free.
The rod configurations are provided in Figures 7-2 to 7-4.
Additional spread criticals were analyzed at BOC 10 and 11.
The rod configurations are shown in Figures 7-5 and 7-6.
The results of the analyses are provided in Table 7-1.
QUESTION #8:
In Tables 4.1 through 4.3, indicate the period and temperature for each of the criticals. What is the relttion between the criticals having the same numerical designation and positive and negative periods? At what core average exposures were the cold criticals shown in Tables 4.2 through 4.3 performed?
RESPONSE
The period and temperature for each critical is provided in response to Question 5.
The criticals are performed to calibrate the worth of a control rod diagonally adjacent to the control having minimum shutdown margin (SDM) to demonstrate required shutdown margin.
The rod having minimum SDM and the rod to be calibrated are fully withdrawn (Notch 48).
The reactor is brought to a slightly supercritical configuration with other rods and the period is measured and reactivity calculated. The rod to be calibrated is then inserted 3
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until the~ reactor is on a negative period greater than 100 seconds.
The period is measured and reactivity calculated.
The reactivity for the positive and negative period is added and is the worth of the rod being moved from its original position to the notch which gave the negative period. The process is repeated until the rod is calibrated from 48 to 00 (fully inserted).
The first positive period is designated 1 POS and the first negative period 1 NEG and so on.
The exposures for the cold criticals in Cycles 8, 9, and 10 is 10.26 GWD/MT, 9.83 GWD/MT and 9.54 GWD/MT respectively.
The Cycle 11 cold critical was at a core exposure of 8.68.GWD/MT.
QUESTION #9:
i Results of shutdown margin measurements for Cycle 8 show a much larger bias than observed in Cycles 9 and 10.
Please explain.
RESPONSE
The higher bias in the Cycle 8 criticals is attributed to a combination of factors.
The critical control rod patterns had two control rods on the edge of the core that were fully withdrawn for most of the criticals. There is a larger uncertainty in rod worth for the control rods on the core periphery due to leakage effects. There was also a larger uncertainty in the exposure of the peripheral bundles.
This was due to the fact that the type III-E bundles were located on the periphery for the two previous cycles.
This was an unusual core residence history necessitated by the shortening of Cycle 5 operation to 256 GWD when it was designed for 500 GWD of operation.
It became necessary to put the low burnup bundles on the core periphery for Cycle 6-2659C m
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operation.
The bundles remained on the' core periphery fc-the next two cycles and the higher uncertainty in power-distribution was compounded.
l QUESTION #10:
,Hhat was the cycle length of Cycles 8, 9, and 10?
RESPONSE
CYCLE LENGTH, GHD/MT (EFPD) 8 6.21 (322) 9 9.38 (486) 10 7.11 (368)
QUESTION 11:
Explain why the calculated core average axial power distributions underpredict the bottom of the core at BOC-8 and overpredict the bottom at EOC-8.
RESPONSE
As explained in response to Questions 2 and 3, the Cycle 8 depletion with NODE-B used an optimized set of normalization factors which were based on an averaging technique over the cycle.
- uch the use of an averaged set of parameters results in the underprediction at BOC-8 and overprediction at EOC-8 in the core average axial power distribution.
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QUEETION #12:
Has any bias been observed in the radial power distributions for Cycles 8 through 10? Please provide radial power distributions to supoort you conclusions.
RESPONSE
The response to Question 12 will be forwarded in a later transmittal.
QUESTION #13:
Describe the procedure for generating four-bundle constants and conversion factors used in the calculation of instrument readings.
RESPONSE
The conversion factors used to calculate the instrument readings from bundle power are obtained directly from CPM. CPM calculates a detector reaction rate for a power range monitor located in the corner of the narrow water gap. CPM normalizes the reaction rates to average bundle power.
This value is used as the conversion factor in N0DE-B.
The conversion factors are obtained as a function of exposure, and coolant density with and without control rods.
N00E-B does not use four-bundle constants, but averages the cont'ribution of each of the four surrounding bundles.
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QUESTION #14:
Are power distribution comparisons for Oyster Creek Cycle 10 (similar to data given in Tables 4.4 through 4.7 for Cycles 8 and 9) available? If so, please supply the data.
RESPONSE
Data is available for Oyster Creek Cycle 10 from the online PSMS nodal code.
