ML20203K412
| ML20203K412 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 02/28/1998 |
| From: | Kam F, Remec I OAK RIDGE NATIONAL LABORATORY |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| CON-FIN-W-6164 NUREG-CR-6453, ORNL-TM-13204, NUDOCS 9803050078 | |
| Download: ML20203K412 (58) | |
Text
- - _
- l NUREG/CR-6453 0RNL/TM-13204 H. B. Robinson-2 Pressure Vessel Benchmark II7Eu, ink.Kam Oak Ridge National Laboratory Prepared for U.S. Nuclear Regulatory Commission _. O j
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DISCLAIMER NOTICE This report was prepared as an account of work sponsored by an agency of the United States Govemment.
Neither the United Ststes Govemment norany agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product. or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.
ORNLfrM ?3204 i
Ho B. Robinson-2 Pressure Vessel Benchmark Manuscript Completed: October 1997 Date Published: February 1998 Prepared by
- 1. Remec, F. B. K. Kam Oak Ridge NationalLaboratory Managed by Lockheed Martin Energy Research Corp.
Oak Ridge NationalLaboratory Oak Ridge, TN 37831-6363 C. J. Fairbanks, NRC Project Manager Prepared for Division of Engineering Technology Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC Job Code W6164 pa %q g"%...../
_______m
ABSTRACT The H. B. Robinson Unit 2 Pressure Vessel Benchmark (HBR-2 benchmark) is described and analyzed in this report. Analysis of the HBR-2 benchmark can be used as panial fulfillment of the requirements for the qualification of the methodology for calculating neutron fluence in pressure vessels, as required by the U.S. Nuclear Regulatory Commission Regulatory Guide DG-1053, Calculational and Dosimetry Methodsfor Determining Pressure Vessel Neutron Fluence.
Section 1 of this repon describes the HBR-2 benchmark and provides all the dimensions, material compositions, and neutron source data necessary for the analysis. The measured quantities, to be compared with the calculated values, are the specific activities at the end of fuel cycle 9. The characteristic feature of the HBR-2 benchmark is that it provides measurements on both sides of the pressure vessel: in the surveillance capsule attached to the thermal shield and in the reactor cavity.
In Section 2, the analysis of the HBR-2 benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed with three multigroup libraries based on ENDF/B-VI: BUGLE-93, SAILOR-95 and BUGLE-96. The average ratio of the calculated-to-L measured specific activities (C/M) for the six dosimeters in the surveillance capsule was 0.90
- 0.04 l for all three libraries. The average C/Ms for the cavity dosimeters (without neptunium dosimeter)
I were 0.89
- 0.10,0.91
- 0.10, and 0.90
- 0.09 for the BUGLE-93, SAILOR-95 and BUGLE-96 libraries, respectively.
It is expected that the agreement of the calculations with the measurements, similar to the agreement obtained in this research, should typically be observed when the discrete-ordinates method and ENDF/B-VI libraries are used for the HBR-2 benchmark analysis.
iii NUREG/CR-6453
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l CONTENTS AB STRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii FIG URE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . .. . . . vii T AB LE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ix
' ACKNOWLEDGMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xi 1 BENCHMARK DEFINITION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I l .1 INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I
- 1. 2 DES C RIPTION . . . .. . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 l.3 CORE POWER DISTRIBUTION AND POWER HISTORY , . . . . . . . . . . . . . , , . 3 1.4 DO S IMETRY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
- 1. 5 REFERENC ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2 B ENCHMARK ANALYSI S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 2.1 METHODOLOGY . . . . . . . . . . . . . . . . . . . . . . .. .......... ......... 26
, 2.2 RESULTS AND DISCUSSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29
- 2. 3 REFEREN C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3
, 3 CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . ............ ............. 34 APPENDIX A COMPARISON OF APPROXIMATIONS FOR MODELING THE REACTION RATE VARIATIONS DUE TO CORE POWER REDISTRIBUTION
'r AND COMPARISON OF RESULTS OBTAINED WITH ENDF/B-IV AND ENDF/B-VI CROSS SECTIONS . . . . . . ....... ..........................35 APPENDIX A REFERENCES ........ ...., ........... .... . ....... 42 l
APPENDIX B CALCULATED NEUTRON SPECTRA AT THE DOSIhETRY LOCATIONS . . . . . . 43 e
i-y NUREG/CR-6453 2
7 4
d
FIGURES 1,1 Horizontal cross section of the HBR-2 reactor . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . 1,2 Schematic sketch of the axial geometry . . _, . . . . . . . . ._ . . . . . . . . . . . . . . . . . , . . . . . 13 1.3 Core baffle geometry . . . . . . _ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 1.4 - Sketch of the surveillance capsule mountirig on the thermal shiel.d . . . . . . . . . . . . . . . , ,15 1.5 The numbering of the fuel elements in the HBR-2 core . . . . . . , . , , . . . . . . . . . . . . . 16 1.6 - Content and format of the FILEl .DAT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 1,7 Content and format of the FILE 2.DAT , . . . . . . . . . . . . . . . . . . . . .. . , . . . . . . . . . . . 18 1.8 Content and format of the FILE 3.DAT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 1.9__ - Content and format of the FILE 4.DAT . . . . . . . . . . . . . . . . . . , . .. . . . . . . . , . . . . . . . . -
1 Content and format of the FILES.DAT , . , , , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 1.11 - Content and format of the FILE 6.DAT , . . . . . . . . . . . . . . . . . . . . . , . . . . .. . . . , , , . 22 1.12 Content and format of the FILE 7.DAT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . 23 -
1.13 Content and format of the FILE 8.DAT . . . . . . . . . . . . . . . . . , , , , . . . . . . . . . . . 24 1.14 Schematic drawing of the axial positions of the cavity dosimeters . . . , , . . ... ... 25.
