ML18054A634

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Rev 2 to Analysis of Fast Neutron Exposure of Palisades Reactor Pressure Vessel.
ML18054A634
Person / Time
Site: Palisades Entergy icon.png
Issue date: 03/31/1989
From: Shaun Anderson, Lippincott E
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML18054A630 List:
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NUDOCS 8904120337
Download: ML18054A634 (30)
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  • E. P. Lippincott S. L. Anderson Westinghouse Electric Corporation Nuclear and Advanced Engineering Division March 1989 0165I:EPL-013189 8904 f2<5:=:3? 89c146-:::- - - -- "

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  • I' 1.0 Introduction This report presents the results of a series of two-dimensional discrete ordinates neutron transport calculations that were performed to determine the current fast neutron exposure of the Palisades reactor pressure vessel and to assess the degree of exposure reduction that could be achieved by the introduction of shielded fuel assemblies in future fuel cycles. The transport computations carried out in R,9 geometry were designed to provide exposure rates averaged over fuel cycles 1 - 1 (standard fuel management) as well as for the cycle 6 design using shi_elded fuel assemblies. In both cases, exposure assessments were made in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) initiated by neutrons above 0.111 MeV .
  • The exposure data generated from these transport calculations may be used in conjunction with damage trend curves to predict the degree of stee.l embrittlement in terms of RTPTS for current and future vessel operation. In addition to pressurized thermal shock assessments, the gradient information developed during this study may be employed to determine pressure-temperature limitations for normal plant operation as well as to evaluate the effect of various heatup/cooldown transients on vessel conditions.

In subsequent sections of this report, the methodology used in the analysis is discussed in some detail, results of the analyses are presented, and a discussion of the adequacy of the analytical approach is provided .

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2.0 Method of Analysis The neutron flux distribution for the Palisades Reactor was calculated (1)  !

using the DOTIIIW Sn transport code. Two.calculations were perfonned 1n R-8 geometry: the first calculation modeled a configuration representative of an average over cycles 1 to 1. and the second calculation modeled the proposed cycle 8 configuration. The model for the cycle 1 - 7 geometry is shown in Figure 2-1. A 45° octant is shown and the model includes a surveillance wall capsule located at 2.0° from the 0° axis (the x axis in the figure). Regions included in the model are the outer part of the fuel zone, baffle plate, barrel, reactor vessel, insulation at the outside of the vessel and next to the concrete shield, and part of the concrete shield. The cycle 8 geometry is shown in Figure 2-2. It is identical to the cycle l - 1 model except for the replacement of fuel by stainless steel rods in four outer rows of two assemblies. These stainless steel rod regions are modeled as a homogenous mixture of stainless steel (47%) and water. (53%).

The DOT model is an updated version of that used in a previous calculation,< 2> with the addition of additional mesh points to better model the shroud. Also, the regions outside the vessel were added to allow evaluation of the neutron exposure in the cavity. The two insulation regions were modeled as a mixture of stainless steel (l.85% in inner insulation and 1.54% in outer), aluminum (0.21% and 0.18%), and air.

Macroscopic cross-sections were calculated for each region in the model using the code GENESIS. Atom densities supplied by Consumers Power are shown in Table 2-1.< 3> Appropriate modifications to these densities were made for the water and insulation regions. Cross-sections were derived from the SAILOR< 4> library except for N and Mo which were obtained from the BUGLE-80( 5) library.

The neutron source input to the DOT calculations were determined from core power distributions provided by Consumers Power.< 3*6 *7 *8> The distributions were processed by the code SORCERY which transforms the 704Bq/js

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power calculated for each pin location into the neutron group by group source for the R-9 DOT geometry. Input to SORCERY included detailed pin by pin power distributions for the outer assemblies< 6 *7> and assembly power distributions as shown in Figures 2-3 and 2-4 for cycle 1

- 7 and cycle 8, respectively.CB) The power input to SORCERY was assumed to be the axial maximum for each assembly, so the assembly average power distributions were increased by the peaking factor for each assembly. These factors are also given in Figures 2-3 and 2-4. For assemblies for which a peaking factor was not given, a value equal to the average of the non-edge assemblies was ~sed. The core average axial peaking factor (1.125 for both cases) was determined by averaging all the assemblies. Based on the fuel cycle averaged burnup of individual fuel assemblies, composite fission neutron spectra, average number of neutrons released per fission, and values of energy release per fission were developed using basic nuclear data from the ENDFB-V data files.

Plutonium - Uranium fission fraction computations were based on the assembly burnups and a nominal 3.1% enrichment in feed assemblies. The resulting fission spectra {x(E)}, v, and K were then input to the SORCERY code to provide appropriate neutron source distributions for the subsequent discrete ordinates calculations.