This code is the same as the offline version except for the LPRM feedback capability. The difference in the results between the online results without LPRM feedback and the offline results is in the exposure update increments.
The online code updates exposure on a daily basis while the offline code is updated on a weekly basis (longer for steady state operation, shorter for non-steady state operation).
The E and V array for the reload analysis using the offline code will come from the online code since it provides more accurate information and better results.
Table 14-1 contains the key information for each statepoint and Table 14-2 contains the K-effective nodal TIP uncertainty for the uncorrected model.
Table 14-3 shows the same data using optimized parameters to reburn the cycle.
QUESTION #15:
In addition to the Oyster Creek and Hatch 1 Cycle 1 data, have any other BHR cores been analyzed with the GPUN code?
RESPONSE
GPUN has not analyzed any other BWR cores besides Oyster Creek and Hatch 1 Cycle 1.
However, EPRI has analyzed Quad Cities Cycle 2 as part of the PSMS benchmarking (Ref. 4 of TR-021).
R(CDXC
E QUESTION #16:
Have any cold critical evaluations been made for Hatch 1 Cycle l?
RESPONSE
No.
QUESTION #17:
In Table 4.11, why are the residuals for some nodes larger than the standard deviations? What is.the cause of the large residual for node-21?
. RESPONSE:
The residuals in Table 4.11 are the average residual for the 106 bundles at each axial elevation.
The sum of the residuals in Table 4.11 is zero.
The residuals are not normalized for each elevation and reflect the error relative to the average of all measurements.
The standard deviation presented is the one sigma range for the residual at that elevation.
For example, the error at node 21 was -13.40%.
This represents the average error at that elevation.
The one sigma deviation around that error is 6.10%.
The statistical definitions used for the TIP and gamma-scan comparisons are presented in Appendix A to TR-021.
The Hatch plant had volds occurring in the bypass region due to plugging holes in the lower grid plate.
The lack of a bypass void model in N0DE-B limited the ability of N0DE-B to predict power distribution, especially in the top of the core where bypass voiding would be the greatest.
This, in combination with the peak power occurring at node 19 and with power dropping sharply in nodes 21 and 23, resulted in being unable to predict the peak node and 2659C u
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correctly matching the power in node 21. A bypass void model is not necessary for Oyster-Creek since bypass voiding is not significant.
QUESTION #18:
How are the y-scan measurement data related to the Barium / Lanthanum distributions? Since the y-scan met.surement is a measurement of " power histo ~y", how are these measured / predicted Ba-140 comparisons related to the bundle power uncertainties for a specific statepoint?
RESPONSE
The gamma scan measures the activity of LA-140 which has a 40.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> half-life.
LA-140 is produced in the fuel both by direct fission and by beta decay of BA-140, another direct fission product, which has a 12.8 day half-lite.
A few days after the shutdown of the reactor, the LA-140 produced by fission decay to insignificant levels, and the BA-140 and LA-140 concentrations equilibrate. At this point, the LA-140 concentration decays at a rate determined by the BA-140 half-life. Thus, the measured LA-140 distribution is equivalent to the BA-140 distribution.
Because of its 12.8 day half-life, the BA-140 distribution at the end of cycle represents the core power distribution integrated over approximately the last three months of the i
cycle.
It is not a direct indication of a specific statepoint.
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~6-9 The BA-140 distribution calculated from NODE-B used the last five statepoints in the Hatch Cycle I step out (Reference 3.1)~.
STATEPOINT DAYS
- CYCLE EXP' POWER 12/19/76 60 9403 MHD/MT 2231 MW 01/20/77 12 9829 2131 01/25/77 30 9883 2153 02/23/77 5
10119 2208 03/07/77 13
-10311 2114 Time step (AT)' used in BA-140 density calculation-The end of Cycle BA-140 Density is calculated by:
N (tn) - yep (tn) + [N (tn-1) - yep ( tn)] e-**'
A A
l where N (tn)
- BA-140 atom density at time step tn A
- BA-140 decay constant - 0.05419 day-'
n
-1,2,3,4,5 P
- Power Density r, MW/CM' AT
= Time step length in days I
Ye
- Effective yield of BA-140 from fission function of=
exposure 2659C f
QUESTION #19:
What fuel types does GPUN intend _to use at Oyster Creek that are not included in the verification data?
