ML18053A677
| ML18053A677 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 11/30/1988 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML18053A676 | List: |
| References | |
| NUDOCS 8812050233 | |
| Download: ML18053A677 (62) | |
Text
ATTACHMENT Consumers Power Company Palisades Plant Docket 50-255 (PRELIMINARY)
ANALYSIS OF THE REACTOR PRESSURE VESSEL FAST FLUENCE FOR THE PALISADES NUCLEAR PLANT THROUGH END OF CYCLE 8 November 1988 Performed by the Reactor Engineering Department Palisades Nuclear Plant Consumers Power Company 61 Pages 0Cl188-0221-NL04
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TA.BLE OF CONTENTS SECTION TITLE PAGE
1.0 INTRODUCTION
1 2.0
SUMMARY
3 3.0 METHODOLOGY 7 3.1 OVERVIEW 7 3.2 FUEL MANAGEMENT 7 3.3 GEOMETRY 8 3.4 MATERIAL CROSS SECTIONS 8 3.5 NEUTRON SOURCE 8
. 3.6 NEUTRON TRANSPORT ANALYSIS 9 3.7 FLUENCE LIMITS 9 3.8 FLUENCE EXTRAPOLATION 10 4.0 RESULTS 20 4.1 FLUENCE DISTRIBUTION 20 4.2 VESSEL LIFETIME 20 4.3 CALCULATION UNCERTAINTY 20 5.0 DISCUSSION 25 5.1 IMPACT OF RESULTS 25 5.2 FUTURE ACTION 25
6.0 REFERENCES
27 1.0 APPENDICES 28 7.1 WESTINGHOUSE FLUENCE ANALYSIS REPORT 1.2 CPCO FLUENCE ANALYSIS PROGRAM MI0887-0055A-OP03 i
LIST OF TABLE~
TABLE TITLE PAGE 2.1 Sunmary of Reactor Vessel Fluence Limits 5 2.2 Fast Neutron Flux Reduction Achieved with Cycle 6 8 Loading Pattern 3.1 Target Fluence Limits for Pressure Vessel Welds 11 and Base Metal Based on 10CFRS0.61 3.2 . Target Fluence Limits for Pressure Vessel 12 Welds and Base Metal Based on Regulatory Guide 1.99, Rev 2.
4.1 Inner Vessel Wall Flux and Fluence (E>l.O MeV) 21 4.2 Sunmary of Fluence Limit Violation Dates 22 with Current Fuel Management to End of Life 4.3 Sunmary of Fluence Limit Violation Dates 23 with Cycle 8 Fluence Reduction to End of Lif~
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,,J LIST OF FIGURES FIGURE TITLE PAGE 3.1 Cycle 8 Peripheral Loading Pattern 13 3.2 Palisades Core Geometry Cycles 1 - 7 14 3.3 Palisades Core Geometry Cycle 8 15 3.4 Plan View of Palisades Wall Capsules 16 3.5 Assembly Radial Power Distribution for Cycles 1 - 7 3.6 Assembly Radial Power Distribution for Cycle 8 18 (Preliminary) 3.7 Assembly Radial Power Distribution for Cycle 8 19 (As-Loaded) 4.1 Azimuthal Reactor Vessel Fast Fluence Distribution 24 7.1 Ex-Vessel Dosimeter Locations MI0887-0055A-OP03 iii
1
1.0 INTRODUCTION
The final rule concerning fracture toughness requirements for protection against pressurized thermal shock events was published in July, 1985 as 10CFR50.61. The rule required a submittal of projected values of the reference temperature for pressurized thermal shock (RTpTs) for reactor vessel beltline materials at the vessel inner surface as calculated using the method specified in 10CFR50.61, paragraph b.2. Consumers Power's initial response to the rule
[l] indicated that no concern existed for Palisades as the projected RTPTS values were at or within the given screening criteria.
In a NRC response to the initial submittal (2], Consumers was requested to provide .additional information and was advised that the initial conclusion that the screening criteria were met was incorrect. An incorrect margin term had been incorporated into the calculation of the base metal RTpTs; use of the correct term caused the screening criteria of 270°F to be exceeded. Consumers' response [3] provided the requested information and also outlined a fluence reduction program to be initiated with the next fuel cycle (Cycle 8). In addition, the response conmitted to re-evaluating the current accumulated pressure vessel fast fluence (E>l.O MeV) and providing a report to the NRC describing the expected results from implementation of a fluence reduction program for Cycle 8 and beyond. By letter [4], the NRC accepted Consumers' proposed milestone schedule.
Prior to submittal of Consumers' fluence reduction program report, discussions with the NRC in late 1987 indicated that the reference temperature (RT) corre-lation in 10CFR50.61 would probably be changed to the correlation used in the draft Regulatory Guide 1.99, Rev 2. It was recommended that Consumers examine the impact of this change on the fluence reduction program under development; a new schedule for submitting the program report ~as submitted to the NRC in Sep-tember, 1987 [5]. Evaluation of the proposed RT correlation change revealed that the limiting beltline material changed to the vessel axial welds, with more restrictive fluence levels corresponding to the PTS screening criteria.
Based on the results of the evaluation, Consumers elected to redesign the planned Cycle 8 loading pattern to incorporate thrice-burned fuel assemblies MI0887-0055A-OP03
I with stainless steel shielding rods located near the axial weld locations to maximize the attainable fluence reduction.via fuel management. A revised milestone schedule for submitting the fluence reduction report was provided to the NRC [6]. The information contained herein is intended to address the fluence evaluation and reduction program as developed for Cycle 8.
The data base utilized in this study consisted of power distribution informa-tion developed from core monitoring system data and fuel vendor-supplied core simulator models, and plant-specific material compositions. The modeling of the vessel and fluence analysis were performed by Westinghouse (Section 7.1 of this report), using the DOTIIIW discrete ordinates transport code and the SAILOR and BUGLE-BO cross section libraries.
Results presented in this report address the accumulated vessel fluence through the end of Cycle 7 as well as the flux reduction obtained with the Cycle 8 low-leakage loading pattern. Vessel fluence limits based on the 10CFRS0.61 PTS screening criteria and both the 10CFRS0.61 and Regulatory Guide 1.99, Rev 2 RT correlations are calculated based on the vessel material chemistries. Vessel lifetimes are calculated relative to the fluence limits assuming both standard and flux-reduction fuel management. In addition, appendices are included that provide the_ complete Westinghouse fluence analysis report and the status of Consumers' in-house fluence analysis program.
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\
3 2.0
SUMMARY
Pressure vessel fluence limits based on the PTS screening criteria of 10CFRS0.61 were calculated using the reference temperature (RT) correlations of both 10CFRS0.61 and Regulatory Guide 1.99, Rev 2 using the vessel chemistries provided in Reference 3. The results are summarized in Table 2.1 and show the dramatic reduction in the vessel weld fluence limits with the use of the Regulatory Guide 1.99 RT correlation. The axial welds become the limiting material as opposed to the base metal which was the limiting material under the 10CFRS0.61 RT correlation.
