ML18065A919

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Responds to RAI Re Updated Reactor Vessel Fluence Values
ML18065A919
Person / Time
Site: Palisades 
Issue date: 09/09/1996
From: Bordine T
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9609180009
Download: ML18065A919 (31)
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Text

consumers Power l'llWERIN&

llllClllliAN"S l'lllllillUS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 September 9, 1996 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT Thomas C. Bordlne Manager. Licensing UPDATED REACTOR VESSEL FLUENCE SUBMITIAL - PARTIAL RESPONSE TO,

  • ADDITIONAL QUESTIONS On April 4, 1996, Consumers Power Company (CPCo) submitted a reevaluation of the Pali.sades fluence data. The reevaluation contained a new estimate of when the limiting reactor vessel material will reach the Pressurized Thermal Shock (PTS) screening criteria. On May 15, 1996, CPCo met with the NRC staff to discuss the updated reactor vessel fluence values. On June 12 and 21, 1996, CPCo responded to the NRC questions from the May 15, 1996 meeting. On August 14, 1996, the NRC staff and CPCo met to further discuss the updated reactor vessel fluence submittal. At that, meeting, the NRC staff asked additional questions. On August 27, 1996, CPCo ii submitted the first part of the response to those questions. This submittal is the second

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of several submittals plarff1ea~to respond to the NRC questions from the --

August 14, 1996 meeting. contains the NRC questions and the CPCo response.

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A CMS' ENER6YCOMPANY

2 provides information requested in a August 29, 1996 teleconference between the NRC, CPCo, and Brookhaven National Laboratory concerning rated power and effective full power days (EFPD's).

SUMMARY

OF COMMITMENTS This letter contains no new commitments and no revisions to existing commitments.

Thomas C. Bordine Manager, Licensing CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector - Palisades Attachments t

t'

ATTACHMENT 1 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 CPCO RESPONSE TO NRC QUESTIONS FROM AUGUST 14, 1996 MEETING 24 Pages

Updated Fluence Submittal Request for Additional Information 2.1 In view of the substantial differences between the in-vessel and cavity dosimetry, why is the CIM bias determined for the cavity fluence applicable for predicting the inner wall fluence? Discuss the statistical basis for combining the in-vessel and cavity measurements to determine the calculation-to-measurement bias used for adjusting the vessel fluence. Are the two populations statistically compatible?

CPCo Response

/

Comparisons of the least squares adjusted best estimate and calculated <t> (E > 1.0 MeV) at each measurement location were provided in Table 7.1-1 of WCAP-14557, Revision 1, "Consumers Power Company_ Reactor Vessel Neutron Fluence Measurement Program for Palisades Nuclear Plant - Cycles 1 through 11", J. D.

Perock, et al., March 1996. This comparison table is reproduced here as Table 2.1-1.

In regard to Table 2.1-1, it should be noted that comparisons are presented in terms of Measurement/Calculation (M/C) ratios rather than the inverse (C/M) noted in Request 2.1. Since all comparisons provided in the Palisades submittal are presented in terms of M/C ratios, thi_s same approach will be used in the responses to the current RAI.

The total data base of M/C comparisons provided in Table 2.1-1 can be broken down into two subsets representing the in-vessel and cavity results. Each of these subsets is summarized as follows:

Average Standard Fractional Number M/C BIAS Deviation Difference Samples Total Data Base 0.831 0.044 17 In-Vessel Capsules.

0.835 0.015 0.005 4

Cavity Capsules 0.830 0.050

-0.001 13 In the above tabulation, the M/C bias derived from each of the data subsets is given along with the associated standard deviation and the fractional difference relative to the total data-base average. **The fractional differences-listed above are obtained from the following equation:

1

F = Subset M/C - Data Base M/C

. Data Base MIC In the comparison shown above, the standard deviation of the respective sample sets was based on the random variation of the M/C ratios as listed in Table 2.1-1. An examination of the above data table shows that the in-vessel and cavity data subsets yield essentially, identical M/C bias results. There are no substantial differences in the data subsets. The use of an M/C bias factor derived solely from in-vessel data or solely from ex-vessel data would result in essentially the same best estimate fluence projection as that derived from the total M/C data base.