RESPONSE
Oyster Creek Cycle 11 operation has begun with two new fuel designs.
These are 3.19 U-235 enriched lattice'(2.99 average enrichment) with axial varying gd. There is 4 and 5% gd design and a 3 and 4% gd design. _ Future cycles may include enrichments up to an average 3.4%.
The gd loading is not expected to exceed 5%.
QUESTION #20:
Please supply hot core reactivity data by statepoint for Cycles 8, 9 and 10.
RESPONSE
Hot core reactivity data was supplied for Cycles 8 and 9 in TR-021 and for Cycle 10 in response to Question 14.
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Table 2-1 OPTIMIZED PARAMETERS FOR OYSTER CREEK Optimized Parameter Values TIP Uncertainty RMS N
f B
"1 "2
9 9
Eigenvalue Before Opt.
After Opt.
Statepoint a
V H
1/09/79 0.01
-0.37 0.17 0.71 0.68 0.39 0.9848 6.49 6.02 2/16/79 0.02
-0.53 0.15 0.79 0.83 0.40 0.9819 6.69 5.73 3/01/79 0.01
-0.44 0.17 0.76 0.81 0.42 0.9817 6.12 5.67 3/20/79 0.03
-0.40 0.23 0.44 0.71 0.32 0.9833 5.63 5.62 4/26/79 0.02
-0.34 0.25 0.47 0.63 0.38 0.9844 5.72 5.68 6/08/79 0.01
-0.29 0.22 0.41 0.74 0.37 0.9832 6.52 5.92 7/26/79 0.02
-0.42 0.24 0.49 0.67 0.22 0.9862 7.24 6.51 8/30/79 0.02
-0.45 0.24 0.45 0.75 0.26 0.9862 6.56 6.23 9/14/79 0.01
-0.40 0.25 0.47 0.75 0.32 0.9869 6.52 6.15 10/11/79 0.01
-0.41 0.33 0.52 0.85 0.48 0.9869 6.39 5.75 11/23/79 0.22
-0.35 0.47 0.43 0.93 1.06 0.9885 6.35 5.67 12/14/79 0.10
-0.36 0.40 0.65 0.72 0.84 0.9887 5.96 5.77 Mean 0.04
-0.40 0.26 0.55 0.76 0.45 0.9852 6.35 5.89 110 10.06 10.06 10.09 10.13 10.08 10.24 10.0023 10.42 10.27 1
1 9
FIGURE 4-1 OYSTER CREEK CYCLE 8 QUARTEF CORE FUEL LOADING
- III-F III-E III-F III-E III-F FUEL TYPE
- 15.1 15.8 15.7 19.5 16.1 EXPOSURE, GWD/MT III-F III-E III-E III-E 17.0 0.0 16.1 0.0 18.1 21.4 III-F III-E IIU-T 6.8 8.5 0.0 6.8 0.0 8.3 17.0 19.0 22.4 III-F III-F III-F 16.3 0.0 18.4 0.0 17.3 0.0 8.7 0.0 8.7 22.1 III-F III-E III-E 8.3 17.6 0.0 17.8 0.8 8.6 0.0 19.1 0.0 8.8 22.5 III-E III-E 16.3 0.0 19.0 0.0 17.5 0.0 17.6 0.0 18.9 0.0 18.9 III-F 8.8 8.7 9.0 9.1 8.8 16.6 0.0 17.3 0.0
'9.0 18.5 III-E 16.4 0.0 19.0 0.0 18.3 0.0 16.6 0.0 8.7 0.0 8.4 21.6 III-F III-E III-F 8.9 18.7 0.0 19.1 7.2 18.4 8.8 17.5 0.0 17.6 0.0 18.0 16.2 III-F III-E 16.0 0.0 18.4 0.0 19.0 0.0 9.0 0.0 17.2 0.0 6.7 0.0 19.4 III-F III E III-F 8.2 15.6 7.1 18.7 0.0 19.0 9.0 19.1 0.0 18.5 0.0 16.2 15.7 III-E 14.0 0.0 16.0 0.0 19.0 0.0 8.8 0.0 17.7 0.0 8.7 0.0 16.8 III-F III-F III-F 19.5 14.1 2.2 16.2 8.8 16.5 8.8 16.2 8.3 16.3 6.8.