B.l . Multigroup neutron spectrum, calculated with BUGLE 96 library, in the surveillance capsule . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 4 9 B.2 - - Comparison of multigroup neutron spectra, calculated with different cross-section libraries, in the surveillance capsule . . . . . . . . . . . , , . . . . . ., . . , , . . 49 B.3 Multigroup neutron spectrum, calculated with BUGLE-96 library, at the position of cavity dosimeters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50
~ B.4 Comparison of multigroup neutron spectra, calculated with different cross-section libraries, at the position of cavity dosimeters . , , , , , , . . . . . . . . . . . . . 50 ,
vii NUREG/CR-6453
TABLES 1.1 - Selected general data and dimensions of the H. B. Robinson Unit 2 . . . . . . . . . . . . . . . . . 6 1.2 Materials of the components and regions . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 1.3 Densities and chemical compositions of reactor component materials . . . . . . . . . . . . . . .- 10 1.4 Measured specific activities of the dosimeters from the surveillance capsule and
- from the cavity, at the end of cycle 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 2.1 Reaction rates calculated for the cycle-average power distribution and core power of 2300 MW (100% of nominal power), with different cross-section i libraries for transport calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . 29 -
2.2 Calculated specific activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 2.3 Ratios of calculated-to-measured (C/M) specific activities . . . . . , , . . . . . . . . . . . . . . . . 31 A.1 Ratios of calculated-to-measured (C/M) specific activities obtained with-
- different approximations for the time-6 pendent variations of reaction rates , . . . . . . . . 40
- A.2 Comparison of the C/M ratios of specific activities from the present analysis with the values from the previous analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 41 B.1 Calculated multigroup neutron fluxes in the surveillance capsule . . . . . . . . . . . . . . . . . 45 B.2 Calculated multigroup neutron fluxes at the location of cavity dosimeters . . . . . . . . . . . 47
)
ACKNOWLEDGMENTS The authors wish to thank C. S. Hinnant, R. M. Krich, and W. K. Cantrell of Carolina Power and Light Company, who provided the core depletion calculations for cycle 9 of the H. B. Robinson Unit-2 to supply us with the core power distributions. This work could not have been completed
. without their kind cooperation. The authors wish to express their appreciation to the reviewers, J.
V, Pace III from the Oak Ridge National Laboratory and W. K. Cantrell of Carolina Power and Light Company, for their comments and suggestions. Special thanks go to D. M. Counce, C. I, Moser, and ,
C. H. Shappert for providing the editorial review, and to T. R. Henson and M. R. Whittenbarger for !
the preparation of this report. Finally, the authors gratefully acknowledge the programmatic support and encouragement from A. Taboada, C. J. Fairbanks, and M. E. Mayfield of the Nuclear Regulatory Commission Their support was essential for accomplishing our objectives.
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xi NUREG/CR-6453 i
I l
1 BENCHMARK DEFINITION
1.1 INTRODUCTION
This section defines the benchmark for analysis of a power reactor pressure vessel surveillance dosimetry based on data from the H. B, Robinson Unit 2 (HBR-2) power plant This benchmark will be referred to as the HBR-2 benchmark. Analysis of the HBR-2 benchmark can be used as partial
- fulfdiment of the requirements for the quali6 cation of the methodology for calculating neutron fluence Lin pressure vessels, as required by the U.S. Nuclear Regulatory Commission Regulatory GuideL DG-1053."
The scope of the HBR-2 benchmark is to validate the capabilities of the calculational methodology to predict the specific activities of the radiometric dosimeters irradiated in a surveillance capsule location (in-vessel) and in a cavity location (ex-vessel), starting from the data that are typically available for an analysis of a power reactor pressure vessel surveillance dosimetry.
The input data provided consist of reactor geometry, material composition, core power distribution, and power history for the time ofirradiation. The data given in Section 1 of this document and on the floppy disk accompanying this report are sufficient for the HBR-2 benchmark analysis,t. References to other documents are provided but are not necessary for the benchmark calculation.
Experimental data provided are the measured (M) specific activities of the radiometric monitors at the end ofirradiation. The dosimeters were irradiated during cycle 9 on the midplane of the HBR-2 ,
core in the surveillance capsule and in the cavity location, The principal results required from the benchmark analysis are the calculated (C) specific activities at the end of the cycle and the C/M ratios, for all the measurements provided. The reaction rates as
- obtained from the transport calculations should'also be given. Short descriptions of the method and
. model used should accompany the numerical results.