A loading pattern was designed for Cycle 8 that provided substantial flux re-duction at the axial weld locations. The associated flux reductions for the primary vessel materials are shown in Table 2.2. Fast flux (E>l.O Mev) reduc-tions of approximately 50% were obtained at the axial weld locations, with modest reductions of about 12% at the circumferential and base metal peak locations. The Cycle 8 loading pattern used in this analysis was not the final loading pattern used to load the reactor; the final loading pattern power distributions indicate slightly lower power levels in the shielding assemblies near the axial weld locations with slightly higher powers in the other peri-pheral assemblies. The minor differences between the two loading patterns are not expected to drastically alter the final results.
Vessel lifetime based on when the PTS screening criteria are met were deter-mined for fuel management schemes with and without flux reduction for Cycle 8 and beyond. The fluence limits were based on the more restrictive Regulatory Guide 1.99, Rev 2 RT correlation and assumed that Cycle 8 concluded in March, 1990, with operation beyond end of Cycle 8 at a capacity factor of 75%. W1th no flux reduction utilized, the PTS screening criteria would be exceeded at the axial welds in approx~mately June, 1995; with flux reduction incorporated in Cycle 8 and beyond, the PTS limit would be exceeded at the axial welds again, but not until about Septe~ber, 2001. Both predicted dates are far short of the assumed nominal plant operating license expiration date of December, 2011 *
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)
4 While the flux ,reduction obtained in Cycle 8 substantially reduced the axial weld flux levels, the reduction is insufficient to remain within the PTS screening criteria through the minimum plant life (nominal end of operating license). Further flux reduction will be possible through more aggressive low-leakage fuel management in Cycle 9 and beyond, and possibly through reanalysis of Cycle 8 using the final, refined loading pattern model, but the incremental improvement believed possible will probably still be insufficient.
In order to allow plant operation at least until the nominal license expiration date, additional PTS-addressing measures will have to be implemented (eg, Regula-tory Guide 1.154 analysis, vessel ann~aling, etc).
Results of further flux reduction obtained with more aggressive low-leakage fuel management in Cycle 9 and beyond, including reanalysis utilizing the final Cycle 8 loading pattern model, will be provided in a report to the NRC in 1989
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5 TABLE 2.1
SUMMARY
OF REACTOR VESSEL FLUENCE LIMITS PTS Screening Fluence Limit (1Ql9n/cm2)
Material Criterion (°F) 10CFR50.61 RG 1.99, R/2 Axial Weld 270 . 7 .971 1.634 Circumferential Weld 300 11.984 3.495 Base Metal 270 4.134 6.046 MI0887-0055A-OP03
6 TABLE 2.2 FAST NEUTRON FLUX REDUCTION ACHIEVED WITH CYCLE 8 LOADING PATTERN Neutron Flux (n/cm 2 sec, E>l.O MeV)
Material Cycles 1 ~ 7 Cycle 8 Flux Reduction (%)
Axial Weld 4.31 X 1010 2.02 x 1010 -53.1 30° 4.28 x 1010 2.21 x 1010 -48.4 Circumferential Weld 5.86 x 1010 5.15,x 1010 -12.1 Base Metal 5.86 x 1010 5.15 x 1010 -12.1 MI0887-0055A-OP03
7 3.0 METHOIX>LOGY 3.1 Overview The pressure vessel fast neutron fluence levels (E>l.O MeV) were calculated utilizing available historical and predictive fuel cycle information. Flux calculations were performed by Westinghouse (See Section 7.1).
The primary analytical model was based on a two dimensional (R,e) discrete ordinates code IX>TIIIW representation of the Palisades reactor vessel config-uration. The representation includes a model of the core/vessel geometry, the neutron source djstribution, and nuclear interactions as represented by cross section data. Two calculations were performed. The first calculation modeled a configura~ion representative of an average over cycles 1 through 7. The second calculation modeled a Cycle 8 configuration having thrice-burned fuel assemblies with stainless steel shielding rods near the axial weld locations.
3.2 Fuel Management Palisades followed a standard OUT-IN fueling scheme through Cycle 7. In this scheme, only fresh fuel was placed around the core periphery. This approach results in the maximum overall core neutron leakage and fast flux to the reactor vessel, but minimizes power peaking and generally provides the greatest thermal*. margin.
With the issuance of Regulatory Guide 1.99, Rev 2, it was found that the axial welds would be responsible for limiting the life of the Palisades reactor vessel. It was decided to alter the fuel management strategy to distribute the power away from these critical weld locations for Cycle 8 operation. A low leakage loading pattern was adopted to improve the neutron economy and to reduce the fluence levels at the axial welds. Since the reload fuel for Cycle 8 had already been built, the flexibility to alter the fuel management was limited due to the predetermined new fuel assembly enrichment and quantity. A total of 16 thrice-burned stainless steel shielded assemblies were installed at the MI0887-0055A-OP03
8 core periphery. In addition, 8 twice burned assemblies were placed on the core periphery. The remaining 24 peripheral locations were filled with fresh fuel assemblies (Figure 3.1). With this arrangement, it is anticipated that the reduced power in the peripheral assemblies would reduce the primary source of fast neutrons reaching the reactor vessel axial welds.
3.3 Geometry The Palisades reactor exhibits one-eighth (1/8) core synmetry, thus only a zero to 45 degree sector has been included in the IX>T models (Figures 3.2 and 3.3).
Figure 3.2 is typical for Cycles 1 - 7 geometry. Figure 3.3 represents the core modification for Cycle 8, with the replacement of fuel by stainless steel rods in four outer rows of two assemblies. In these models a surveillance capsule at the 20° azimuthal location is also shown. This capsule corresponds to the W-290 capsule (Figure 3.4), which was removed at the end of Cycle 5 and analyzed for measured flux values (7).
3.4 Material Cross Sections The DOT model analysis employed a P3 expansiqn of the scattering cross sec-tions. The microscopic cross sections used in the analysis were obtained from the SAiLOR and BUGLE-80 cross section libraries. Macroscopic cross sections were calculated for each region in the model. Plant specific material composi-tions and the corresponding atomic densities were used for this analysis.
3.5 Neutron Source The neutron source inputs to the IX>T calculations were determined from the core power distributions (8). Detailed pin-by-pin power distributions for the outer assemblies was used. Normalized radial assembly power distributions based on assembly energy generation are shown in Figures 3.5 and 3.6 for Cycles 1 - 7 and Cycle 8, respectively. Figure 3.7 shows the normalized radial power distribution for the actual, as-loaded Cycle 8 loading pattern *
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9 3.6 Neutron Transport Analysis The spatial distribution of neutron flux in the reactor was calculated using the DOT computer code. DOT program solves the Boltzman transport equation in two-dimensional geometry using the method of discrete ordinates. Third order scattering ~P3) and Sa angular quadratures were used. Fluence levels were obtained by multiplying the flux by the effective full power seconds at 2530 MWT.