The DOT model used 150 radial mesh points and 59 angular points. Extra points were used to mock up the baffle geometry in detail and to obtain precise values at specific angles of interest including 0°, 17°, 30°, and 33°. The calculation was carried out using an s8 angular quadrature and a P3 cross-section expansion. The 47 energy group SAILOR set (Table 2-2) was used but only the upper 26 neutron groups were converged to obtain the flux above 0.111 MeV and dpa. The entire geometry was run for the cycle 1 - 7 case but this was found to result in excessive computer usage so the cycle 8 case was divided in half with an overlap region. This was found to introduce negligible error at points of interest .

TABLE 2-1 ISOTOPIC NUMBER DENSITIES (ATOHS/B-CH)

NUCLIDE CARBON CONCRETE STAINLESS HOMOGENIZED WATER* AIR STEEL STEEL CORE (DENSITY lCH/CC)

(SA-302B) (TYPE 304)

H ANSI Standard 2.825-2 6.679-2 B-10 Type 04@2.31 2.692-5 6.649-6 c 9.84-4 2.581-5 N 4.271-5 0 2.697-2 3.340-2 1.129-5 All 3.560-4 I Si 7.589-4

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I Ca Cr 1.853-2 1.608-5 Hn 1.391-3 .. 1.752-3 Fe 8.183-2 5.807-2 2.205-5 Ni 8.574-3 3.877-5 Zr 5.448-3 Ho 3.148-4 U235 1.12-4 U238 6.03-3 PU239 2.20-5 PU240 7.12-6

  • By-pas1 water is at 560.F and 2010 psia and inlet water is at 536°F and 2010 psia.

Appropriate density values should be used for these regions.

OC0188-00058-0P03

TABLE 2-2 ENERGY GROUP STRUCTURE USED IN TRANSPORT ANALYSIS Lower Energy Relative Fission Spectra Group CMeV) Cycle 1-7 Cycle 8 l 14. 19* 4.4841 (-5) 4.6932 (-5) 2 12.21 1.8826 (-4) 1.9415 (-4) 3 10.00 1.1058 (-3) 1.1268 (-3) 4 8.61 2.6187 (-3) 2.6574 (-3) 5 7 .41 5.8744 (-3) 5. 9344 (-3) 6 6.07 1.6720 (-2) 1.6827 (-2) 7 4.97 3.1855 (-2) 3.1976 (-2) 8 3.68 8 *.3549 c-2> . 8.3794 (-2) 9 3.01 . 7.8003 (-2) 7.8145 (-2) 10 2.73 4.4029 (-2) 4.4088 (-2) 11 2.47 4.6126 (-2) 4.6174 (-2) 12 2.37 1.9759 (-2) 1.9776 (-2) 13 2.35 3.8327 (-3) 3.8355 (-3) 14 2.23 2.3995 (-2) 2.4011 (-2) 15 1.92 7 .1960 (-2) 7.1983 (-2) 16 1.65 7.0237 (-2)' 7.0220 (-2) 17 1.35 8.8042 (-2) 8.7953 (-2) 18 1.00 1.1320 (-1) 1.1296 (-1) .

19 0.821 6.1965 (-2) 6.1777 (-2) 20 0.743 2.6821 (-2) 2.6731 (-2) 21 0.608 4.6036 (-2) 4.5868 (-2) 22 0.498 3.6937 (-2) 3.6793 (-2) 23 0.369 4.1396 (-2) 4.1238 (-2) 24 0.298 2 .1535 (-2) 2 .1462 (-2) 25 -0.183 3.0971 (-2) 3.0939 (-2) 26 0.111 1. 6332 (-2) 1.6322 (-2)

  • *The upper energy of group 1 is 17. 33 MeV.

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Palisades Geometry Cycles 1 - 7 2H

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  • Figure 2-3 Normalized Radial Power Distribution by Assembly for Cycles 1 - 7 oo 0.892 0.874 0.695 1.138 1.147 1.152 0.981 1 .074 1.133 0.956 0.611 1.116 1.127 1.144 1.145 1.153 1.032 1.002 0.923 1.059 0.972 ,45° 1.109 1.122 1.116 l.127 1.142 ,/

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  • 1.103 l .005 l .015 1.141 1.171 1.069 0.938 1.108 0.972 1.175 1.144 1.118 1.047 1.074 1.126 0'.937 0.965 0.892 Normalized Power 1.138 Axial Peaking Factor for Outer Assemblies 7048Q/js I # * ~