+

A relative frequency histogram of the 17 sample set comprising the total MIC data base is depicted in Figure-2.1-1. The individual bin sizes chosen for the construction of the

  • histogram correspond to the standard deviation of the data base. Superimposed on the relative frequency. histogram is a normal distribution function with a mean of 0.831 and a standard deviation of 0.044. The relative freq1.,1ency histogram compares favorably with the expected normal'distribution curve.

Each of the 17 samples comprising the least squares adjusted cp (E > 1.0 MeV) M/C

- data base are also shown in Figure 2.1-2 in terms of individual M/C ratios and associated 1o uncertainties. An examination of Figure 2.1-2 indicates that all ofthe

  • data, both reactor cavity and internal surveillance capsule, can be considered to belong to the same general population.

Based on the comparisons illustrated in the relative frequency histogram and the M/C ratio comparison.. the in-vessel and ex-vessel data are statistically compatible and: can be combined to determine the overall M/C bias factor for use in the determination of the

2

TABLE 2.1-1 COMPARISON OF MEASURED AND CALCULATED EXPOSURE RA TES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS lt>(E > 1.0 MeV) [n/cm2-sec]

Calculated Measured M/C Internal Cagsules A240 (30°)

6.29e+11 5.36e+11 0.852 W290 (20°)

6.69e+10*

5.63e+10 0.842 W290-9 (20°)

3.82e+10 3.12e+10 0.818 W110 (20°)

6.12e+10

,5'.06e+10 0.826 6° Gavit¥ Cycle 8 Cycle 9 1.11e+09 9.57e+08 0.863 Cycle 10/11

  • 6.97e+08 6.43e+08 0.922 16° Gavit¥ Cycle 8 1.52e+09 1.34e+09 0.883 Cycle 9 1.07e+09 8.56e+08 0.801 Cycle 10/11 7.65e+08 6.51e+08.

0.851 24° Gavit¥ Cycle 8 Cycle 9 Cycle 10/11 6.84e+08 5.46e+08 0.798 26° Gavit¥ Cycle 8 1.19e+09 9.97e+08 0.835 Cycle 9 9.06e+08 7.83e+o8 0.864 Cycle 10/11

. 7.06e+08 6.05e+08 0.856 36° Gavit¥.

Cycle 8 Cycle9 Cycle 10/11 6.16e+08 4.89e+08 0.794 39° Gavit¥ Cycle 8 8.60e+08 6.94e+08 0.807.

Cycle 9 6.70e+08 4.87e+08 0.727 Cy9le 10/11

  • .5.84e+b8 4.64e+08 0.794 Average Bias Factor (K) 0.831 Standard Deviation ( 1 o)

+/-0.067 3

FIGURE 2.1-1 RELATIVE FREQUENCY DISTRIBUTION OF THE MIC RA TIO BASED ON THE LEAST SQUARES ADJUSTED <J> (E > 1.0 MeV) 0.5 --.----------------------------.,

0.4 -

0.3 -

0.2 -

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0.65 0.70 0.75 0.80 0.85 0.90 0.95 1.00 MIC Ratio 4

1.05 1.00 0.95 0.90

.2 0.85 QI:;

u --

0.80

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0.75 0.70 0.65 0.60 FIGURE 2.1-2 COMPARISON OF THE M/C RATIOS BASED ON THE LEAST SQUARES ADJUSTED d> (E > 1.0 MeV)

I

  • Accelerated Capsule 0 Wall Capsules
  • Cavity Dosimetry
  • Average 5

Reguest for Additional Information 2.2 Provide a comparison of the unadjusted calculation-to-measurement (CIM) bias based on:

  • 1) the accelerated capsule.
2) the inner-wall capsules.
3) the cavity capsules
4) the long-lived Cs-137 fission product.

CPCo Response Unadjusted comparisons of measured and calculated individual sensor reaction rates were provided in Table 7.2-1 of WCAP-14557, Revisio~y 1, "Consumers Power

  • Company Reactor Vessel Neutron Fluence Measurement Program for Palisades Nuclear Plant - Cycles 1 through 11", J. D. Perock, et al., March 1996. This table is reproduced here as Table 2.2-1.

In regard to Table.2.2-1, it should be noted that comparisons are presented in terms of MeasuremenUCalculation (M/C) ratios rather than the inverse (C/M) noted in Request 2.2. Since all comparisons provided in the Palisades submittal are presented in terms

,_<?f M/C ratios, this same approach will be used in the responses to the current RAI.