The cross-section sets, modeling techniques, and approximations to be used in the HBP.-2 benchmark analysis will be selected by the analyst; however, they are essential components of the qualified -
methodology and must be used in a consistent way.
'U.S. Nuclear Regulatory Commission, Calculational and Dosimetry Methodsfor Determining Pressure Vessel Neurrun Fluence. DraA Regulatory Guide DG 1053, to be published. .
i The description of the core power distribution requires a large amount of data, w hich are provided on the floppy disk.
.~ e
1.2 DESCRIPTION
HBR-2 is a 2300-MW (thermal) pressurized light-water reactor (PWR) designed by Westinghouse and placed in operation in March of 1971. It is owned by Carolina Power and Light Company. The !
data- presented in this section were obtained from Refs. I and 2, and fram personal Communications.I'"
The core of the HBR-2 reactor consists of 157 fuel elements and is surrounded by the core baffle, core barrel, thermal shield, pressure vessel, and biological shield. Selected general data and dimensions of the HBR-2 reactor are given in Table 1.1.- An octant of the horizontal cross-section of the reactor is shown schematically in Fig.1.1, which also shows the locations of the capsule and -
cavity dosimeters. Axial geometry and dimensions are given in Fig.1.2. The core baffle geometry is further specified in Fig.1.3. Surveillance capsules are located in the downcomer region and are attached to the thermal shield. The details of the capsule mounting are shown in Fig.1.4.
The reactor cavity is 17.10 cm (6.73 in.) wide, measured from the pressure vessel outer radius to the inner radius of the cylindrical biological (concrete) shield. A 7.62-cm (3-in.) thick insulation is installed in the cavity, leaving a 1.31-cm (0.52-in.) air gap between the pressure vessel and the insulation and an 8.18-cm (3.22-in.) air gap between the insulation and the concrete shield. The insulation consists of three steel sheets and eight steel foils with air gaps between them. The total thickness of the insulation steel sheets and foils is 0.2286 cm (0.090 in.). There are two relatively wide (38 cm, or 15 in.) and deep (80.645 cm, or 2 ft,7.75 in.) detector wells at 0* and 45* azimuthal locations. In each well is a vertical cylinder with a 19.05-cm (7.5-in.) outer diameter and 0.635-cm (0.25-in.)-thick steel wall. The vertical axis of the cylinder is at 252.174 cm (8 ft, 3.28125 in.) from the core center. The concrete surfaces of the detector well are covered with a 0.635-cm (0.25-in.)-
thick steel liner. Other concrete surfaces are bare.
The material composition of the reactor components (e.g., pressure vessel, thermal shield, etc.) is -
given in Table 1.2. Some components (e.g., fuel elements), have an elaborate design, but they are usually approximated as homogenized regions in the transport calculations of the out-of-the-core neutron field. To reduce the amount of data needed for such regions, the volume fractions of the materials are given in Table 1.2. The regions given in Table 1.2 correspond to the ones shown in Figs.
1.1 and 1.2. The core-average water temperature during cycle 9 was ~ 280 C (536 F), and the temperature of the water in the downcomer was approximately 267 C (512*F).* The pressure was 15.513 MPa (2250 psia). The cycle average boron concentration in the coolant was approximately 500 ppm. The corresponding water densities in different regions are also given in Table 1.2. The 8S. L Anderson, Westinghouse Electric Corporation, personal communication to 1. Remec, Oak Ridge National Laboratory,19%.
"R. M. Kirch,11. B. Robinson Steam Electric Plant. Unit No. 2, response to request for infonnation regarding operating cycle 9, personal communication to J. V. Pace, Oak Ridge National Laboratory, Oct.1,19%.
densities and chemical compositions of the other materials are given in Table 1.3. The concrete of the bio!ogical shielding is assumed to be type 02-B ordinary concrete (Ref. 3) with water content reduced to 4.67% by weight and iron concentration increased to reflect an estimated 0.7% by volume addition of rebar (Ref.1),
1.3 CORE POWER DISTRIBUTION AND POWER IIISTORY The fuel assemblies in the core are numbered as shown in Fig.1.5. These numbers are used in the description of the core power distribution during cycle 9. The data files referred to in the following discussion are provided as ASCII files on the floppy disk.
For each assembly in the core, the mass of uranium, bumup at the beginning of cycle life (BOL) and end ofcycle life (EOL), bumup increment in cycle 9, and cycle-average relative power are listed in the data fde FILEl.DAT. Part of the 61e is shown in Fig.1.6. These data were taken from the TOTE output, except for the cycle-average assembly power. It was calculated from the BOL and EOL assembly-average burnup, taking into account the assembly uranium content. Assembly powers are normalized to the core-wise average of 1.00.