Figure 2-4 Normalized Radial Power Distribution by Assembly for Cycles o*

I rI 0.356 0. 756 : 0.641 1.084 1.135 j 1.145

1. l 01 1.1061 1.133 0.576 0.221
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1. 219 1.043 1.299 1.076 1.239 1.115 1.166 l .009 1.286 1.005 1.133 l .200 0.997 1.100 1.166 1.128 0.931 0.356 Normalized Power 1.084 Axial Peaking
  • Factor for Outer Assemblies 7048q/js
  • 3.0 Results of Analysis Results of the neutron transport analysis of the Palisades reactor are surrmarized graphically in Figures 3-1 through 3-4 and in tabular form in Tables 3-1 through 3-7. These data, applicable to the peak in the axial exposure distribution, represent the absolute exposure rates calculated for operation at a core thermal po~er level of 2530 MW. Again, d~e to symnetry considerations, data are presented only for a 0° - 45° sector.

In Figure 3-1 and Table 3-1, the azimuthal distribution of fast neutron flux (E > 1.0 MeV) at the pressure vessel clad-base metal interface is presented for both cycles 1 - 7 and cycle 8 core power distributions.

Similar data illustrating the azimuthal distribution of iron atom displacement rate (dpa/sec) are given in Figure 3-2 and Table 3-2. The dpa rate is determined by the DOT code by multiplying the group dpa cross sections by the group fluxes at each location and su11111ing the contributi.on

  • for each of the 26 groups. The dpa cross section in the 47 group structure is derived from ASTM Standard E693.C 9> Contributions to the dpa rate by neutrons below 0.111 MeV have been neglected. At typical locations this omission is less than 3% of the integral cross section. An examination of Figures 3-1 and 3-2 clearly indicates the impact of the introduction of the shielded fuel assemblies into the cycle 8 reload design.

In Figure 3-3 and 3-4 the relative radial distributions of fast neutron flux (E > 1.0 MeV) and iron atom displacement rate are presented for cycles l - 7 and cycle 8, respectively. These graphical representations are representative of the peak location in the azimuthal exposure rate distributions (16.44°). In Tables 3-3 through 3-6 relative radial distribution data are given for several azimuthal locations in addition to the peak location illustrated in Figures 3-3 and 3-4. The data in Tables 3-3 through 3-6 have been normalized to the absolute exposure rates calculated at the pressure vessel clad/base metal interface.

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  • Therefore, the azimuthal data listed in Tables 3-1 and 3-2 and the relative radial data given in Tables 3-3 through 3-6 can be used in a multiplicative fashion to create R,9 maps of both fast neutron flux (E > 1.0 MeV) and iron atom displacement rate within the pressure vessel wall.

In Table 3-7 updated exposure parameters and spectrum averag,ed neutron dosimetry cross-sections are provided for the 20° surveillance capsule modeled in the analysis. These data may be used in the evaluation of lead factors based on cycle specific analyses .

  • 704Bq/js TABLE 3-1 FAST NEUTRON FLUX {E>l.O MEV} AT THE CLAD/BASE METAL INTERFACE - AXIAL MAXIMUM e * (n/cm2 - sec) e f (n/cm 2 - sec)

(deg.) Cycles 1-7 Cycle 8 (deg.) Cycles 1-7 -_ Cycle 8 0.125 4.23 x 1010 2.07 x 1010 22.75 4. 71 x 10-10 3.70 x 1010 0.500 4.23 x 1010 2.06 ~ 1010 23.25 4.61 x 1010 3~56 x 1010 1.25 4.23 x 1010 2.07 x 10 10 23.75 4.51 x 1010 3.41 x 1010 2.25 4.26 x 10 10 2.14 x 10 10 24.50 4.36 x 1010 3.19 x 10 10 3.25 4.32 x 1010 2.26 x 1010 25.50 4.23 x 1010 2.95 x 1010 4.25 4.38 x 10 10 2.41 x 1010 26.50 4.17 x 1010 2. 74 x 1010 5.37 4.47 x 10 10 2.60 x 10 10 27.50 4.14 x 1010 2.55 x 1010 6.62 4.60 x 10 10 2.86 x 10 10 28.50 4.14 x 1010 2.39 x 1010 7.62 4.71 x 10 10 3.10 x 1010 29.44 4.16 x 1010 2.27 x 10 10 8.50 4.83 ~ 1010 3.32 x 10 10 30.00 4.17 x 1010 2.21 x 10 10 9.50 4.96 x 10 10 3.57. x 10 10 30.56 4.18 x 1010 2.14 x 10 10