The M/C comparisons requested in 2.2 represent subsets of the data provided in Table 2.2-1. Each of these subsets is summarized as follows:

M/C

SUMMARY

INCLUDING ALL REACTIONS Average Standard Fractional Number of.

M/C BIAS Deviation Difference Samples Total Data Base 0.879.

0.072 96 Accelerated Capsule 0.954 0.092 0.085 4

Inner-Wall Capsules 0.906 0.071 0.031 15 Cavity Capsules 0.870 0.068

-0.010 77 Cs-137 Fission Product 0.863 0.090

-0.018 29 In the above tabulation, the M/C bias derived from each of the data subsets is given along with the. associated standard dev_iation and the fractional difference relative to the total data base average. An examination of the fraetiona1* difference column indicates that the average M/C bias derived from all of the data subsets falls very close to or within 1 standard deviation of the MIC bias determined from the total data base. The fractional differences listed above are obtained from theJollowing equation:

6

F = Subset M/C - Data Base M/C Data Base MIC In the response to Request 2.10, to be provided in a future submittal, it will be shown that a bias seems to exist between the high energy threshold Cu-63 (n,a) and Ti-46 (n,p) M/C ratios and the M/C ratios determined from the remainder of the reaction rate database. This observation is particularly clear in the case of the in-vessel capsules.

Further, the least squares adjustment evaluations imply that, due to the sparsity of neutrons at the high energies influencing the response of these sensors, the Cu-63 (n,a) and Ti-46 (n,p) reactions have a minor impact on the derived best estimate results. Therefore, as a part of this response, the following. comparison of the various subsets of M/C ratios is provided excluding the results based on the Cu-63 (n,a) and Ti-46 (n,p) reactions.

M/C COMPARISONS EXCLUDING Cu-63 (n,a) and Ti-46 (n,p) REACTIONS Average Standard Fractional Number of MIC BIAS Deviation Difference Samgles Total Data Base

- 0.851 0.066 62 Accelerated Capsule 0:882' 0.026 0.036 2

Inner-Wall Capsules 0.852 0.022.

0.001 9

Cavity Capsules 0.849 0.072

-0.002 51 Cs-137 Fission Product.0.863 0.090 0.014 29 The tabulation excluding. the Cu-63 (n,a) and Ti-46 (n,p) reactions shows improved agreement for all of the subsets of data comprising the overall reaction rate data base.

All of the data subsets fall well within the standard deviation of the total data base.

Furthermore, the average reaction rate M/C ratio for the total data base (0.851 +/- 0.066) is in excellent agreement with the average M/C ratio from the least squares adjustment procedure (0.831 +/- 0.044) discussed in Response 2.1.

The reaction rate data comparisons, both with and without the Cu-63 (n,a) and Ti-46 (n,p) M/C ratios are illustrated graphically in Figure 2.2-1.

7

TABLE 2.2-1 COMPARISON OF MEASURED AND CALCULATED NEUTRON SENSOR REACTION RATES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS Cu63(n,a)

Ti46(n,Q)

Fe54(n,Q)

Ni58(n,Q)

U238(n,f)

Ng237(n,f) 1'nternal A240 (30°)

0.982 1.069 0.900 0.863 W290 (20°)

0.979 0.962 0.856 0.878 0.858 W290-9 (20°)

0.978 1.009 0.814 0.860 0.871 0.817.

W110 (20°)

0.997 0.993 0.852

. 0.865 6° Cavi~

Cycle 8 Cycle 9 0.938 0.964 0.868 0.844 0.875 0.932 Cycle 10/11 0.970 0.946 0.885 0.881.

1.049

. 0.874 16° Cavi~

Cycle 8 0.904 0.948 0.854 0.852 0.833 1.083 Cycle 9 0.883

'0.911 0.821 0.830 0.753' 0.900 Cycle 10/11 0.934 0.930 0.853 0.874 0.868 0.864 24° Cavi~

Cycle 8 Cycle 9 Cycle 10/11 0.863 0.886 0.798 0.811 0.916' 0.713 26° Cavi~

Cycle 8 0.887 0.933 0.830 0.825 0.797 0.992 Cycle 9 0.896 0.919 0.815 0.834 0.873 0.982 Cycle 10/11

  • 0.910 0.943 0.849 0.838 0.885 0.88!3 36° Cavi~.