3.7 Fluence Limits Target fluence limits for pressure vessel welds and base metals are calculated using the 10CFRS0.61 correlation for RTPTS and the vessel material PTS screening criteria. The reference temperature correlation is given as:
RTpTs =I + M +(-10 + 470Cu + 350CuNi)f0.27 Where RTprs is the adjusted reference temperature for pressurized thermal shock considerations (°F)
I is the initial reference temperature (°F)
M is the margin term (°F)
Cu, Ni are the copper and nickel content (in weight percent), respectively f is the accumulated fluence (E>lMeV) in units of 10l9n/cm2 The corresponding fluence limits are determined by solving the RT correlation for the fluence value. The target limits are shown in Table 3.1.
Target fluence limits for pressure vessel welds arid base metals are also calculated using Regulatory Guide 1.99, Rev 2 reference temperature correlation and the 10CFR50.61 PTS screening criteria.
MI0887-0055A-OP03
10 The* adjusted reference temperature for each material in the belt line is given as:
RTNDT =I + H +(CF*f(0.28-0.10 log f))
I, Hand f have the same meaning as above. CF, the chemistry factor, depends on the contents of copper and nickel in the belt line materials. This factor is provided in Regulatory Guide 1.99, Rev 2. The corresponding fluence limits are determined by solving the RT correlation for the fluence value and are shown in Table 3.2.
3.8 Fluence Extrapolation Based upon the Cycles 1 - 7 averaged fuel management scheme, the estimated end of plant life dates are calculated relative to the fluence limits of 10CFRS0.61 and Regulatory Guide 1.99, Rev 2 for the axial welds, circumferential weld and base metal. Similar calculations are performed for the Cycle 8 fuel management scheme. It is assumed that the Palisades Plant will be operating at 75%
- capacity factor after EOC8, which is currently scheduled to end in March, 1990.
The nominal projected end of plant life is estimated by assuming that the plant would operate 40 years beyond initial criticality under its existing provi-sional license to December, 2011.
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11 TABLE 3.1 TARGET FLUENCE LIMITS FOR PRESSURE VESSEL WELDS AND BASE METAL BASED ON 10CFR50.61 Chemistry 10CFRS0.61 Parameters Fluence Screening Limit Material Cu(w/o) Ni(w/o) I(°F) M(°F) Criterion(°F) (10l9n/cm2)
Axial Weld .19 1.10 -56 59 270 7 .971 Circumferential Weld .20 .97 -56 59 300 11.984 Base Metal .25 .54 - 5 48 270 4.134 MI0887-0055A-OP03
12 TABLE 3.2 TARGET FLUENCE LIMITS FOR PRESSURE VESSEL WELDS AND BASE METAL BASED ON REGULATORY GUIDE 1.99, REV 2 Chemistry Reg Guide 1.99 Parameters Fluence Screening Limit Material Cu(w/o) Ni(w/o) I(°F) M(°F) Criterion(°F) (10l9n/cm2)
Axial Weld .19 1.10 -56 66 270 1.634 Circumferential Weld .20 .97 -56 66 300 3.495 Base Metal .25 .54 - 5 34 270 6.046 Chemistry Factors: Axial Weld - 229.00°F Circumferential Weld - 218.65°F Base Metal - 167.60°F MI0887-0055A-OP03
13 Figure 3.1 Cycle 8 Peripheral Loading Pattern Stainless Steel Shield Rods Number of Accumulated Cycles 150°
- 0 900_-1.....~Mo----;.....- -..............__........~i--..........- - - -....----+--+--3.. _.,,..~...~- 270° 3
0 3 3 0 Axial Weld
- Azimuthal Location MI0887-0055A-OP03
14 Figure 3.2 Palisades Core Geometry Cycles l - 7 Ill
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MI0887-0055A-OP03
15 Figure 3.3 Palisades Core Geometry Cycle 8 IH HI Ill Ill 111 H
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1 .....~~--~--~-+~~---+-------""+----.........._~~.-.-+--------+-~------+-------1---~~-+--..____
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16 Figure 3.4 Plan View of Palisades Wall Capsules Acmler118d - - - - - - - -
A-to
w1u W-10 Aesior V11Mt - - - - ~-------wau W-100
~-....~-.,..~~~~----w111 W-110 w**-----
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~m MI0887-0055A-OP03
17 Figure 3.5 Assembly Radial Power Distribution for Cycles l - 7 o*
'I I
0.892 0.874 0.695 1.138 1. 14 7 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 1.109 1.122 1.116 1.127 1.142 1.103 1 .015 1.171 0.938 1.175 1.108 1. 144 1.005 1.141 1.069 0.972 1.118 1.047 1.074 1.126 0.937 0.96S 0.892 Norma11zed Power 1.138 Axial Peaking Factor for Outer Assemblies MI0887-0055A-OP03
18 Figure 3.6 Assembly Radial Power Distribution for Cycle 8 (Preliminary) o*
I I
0.356 0.756: 0.641 1.084 1. 135 ' 1. 145 I
- 1. 101 1. l 061 1. 133 0.576 0.221
- 1. 104 1. 108 : 1. 139 1.114 l.12Q 1 .206 1.259 1. 008 1. 165 .0.907 1.098 1. 103 1. 103 1
- 155 1. 164 1.219 1. 043 1.299 1.076 1.239 1.115 1.166 1.009 1.286 1.005 1 .133 .
1.200 0.997 1. 100 1.166 1. 128
. . j 0.931 0.356 Normalized Power 1.084
- Axial Peaking Factor for Outer Assemblies MI0887-0055A-OP03
19 E'igure 3.7 Assembly Radial Power Distribution for Cycle 8 (Aa Loaded) o*
I I
0.271 0.812 0.656 l.139 l.168 l.087 0.621 0.156 l.198 1.193 l.003 l.225 0.932 l.209 l.034 1.260 l.033 l.214 l.021 l.2S9 l.021 l.126 1.207 1.013 l.118 l.193 1.154 0.966 MIOll7-0055A-OP0l
20 4.0 RESULTS 4.1 Fluence Distribution The azimuthal distribution of inner vessel wall fluence is given in Figure 4.1.
For cycles 1 through 8, the maximum peak flux/fluence occurs at the nominal 16° location on the vessel wall. Fluence levels at the two axial weld locations 0° and 30° are marked by vertical lines. Fluence limits corresponding to axial welds, circumferential welds and base metal are shown as horizontal lines.
These limits are calculated per Reg Guide 1.99, Rev 2. Table 4.1 provides the flux/fluence levels for 0°, 16° and 30° azimuthal locations. In this table, other parameters used for the calculations of the fluence from the flux (Sec 7.1) are also provided.