Cycle-average, assembly-wise axial power distributions are given in FILE 2.DAT. Part of FILE 2.DAT is shown in Fig 1.7. Each assembly is divided vertically into 12 equal length segments, covering the active length of the fuel, with the first segment on the top and the twelfth segment at the bottom. Cycle-average relative power for each segment is given. Assembly segment powers are normalized to the average value 1.00. Relative powers of the segments were calculated from the relative cumulative axial bumup distributions given in the TOTE output for each assembly.
l The cycle-average assembly-pin-power distributions are given in FILE 3.DAT. The content of the file is illustrated in Fig.1.8. Distributions are given for the assemblics in the top right quadrant of the core (e.g., assemblies 2, 3, 7, 8, 9,10, . . ,79, 80, 81, 82, 83, 84, 85, 86) only. For each assembly, an array of 15 x 15 relative pin powers is given. Pin powers are normalized so that the average of the fuel-pin j powers (e.g., 204 per assembly) is 1.00. The pin powers are ordered in rows: the first value corresponds to the pin in the top left corntr of the assembly, the last value in row I to the pin at the top right corner of the assembly, and the last value in row 15 to the pin at the bottom right corner of the assembly. The orientation of the assembly in the core is as shown in Fig.1.5. The cycle-average pin powers were obtained by weighting the pin powers which were given at eight core burnup steps during the cycle. The weight assigned to the power distribution at the I-th burnup step was proportional to the bumup increment from the midpoint of the (I- 1)-th and I-tn burnup step and 1-th and (I+1)-th burnup step.
For cycle 9, a low-leakage core loading pattem was used in which 12 previously burned fuel elements (i.e., elements number 1, 2, 3, 57, 71, 72, 86, 87,101,155,156, and 157) were put on the core periphery. During cycle 9, the relative powers of the outer assemblies changeu *mificantly. This effect, which is often referred to as power redistribution, is caused by the fuel burnout and gradual changes of the boron concentration in the coolant during the cycle. The power redistribution affects 3 NUREG/CR-6453
- the core neutron leakage and consequently the dosimeter reaction rates. For this reason, the cycle-average core power distribution data, described previously, are :,upplemented by the power distribution data at several burnup steps during the cycle. At the core-average cycle burnups of 147,
~
417,1632, 3363, 5257, 7595, 9293, and 10379 megawatt days per metric ton of uranium (mwd /MTU) the - following information is provided: average assembly powers (FILE 4.DAT, see Fig.1.9), assembly burnups (FILES.DAT, see Fig.1.10), pin-power distributions for the assemblies in the upper left quadrant of the core (FILE 6.DAT, see Fig.1.11), and assembly-wise axial power.
distributions in 12 axial segments (FILE 7.DAT, see Fig.1.12),
The core power history for cycle 9 is given in the FILE 8.DAT as is illustrated in Fig.1.13_.
Descriptions of the contents and formats of the files are given at the end of each file and are shown in Figs.1.6-1.13, 1.4 DOSIMETRY During cycle 9, comprehensive sets of dosimeters were irradiated in the surveillance capsule position -
and in several locations in the reactor cavity (Ref. 2). For the benchmark, a subset of the measurements was chosen. The selected subset consists of the threshold radiometric monitors from the surveillance capsule at the azimuthal angle of 20' and from the cavity dosimetry located at the azimuthal angle of 0*.
A specially built surveillance capsule containing no metallurgical specimens, but otherwise identical to a standard Westinghouse capsule, was placed in a previously used holder ai the 20" azimuthal .
1 angle location in the downcomer. The region that usually contains metallurgical specimens was filled
- with c srbon steel, and the dosimeters were installed in the holes drilled in the steel. Specific activities given in Table 1.4 are for the core-midplane set." Radially, the dosimeters were installed at the capsule centerline at the radius of 191.15 cm (see Fig.1.4). The specially built capsule was irradiated during cycle 9 only.
"Dommetry sets were installed in the capsule at the core midplane and approximately 28 cm (11 in.) above and below tae midplane. The measured activities showed axial variations of only ~ 3%. which is not considered important, and therefore only the results for the midplanc set are given.
Speci6c activities of the cavity dosimeters irradiated at O' azimuth, on the core midplane,88 are also given in Table 1.4. The dosimeters wwe irradiated in an aluminum 6061 holder 5.08 cm (2 in.) wide, 1.422 cm (0.56 in.) thick, and 15.240 cm (6 in.) long. Aluminum was selected as the holder material in order to minimize neutron flux perttubations at the dosimeter locations. The holder was supported by a 0.813 mm (0.032 in.)-diam. stainless steel gradient wire mounted vertically in the gap between the lasulation and the biological shield at a radius of 238.02 cm (93.71 in.). The sketch of the O' azimuth cavity dosimetry axial locations is given in Fig.1.14.
Specific activities listed in Table 1.4 are as-measured with no corrections (e.g., for impurities or photofission). The corrections, which were estimated and used in a previous analysis (Ref.1) are given in the footnotes to Table 1.4; however, their use is left to the analyst. The specific activities are given for the end of HBR-2 cycle 9 (January 26,1984, at 12 P.M.).
1.5 REFERENCES
- 1. R. E. Macrker, "LEPRICON Analysis of the Pressure Vessel Surveillance Dosimetry Inserted into H. B. Robinson-2 During Cycle 9," Nuc. Sci. Eng.,96:263 (1987).