17.1 15.2 Unless otherwise noted, fuel type is ENC V-B.
l I
FIGURE 4-2 OYSTER CREEK CYCLE 9 QUARTER CORE FUEL LOADING e
III-F III-F III-F FUEL TYPE
- 22.5
.6 17.8 18.3 18.0 EXPOSURE, GWD/MT 5.9 0.0 7.8 0.0 14.0 23.7
)
7.6 7.7 0.0 15.0 0.0 7.0 22.9 24.0 23.1
' l i
15.4 0.0 16.1 0.0 8.0 0.0 7.7 0.0 7.6 24.0 l
7.6 7.7 0.0 7.8 0.0 8.0 0.0 16.3 6.1 7.7 23.1 15.3 0.0 22.0 0.0 22.3 0.0 16.1 0.0 16.6 0.0 24.0 7.5 7.8 6.0 7.8 6.2 15.9 7.3 16.0 0.0 7.6 22.9 15.0 0.0 16.0 0.0 15.5 0.0 15.9 0.0 8.0 0.0 7.2 Z4.o III-F 7.5 13.5 0.0 15.9 7.2 L5.5 6.2 22.3 0.0 7.9 0.0 13.9 18.0 14.0 0.0 20.6 0.0 15.8 0.0 7.8 0.0 7.7 0.0 15.0 0.0 22.6 i
III-F 7.4 13.4 7.5 20.5 0.0 L6.0 6.0 22.1 0.0 16.1 0.0 6.6 l
18.4 13.3 0.0 13.4 0.0 13.6 0.0 7.9 0.0 7.6 0.0 7.7 0.0 22.5 V
III-F 25.1 13.3 7.5 14.1 7.4 L5.0 7.5 15.3 7.5 15.3 7.6 5.9 17.9 i
i s.
i Unless otherwise noted fuel type is ENC V-B.
i
l FIGURE 4-3 l
l OYSTER CREEK CYCLE 10 QUARTER CORE FUEL LOADING FUEL TYPE
- 16.7 22.5 IS.6 23.6 24.0 EXPOSURE, GWD/f4T GE239 GE239 13.6 0.0 8.8 0.0 9.5 11.2 GE239 GE265 GE265 0.0 10.9 0.0 14.9 0.0 9.6 18.6 9.6 15.8 GE239 GE265 GE265 GE265 10.9 0.0 15.6 0.0 11.2 0.0 15.8 0.0 9.0 18.6 GE239 GE239 GE265 GE265 10.5 17.2 0.0 18.4 0.0 11.3 0.0 15.4 0.0 9.0 13.3 GE239 GE239 GE239 CE239 GE265
)
11.2 0.0 11.5 0.0 17.6 0.0 10.9 0.0 15.4 0.0 10.8 t
ENCVB GE239 ENCVS GE265 10.7 18.1 0.0 18.3 0.0 17.8 0.0 10.9 0.0 15.9 18.7 GE239 GE265 GE239 GE239 GE265 i
11.0 0.0 15.8 0.0 18.4 0.0 17.9 0.0 11.3 0.0 9.6 11.2 ENCVS ENCVB GE239 GE239 GE265 10.4
_18.2 0.0 18.3 0.0 18.4 0.0 17.6 0.0 11.3 0.0 9.6 23.6 GE239 GE239 GE265 GE239 GE265 GE239 t
0.0 18.3 0.0 18.3 0.0 18.4 0.0 18.5 0.0 14.9 0.0 23.4 3g,4 ENCV8 ENCVS ENCVS GE239 GE265 11.1 11.2 0.0 18.3 0.0 15.8 0.0 11.3 0.0 15.6 0.0 8.8 18.7 GE239 GE239 GE239 GE239 GE239 GE239 11.1 0.0 11.2 0.0 18.2 0.0 18.0 0.0 17.3 0.0 10.9 0.0 22.7 GE239 18.6 11.1 11.1 16.3 10.4 11.0 10.7 11.2 10.5 10.9 0.0 14.$
16.8 s
Unless otherwise noted, fuel type is ENC V-B.
0 l
a I
i i
FIGURE S -l CYCf & 8
.e t
l
}
,- =
l 12.
48 FP* - u gg ya
.?... ~ u e
i 48 48 1
j j
qa og
.!........ =
i 84 8 od.
'J,.....