10. 50 . 5.11 x 1010 3.84 x 1010 31.50 4.19 x 1010 2.04 x 10 10 11.50 5.23 x 10 10 4.08 x 1010 32.44 4.16 x 1010 1.94 x 1010 12.50 5.36 x 10 10 4.31 x 10 10 33.00 4.09 x 1010 1.91 x 1010 13.50 5.46 x 10 10 4.50 x 10 10 33.56 4.07 x 10 10 1.86 x 1010 14-.50 5.54 x 10 10 4.65 x 10 10 34.50 . 3.98 x 1010 1. 79 x 1010 15.50 5.59 x 10 10 4.75 x 10 10 35.50 3.86 x 1010 1. 74 x 10 10 16.44 5.62 x 10 10 4.81 x 1010 36.25 3.73 x 10 10 1. 71 x '10 10 17.00 5. 59 x 1010 4.78 x 1010 36.75 3.65 x 1010 1.69 x 10 10 17.56 5.56 x 1010 4.79 x 1010 37.38 3.56 x 1010 1.67 x 1010 18.25 5.25 x 1010 4.50 x 1010 38.12 3.43 x 1010 1-.66 x 1010 18.84 5.10 x 10 10 4.36 x 1010 38.87 3.30 x 1010 1.66 x 1010 19.22 4.96 x 1010 4.22 x 1010 39.62 3.16 x 1010 1.66 x 10 10 19.51 4.84 x 10 10 4.10 x 1010 40.50 3.01 x 1010 1.65 x 1010 20.01 4.70 x 10 10 3.95 x 1010 41.37 2.89 x 1010 1.64 x 10 10 20.50 4.66 x 1010 3.87 x 1010 42.25 2.78 x 1010 1.63 x 1010 20.78 4.67 x 1010 3.87 x 1010 43.12 2.71 x 1010 1.63 x 10 10 21.11 4.69 x 10 10 3.85 x 1010 43.88 2.67 x 10 10 . 1.63 x 10 10
21. 71 4.66 x 1010 3. 76 x 1010 44.63 2.64 x 1010 1.62 x 10 10 22.25 4.76 x 1010 3.80 x 10 10

. 761 le:ld/030789 TABLE 3-2 IRON ATOM DISPLACEMENT RATE {dEa/sec~ AT THE CLAD/BASE METAL INTERFACE - AXIAL MAXIMUM e dpa/sec e dpa/sec (deg.) Cycles 1-7 Cycle 8 (deg.) Cycles 1 cxcle 8 0.125 6.32 x 10- 11 3.13 x 10-11 22.75 7.10 x 10- 11 5. 57 x 10- 11 0.500 6.32 x 10-ll 3.12 x 10- 11 23.25 6.92 x 10- 11 5.-34 x 10- 11 1.25 6.32 x 10- 11 3.13 x 10- 11 23.75 6.75 x 10- 11 5.10 x 10- 11 2.25 6.37 x 10- 11 3.23 x 10- 11 24.50 6.53 x 10- 11 4.78 x 10- 11 3.25 6.44 x 10- 11 3.40 x 10- 11 25.50 6.33 x 10- 11 4.42 x 10- 11 4.25 6.42 x 10- 11 3.62 x 10- 11 26.50 6.22 x 10- 11 4.10 x 10- 11 5.37 6.54 x 10- 11 3.90 x 10- 11 27.50 6.17 x 10- 11 3.82 x 10- 11 6.62 6. 66 x 10- 11 4.27 x 10- 11 28.50 6.17 x 10- 11 3. 59 x 10- 11 7.62 6.85 x 10- 11 4.62 x 10- 11 29.44 .6.19 x 10- 11 3.40 x 10- 11 8.50 7.02 x 10- 11 4.93 x 10- 11 30.00 6.20 x 10- 11 3.31 x 10- 11 9.50 1 .18* x 10- 11 5.31 x 10- 11 30.56 6.21 ~ 10- 11 3.20 x 10- 11 10.50 7.38 x 10- 11 5.68 x 10- 11 31.50 6.21 x 10- 11 3.06 x 10- 11 11.50 7.64 x 10- 11 6.04 x 10- 11 32.44 6.16 x 10- 11 2.93 x 10- 11 12.50 7.76 x 10- 11 6.37 x 10- 11 33.00 6.06 x 10- 11 2.87 x 10- 11 13.50 7.93 x 10"" 11 6.64 x 10*- 11 33.56 6.03 x 10- 11 2.79 x 10- 11 14.50 8.09 x 10- 11 6.86 x 10- 11 34.50 5.89 x 10- 11 2.69 x 10-ll 15.50 8.21 x 10- 11 . 7.02 x 10- 11 35.50 5. 70 x 10- 11 2.61 x 10- 11 16.44 8.30 x 10- 11 7.11 x 10- 11 36.25 5. 52 x 10- 11 2.57 x 10- 11 17.00 8.35 x 10- 11 7.10 x 10- 11 36.75 5.41 x 10- 11 2.54 x 10-ll 17.56 8. 33 x 10- 11 7.15 x 10- 11 37.38 5.28 x 10- 11 2.51 x 10- 11 18.25 8.35 x 10- 11 6. 77 x 10- 11 38.12 5.10 x 10- 11 2. so x 10- 11 18.84 7.93 x 10-i-i* 6.61 x 10- 11 38.87 4.92 x 10- 11 2.49 x 10-ll 19.22 ~1. 59 x 10- 11 6.43 x 10- 11 39.62 4.63 x 10- 11 2.49 x 10-ll 19.51 7.41 x 10- 11 6.25 x 10- 11 40.50 4.50 x 10- 11 2.48 x 10- 11 20.01 7.22 x 10- 11 6.04 x 10- 11 41.37 4.34 x 10- 11 2.47 x 10- 11 20.50 7.15 x 10- 11 5.92 x 10- 11 42.25 4.17 x 10- 11 2.46 x 10- 11 20.78 7.17 x 10- 11 5.91 x 10- 11 43.12 4.08 x 10- 11 2.45 x 10- 11 x 10- 11 . 5.89 x 10- 11 x 10- 11 21.11 7.18 43.88 4.01 2.45 x 10-ll