- 2. E. P. Lippincott et al., Emluation of Surveillance Cap.svle arul Reactor Cavity Dosimetry from H. B. Robinson Unit 2, Cycle 9, NUREG/CR-4576 (WCAP-11104), Westinghouse Corp.,
Pittsburgh, Pa., February 1987.
- 3. Reactor Physics Constants,2nd ed., ANL-5800, p.600, Argonne National Laboratory,1%3.
"In the present benchmark, only the midplane measurements are considered. However, at the O' azimuth multiple dosimeter sets were irradiated at the midplanc and at 213 cm ( 7 ft) and 107 cm (3.5 ft) above and below the midplane; and activities of the gradient wire ("Fe(n.p)"Mn and 5'Ni(n.p)"Co reactions) were measured at several positions between the foil locations. Adding these measurements to the benchmark would enlarge the scope of the benchmark to include verification of the calculational methodology for off-midplane locations.
Table 1.1 Selected general data and dimensiones of the H. B. Robinson Unit 2 Plant Location South Carolina, Hartsville Owner Carolina Power and Light Beginning of operation March 1971 Reactor Vendor Westinghouse Type PWR Coolant HO2 Number ofloops 3 Thermal power 2300 MW Core Number of fuel assemblies 157 Pitch 21.504 cm 8.466 in.
Fuel Element Type 15 x 15 array of fuel pins Fuel pins per element 204 Horizontal cross section rectangular, (including gap) 21.504 cm x 21.504 cm Ileight of fuel 365.76 cm 12 ft.
Core Bame' Dimensions See Fig.1.3 Thickness 2.858 0.013 cm 1.125
- 0.005 in.
Core Barrelt inner radius 170.023
- 0.318 cm 66.938
- 0.125 in.
Thickness 5.161 0.107 cm 2.032
- 0.042 in.
Thermal Shield Inner radius 181.135
- 0.318 cm 71.313 0.125 in.
Thickness 6.825
- 0.160 cm 2.687
- 0.063 in.
Pressure Vessel Cladding Inner radius 197.485
- 0.076 cm 77.750 0.030 in.
Thickness (minimal) 0.556 cm 7/32 in. (0.219 in.)
Base metal Inner radius
- 198.041 cm 77.969 in.
Thicknesi' 23.614
- 0.041 cm 9.297 0.016 in.
Total thickness 24.170 c:n 9.516 in.
(wall + cladding )
Table 1.1 (continued)
Pressure Vessel Thermal Insulation Inner radius 222.964 cm 87.781 in.
Totalinsulation thickness 7.620 cm 3.0 in, (including voids)
Insulation steel components I steel sheet 0.079 cm 0.031 in.
I steel sheet 0.046 cm 0.018 in.
I steel sheet 0.064 cm 0.025 in.
8 steel foils 0.005 cm 0.002 in.
Total thickness of the steel 0.229 cm 0.090 in.
in the insulation Pressure Vessel Cavity See Fig.1.1 Dimensions Vessel-to-insulation gap 1.31 cm 0.52 in.
Insulation 7.62 cm 3.00 in.
Insulation-to-concrete gap 8.18 cm 3.22 in.
Total width of the 17.10 cm 6.73 in.
cylindrical patt Biological Shield Dimensions See Fig.1.1 Inner radius of cylindrical 238.760 cm 7 ft10 in.
surfaces
- The taffle units are positioned symmetrically about the core center within 0.025 cm (0.010 in.) measured at the top and bottom former elevations.
t The annular gap between the core barrel outer radius and the thermal shield inner radius is maintaired uniform within 0.381 cm(0.150in.).
t The pressure vessel base metal inner radius is obtained as the cladding inner radius plus specified nummum cladding thickness of 0.556 cm (7/32 in.),
" The pressure vessel thickness is based on a single measurement of the lower shell and three measurements of the intermediate shell (S. L Anderson Westinghouse Electric Corporation, personal communication to I. Remec, Oak Ridge National Laboratory,19%).