... m i
i I
i i
e e
e a
a e
o e
e e
e a
e e
CRITICAL *
[ 905 CRITICAL #
/ NEG-TEMP.,
- F Q6.4 TE".P.,
- F 16.6
?ERIGC, SEC.
- 2. 8. 8 'i PERIOD, SEC.
2/8.S R E ACTI*/ITY, rk
- f. Z 2-REACTIVITY, Ink.0.3 9
........... a
.... e
-....e
... s
... s
.n
.w
-w
-m i
e a
s
.s 82.
qb
=?==*
t
.?**~*
e i
i He
.,..l...
48 30
-t.....
3 e
e e
e a
e e
e e
e a
a e
e CR*TICAL =
a Pc5 CRITICAL #
2 ^/6 G TEMP.,
- F 96.6 TE:P.,
'F 96.6 PEPICD, SEC.,
29.37 PE2 ICE, SEC.,
/.5 7. 8
- i. - l c: / : Y,
v.
/. 21 3EACTIVITY, mk -0.75 i
FIGURE 6-l coa 't.
........... =
............ =
.=
... n
.=
i 4
i
- 2 i
12.
og g
.g...
u og 43
.g...
a 48 48 l
i j
y8 48
-.l......!.... =
i i
48 48
~ l......i.... =
d i
i I
'I i
i
)
1 e
e e
a m
CRITICAL
- 3 Pos CRITICAL
- 3 A/4G TEMP.,
- F Q6.6 TE"P.,
- F
- 6.6 PERIGC, SEC.
29.20 PERIOD, SEC.
/42. /
REACTIVITY, rk /. 2. 2-REACTIVITY, mk.O.70
........... =
............ =
.... =
... s
... m
.n t
i i
I s
48 48 s
oi 48
.t u
oo 48
-t
- S l
48 04 48 oi l
l l
48 4B
- +....
....e 48 48 t....
l e
e e
e a
e o
e e
e a
a e
m
.rt.TICAL =
4 koS CRITICAL 4 6 fos TE:'P.,
- F 96.6 T E;;p.,
- F 7$.6 PERICD, :iEC.,
- B.09 EE2ICC, SEC.,
/ 22.. /
b r J CT !? :'"Y,
. O 89 REACTIVITY, mk O.32.
FIGURE lo-l.
CYCLE q
.=
09 48 o4 si qs
.=
ys
=
98
%8
~.
~.
.g...
a
.g...
a
.i i l
1 s,.....
....e s,.....
l t.
e e
e a
a e
a e
e e
e a
e a
CRITICAL 8
/ POS CRITICAL #
/NEG TEMP.,
- F 90.8 TE'P.,
- F 90.O PERIOD, SEC.
SV.S PERIOD, SEC.
2 21.1 REACTIVITY,r.k O. 8 2.
REACTIVITY, mk.O.39
............ =
'... =
e4 14 al 10 48 46 4
48 4
48
-e l
i l
. g..... a
.g...
a 1
I
.s
...e
- 9....
...e l
1 l
e e
e a
a e
e e
e e
a m
e e
CRITICAL 8 2 80)
CRITICAL #
- 2. 46G TE:'P.,
- F 41.4 TE:!P.,
- F 91.4 PERICD, SEC.,
59.6 P:2 ICE, SEC.,
/f= 8.9 i
- i. t l,CT IV ;TY,
t 0.77 REACTIVITY, mk -0.57 i
FIGURE [p - /. Co d
... =
.=
ot to 02 43
.e qs
.=
10 48 to 46
~ =
l i
i i
i
.g...
u
. t... ~ u
..'......m s,........ e s,
I i
I i
i i
i e
e e
a a
e e
e a
e a
a e
e CRITICAL
- 3 Ib5 CRITICAL e 3 4/ec, TEMP.,
"F Ql.8 TE".P.,
"F 9/. 8 PERIOD, SEC.
4/.6 PERIOD, SEC.
//3. 2.
REACTIVITY, nk O.%
REACTIVITY, mk - /.0T 1
l i
l
........... =
i
............ =
j
=
oz e6 qg
.n
=
~
2.L 48 l
ll p g... - a
.g...
a i
I 1
1.....
l l
s,.....
g e
e e
e n
e e
e a
m CRITICAL e y Pos CRITICAL 4 4 dec, TE:.P.,
"F 92.
TE!!P., *F 92.