21. 71 7.09 x 10- 11 . 5. 72 x 10- 11 *44.63 3.98 x 10- 11 2.45 x 10-ll 22.25 7.21 x 10- 11 5.74 x 10-ll 7611e:1d/030789
  • TABLE 3-3 RELATIVE RADIAL DISTRIBUTION OF FAST NEUTRON FLUX

~E > 1.0 MeV} WITHIN THE PRESSURE VESSEL WALL CYCLES 1-7 Radius Relative Neutron Flux (cm) 0.125° 16.44° 33° 44.63° 219.07(l) 1.000 1.000 1.000 1.000 219.52 0.974 0.972 0.973 0.976 220.40 0.895 0.889 0.894 0.900 221.29 0.812 0.800 0.810 0.817 222.17 0.730 o. 716 0.726 0.736 223.06 0.653 0.637 .0.648 0.660 224.03 0.576 0.558 0.570 0.583 225.11 0.499 0.480 0.492 0.506 226.22 0.430 0.410 0.422 0.437 227.33 0.369 0.349 0.361 0.376 228.44 0.315 0.297 0.308 0.322 229.56 0.269 0.252 0.262 0.276 230.68 0.229 0.213 0.223 0.236 231.79 0.195 0.181 0.189 0.202 232.90 0.166 0.153 0.160 0.172 234.01 0.141 0.129 0.136 0.147 235.12 0.119 0.108 0.115 0.125 236.24 0.101 0.0907 0.0963 0.106 237.35 0.0844 0.0756 0.0806 0.0895 238.46 0.0702 0.0624 0.0669 0.0753 239.53 0.0581 0.0510 0.0553 0.0636 240.35 0.0493 0.0426 0.0464 0.0552 240.67( 2) 0.0474 0.0408 0.0446 0.0537 (l)Reactor vessel inner radius (Clad-Base Metal Interface)