Table 1.2 Materials of the components and regions
. Region Material
- Volume fraction UO 2, enriched to 2.9%, 0.2997 density 10.418 g cm'3 Zircaloy-4 0.1004 Reactor core Inconel-718 0.00281 Stainless steel SS-304 0.00062 Water 0.5886 density 0.766 g cm'8 Core bame Stainless steel SS-304 1.00 Bypass region Water See Fig.1.2 density 0.776 g cm'8 Core barrel Stainless steel SS-304 1.00 Downcomer Water 1.00 region No I density 0.787 g cm
Thermal shield Stainless steel SS-304 1,00 Surveillance capsule Mounting Stainless steel SS-304 1.00 Content Steel A533B 1,00 Downcomer Water 1.00 region No. 2 density 0.787 g cm'3 Pressure vessel Stainless Steel SS-304 1.00 cladding Pressure vessel Steel A533B 1.00 Insulation Stainless steel SS-304 0.03 Air 0.97 Reactor cavity Air - 1.00 Biological shield Concrete 1.00 NUREG/CR-6453 8
Table 1.2 (continued)
Region Material Volume Fraction Core support Stainless steel SS-304 0.049 Water 0.951 density 0.787 g cm '
Lower core plate Stainless steel SS-304 0.668 Water 0.332 density 0.787 g cm-3 Nozzle legs Stainless steel SS-304 0.070 Water 0.930 density 0.787 g cm'8 l
Bottom nouie plate Stainless steel SS-304 0.717 Water 0.283 i
density 0.787 g cm
Water gap No.1 Stainless steel SS-304 0.007 Water 0.993 density 0.787 g cm'8 End plugs Stainless steel SS-304 0.300 Water 0.700 density 0.787 g cm
Fuel plenum Stainless steel SS-304 0.224 Water 0.560 density 0.745 g cm'3 Water gap No.2 Stainless steel SS-304 0.017 Water 0.983 density 0.745 g em-3 Top nonle Stainless steel SS-304 0.275 Water 0.725 density 0.745 g cm'3 Formers Stainless steel SS-304 0.900 Water 0.100 density 0.766 g cm'3
- The boron concentration in the coolant was approximately 500 ppm (cycle average).
i Table 1.3 Densities (gem 8) and chemical compositions (wt %) of reactor component materials l
Carbon steel Stainless steel l A533B SS-304 Inconel-718 Zircaloy-4 Concrete
- Density 7.83 8.03 8.3 6.56 2.275 Element Fe 97.90 69.0 7.0 0.50 3.82 j Ni 0.55 10.0 73.0 i Cr 19.0 15,0 Mn 1.30 2.0 ,
C 0.25 0.10 Ti 2.5 Si 2.5 34.09 Zr 97.91 Sn 1.59 Ca 4.40 K 1.31 Al 3.43 Mg 0.22 Na 1.62 0 50.50 H 0.51
- The concrete is assumed to be type 02-B ordinary concrete (Ref. 3) with water content reduced to 4.67% by weight and iron concentration increased to reflect an estimate 0.7% by volume addition of rebar (Ref.1).
1 NUREG/CR-6453 10
Table 1.4 Measured specific activities of the dosimeters from the surveillance capsule and from the cavity, at the end of cycle 9 (1/26/1984). Specific activities (Bq/mg) are given per mg of Ni, Fe, Ti, and Cu material with naturally occurring isotopic composition, and per mg of 23'Np and 23:U isotopes.
Surveillance capsule Cavity Dosimeter (core midplane,20' azimuth, (core midplane,0* azimuth, radius 191.25 cm)* radius 238.02 cm)t 2"Np(nf) 8"Cs 3.671 x 102 2.236 x 10 8 23:
U(nf)'"Cs 5.345 x 10' 8.513 x 10 8 5'Ni (n.p) 5'Co 1,786 x 10 d8 1.959 x 10 2 "Fe (n,p)"Mn 9.342 x 102 : g,73 3 "Ti (n,p)"Se 3,500 x 102 " 3.310 4 "Cu (n, a)'#Co 2.646 x 10'" 2.645 x 10-8
- Dosuneters in the capsule were irradiated under 0.508 mm (0.020 in.) Od cover, except where noted difTerently. Ref. I estimates that in order to compensate for the photofission contribution, the "'Cs activity in "'Np and "*U should be reduced by 2.5% and 5%, respectively; and "Co activity in "Cu should be reduced by 2.5% to compensate for the contribution from the "Co(n. y)"Co reaction on the Co impurities in the Cu dosimeter, t in the cavity, the "'Np.2"U, and Ni dosuneters were in adiated under 0,508 mm (0.020 in.) Cd cover.
Ti and Fe dosimeters were irradiated bare. Activity of the Fe is an average of four measurements. The Cu activity is an average of one bare dosimeter and one dosimeter irradiated in Cd cover. Ref.- 1 esumates that in order to compensate for the photofission contribution, the "'Cs activity in "'Np and 2"U should be reduced by 5.0% and 10.0%, respectively', and "Co activity in Cu should be reduced by 2.5% to compensate for the contribution from the "Co(n y)"Co reaction on the Co impurities in the Cu dosimeter.
Average of five dosimeters, three inside Gd and two outside.
" Average of two measurements.
1I NUREG/-CR-6453
Regions: 8 1.. reactor core core baffle pr.,
2 l g
- 3. bypass region #2 D 3 4 core barrel
? downcomer reg. #1 T 5 0 6. thermal shicId 7 . downconer reg. #2 g 8 pressure vessel 9,1 I cavity (air) I3 10 vesselinsulation 88 12 biological shield 13 . steel-wallcylinder g
I4 detector well 3 x . capsule and cavity l I ;
dosimeters locations &
5 c5 l 2d* c 4 i 3
G i
N. ,
/
V /
// 7
/
}
[
/ / g O -- 0-degrees Fig.1.1 IIorizonini cross section of the IIBR-2 reactor. One octant of the core is shown
EttAta , __
ll 30L713
~
sonAss _ WATER QAP #2.