PERIOD, SEC.,
53b PraIco, SEC.,
258.6 i t.lsCT I'/ ITY, :9k O.83 REACTIVITY, mk -0.32.
i
FIGURE 63 CYC(( /0 l
............ =
..........m 48 V8 i
gg qs 12.
48 Li 48 I4 24 48 24 48 48 48 48 Yi 34
.e
.=
i i
i 1
.p..
u
.p..
u j
i i
9....7....
s,......:....
e a
e a
m
.e a
e e
e a
a e
e 4
CRITICAL #
/ Moj CRITICAL #
/ A/6G TEMP.,
'T 4.3 TE".P.,
'F J3 PERIOD, SEC.
(,2. 6 PERIOD, SEC.
/'v 8. 9 REACTIVITY,r,k O.75 REACTIVITY, mk -0. 66
........... m 48 48 gz q8
. oe 4s V8 (8
18 48
'48 48 j
48 V8 48
'48 06 48 08 48 a
zy 2.4 i
i i
.p..
a
.p..- u j
1 9... 7... e s,........ e i
i I
I I
i i
e e
e e
a e
e e
e e
a a
e a
CRITICAL 8 2 805 CRITICAL #
2 A/d G-TE 'P.,
- F 93 TEMP., *F 93 PERICD, SEC.,
4/V. 6 PE3ICC, SEC.,
J W.0 i-F.JsCT IV ITY, 9 O.9'l REACTIVITY, mk -043
..,,,.,,._..,,,n..
- ~.,,
e FIGURE 4-3 Con #I.
l
=
46 48 08 48 04 48 4B 48 18 48 48 40 48 46 48 48 Z8 48 Z8 48 Zi Z4
~=
i I'
.p..
u
.p... u
!-1 I
D- -
.....7....
+.........
I e
a e
a a
e e
e e
e a
a e
a 3 905 CRITICAL #
3 #EG CRI f f Q -
- TEMP., +1 93 TE".P.,
- F 43 PERIOD, SEC.
7/.9 PERIOD, SEC. 245.9 REACTIVITY, ok O.68 REACTIVITY, mk -0.34 1
............ m
'............ m 48 og 93 00 qg 48 48 48
- 8 V8 48 ga 48
.=
48 48
... =
48 48 Zt 48 18 22.
19 48 a
li 48 l
~.
t i
i
.y.... m
.y.. a
~.1......:... =
- +....... =
l 3
e a
e e
a e
a e
o e
e a
e e
CR!TICAL.
4 Po3 CRITICAt.
<< as o.
TEP.P.,
- F T3 TEMP., 'F
_93 PERIOD, SEC.,
68.8 PE2 ICD, SEC.,
//62.1 i t.liCTIVITY, :tk O.70 REACTIVITY, mk. O.06
e FIGURE 6*N I
c.YC LE ll
............ m
.m
.m i
12 48 46 -*
i
'48 18 i
.L
.p..
a
.p... a
+....... e l
-+...7...m I.
e a
e e
a e
e e
o e
a a
e a
l CRITICAL #
/ [05 CRITICAL #
TEMP.,
- F 86 TE".P.,
'F PERIOD, SEC.
/00.0 PERICD, SEC.
REACTIVITY, nk d.66 REACTIVITY, mk 1
a
............ =
.... e
... e
.m
.m
.e
.=
W m
L
.p..
a
.p..
a j
- +........
+........ =
l i
i I
I i
i i
i e
e e
e a
e e
e o
e e
a e
a CRITICAL =
CRITICAL #
TE.M P.,
- F TEMP., 'F PERIOD, SEC.,
PE2IOD, SEC.,
'k REACTIVITY, mk
' t.l.CT I'/ :TY,
e
. ~....
2 TABLE 7-1 IN SEQUENCE CRITICALS CYCLE EXPOSURE PERCENT' CONTROL REACTOR TEMP PERIOD CODE CYCLE GWD/MT-ROD DENSITY
- F-SEC-BIAS 9
.0.0 75.70-190.0 100
-0.00728.