Cycle 8 Cycle 9 Cycle 10/11 0.892' 0.859 0.782 0.835 0.798 39° Cavi~

Cycle 8 6.922 0.940 0.840 0.822 0.771 0.934 Cycle 9 0.891 0.896 0.794 0.798 0.708 0.752 Cycle 10/11 0.845 0.908 0.808 0.810 0.814 0.793 Average 0.922 0.942 0.836.. --

"0.843

-0,847 0.880 Std. Dev. (1o) 0.046 0.049 0.033 0.026 0.079 0.101 Average Bias Factor (K) 0.879 Standard Deviation ( 1 o)

+/-0.072 8

FIGURE 2.2-1

SUMMARY

OF MIC RATIOS BASED ON SENSOR REACTION RA TES PRIOR TO APPLICATION OF LEAST SQUARES ADJUSTMENT PROCEDURES I.I --...-------------------------.

  • Total Data Base
  • Excluding Cu & Ti 1.0

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0.8 Accelerated Wall Cavity Cs-137 Total 0.7 9

Request for Additional Information 2A

1) Describe how the shift in the cavity support was determined.
2) Does the shift result in an increase or decrease in the inferred vessel inner-wall fluence?
3) What is the effect of this shift on the capsule measurements and on the CIM bias?
4) What is the uncertainty in this shift and how has this been included?
5) Does the FERRET adjustment calculation account for this uncertainty?

CPCo.Response The following brief comments are in response to Requ~st 2.4. An extended discussion of each of the issues included in Request 2.4 is also provided in the response to Request 3.10.

1)

The radial shift in the bar supporting 'tfie reactor cavity multiple foil sensor sets was determine~ by a series of measurements taken coincident with the Cycle 10 dosimetry installation. The azimuthal shift was determined by* a measurement made during the initial installation for Cycle 8. -Measured azimuthal gradients from the multiple foil sets were also used to check the dosimeter positioning for.

consistency with the expected azimuthal shape of the neutron flux. *

2)

. The shift in the cavity support bar, if accounted for properly in the determination of M/C ratios, should not impact the magnitude of the inferred vessel inner-wall fluence. The uncertainty in the bar position does, however, contribute to the overall uncertainty in the neutron fluerice at the pressure vessel inner wall.

, 3)

The shift in the cavity dosimetrY support bar does not have an *effect on the*

measurement process. The measured sensor reaction rates are representative

. of the true location-of the sensor sets. The re~ultant M/C comparisons are impacted orily if the true positions of the sensor sets are not known.

4)

Based on post-installation measurements of the support bar location, the positioning uncertainty for the multiple foil sensor sets is estimated to be 2 inches in any directionfrom the nominal location. This positioning uncertainty is included as one component of the overall uncertainty associated with the vessel inner wall fluence.

5)

The uncertainty in the M/C ratios due to the positic;ming of the sensors is not included in the least squares adjustment; and, therefore, the bias factor uncertainty does not include this component. The uncertainty in sensor set position is an additional component to the overall uncertainty in the neutron 10

fluence at the vessel inner-wall and is included subsequent to the completion of the least squares adjustment procedure.

11

Reguest for Additional Information 2.7 The standard deviation in the overa/117% MIC fluence bias given in Table 7.1-1 appears to be larger than what would be expected from the numbers in the table.

Please indicate how the standard deviation in the f/uence MIC bias was determined. What statistical tests of the MIC data were performed to test for normality and justify the calculation of the standard deviation? Why was the log-normal least squares adjustment chosen?

CPCo Response Based* on the response to request 2.1, the average M/C ratio and the standard deviation based on the random*variations within the data base is 0.831 +/- 0.044. The average value and associated uncertainty quoted in Table 7.1-1 of WCAP-14557, Revision 1 is 0.831 +/- 0.067. The additional uncertainty included in Table 7.1-1 was due.

tO the inclusion of uncertainties associated with each of the individual measurement*

points comprising the overall M/C data base.