<2>Reactor vessel outer radius 7611e:1 d/033188

TABLE 3-4 RELATIVE RADIAL DISTRIBUTION OF FAST NEUTRON FLUX

{E > 1.0 MeV} WITHIN THE PRESSURE VESSEL WALL CYCLE 8 Radius Relative Neutron Flux (cm} 0.125° 16.44° 33° 44.63° 219.07(l) 1.000 1.000 1.000 1.*000 219.52 0.968 0.967 0.969 0.969 220.40 0.891 0.885 0.891 0.892 221.29 0.810 0.796 0.809 0.809 222.17 0.730 o. 711 0.728 0.728 223.06 0.655 0.632 0.651 0.652 224.03 0.579 0.553 0.574 0.575 225.11 0.503 0.476 0~498 0.499 226.22 0. 43_4 0.406 0.429 0.429 227.33 0.374 0.345 0.368 0.368 228.44 0.321 0.293 0.315 0.316 229.56 0.275 0.248 0.269 0.269 230.68 0.236 0.210 0.230 0.230 231.79 0.202 0.177 0.196 0.196 232.90 0.172 0.150 0.167 0.167 234.01 0.147 0.126 0.142 0.142 235.12 0.125 0.106 0.120 0.121 236.24 0.106 0.0883 0.102 0.102 237.35 0.0897 0.0733 0.0858 0.0866 238.46 0.0754 0.0601 0.0720 0.0730 239.53 0.0634 0.0487 0.0604 0.0617 240.35 0.0547 0.0402 0.0522 0.0539 240.67_( 2) o. 0513 0.0370 0.0491 0.0511 7611e:1d/011689 TABLE 3-5 RELATIVE RADIAL DISTRIBUTION OF IRON ATOM DISPLACEMENT RATE WITHIN THE PRESSURE VESSEL WALL CYCLES 1-7 Radius Relative Displacement Rate (cm) 0.125° 16.44° 33° 44.63° 219.01< 1> 1.000 1.000 1.000 1.'000 219.52 0.978 0.978 0.977 0.979 220.40 0.913 0.910 0.911 0.917 221.29 0.844 0.837 0.841 0.850 222.17 o. 776 0.765 o. 771 0.783 223.06 o. 711 0.699 0.705 0.721 224.03 0.645 0.631 0.639 0.657 225.11 0.579 0.563 0.571 0.592 226.22 0.518 0.501 0.509 0.532 227.33 0.463 0.445 0.454 0.477 228.44 0.414 0.395 0.404 0.429 229.56 0.369 0.351 0.360 0.385 230.68 0.330 0.312 0.320 0.345 231.79 0.294 0.277 0.285 0.310 232.90 0.262 0.245 0.253 0.278 234.01 0.233 0.216 0.224 0.249 235.12 0.206 0.190 0.198 0.222 236.24 0.181 0.166 0.174 0.197 237.35 0.158 0.143 0.152 0.175 238.46 0.136 0.122 0.131 0.153 239.53 0.116 0.102 0.111 0.134 240.35 0.101 0.0867 0.0964 0.119 240.67( 2) 0.0970 0.0832 0.0931 0.116

(!)Reactor vessel inner radius (C1ad-Base Meta1 Interface)

( 2)Reactor vessel outer radius

-i6-7611e:1d/033188

TABLE 3-6 RELATIVE RADIAL DISTRIBUTION OF IRON ATOM DISPLACEMENT RATE WITHIN THE PRESSURE VESSEL WALL CYCLE 8 Radius Relative Displacement Rate (cm) 0.125° 16.44° 33° 44.63° 219.07(l) 1.000 1.000 1.000 1.000 219.52 0.975 0.974 0.975 0.974 220.40 0.912 0.905 0.910 0.910 221.29 0.845 0.831 0.842 0.840 222.17 0.780 0.759 0.775 o. 772 223.06 o. 717 0.692 o. 711 0.708 224.03 0.654 0.624 0.647 0.643 225.11 0.590 0.556 0.581 0.577 226.22 0.530 0.493 0.520 0.516 227.33 0.477 0.437 0.466 0.462 228.44 0.429 0.387 0.417 0.414 229.56 0.385 0.343 0.374 0.370 230.68 0.346 0.303 0.334 0.331 231.79 0.311 0.268 0.299 0.297 232.90 0.278 0.237 0.268 0.265 234.01 0.249 0.208 0.239 0.237 235.12 0.222 0.182 0.212 0.212 236.24 0.197 0.158 0.188 0.188 237.35 0.174 0.136 0.166 0.167 238.46 0.152 0.115 0.145 0.147 239.53 0.132 0.0952 0.126 0.129 240.35 0.116 0.0795 0.111 0.116 240. 57_( 2) 0.110 0.0732 0.106 0.111 7611 e:1d/011689 TABLE 3-7 FAST NEUTRON EXPOSURE PARAMETERS AND SPECTRUM AVERAGED DOSIMETRY REACTION CROSS-SECTIONS AT THE CENTER OF.

A SURVEILLANCE CAPSULE LOCATED AT 290° Cycle 1-7 Cycle 8

(E > 1.0 MeV) 7.47 x 1010 6.29 x 1010 dpa/sec 1.08 x 10- 10 9.01 x 10- 11 a Cu-63 (n,a) 0.00102 0.00101 a Ti-46 (n,p) 0.0201 0.0199 a Fe-54 (n,p) 0.119 0.118 a Ni-58 (n,p) 0.153 0.152 a U-238 (n,f) 0.426 0.425 a Np-237 (n,f) 2.20 2.20 Note
. The average

= o(E) ;(E) dE where a is the cross section cross section, a, 0  !