197.13:
FUEL PLENUM ,
teLast _
18 Lees _
ima _ /
141.388 j
FORMERS
\ j I
E eLaas es,141 _
9 lI g
0 g
h l
REACTOR *
""- CORE 4 d Ies.sas- 11.7en
- a. _
fj l g qL
'[
/
.is.ase j
/
/
. eras 1 FORMERS :
N
$ .11s.15s s
11La _
g 172.112 _
+17EA04 ,,
g
.teLego tas. ass END PLUGS
.tasais , WATER QAP #1
.tes.7a BOTTOM Mam a PLATE
. tea.no MOZZLE LEGS
.tes.ata ,, LOWm.R CORE PLATE CORE SUPPORTS
,,, m ,
Fig.1.2 Schematic sketch of the axial geometry (not to scale). Dimensions are in centimeters ,
ORNL 97 5341 hec
- -A-+l j: B :
- -C :I J: D :L
- E :
l - -
I : F :
1 I
i l
I I DIMENSION A 64.607
- 0.036 cm (25.435
- 0.014 in.)
B 150.663 0.046 cm (59.316
- 0.018 in.)
C 193.680
- 0.056 cm (76.252
- 0.022 in.)
D 236.698 0.066 cm (93.188
- 0.026 in.)
E 279.715
- 0.076 cm (110.124
- 0.030 in.)
F 322.684
- 0.084 cm (127.041
- 0.033 in.)
Fig.1.3 Core ballie geometry. Nominal dimensions are given for the core-side surfaces of the bafile plates. Deviations from nominal are for the maximum and minimum dimensions (e.g., for A the nominal dimension is 64.607 cm, with the maximum value of 64.643 cm and the minimum value of 64.571 cm)
E g4 E sQ f 3 g R
. =
g .
,,, ,ir o o RADIUS (cm)
+- - 193.212
+--- 192.742 MOUNTING (STAINLESS STEEL) - 191.155
/
CO T ~ ~
- 189.587 (STEEL A5338) f1 gg,ggg COOLANT
+-- '187.980 THERMAL SHIELD Fig.1.4 Sketch of the surveillance capsule mounting on the thermal shield (not to scale). The capsule centerline is at 191.155
- 0.152 cm (75.258 i 0.060 in.) (S, L.
Anderson, Westinghouse Electric Corporation, personal communication to I. Remec, Oak Ridge National Laboratory,1996) 15 NUREG/CL-6453
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 107 108 109 110 111 112 113 114 115 116 117 118 119 120 121 122 123 124 125 126 127 128 129 130 131 132 133 134 135 136 137 138 139 140 141 142 143 144 145 146 147 148 149 150 151 152 153 154 155 156 157 Fig.1.5 The numbering of the fuel elements in the HBR-2 core NUREG/CR-6453 lu
I ASS. # MTU BU-BOL BU-EOL dBU Ave. P 1 0.4285 29488. 33102. 3614. 0.337 2 0.4269 22091. 26798. -4707. 0.438 3 0.4288 29122, 32688. 3566. 0.333 155 0.4285 29057, 32748. 3691. 0.345 156 0.4281 22727, 27443. 4716. 0.440 157 0.4291 29270. 32842. 3572. 0.334 EOF Legend:
ASS. H . . . . . assembly number as in Fig. 1. 5.
MTU........ mass of uranium in the assembly in metric tonnes.
BU-BOL..... assembly burnup in mwd /MTU at the beginning of cycle (BOL).
BU-EOL..... assembly burnup in mwd /MTU-at the end of cycle (EOL).
dBU........ assembly burnup increase in cycle 9 in mwd /MTU.
Ave. P..... cycle-average assembly power.
Format: (I3,F9.4,3F9.0,F9.3)
Fig.1.6 Content and format of the FILEl.DAT. Beginning and end of file are shown 17 NUREG/CR-6453
ASS. 4 Cycle-average, assembly axial-segment powers 1 0.696895 0.995964 1.064975 1.054674 1.076852 1.063509 1.087683 1.073097 1.090714 1.071349 1.012495 0.711793 2 0.690069 0.992619 1.064409 1.053981 1.080842 1.062351 1.092513 1.074481 1.088941 1.070761 1.012128 0.716907 3 0.597160 0.942708 1.056286 1.058032 1.084421 1.075771 1.101560 1.090651 1.100337 1.093457 1.032470 0.767144 155 0.766128 1.028479 1.076598 1.086410 1.084673 1.0772"3 1.102337 1.077251 1.093745 1.056529 0.955013 0.593564 156 0.695985 0.991515 1.066876 1.080861 1.088512 1.080346 1.104275 1.084720 1.098602 1.071259 0.984913 0.652136 157 0.572353 0.931786 1.054233 1.069350 1.094944 1.084444 1.106110 1.094919 1.103209 1.094470 1.029888 0.764293 EOF Legend:
Ass. # ........ assembly number as in Fig. 1.5.