In performing the least squares adjustment for the 17 multiple foil sensor sets comprising the overall M/C data base, an uncertainty associated with the derived best estimate value of

1.0 MeV) is computed for each data point. These individual uncertainties are.reported in Sections 5.0 and 6.0 of WCAP-14557, Revision 1 and are summarized as follows. In-Vessel A240 (30°) 11 % W290 (20°) 9% W290'-9 (20°) 7% W110(20°) 11% 6 ° Cavity Cycle.g 7% Cycle 10/11 . 8% 16 ° Cavity Cycle 8 _7% Cycle 9 8% Cycle 10/11 8% 24 ° Cavity. Cycle 10/11 8% -25 ° Cavity *cycle 8

  • 7%*

Cycle 9 8% Cycle 10/11 8% 12 36 ° Cavity Cycle 10/11 39 ° Cavity Cycle 8 Cycle 9 Cycle 10/11 8% 7% 8% 8% These individual uncertainties reflect the best fit in the least squares adjustment including the combination of uncertainties in the calculated neutron spectrum, the measured reaction rates, and the dosimetry cross-sections. The combination of these uncertainties in the individual M/C ratios results in the slightly higher uncertainty in the average M/C value given inTable 7.1-1. The least squares adjusted <t> (E > 1.0 MeV) M/C data b,ase shown in Table 2.1-1 passes the W-test for normality described in ANSI N 15~ 15-197 4, "Assessment ofthe. Assumption of Normality (Employing Individual Observed Values)". The reasons for the choice of the log-normal least squares adjustment will be provided in a subsequent submittal. 13 . Reguest for Additional Information 3.1 O The reliability of the MIC f/uence bias and the FERRET adjustment procedure depends on reasonable agreement between the measured and calculated reaction rates. However, the measured reaction rates are sensitive to the capsule location and the position of the dosimeters inside the capsule, and the as-built positions of the dosimeters (relative to the core) typically include a substantial degree of uncertainty.

1)

Provide an estimate of the uncertainty in the dosimeter locations and the resulting uncertainty in the measured reaction rate.

2)

Describe how this uncertainty is included in the FERRET analysis.

3)

How does this uncertainty compare with the uncertainty in the calculated bias? CPCo* Response As noted in Section 8.1 of WCAP-14557, Revision 1, the best estimate neutron fluence at the Palisades reactor pressure vessel wall is determined from the following relationship: where:, <!>Best Est K <l>calc. = = = The best estimate fast neutron exposure at the location of interest. The plant specific measurement/calculation (M/C) bias factor derived from all available surveillance capsule and reactor cavity dosimetry data.. The absolute calculated fast neutron exposure at the location of interest. The bias factorK is-determined-from-a-comparison of the results of the least squares adjustment evaluation of all available dosimetry sets with the corresponding analytical predictions at the nominal location of the multiple foil sensor sets. The uncertainty associated with the least squares adjustment results includes a combination of the uncertainties in the measured reaction rates, dosimetry cross-sections, and the 14 calculated neutron spectrum at the nominal location of the dosimetry. The least squares adjustment does not include any component due to positional uncertainty of the dosimetry. Therefore, this additional componentis not reflected in the uncertainty associated with the bias factor, K. Note, from Table 7.1-1 of WCAP-14557, Revision 1, that the uncertainty in the bias factor was estimated to be approximately 8% [(0.067 /0.831)*100). As noted in Section 8.2 of WCAP-14557, Revision 1, the overall uncertainty in the best estimate fast neutron exposure of the pressure vessel wall includes, in addition to the uncertainty in the bias factor, K, components to account for the relative locations of the measurement points and the pressure vessel wall, the potential variation in the dimensions of the reactor internals, and the potential variations of water density within the reactor. The magnitude of these additional uncert~~nty component~ i_s _determined from analytical sensitivity studies carried out for the* Palisades reactor geometry. In addition, an added uncertainty of 5% is included to account for minor components of the overall uncertainty that are not specifically addressed in either the measurement to calculation comparisons or in the analytical sensitivity studies. Note, from Section 8.2 of WCAP-14557, Revision 1, that combining these added uncertainty components with the uncertainty in the bias factor obtained from the least squares adjustment M/C data base results in a net uncertainty of 14.5% in the best estimate projections of the fast neutron (E > 1.0 MeV) exposure of the Palisades reactor pressure vessel.