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Figure 3-1 Fast Neutron {E 1.0 MeV) as a Function of Azimuthal Angle at the Clad/Base Metal Interface

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218 .220 222 224 226 228 230 232 234 236 238 240 242 Radius (cm)

Figure 3-3 Relative Radial Variation of Fast Neutron Exposure within the Vessel Wall Cycles 1 - 7

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Figure 3-4 Relative Radial Variation of Fast Neutron Exposure within the Vessel Wall ~ Cycle 8 4.0 Summary In the preceding section, best estimate calculated fast neutron exposure rates in terms of both neutron flux (E > 1.0 MeV) and atom displacement rates {dpa/sec) have been provided for the Palisades reactor pressure vessel. The calculated values applicable to operation during cycles 1 - 7 have been used to establish the current exposure of the vessel at several key locations along the clad/base metal interface. These evalLlations may be summarized as follows:

End of Cycle 7

¢ (E > l.0 2MeV) dpa (n/cm 00 9.49 x 1018 .0142 16.44° 1.26 x 1019 .0187 33° 9.18 x 1018 .0136 45° 5.92 x 1018 .00894 These exposure values are based on the total power generation through cycle 7 of 2597.2 effective full power days at 2530 MWT. The modeling accuracy depends on both .as-built dimensions and cycle operating characteristics such as water temperature fluctuations and variations in axial or radial core power distributions.

The accuracy of the calculated neutron exposures is dependent on uncertainties in the data input to the calculation and on approximations made to solve the neutron transport equations with a limited number of spatial points, angles, and energy groups. Uncertainties exist in the time-averaged core power distribution, the cross sections, and in the geometric modeling of the reactor.

The transport methodology using the SAILOR cross-section library has been benchmarked against neutron dosimetry data obtained at the ORNL PCA facility. (lO) Extensive comparisons of analytical predictions with measurements from power reactor surveillance capsules and reactor cavity 0165l:EPL-013189

dosimetry programs have also been made. When plant specific core power distributions are employed in the analyses, experience has shown that fluence predictions are within +/-15% of measured values at surveillance capsule locations. Calculations applicable *to reactor cavity locations tend to be biased low by approximately 20% depending on the thickness of the pressure vessel.

Work is continuing to better understand the importance of each of the contributors listed above to the 15% uncertainty and to the cavity bias.

The cavity result is especially affected by the iron inelastic cross section (for the penetration through the vessel steel) and by the 2-dimensional modeling of the cavity region. Further benchmark measurements at the NESOIP facility may provide additional definition of these effects.Cll) 7048q/js

' i References

1. Soltesz, R. G. et al., *Nuclear Rocket Shielding Methods, Modification, Updating, and Input Data Preparation - Volume 5 - Two-Dimensional Discrete Ordinates Transport Technique", WANL-PR-(LL)-034, August 1970 ..
2. M. K. Kunka and C. A. Cheney, Analysis of Capsules T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program, WCAP-10637, September 1984.
3. Telecopy, John Ho to Stan Anderson, February 8, 1988.
4. G. L. Simmons and R..W. Roussin, SAILOR-Coupled. Self-Shielded. 47 Neutron. 20 Gamma Ray, P3. Cross Section Library for Light Water Reactors, RSIC-DLC-76, March 8, 1983, ORNL Radiation Shielding Information Center
5. R. W. Roussin, BUGLE Coupled. 47 Neutron, 20 Gamma Ray, P3 Cross-Section Library for LWR Shielding Calculations, RSIC-D_LC-75, June 1980, ORNL Radiation Shielding Information Center.
6. Letter~ J. C.. Ho to S. L. Anderson, "DOT Neutron Source Database for Palisades PTS Evaluation", February 4, 1988.
7. Letter, J. C. Ho to S. L. Anderson, "DOT Neutron Source Database for Palisades PTS Evaluation", February 15, 1988.
8. Letter, J. C. Ho to E. P. Lippincott, "DOT Calculations with Revised Neutron Source for Cycle 811 , December 19, 1988.
9. ASTM Designation E 593-79, Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa) 11

, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1987.

10. WCAP-11428, "Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology - PCA Evaluations", S. L. Anderson and K. C. Trari, to be published.
11. J. Butler, et al., "Rev1ew of the NESTOR Shielding and Dosimetry Improvement Program NESDIP", Proceedings of the 6th ASTM-Euratom Symposium on Reactor Dosimetry, ASTM STP-1001, to be published.

0165I:EPL-013189

7.2 CPCo Fluence Analysis Program 7.2.1 DOT Methodology In-house efforts have been established to perform the flux/fluence calculations utilizing the DOT4.3 discrete ordinates transport codes. In 1987, three engineers received work-study training from Combustion Engineering. The scope of the training included DOT4.3 code usage, model development and result evalua-tions. In-house computer codes have been developed for the pre and post processing of data for the DOT code. Model development has been performed using plant-specific operating data and the reactor design drawings for the geometry.