4.2 Vessel Lifetime Table 4.2 provides the fluence limit violation dates assuming the standard fuel management scheme, typical of cycles 1 through 7, for Cycle 8 and beyond.
Calculated dates are given both for the 10CFRS0.61 PTS screening criteria and the Reg Guide 1.99, Rev 2 RT correlations (based upon the known vessel material chemistries). Table 4.3 is presented for the flux-reduction fuel management scheme assuming the Cycle 8 loading pattern, with thrice-burned fuel assemblies with stainless steel shielding rods at the selected core locations (Figure 3.1), for Cycle 8 and beyond.
4.3 Calculation Uncertainty Fuel loading patterns for Cycles 1 through 7 are very similar to each other.
Measured flux for the W-290 capsule at the EOC 5 is 6.73 X 10 10 n/cm 2 -Sec [8].
This compares well with the calculated flux at the center of W-290 capsule of 7.70 X 1010 n/cm 2 -Sec (Section 7.1, Table 3 - 7) for Cycles 1 - 7. The calcula-ted flux is about 15% higher than the measured value.
A number of factors contribute to the uncertainty in the projected peak fast fluence at the reactor vessel wall. These factors are due to the conversion of measured activity data to the flux, material composition and neutron cross sections, power distribution and cycle-by-cycle variation in the lead factors.
An uncertainty of +/-25% is estimated in the calculated vessel wall fluence.
MI0887-005SA-OP03
21 TABLE 4.1 INNER VESSEL WALL FLUX AND FLUENCE (E>l.O MeV)
Azimuthal CYCLES 1 - 7 CYCLE 8 CYCLES 1 - 8 Location Average Flux Total Fluence Average Flux Total Fluence (X 1010 n/cm 2 -sec} (X 10 19 n/cm 2 ) (X 1010 n/cm 2 -sec) (X 1019 n/cm2) oo Axial Weld 4.31 0.969 2.02 1.031 16° Base Metal 5.86 1.318 5.15 1.475 30° Axial Weld 4.28 0.962 2.21 1.030 NOTES: 1. The rated full power is 2530 MW.
- 2. Flux values are computed at 0.125°, 16.44° and 30° for the nominal angles of 0°, 16° and 30° respectively.
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TABLE 4.2
SUMMARY
OF FLUENCE LIMIT VIOLATION DATES WITH CURRENT FUEL MANAGEMENT TO END OF.LIFE Material Fluence Levels (1019 n/cm2) 10CFR50.61 Reg Guide 1.99 1 R/2 EOC7 (8/88) EOC 8 (3/90) Increment Time to Limit Date of Limit Time to Limit Date of Limit Axial Weld oo .969 1.101 .136/EFPY >50 yrs 5.23 yrs - 6/1995 30° .962 1.093 .135/EFPY >50 yrs 5.35 yrs - 7/1995 Circumferential Weld Base Metal 1.318 1.318 1.498 1.498
.185/EFPY
.185/EFPY
>50 yrs - 14.39 yrs 8/2004 19 yrs - 3/2009 32.77 yrs - 12/2022 All times and dates based on 75% capacity factor after EOC8 (3/90).
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- 23 TABLE 4.3
SUMMARY
OF FLUENCE LIMIT VIOLATION DATES WITH CYCLE 8 FLUENCE REDUCTION TO END OF LIFE Material Fluence Levels (1019 n/cm2) 10CFR50.61 Reg Guide l.99t R/2 EOC7 (8/88) EOC 8 (3/90) Increment Time to Limit Date of Limit Time to Limit Date of Limit Axial Weld oo .969 1.031 .064/EFPY >50 yrs 12.56 yrs - 10/2002 30° .962 1.030 .070/EFPY >50 yrs 11.51 yrs - 9/2001 Circumferential Weld 1.318 1.475 .162/EFPY >50 yrs 16.63 yrs - 11/2006 Base Metal 1.318 1.475 .162/EFPY 21.88 yrs - 2/2012 37.63 yrs - 11/2027 All times and dates based on 75% capacity factor after EOC8 (3/90).
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FI 4. 1 AZIMUTHA~ EACTOR. VESSEL FAST FLUENCE DISTRIBUTION 6.50 BASE MEI AL REG GUIDE J.99 REV 2 LlHlT* 6.046 E+lQ s.oo s.so
>w 5.00
":i 0 4,50 A
w 4.00 CLRCUMFERENTLAL WELD N REG GUIDE J.99 REY 2 LlHlT* J.495 E+J9
- ii: J.50
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+
w 2.50
~
w 2.00
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1 .oo fOC 8 fO(' 7 a.so 0 AND 30 DEGREES ARE AXIAL WELD LOCATIONS o.oo 35 40 20 25 30 5 10 15 0
AZIMUTHAL ANGLE COEGREES)
25 5.0 DISCUSSION 5.1 Impact of Results The utilization of the Reg Guide 1.99, Rev 2 reference temperature (RT) correla-tion with the current 10CFR50.61 PTS screening criteria substantially reduces the allowable fluence at which the screening criteria are met. In addition to lowering the allowable fluence limits, the limiting material was relocated to the axial welds as opposed to the base metal.
The flux reduction obtained with the Cycle 8 loading pattern of about 50% at the axial welds and about 12% at the circumferential weld/base metal peak locations delays exceeding the PTS screening criteria to about September~ 2001, as opposed to June, 1995 if no additional flux reduction measures were taken.
The flux reduction is insufficient, however, to allow operation of the plant within the PTS screening criteria until the minimum expected plant life, corresponding to the expiration of the current operating license in December, 2011.
In order to maximize vessel lifetime, further measures must be taken in the areas of greater flux reduction, Reg Guide 1.154 analysis to properly define the real PTS screening criteria, and possible vessel annealing/shielding actions to reduce the accumulated vessel embrittlement.
5.2 Future Action The most straightforward method of reducing the vessel fast flux levels is reduction of the source itself, which has been initially addressed with the incorporation of low-leakage fuel management and stainless steel shield rods in Cycle 8. Two more refinements of the source reduction technique involve reanalyzing the Cycle 8 flux levels using the final loading pattern and utiliz-ing more aggressive low-leakage fuel management in Cycle 9. Reanalyzing Cycle 8 is not expected to provide a substantial improvement in the vessel fluxes; utilizing more burned fuel on the periphery for Cycle 9 could result in flux levels lower than Cycle 8 on the order of several percent at the axial weld locations, which would possibly add one to two years to the date at which the screening criteria are exceeded. Both of these issues will be addressed in the fluence program report to be provided to the NRC in mid-1989.
MI0887-0055A-OP03
26 Since the planned flux reduction measures do not appear to fully solve the vessel fluence issue relative to PTS, investigations have been initiated to possibly implement a Reg Guide 1.154 evaluation to more realistically define the vessel PTS temperature limits. The schedule for performing the evaluation will be more fully defined once the results from the 1989 fluence reduction report are available. Also; discussions have been held with NSSS vendors on the possibility of performing vessel annealing and critical material area shielding.