  • The following discussion pertains to the location and positional uncertainty associated*

with the* rea.ctor cavity and internal surveillal")ce capsule multiple foil sensor sets.. Reactor*Cavity Sensor Locations The multiple foil-sensor sets irradiated in the Palisades reactor cavity during fuel cycles 8 through 11 were suspended from an aluminum support bar that was originally intended to be positioned at a nominal radius of 108 inches relative to the core centerline. The azimuthal positioning of the support chains supporting the aluminum bar ~as determined relative to reactor coordinates stamped on the reactor vessel head at the 0°, 90°, 180°*, and 270° cardinal axes and to seal ring bolt holes equally spaced in the reactorvessel seal ledge on a 213-11/16 inch bolt circle. During a walkdown of the initial installation, it was noted that the aluminum support bar . was displaced azimuthally due to the passage of a support chain located at 330° over an angle iron located in the reactor cavity. This displacefllent also resulted in a radial skew of the -slipporfl5a"fwith -thef 270 ° end of the b~r moving closer to the reactor pressure vessel. At that time, an estimate of the azimuthal displacement was made using a plumb line dropped at 300° as a reference point. The bar was initially estimated to have been translated azimuthally by approximately 5 °. Consistency 15 \\ "t checks using measured and calculated azimuthal neutron flux gradients indicated that the azimuthal shift was actually 6 °. During the dosimetry installation prior to the onset of Cycle 10, a more detailed set of measurements were made to determine the as-built radial location of the dosimetry bar in-the reactor cavity relative to the outside surface of ttie mirror insulation on the reactor vessel. The distances measured were from the surface of the vessel insulation to the center of the LI-tubes on the dosimetry support bar. These as-built locations are depicted in Figure 3.10-1. Also shown on Figure 3.10-1 are bands representing a potential +/- 3 inch displacement of the dosimetry sets in both the azimuthal and radial directions. Note that for a radius of 108 inches, a translation of+/- 3 inches corresponds to an angular displacement of+/- 1.6 °. Although the uncertainty in sensor set positioning is esfi~ated to be +/- 2 inches based on the measurements obtained after installation of the dosimetry suppo~ bar, comparisons are provided here for a range of+/- 3 inches in order to better illustrate the behavior of the neutron field in the cavity. In performing these comparisons it was assumed that translation of the sensor sets could take place in any direction relative to the nominal position. All comparisons provided herein are based on gradient

  • information extracted from the Cycle 9 transport calculation.

An examination ofthe neutron flux and reaction rate gradients within the reactor cavity indiGate that variations in the azimuthal direction are less than the variations in the radial direction. Therefore, the largest impact of the uncertainty in dosimetry location - can be determined by an evaluation of the radial gradient data at each azimuthal location containing dosimetry. Gradient data describing _the radial variation of cp (E > 1.0 MeV) at the first octant equivalent(FOE) 6°, 16°, 24°, 26°, 36°, and 39° dosimeter locations ar~ provided in' Figures 3.10-2 through 3.10-7. The gradient information. shown on Figures 3.10-2 through 3.10-7 is provided in terms of fractional change in calculated response as a function of distance from the nominal sensors.et position. Over the range of interest, the gradient data for all fast neutron reaction rates are. similar to those observed for cp (E > 1.0 MeV). - An examination of Figures 3.10-2 through 3.10-7 shows that within the 3 inch range from nominal position the fractional change in calculated flux ranges from +/- 0.5% for the 39° azimuth to+/- 5% at-.the 16° location. Therefore, the maximum bias that could be introduced into the M/C comparisons due to mis-positioning of the cavity dosimetry sensor sets would be expected to be no greater than 5% with an average uncertainty for all cavitrs-ensor sets-substantially less than-the-5%-_ value. 16 I I Internal Capsule Locations Multiple foil sensor sets irradiated in the internal surveillance capsules are positioned laterally at a constant radius from the core center. The lateral gradients across the capsule are small; and as is the case with the cavity dosimetry the maximum impact of positioning uncertainties occurs due to radial rather than azimuthal gradients. However,

  • unlike the reactor cavity where positioning uncertainties result in a translation in air, the positioning uncertainties for the internal capsules result in a translation through either a water or steel medium.

In assessing this component of the uncertainty for the Palisades reactor, dosimeter positioning was assumed to be +/- 0.25 inches and +/- 0.125 inches for the wall and accelerated capsules, respectively. These dimensions were taken from the allowable tolerance specified in the design drawings for the Palis*ades reactor. Based on parametric studies of the variations of capsule positioning within the reactor environment, the following impacts were developed for the in-vessel dosimetry sets. Dosimeter Position (acc.) 11.0%/cm Dosimeter position (wall) 6.0%/cm 3.5% uncertainty 3.8% uncertainty It should be noted that these uncertainties, while representing a maximum based on tolerance limits, were treated as 1 o values* in the overall,uncertainty associated with the

  • neutron exposure of the pressure vessel wall.