Presently, the in-house DOT geometry model consists of a 1/8 core configuration in the two dimensional (R-8) representation. This model consists of different regions: fuel, core shroud, bypass and inlet water, core support barrel, vessel with stainless steel cladding and three surveillance capsules - one at the core support barrel and two at the vessel wall. There are 89 radial and 83 azimuthal intervals in the DOT model. For Cycle 8 calculations, the shielded stainless steel assemblies would be included in the DOT geometrical model.

Ctoss-sections for the Palis~des core and materials are calculated using the GIP (Group-organized cross se~tion input program) code. One of the features of the GIP code is that it accepts nuclide organized microscopic cross section data either from the card image or from a data library (eg, SAILOR or CASK-81).

Macroscopic cross sections of mixtures as required by the DOT model can be prepared by the use of the isotopic densities. Isotopic densities for the plant specific core and the structure materials have been computed. AP 3

Legendre expansion for scattering is used for the material cross section calculations.

Fuel vendor supplied pin power distributions are derived from discrete PDQ model calculations. The assembly-wise radial power distributions are appro-priately adjusted using the existing data from the Palisades incore monitoring system (INCA) *. Axial power information. is obtained from the INCA or the XTG (nodal simulator) models.

MI0887-0055A-OP03

A neutron energy group spectrum corresponding to the SAILOR library has been developed. Presently, attempts are being made to include the effect of fuel depletion on the core neutron source. The contribution of the individual fissile isotopes to the core neutron source changes would be included. Since the fission spectra and effective neutron yield differ for the various fissile isotopes, the core neutron source and vessel flux will change with the fuel depletion.

Other features of the in-house DOT transport calculation methodology are very similar to that as used in the industry. It is intended that in-house methodology will be utilized for the Cycle 9 and beyond analysis due in mid-1989 and on a ~ycle-by-cycle basis in the future to monitor vessel fluence levels.

7.2.2 Supplemental Dosimetry Program Palisades has established a program for supplemental vessel dosimetry. This,.

program serves two main purposes:

  • l) it provides plant specific measured flux/fluence data which can be used for benchmarking the vessel flux/fluence calculations; and, 2) it provides cycle-specific results which help better control fuel management schemes.

An ex-vessel dosimetry program was developed by Westinghouse with installation of the dosimeters completed during the EOC7 refueling outage. The dosimeters are sensitive to a neutron energy spectrum over a wide range from thermal energy to 8 MeV. The radiometric monitors include cadmium-shielded foils of iron, nickel, copper, titanium, niobium and cobalt-aluminum. Cadmium-shielded fast fission detectors include U-238 in paired uranium detectors (PUD) and Np-237 in vanadium encapsulated neptunium oxide detectors. Bare iron and cobalt monitors are also included.

Gradient chains are of stainless steel which has iron, nickel, and cobalt materials and serve also as continuous dosimeters. The dosimeters are installed outside the vessel wall just beyond the reactor vessel wall insula-tion (Figure 7 .1). At seven locations only gradient chains are installed: 30°,

90°, 150°, 210°, 260°, 330°and 340°. At another five locations both .gradient MI0887-0055A-OP03

  • .( ... :.~. ' ,., ~
  • chains and discrete dosimeter capsules are installed at 270°, 280°, 290°, 300°,

and 315° at an elevation corresponding to the core mid-plane.

the bottom of the core.

At the 270° and 290° locations, dosimeters are also installed at the height corresponding to This dosimetry will provide a detailed azimuthal and axial mapping of the reactor vessel flux.

It is intended to change out these dosimeters each cycle to provide cycle-specif ic measured fluxes outside the vessel wall.

In addition to the ex-vessel dosimetry program, Combustion Engineering was contracted to fabricate and install a replacement in-vessel dosimetry capsule to be inserted into the W-290 capsule holder vacated at the EOC 5. This in-vessel capsule has three sets of dosimeters corresponding to the top, mid and bottom of the core plane. Type and neutron energy range of .these dosi-meters are very similar to the ex-vessel dosimeters. Installation attempts were not successful during the EOC 7 refueling outage. However, this capsule will be installed during the next refueling outage. When installed, this

  • capsule will provide an excellent through-wall correlation with the ex-vessel dosimetry.

MI0887-0055A-OP03

Figure 7.1 Ex-Vessel Dosimeter Locations 180° 90°

() Location of Gradient Chains CJ Location of Core Mid-9lane Dosimeters 0 Location of Core Mid-plane and Bottom of Core Dosimeters H10887-0055A-OPOJ

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