In order to benchmark vessel fluence calculations, an upgraded vessel dosimetry program has been initiated to supplement the existing surveillance capsule program. An ex-vessel dosimetry program was developed by Westinghouse and hardware install~tion occurred during the end of Cycle 7 refueling outage. The dosimetry installed will provide detailed azimuthal and axial mapping of the 270-360 degree vessel quadrant, with gradient chains installed in the other three quadrants to provide accurate axial and cross-quadrant mapping. It is intended to change out the dosimetry each cycle to provide accurate
- cycle-specific results. The dosimetry will provide measured data for use in vessel wall and vessel support fluence evaluations. In addition to the ex-vessel 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 following Cycle 5. Installation attempts were unsuccess-ful during the end of Cycle 7 refueling outage; the capsule will be installed during the next refueling, however. When installed, this capsule will provide an excellent-through-wall correlation with the ex-vessel dosimetry installed in the same quadrant.
In addition to implementing the supplemental dosimetry program, efforts have been underway to develop in-house capability to perform vessel fluence analyses using the industry-standard DOT methodology. Initial training in model development and analysis methodology were obtained from Combustion Engineering; in-house capabilities will allow convenient and economical evaluation of vessel fluence accumulation on a cycle-by-cycle basis without depending on outside vendors. Further details on the supplemental dosimetry program and the in-house calculational program development are included in Appendix 7.2.
MI0887-0055A-OP03
27
6.0 REFERENCES
- 1. Letter from KW Berry (CPCo) to NRC, "Response to 10 CFR 50.61 - Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," January 23, 1986
- 2. NRC letter from T V Wambach to KW Berry, "Pressurized Thermal Shock Rule (PTS), 10 CFR 50.61, Response for Palisades Plant," May 6, 1986
- 3. Letter from KW Berry to NRC, "Response to Request for Additional Informa-tion - Pressurized Thermal Shock (PTS) Rule 10 CFR 50.61, 11 August 7, 1986
- 4. NRC Letter from A C Thadani to K W Berry, "Palisades Plant - Fracture Toughness Requirements for Protection Against Press~rized Thermal Shock Events, 10 CFR 50.61," December 5, 1986
- 5. Letter from R W Smedley (CPCo) to NRC, "Compliance with Pressurized Thermal Shock Rule 10CFRS0.61 - Additional Information," September 29, 1987
- 6. Letter from R W Smedley (CPCo) to NRC "Revised Schedule for Compliance with Pressurized Thermal Shock Rule - Draft Regulatory Guide 1.99 Revision 2, and 10CFRS0.61," March 9, 1988
- 7. WCAP - 10637, Analysis of Capsules T-330 and W-290 from the Palisades Reactor Vessel Radiation Surveillance Program, M K Kunka and C A Cheney, September, 1984.
- 8. Engineering Analysis Package for PTS study, Reactor Engineering Department, Palisades Plant (1987-88).
MI0887-0055A-OP03
28 7.0 APPENDICES 7.1 Westinghouse Fluence Analysis Report Included in this section is the entire Westinghouse fluence analysis report performed in early 1988.
MI0887-0055A-OP03
ANALYSIS OF FAST NEUTRON EXPOSURE OF THE PALISADES REACTOR PRESSURE VESSEL E. P. Lippincott
- s. L. Anderson Westinghouse Electric Corporation Nuclear Technology Systems Division July 1988
- 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 ov*r fuel cycles 1 - 7 (standard fuel management) as well as for the cycle 8 design using shielded 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 MeY.
The exposure data generated from these transport calculations may be used in conjunction with damage trend curves to predict the degree of steel embrittlement in terms of RTPTS for current and future vessel operation. In addition to pressurized thermal shock assessments, the gradient 1nfonnation 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 deta;1, results of the analyses are presented, and a discussion of the adequacy of the analytical approach is provided.
7048Q/js
2.0 Method of Analysis
- The neutron flux distribution for the Palisades Reactor was calculated using the OOTIIIW Sn transport code. (l) T~o calculations were performed in R-0 geometry: the first calculation modeled a configuration representative of an average over cycles 1 to 7 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 20° 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 - 7 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 (47S) and water (531) .
- 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.851 in inner insulation and 1.541 in outer), aluminum (0.21S and 0.18S), 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.(l) Appropriate modifications to these densities were 111d1 for the water and insulation regions. Cross-sections were derived from the SAILORC 4> library except for N and Mo which were obtained from the BUGLE-ao< 5> library.
The neutron source input to the DOT calculations were determined from core power d1stribut1ons provided by Consumers Power.< 3*6* 7*8> The distr1but1ons were processed by the code SORCERY which transfonns the 7048q/js power ~alculated for each pin location into the neutron group by group
- source for the R-9 DOT geometry. Input to SORCERY included detailed 7
pin by pin power distributions for the outer assemblies( 6 * ) and assembly power distributions as shown in Figures 2-3 and 2-4 for cycle
- 7 and cycle 8, respectively.(B) 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 used. 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, compos1te 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.lS enrichment in feed assemblies. The resulting fission spectra {x(E)}, v, and r 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 o*, 11*, 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 l - 1 case but this was found to result in excessive computer usage so the cycle 8 case was d1v1ded 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) tlUCLIDI CARBON CONCRETE STAINLESS HOHOGENIZED WATER* AlR STEEL STEEL CORE (DENSITY lGH/CC)
(SA-1028) (TYPE 304)
H ANSI Standard 2.825'-2 6.679-2 B-10 T1pe 04@2.ll 2.692-5 6.649-6 c 9.84-4 2.581-S II 4.271-5 0 2.697-2 3.340-2 l.129-'.l AR l.560-4 I Si 7.589-4
~
I Ca Cr l.8H-2 l.608-S 1.391-3 .. 1.752-3
""'Ni Fe 8.181-2 5.807-2 8.574-3 2.205-5 3.871-5 Zr 5.448-l Ho 3.148-4 UilS 1.12-4 U2l8 6.03-l PU239 2.20-5 PU240 1.12-6
- By-pasa water i1 at 560.F and 2010 psia and inlet water is at 536.F and 2010 psia.
Appropriate density values should be used (or these regions.