17 FIGURE 3.10-1 LOCATION OF REACTOR CAVITY MULTIPLE FOIL SENSOR SETS 350 340-Bioshield Wall . 330 320 310 -s 300 u - Vl 290

  • =

"Cl* C'll 280. c:i::: + 270 + + + 260 250 240 Pressure Vessel 230 0 10 20 30 40 50 60 70 Azimuthal Angle (deg} 18 I Q ~ ~ =: u --::; .5 ~ ~ = ~ .c: u -; = -~ - ~ ~ ~ FIGURE 3.10-2 FRACTIONAL CHANGE AS A FUNCTION OF RADIAL POSITION 6 DEGREE AZIMUTHAL TRAVERSE $ (E > 1.0 MeV) . 0.05 0.00 -0.05 -4 -3 -2 -I 0 2 3 4 . Distance From Nominal Radius (in) 19 I .~ ~ c:i:: u --::; .5 Q,j* 'l:ll) c ~ .c

  • u c

.~ (j ~ ~ FIGURE 3.10-3 FRACTIONAL CHANGE AS A FUNCTION OF RADIAL POSITION 16 DEGREE AZIMUTHAL TRAVERSE <!> (E > 1.0 MeV) 0.05 0.00 -0.05 -4 -3 -2 -1 o* 2 3 4 Distance From Nominal Radius (in) 20 l FIGURE 3.10-4 FRACTIONAL CHANGE AS A FUNCTION OF RADIAL POSITION 24 DEGREE AZIMUTHAL TRAVERSE ip (E > 1.0 MeV). 0.10 -.--------------------------. -~ - 0.05 ~ =::: u -- ~ .5 ai ~ 0.00 c ~ .c u c Q *.: ~ ~ -0.05.

  • i;i;.

-0. I 0 -l", -..,......--,--,....--....,....----,.-..,......--,--,....--....,....----,.-..,......-..,...-,....--....,....---,~ -4 -3 -2 -1 0 2 3 4 Distance From Nominal Radius (in). 21 I FIGURE 3.10-5 FRACTIONAL CHANGE AS A FUNCTION OF RADIAL POSITION 26 DEGREE AZIMUTHAL TRAVERSE <!> (E > 1.0 MeV) 0.10 -.-----------------------~ . s o.os* ~ u 'i ~ c 0.00 ~ .c u c.s - CJ E -o.os ~ -4 -3 -2 -1 0 I 2 3 4 Distance From Nominal Radius (in) ~ ---- 22 .2 -"' " u ~ .5 Q> Oil c .c u ~ c .2 - c..i lo. ~ FIGURE 3.10-6 f RACTIONAL CHANGE ASP.. FUNCTION OF RADIAL POSITlON 36 DEGREE AZIMUTHAL TRAVERSE ~ (E > 1.0 MeV) 0.10 ~-----:..........;. _______________ . 0.05 . 0.00 -0.05 -0.10 -4 -3 -2 -1 0 2 3 .4 Distance From Nominal Radius (in)

1.

23 Q -.: ~ c:i:: u --::; .5 a.I Oil c ~ .c u ~ c.s - (,,I ~

a.

~ FIGURE 3.10-7 FRACTIONAL CHANGE AS A FUNCTION OF RADIAL POSITION 39 DEGREE AZIMUTHAL TRAVERSE $ (E > 1.0 MeV) 0.05 0.00 -0.05 -4 -3 -2 -1 0 2 3 4 Distance From Nominal Radius (in) 24 ATTACHMENT 2 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 PALISADES RESPONSE TO REQUEST FOR INFORMATION FROM AUGUST 29, 1996 TELECONFERENCE 3 Pages

  • consumers Power POW ERi Nii lllllCHlliAN"S PROliRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 August 29, 1996 Brookhaven National Laboratory Attn: Mr. Arnie Aronson

Dear Arnie:

I have included photocopies of two pages from previous submittals. The first page is Table 6-13 on page 6-27 fr()m WCAP-14014 submitted to the staff on June 12, 1994. This provides the.