9 4.26 75.64 201.0 100
-0.00090 9
5.28 63.99 192.8 100 0.00042
'9 6.14 60.43 189.0 120 0.00058'-
10 0.0 70.04 195.0 75
-0.00341 11 0.0
-73.48 198.0 80
-0.00716 4
9
'2659C m
d.
s FIGURE 7-/
CYCLE i INSEQUENCE CRITICAL 51 4a 98 "I
4s 46 48 48 98 48 35 96 48 40 48 is 48 27 48 48 48 48 48 18 19 ib 48 46 48 48 48 e
i i
I I
I i
-,---- 11 I
48 14 48 48 48 96 I
i 1
I I
l 1
8 i
I e
I i
1 l
96 48 a- - -
0 3 I
i e
t e
i e
i.
e e
I a
02 10 15 26 34 42 50 CORE EXPOSURE, CWD/MT 00 TEMPERATURE,
'F 190 #
PERICD, SECCNDS
/00 CALCULATED K-EFFECTIVE O.99303
---~.--,.,,.,-,r-.
r-
--e
-.,a.r
e FIGURE 7-L CYCLE T INSEQUENCE CRITICAL
'l 48 48 i
)
l 42 98 Se Ss se 98 48 is 35 48 48 48 48 48 48 27 48 48 48 is qs 4g 19 is 46 48 48 48 98 e
I I
1 g
I I
l is 48 48 48 4e 4s
-j---- 11 I
I I
I I
1 6
g e
6 l
48 is
- - - - '- - - - - _i_ _ _ _ 0 2 8
1 8
8 8
i
,e i
8 I
e 02 10 18 26 34 42 50 CORE EXPOSURE, CWD/MT Y. E b TEMPERATURE,
'F 20/.0 PERIOD, SECCNDS
/00 CALCULATED K-EFFECTIVE 0.99 % 4
o o
FIGURE 7-3 CYCLE 7 INSEQUENCE CRITICAL 51 48 48 48 48 "I
46 48 48 48 48 46 48 48 48 48 48 48 4s is 49 35 lb 48 48 48 27 48 48 48 49 48 48 49 4B 48 19 ge ge 48 46 46 48 s
1 48 48 I
e 1
l is 48 48 48 4s 9e
- i- - - - 11 I
I 4B 48 48 l
1 I
s t
i 4.
48
_ _ _ _ _;_ _ _ _ _ _;_ _ _.. o 3 l
i i
i 02 10 18 26 34 42 50 l
CORE EXPOSURE, GWD/MT 8.16 TEMPERATURE,
- F I47. 8 PERICD, SECCNDS
/00 CALCULATED K-EFFECTIVE /.00086
4-y*
FIGURE.7-Y CYCLE T INSEQUENCE CRITICAL i
- - - - _. _ _ _ _ _. -51 qg 4g 48 48
- 3 48 48 48 48 48 48 48 48 48 I
48 49 48 48 48 48 35 48 48 48 48 48 48 48 40 48 48 48 48 27 46 48 48 19 iB 40 46 48 48 46 i
I to 48 48 48 48 I
e I
I 18 48
't B 48 48 48 -i---- 11 I
48 is t8 l
i i
i e
t i
i t
48 48 0 3 l
l l
i i
e i
02 10 18 26 34 42 50 CORE EXPOSURE, GWD/MT 6.19 TEMPERATURE,
- F
/24 I PERICD, SECCNDS
/20 j
CALCULATED K-EFFECTIVE /. o o 0 8 3
4 FIGURE 7"I CYCLE 10 INSEQUENCE CRITICAL 51 qs 48 93 17 I 2.
42 4e 98 48 98 4e f ?-
11 48 48 48 48 48 48 43 35 17-(L 1L 12.
40 4B 46 48 48 48 48 27 17-12.
s4 12.
19 48 48 48 48 48 48
+8 I
12.