OC0188-000~B-OP0l
- l TABLE 2-2 ENERGY GROUP STRUCTURE USED IN TRANSPORT ANALYSIS Lo'tiler Energy Relative Fission SQectra Group (MeV} Cycle 1-7 Cycle 8 1 14. 19* 4.4841 ( -5) 4.&932 (-5) 2 12.21 l. 8826 (-4) l. 9415 ( -4) 3 10.00 l . l 058 (-3) l . 1268 ( -3) 4 8. 61 2. 6187 (-3) 2.6574 (-3) s 7.41 5.8744 (-3) 5.9344 (-3) 6 6.07 l. 6720 (-2) , . 6827 ( -2) 7 4.97 3. 1855 (-2) 3.1976 (-2) 8 3.68 8.3549 (-2) 8.3794 (-2) g 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) l7 1.35 8.8042 (-2) 8.7953 (-2) 18 1.00 1. 1320 ( _,) 1.1296 (-1) 19 0.821 6.1965 (-2) 6.1777 (-2) 20 o. 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 o. 183 3.0971 (-2) 3.0939 (-2) 26 0.111 l.6332 (-2) 1 . 6322 ( -2)
- The upper energy of group l 1s 11. 33 ri.ev.
- 7048Q/js Figure 2-1 Palisades Geometry Cycles 1 - 7
.... ..... .,... ..... ..... ..... .,... . ...------+-.....
, ...______._,;1.------------+-----..M.+---.J..1"""""4----............- -......
, 1n.1 HOIUI ICIU
--+i------"""""4------...... . ... ---'
Figure 2-2 Palisades Geometry Cyc1e 8 IH 11'1 Ill IH Ill H
,.... 111.1 111.1 111.1 ..... .....
HOlilt IClll IU.I IPl.I ..... IH.I
' I Figure 2-3 Normalized Radial Power Distribution by Assembly for Cycles 1 - 7 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 ,
1.109 1. 122 1.116 1.127 1.142 ;
/ "
- 1. 103 1. 015 1.171 0.938 1.175 1.108 1. 144 1.005 1. 141 1.069 0.972 1.118 1.047 1.074 1.126 0.937 0.965 0.892 Nonna11zed Power 1.138 Axial Peaking Factor for Outer Assemblies 7048q/js Figure 2-4 Normalized Radial Power Distribution by Assembly for Cycles o*
I r
I I
0.356 o. 756, 0.641 l .084 1. 135 ' l . 145 I
- 1. 101
- 1. 104 1. 1aa I
- 1. 1061 l. 133
- 1. 139 0.576
- 1. 114 0.221 1.120 1.206 1.259 1. 008 1. 165 0.907 45*
I 1.098 1.103 1. 103 1.155 1. 164 I
/
/
1.219 1.043 l. 299 1.076 1.239 1.115 1. 166 l .009 1.286 1.005 1. 133 '
1.200 0.997 1.100 1., 66 1.128 0.931
~ Norma 1iied Power Axial Peaking
~ Factor for Outer Assemblies
- 7048q/js
' I 3.0 Results of Analysis Results of the neutron transport analysis of the Palisades reactor are sunJnarized graphically in Figures 3-1 through 3-4 and in tabular form in Tables 3-l 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 power level of 2530 MW. Again, due to sy!Tl'!letry 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 1s determined by the DOT code by multiplying the group dpa cross sections by the group fluxes at each location and sunning the contribution 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 31 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 1 - 7 and cycle 8, respectively. These graphical representations are representative of the peak location in the azimuthal, exposure rate d1str1but1ons (16.44*). In Tables 3-3 through 3-6 relative radial d1str1but1on data are given for several azimuthal locations in addition to the peak location illustrated 1n 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.
7048q/js 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 multiplicat1ve fashion to create R,0 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 averaged neutron dos1metry cross-sections are provided for the 20° surveillance capsule modeled in the analys1s. These data may be used in the evaluation of lead factors based on cycle specific analyses.
7048q/js TABLE 3-1 F~ST .NEUTRON FLUX ~E>l.O MEV~ AT THE CLAD/BASE METAL
- e (deg.)
- {n/cm 2 - sec)
Cycles 1-7 INTERFACE - AXIAL MAXIMUM Cycle 8 e
(deg.)
- (n/cm 2 - sec)
Cycles 1-7 Cycle 8 0.125 4.31 x 10 10 2.02 x io 10 22.75 4.89 x lo 10 3.93 x 10 10 0.500 4.31 x 10 10 2.01 x 10 10 23.25 4.79 x lo 10 3.77 x 10 10 1.25 4.31 x 10 10 2.02 x 1010 23. 75 4.68 x 10 10 3.60 x 10 10 2.25 4.35 x 10 10 2.10 )( 10 10 24.50 4.51 x 10 10 3.36 x 10 10 3.25 4.41 x 10 10 2.23 x 10 10 25.50 4.37 x 10 10 3. 08 )( 10 10 4.25 4. 48 x 10 10 2.40 )( 10 10 26.50 4.29 )( 1010 2.83 x 10 10 5.37 4.58 x 10 10 2.62 x 1010 27.SO 4.25 x 10 10 2.62 x 10 10 6.62 4.72 x 10 10 2.91 x 1010 28.50 4.25 x 10 10 2.43 x 10 10 7.62 4.85 )( 10 10 3.18 )( 10 10 29.44 4.27 x lOlO 2.28 x 10 10 8.50 4.98 x 10 10 3.43 )( 10 10 30.00 4.28 x lo 10 2.21 x 10 10 9.50 5.12 x lOlO 3.72 x 10 10 30.56 4.29 ll 1010 2.13 x 10 10 10.50 5.28 x 10 10 4.02 l( 10 10 31.50 4.30 ll io 10 2.02 ll io 10 11.50 5.42 x 10 10 4.30 x 1010 32.44 4.27 ll 10 10 1.92 x io 10 12.50 5.56 )( 10 10 4.56 )( lo 10 33.00 4.20 )( io 10 1.87 x 10 10 13.50 5.67 x 10 10 4.78 x lo 10 33.56 4.18 x 1010 1.81 x lOlO 14.50 5. 77 x io 10 4.96 x lOlO 34.50 4.08 x 1010 1. 74 x 1010 15.50 5.83 x 1010 5.09 x lOlO 35.50 3.95 )( io 10 1.68 )( 10 10 16.44 5.86 x io 10 5.15 ll 10 10 36.25 3.81 )( 1010 1. 65 )( 10 10 17.00 5.83 x 10 10 5.12 x 1010 36.75 3.73 )( io 10 1.63 x io 10 17.56 5.80 x 10 10 5.13 )( 1010 37.38 3.63 x 1010 1.60 x io 10 18.25 5.44 )( 10 10 4.78 l io 10 38.12 3.50 x 1010 1. 59 x lOlO 18.84 5.28 x io 10 4.63 x iolO 38.87 3.36 x lOlO 1. 58 )( 10 10 19.22 5.13 x 1010 4.47 x lOlO 39.62 3.21 )( 10 10 l. 58 )( lo 10 19.51 5.00 l 10 10 4.34 )( 1010 40.50 3.05 x 1010 1. 57 )( 10 10 20.01 4.86 )( 1010 4.18 x 1010 41.37 2.92 x 1010 l.57 x 10 10 20.50 4.81 ll 10 10 4.09 )( 1010 42.25 2.80 x lalO 1.56 x io 10 20.78 4.82 x. 1010 4.08 x lOlO 43.12 2.73 x lolO 1. 56 x 10 10 21.11 4.84 x 10 10 4*.06 )( 1010 43.88 2.68 x 10 10. 1. 56 )( 10 10 21.71 4.81 )( 1010 3.97 l( 10 10 44.63 2.65 x 10 10 1. 56 x 10 10 22.25 4.94 )( io 10 4.02 )( 1010 711te:1d/0331U TABLE 3-2 IRON ATOM DISPLACEMENT RATE {dea/sec~ AT THE CLAD/SASE