EFPD's for cycles 1 through 10 explicitly. All of these values assume a rated power of 2530 MWth. However the plant only ran at 80% (2200 MWth) for this value for the first one and half cycles.

  • *The cycle 11 EFPD value can be calculated using the,data provided on page D-2, Table *D-1 of WCAP.14557 Rev. 1, submitted to the staff on April 4, *i996. Summing the MW-hr for cycle 11 gives 26131033 MW-hr of thermal generation during cycle 11. Dividing by 2530 MWth times

.24 hrs provides a value of 430.4 EFPD's during cycle.11. The values for cycles 1 through 10 can be calculated in the same manner using the values provided at in Appendix A of WCAP 14557 Rev. 1.

If there is any other data that you require' to continue your review of our submittal please let us know so that we can address these request promptly. Our next submittal to the staff will cover*

the dosimetry capsule position and uncertainty.

Sincerely, 12* S:~,_j Ross-D-Snuggerud-*-. -*-- ---- -- - --

  • cc Lambrois Lois Charlie Kozup George Goralski RDS96*15 A CMS ENERGY COMPANY

J f

c,_:,.

~*

  • TABLE 6-13 CALCULATED FLUENCE (E>l.0 MeV) TIIROUGH CYCLE 10 AT TIIE PRESSURE VF.SSEL CLAD-BASE ~AL INTERFACE Cycle Cycle Length Cycle Flux Cycle Fluence Cumulative j

(EFPD)

(D/cm2-sec)

(D/cm2)

Fluence (D/cm2) 0 Deg[ee l

379.4 4.59E+10 l.50E+l8 l.50E+l8 2

449.l 4.59E+l0 l.78E+18 3.28E+l8 3

349.5 4.59E+10 l.39E+l8 4.67E+l8 4

327.6 4.59E+10 l.30E+l8 5.97E+l8 5

394.6 4.59E+l0 l.56E+l8 7.53E+l8 6

333.4 4.87E+10 l.40E+l8 8.94E+l8 7

369.9 4.87E+l0 l.56E+l8 l.05E+l9 8

373.6 2.16E+10 6.97Etl7 Ll2E+l9 9

298.5 2.08E+10 5.36E+l7 Ll7E+l9 10 356.9

.l.51E+l0 4.66E+l7 l.22E+l9 16 Deg[ee*

1 379.4 6.03E+l0 '..

  • 1.98E+l8 l.98E+l8 2

449.1 6.03E+10 2.34E+l8 4.31E+l8 3

349.5 6.03E+10 l.82E+l8 6.13E+l8 4

327.6 6.03E+10 l.71E+l8 7.84E+l8

5.

394.6 6.03E+l0 2.05E+l8 9.89E+l8 6

333.4 6.25E+10 l.80E+l8 l.l 7E+l9 7

.369.9 6.25E+l0 2.00E+l8 l.37E+l9 8

373.6 4.89E+l0 l.58E+l8 l.53E+l9 9

298.5 3.06E+l0 7.89E+l7 l.61E+l9 10

'356.9 2.40E+10 7.40E+l7 l.68E+l9

-* Pressure-vessel iriiier" radius -m.ammum fluence *w.muth.

6-27.

  • ~

c TABLE D-1 IRRADIATION ffiSTORY OF REACTOR CAVITY SENSOR SETS Cycle 10 Cycle 11 Thermal Thermal Generation

  • Generation

~ MW-hr Date MW-hr Mar-92 0

Jul-93 0

Apr-92 620112 Aug-93 0

May-92 1878432 Sep-93 0

Jun-92 1819464 Oct-93" 0

Jul-92 1392552 Nov-93 1242336 Aug-92 1459272 Dec-93 1876608 Sep-92 1260672 Jan-94 1844112 Oct-92 1779079 Feb-94 1004688 Nov-92 1326168 Mar-94 0

Dec-92 1880496 Apr-94 0

Jan-93 1879536 May-94 0

Feb-93 1698408 Juri 666768 Mar-93 1880544 Jul-94 1874208

. Apr-93 1688919 Aug-94 1874448 May-93 862632 Sep-94 1812408 Jun-93 237864 Oct-94 1869960 Nov-94 1813872 Dec-94

1867368, Jan-95 1874304 Feb-95 1655688 Mar-95 1871832 Apr-95 1811689

. ___ May-95..

. 1170744 D-2

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