It i
i I
g 1
48 98 48 4B 48
- j- -- - 11 I
I i
i IL IL 0
I I
I l
i 48 4a 48
- - - _ _;- - _ _ _ _._ _ _ _ 0 3 e
i e
e 8
8 8
e
,e i
I 8
I e
i 02 10 18 26 34 42 50 i
\\
CORE EXPOSURE, GWD/MT
- 0. 0 l
TEMPERATURE, 'F
/98 PERIOD, SECCtJOS 78 CALCULATED K-EFFECTIVE O.9977/
4
+
FIGURE 7-d CYCLE il INSEQUl:.NCE CRITICAL i
51 48 48 oG, M
4 46 48 48 48 48 48 ------'l M
08 48 48 48 48 48 48 35 t
oS o8 Ob 48 4B 46 48 48 48 27 08 08 08 19 48 48 48 48 48 48 i
1 08 08 I
a i
I i
48 48 48 48 48 48 -i---- 11 1
I oc, oG l
l i
i e
i 4
e i
e a
43 48
.',.______03 i
i i
i i
i 02 10 18 26 34 42 50 CORE EXPOSURE, GWD/MT o.0 TEMPERATURE, 'F 198 PERIOD, SECCNDS 80 4
CALCULATED K-EFFECTIVE O.9937l
o
- a f'
TABLE 14-1 KEY INFORMATION FOR OYSTER CREEK STATEPOINTS CYCLE 10 POWER FLOW EXPOSURE ROD
% ROD DATE (MHth)
(M1bs/hr)
(GHD/MTU)
SEQUENCE DENSITY 12/12/84 1522.6 53.39 9.70 A-1 7.91 01/04/85 1807.3 55.17 10.03 A-1 4.87 01/24/85 1881.1 56.50 10.38 A-1 5.60 01/31/85 1849.7 59.86 10.50 A-1 7.06 03/10/85 1680.8 55.59 10.58 B-1 8.45 03/25/85 1852.8 60.66 10.81 B-1 9.49 04/14/85 1872.4 60.56 11.12 B-1 10.10 05/03/85 1888.9 58.75 11.44 B-1 10.40 05/29/85 1911.1 58.86 11.89 A-2 12.53 06/06/85 1918.2 60.90 12.05 A-2 12.71 07/03/85 1902.0 60.43 12.42 A-2 13.02 08/18/85 1896.8 60.41 12.95 A-2 13.02 09/14/85 1913.9 60.73 13.45 B-2 10.40 10/12/85 1849.8 58.35 13.99 B-2 10.04 12/05/85 1908.3 60.27 14.34 B-2 8.82 01/07/86 1903.8 60.36 14.94 B-2 7.00 02/06/86 1899.7 60.63 15.46 B-2 3.35 03/14/86 1747.4 60.36 16.10 B-2 1.03 2659C
.a
,, n a s
- e w
TABLE 14-2 CORE REACTIVITY AND POWER DISTRIBUTION COMFARISON FOR OYSTER CREEK CYCLE 10 (CYCLE.8 & 9 NORMALIZATION PARAMETERS)
TIP N0DAL UNCERTAINTY STATEPOINT CALCULATED Kr,r IN % RMS 12/12/84 0.9913 7.53 1
01/04/85 0.9947 9.40 01/24/85 0.9945 10.62 01/31/85 0.9935 9.64 03/10/85 0.9929 9.00 03/25/85 0.9926 9.17 04/14/85 0.9938 12.45 05/03/85 0.9940 13.97 05/29/85 0.9944 12.87 06/06/85 0.9945 11.76 07/03/85 0.9954 10.19 08/18/85 0.9949 7.53 09/14/85 0.9956 6.70
~
10/12/85 0.9900 8.83 12/05/85 0.9944 9.64 01/07/86 0.9951 10.02 02/06/86 0.9963 12.06 03/14/86 0.9965 16.66 4
MEAN 0.9942 10.45 1 49 2
1 0016 0
1 SIGMA i
2659C 1
n.
74
[ J * %*e -
h*
TABLE 14-3 CORE REACTIVITY AND POWER DISTRIBUTION COMPARISON FOR OYSTER CREEK CYCLE 10-(OPTIMIZED NORMALIZATION PARAMETERS) 1 TIP NODAL UNCERTAINTY STATEPOINT CALCULATED Ker, IN 1. RMS -
12/12/84 0.9897 7.72 01/04/85 0.9931 6.89 01/24/85 0.9927 8.45 01/31/85 0.9918 7.15 03/10/85 0.9936 5.74 03/25/85 0.9923 6.42 04/14/85 0.9908 5.97 05/03/85 0.9911 5.84 05/29/85 0.9910 5.72 06/06/85 0.9913 6.44 07/03/85 0.9921 5.42 08/18/85 0.9930 5.80 09/14/85 0.9935 5.80 10/12/85 0.9932 5.79 12/05/85 0.9930 5.99 01/07/86 0.9941 6.43 02/06/86 0.9955 7.06 03/14/86 0.9946 6.54 MEAN 0.9926 6.40 1 60 0
1.0015 0
1 SIGMA 2659C k