- e (deg.) Cycles 1-7 METAL INTERFACE - AXIAL MAXrMUM dpa/sec Cycle 8 a
(deg.) Cycles 1-7 dpa/sec Cyc1e 8 0.125 6.44 x 10- 11 3.05 x 10-ll 22.75 7.37 x 10-ll 5.91 x 10-ll 0.500 6.44 x 10-ll 3.04 x 10-ll 23.25 7.18 x 10- 11 5.65 x 10- 11 l.25 S.44 x 10-ll 3.06 x 10-ll 23.75 7. 00 )( 10-ll 5.39 x 10-ll 2.25 6 . 50 )( 10 - ll 3.17 x 10-ll 24.50 6.75 )( 10-ll 5.02 x 10- 11 3.25 6.58 x 10- 11 3.36 x 10-ll 25.50 6. 53 x 10-ll 4.61 x 10- 11 4.25 6.69 )( 10-ll 3.61 )( 10-ll 26.50 6.40 )( 10-ll 4.25 x 10- 11 5.37 6 *83 )( 10- 11 3.93 x 10- 11 27.50 6.34 x 10-ll 3.92 )( 10- 11 6.62 7 .03 x 10-ll 4.35 )( 10- 11 28.50 6.33 )( 10-u 3.65 ll 10-ll 7.62 7.22 x 10-ll 4.75 )( 10- 11 29.44 6.35 x 10- 11 3.43 )( 10-ll a.so 7.40 x 10-ll 5.10 )( 10-ll 30.00 6.36 )( 10- 11 3.32 x 10-ll 9.50 7 .62 )( 10-ll 5.s2 x 10- 11 30.56 6.37 )( io- 11 3.20 x 10- 11 10.50 7.83 )( 10-ll 5. 95 x 10-ll 31.50 S.37 x 10-ll 3.03 )( 10-ll
- 11. 50 8. 04 x 10-ll 6.37 lt 10-ll 32.44 6.32 x 10- 11 2.89 x lo- 11 12.50 8.23 x 10-ll 6. 74 J( 10-ll 33.00 6.22 x* 10-ll 2.82 x 10-ll 13.50 8. 40 )( 10-ll 7 .06 )( 10-ll 33.56 6.19 )( 10-ll 2.73 )( 10-ll 14.SO 8.54 )( 10-ll 7.32 x 10-ll 34.50 6.04 x 10-ll 2.62 x 10- 11 15.50 8.64 x 10-ll 7.51 x 10* 11 35.50 s.a4 x io* 11 2.53 x 10- 11 16.44 8. 10 x 10- 11 7. 62 x 10* 11_ 36.25 5.64 x 10-ll 2
- 48 )( l 0- ll 17.00 a. 68 x 10- 11 7 .61 x 10-ll 36. 75 5. 53 x 10-ll 2.46 x 10- 11 17.56 8.70 x 10-ll 1.ss x io* 11 37.38 s.39 x ia* 11 2.41 ll 10~ 11 18.25 8.23 x ia- 11 7.20 x io- 11 38.12 5.20 x 10-ll 2.39 )( 10-ll 18.84 8.04 x 10-ll 7.02 x 10* 11 38.87 5. 00 ll 10 - ll 2.38 x 10-ll 19.22 1.aa x 10* 11 6.82 x 10-ll 39.62 4.79 ll 10-11 2.38 x 10-ll 19.51 7.67 l 10*
11 6.63 x io* 11 40.50 4.56 x 10* 11 2.37 x 10-ll 20.01 7.47 x 10- 11 6.40 x 10-ll 41.37 4.38 x 10- 11 2.37 x 10- 11 20.50 7.39 ll 10-ll s.21 x 10- 11 42.25 4.20 x 10* 11 2.36 ll 10-ll 20.78 7.41 x 10-ll 6.25 x 10* 11 43.12 4.10 x 10- 11 2.36 )( 10-ll 21.11 7.42 x 10- 11 6.22 x 10* 11 43.88 4.03 x io* 11 2.36 x io* 11
- 21. 71 7.33 x 10-ll 6.04 x 10-ll 44.63 4.00 x 10- 11 2.36 x io- 11 22.25 7 .48 x 10-ll s.08 x 10- 11 71111*:1 a/0331 . .
TABLE 3-3 RELATIVE RADIAL DISTRIBUTION OF FAST NEUTRON FLUX
- .Radius
~E > 1.0 Mevi WITHIN THE PRESSURE VESSEL ~ALL CYCLES 1-7 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 22.4. 03 0.576 0.558 0.570 0.583 225.11 0.499 0.480 o.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 0.172 234.0l 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 240.61< 2> 0.0474 0.0408 0.0446 0.0537
{l~Reactor vessel inner radius {Clad-Base Metal Interface)
(Z)Reactor vessel outer radius
- ?111e:1d/0331 II
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 0.971 0.961 0.976 0.975 220.40 0.894 0.882 0.897 0.898 221.29 0.813 0.796 0.815 0.814 222.17 0.732 o. 712 0.733 0.732 223.06 0.657 0.634 0.657 0.656 224.03 0.581 0.556 0.579 0.578 225oll 0.506 0.477 0.502 0.501 226.22 0.437 0.407 0.433 0.431 227.33 0.376 0.346 0.372 0.370 228.44 0.323 0.294 0.318 0.317 229.56 0.277 0.249 0.272 0.271 230.68 0.238 0.211 0.232 0.231 231.79 0.203 0.178 0.198 0.197 232.90 0.174. 0.150 0.169 0.168 234.01 0.148 0.126 0.144 0.143 235.12 0.126 0.106 0.122 0.121 236.24 0.107 0.0885 0.103 0.103 237.35 0.0909 0.0734 0.0872 0.0870 238.46 0.0766 0.0602 0.0732 0.0734 239.53 0.0645 0.0487 0.0616 0.0623 240.35 0.0558 0.0402 0.0534 . 0.0546 240.61< 2> 0.0542 0.0383 0.0517 0.0530 (l)Reactor vessel inner radius (Clad-Base Metal Interface)
( 2)Reactor vessel outer radius 7111e:1d/033111
' ,t 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° 44.53° 219.07(l) 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 o. 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 O. lll 0.134 240.35 0.101 0.0867 0.0964 0.119 240.61< 2> 0.0970 0.0832 0.0931 0.116 (l)Reactor vessel inner radius (Clad-Base Metal Interface)
(Z)Reactor vessel outer radiu; 7111e:1dlm111
TABLE 3-5 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 0.980 0.978 0.977 0.977 220.40 0.917 0.909 0.913 0.912 221.29 0.850 0.834 0.845 0.842 222.17 0.785 0.762 o. 778 0.774 223.06 o. 722 0.695 0. 714 0.709 224.03 0.659 0.626 0.650 0.644 225.11 0.595 0.558 0.584 0.578 226.22 0.536 0.494 0.524 0.517 227.33 0.482 0.438 0.469 0.462
- 228.44 229.56 230.68 231.79 0.433 0.390 0.351 0.315 0.388 0.344 0.304 0.269 0.420 0.377 0.337 0.302 0.414 0.370 0.331 0.297 232.90 0.283 0.237 0.270 0.266 234.0l 0.254 0.208 0.241 0.237 235.12 0.226 0.182 0.215 0.212 236.24 0.201 0.158 0.191 0.189 237.35 0.178 0.136 0.168 0.168 238.46 0.156 0.115 0.147 0.148 239.53 0.136 0.0950 0.128 0.131 240.35 0.120 0.0792 . 0.114 0.118 240.67( 2) 0.117 0.0756 0.111 0.115
(!)Reactor vessel inner radius (Clad-Base Metal Interface)
(Z)Reactor vessel outer radius 7111*:1dl0331 ..
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 Me V) 7. 70 x 10 10 6.69 x 10 10 dpa/sec l.ll x io- 10 9.58 x 10- 11 a Cu-63 (n,a) 0.00102 0.00100 a Ti-46 (n,p) 0.0201 0.0197 a Fe-54 (n,p) 10- 24 cm 2 0.119 0.118
- a Ni-58 {n,p) l0- 24 cm2 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 cross section, a, 0 a J-o(E) ;(E) dE 0
wnere o is the cross section for eacn of the reactions above is defined as * ;(E) dE
- and is the neutron flux at the capsule center location.
1 1.0 MeV
- 7911e:1d/01Qql io, *~~~~~~~
11
.:*:*':X...:~ :.:-: ...~::--
c;c* '-~~
=
. .,_~~-~: .. -,_,
--*-<g:-_:~"-_: ,,_:
-~
1 8
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- Fast Neutron (E > 1.0 MeV) as a Function of Azimuthal Angle at the Clad/Base Metal Interface 8
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- Figure 3-2 Iron Atom Displacement Rate as a Function of Azimuthal Angle at the Clad/Base Metal Interface
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- 4. 0 Sunmar'j In the preceding section, best estimate calc~lated fast neutron exposure rates in terms of both neutron flux (E > 1.0 MeV) and iron atom displacement rates (dpa/sec) have been provided for the Palisades reactor pressure vessel. The calculated values applicable to operation during cycles 1 - 1 have been used to establish the current exposure of the vessel at several key locations along the clad/base metal interface.
These evaluations may be sunmarized as follows:
End of Cycle 7
- (E > 1.0 '4eV) dpa (n/cm2) oo 9.67 x 1018 0.0145 16.44° 1. 31 )( 101 g 0.0195 33* 9.42 x 1018 0.0140 45* 5.95 x l ol 8 0.00898 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 geometr1c modeling of the reactor.
The transport methodology using the SAILOR cross-section 11brary has been benchmarked against neutron dosimetry data obtained at the ORNL PCA fac111ty.(lO) Extensive comparisons of analytical predictions with measurements from power reactor surveillance capsules and reactor cavity 704Bq/js
,I dosimetry programs have also been made. When plant specific core power d1stribut1ons are employed in the analyses, experience has shown that fluence predict1ons are within ~lSS of measured values at surveillance capsule locations. Calculations applicable to reactor cavity locations tend to be biased low by approximately 20S depending on the thicKness of the pressure vessel.
work 1s cont1nu1ng to better understand the importance of each of the contributors 11sted above to the lSS 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 NESDIP facility may provide additional definition of these effects.(ll)
.\
7048q/js
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 Caosules T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program, WCAP-10637, Sept. 1984.
- 3. Telecopy, John Ho to Stan Anderson, Feb. 8, 1988.
- 4. G. L. S1nmons and R. w. Roussin, SAILOR-Coucled. Self-Shielded, 47 Neutron. 20 Gan1na Ray, P3. Cross Section Library for Light Water Reactors, RSlC-OLC-76, March 8, 1983, ORHL Radiation Shielding Information Center.
- 5. R. W. Roussin, BUGLE Coupled. 47 Neutron, 20 Gamna Ray, P3 Cross-Section Library for LWR Shielding Calculations, RSIC-OLC-75, June 1980, ORNL Radiation Shielding Information Center.
- 6. Letter, J. c. Ho to s. L. Anderson, *oar Neutron .Source Database for Palisades PTS Evaluation*, February 4, 198~.
- 7. Letter, J. C. Ho to s. L. Anderson, 'DOT Neutron Source Databas1 for Palisades PTS Evaluation*, February 15~ 1988.
- 8. Telecopy, John Ho to Stan Anderson, March 2, 1988.
- 9. ASTM Designation E693-79, Standard Practice for Character1ting Neutron Exposures 1n Ferrit1c Steels in Terms of Displacements per Atom (dpa)',
in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1987.
- 10. WCAP-11428, *eenchmark Testing of Westinghouse Neutron Transport Analysis Methodology - PCA Evaluations*, A. L. Anderson and K. C. Tran, to be published.
- 11. J. Butler, et al., *Review of the NESTOR Sh1eld1ng and Dosimetry Improvement Program NESDIP*, Proceedings of the &th ASTM-Euratom Symposium on Reactor Dosimetry, ASTM STP-1001, to be published.
7048Q/js 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-e) 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.
Cross-sections for the Palisades core and materials are calculated using the GIP (Group-organized cross section 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. A P3 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
~
( l ~.) t 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 cycle-by-cycle basis in the future to monitor vessel fluence levels.
1.2.2 Supplemental Dosimetry Program Palisades has. established a program for supplemental vessel dosimetry. This program serves two main purposes: 1) 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 five locations only gradient chains are installed: 30°,
90°, 150°, 210° and 330°. At another five locations both gradient chains and MI0887-0055A-OP03
discrete dosimeter capsules are installed at 270°, 280°, 290°, 310°, and 315° at an elevation corresponding to the core mid-plane. At the 270° and 290° locations, dosimeters are also installed at the height corresponding to the' bottom of the core. 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-specific 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
.. 1,.-
- Figure 7.1 Ex-Vessel Doti.meter Locations 180° 90°
() Location of Gradient Chains
[J Location of Core Mid-plane Dosimeters
() Location of Core Mid-plane and Bottom of Core Dosimeters MI0887-0055A-OP03