Cycle-average, assembly axial-segment powers (12 segments per assembly) are relative powers of axial segments of the assembly. Each segment is 30.48 ch. long (1 ft). First segment (after Ass. #) is at the top of active fuel and last segment is at the bottom of the fuel. Normalization is to the average segment power of 1.00 in every assembly (e.g., the sum of segment powers in any assembly is 12).
Format: (13, 6 F9. 6/ 3X, 6F9. 6)
Fig.1.7 Content and format of the FILE 2.DAT. Beginning and end of file are shown
}
CY0Lt.AvtAAGE PIN PNERS ASSEMBLY WUMBER 2 1 .2096 .2963 .3044 .3100 .3162 .3202 .3219 .3224 .3213 .3211 .3113 .3124 .3062 .2985 .2920 2 .4325 .4446 .4600 .4620 .4693 .4190 .4150 .4121 .4164 .4810 .4110 .4650 .4616 .4416 4359 3 .5393 .5615 0000 .5859 .5935 .0000 .6055 .6004 .6061 .0000 .5956 .5889 .0000 .5654 .5433 4 .6241 .6441 .6698 .6191 ,695; 1006 1009 .0000 .1011 .1022 .6901 .6t30 .6139 .6481 .6312 5 .1011 1214 .1561 1154 .0000 1168 1611 .1143 16t0 .1185 .0000 1191 1601 1326 1128 6
1004 .8194 .0000 .6609 .8563 .8424 .4138 .034a .5341 .e441 .3593 .4650 .0000 .6254 1930 1 .0616 .9910 .9323 9441 .9261 .9131 .9222 .9346 .9231 .9151 .929v .9495 .9311 .9910 .41)$
4 .9499 .9653 1.0090 .0000 1.0204 .9916 1.0191 .0000 1.0201 .9991 1.0231 .0000 1.0148 .9115 .9549 9 1.0329 1.0596 1.1016 1.1206 1.0993 1.0834 1.0931 1.1014 1.0941 1.0454 1.1021 1.1255 1.1136 1.0661 1.0391
-10 1.1185 1.1604 .0000 1.2142 1.2061 1.1859 1.1126 1.1136 1.1131 1.1981 1.2096 1.2192 .0000 1.1611 1.1244 11 1.2009 1.2303 1.2149 1.3039 .0000 1.3016 1.2042 1.2951 1.2853 1.3038 .0000 1.3090 1.2811 1.2311 1.2011 12 1.2853 1.3149 1.3(14 1.3165 1.4041 1.4115 1.4101 .0000 1.4110 1.4141 1.4004 1.3815 1.3611 1.3211 1.2914 13 1.3129 1.4101 .0000 1.4660 1.4190 .0000 1.5020 1.4993 . 5030 .0000 1.4826 1.4110 .0000 1.4269 1.3108 14 1. 4609 1. 4 015 1. 5288 1. 5240 1. 5369 1. $44 9 '1. 5400 1. 534 3 1. 54 90 1,5610 1. 5403 1. 52 91 1. 534 9 1. 4 940 1. 4665 15 1.5606 1.5104 1.5848 1.6019 1.6130 1.6223 1.6233 1 6233 1.6243 1.6244 1.6163 1.6066 1.5941 1.5168 1.5660 A$itMBLY NUMBER 86 1 1.5463 1.4414 1.3604 1.2150 1.1929 1.1120 1.0219 .9449 .0650 1881 1084 .6219 .5406 .4339 .2901 2 1.5542 1.4124 1.4063 1.3042 1.2221 1.1559 1.0563 .9635 .8913 .8222 1233 .6414 .5651 .4480 .2992 3 1.5104 1.5130 .0000 1.3520 1.2692 .0000 1.1051 1.0100 .9345 .0000 1616 .6156 .0000 .4664 .3085 4 1.5920 1.5068 1.4531 1.3662 1.2985 1.2104 1.1206 .0000 .94e1 .8663 1825 .6866 .5934 .4691 .3158 5 1.5916 1.51st 1.4666 1.3960 .0000 1.2029 1.0942 1.0222 .9305 .0626 .0000 1050 .6028 .4110 .3221 6 1.5999 1,5461 .0000 1.4018 1.2953 1.1818 1.0825 .9997 .9180 .e491 1951 .1109 .0000 .4993 .3211 1 1.5918 1,5215 1.4868 1.4016 1.2113 1.1692 1.0930 1.0223 .9211 .8414 .1168 1131 .6112 .4061 .3296 8 1.5953 1.5125 1.4136 .0000 1.2895 1.1109 1.2006 .0000 .9415 .5436 1858 .0000 .6139 .4835 .3309 9 1.5951 1.5261 1.4810 1.4021 1.2192 1.1116 1.0959 1.0256 .9312 .8452 1406 1112 .6210 .4895 .3321 10 1.5944 1.5450 .0000 1.4041 1.2991 1.1861 1.09 3 1.0062 .9251 .t565 1935 1181 .0000 .4999 .3320 11 1.5051 1.5164 1.4613 1.3995 .0000 1.2104 1.1011 1.0322 .9411 .9139 .0000 1165 .6131 .4L 6 .3292 12 1.5142 1.5036 1.4541 1.3108 1.3062 1.2210 1 1321 .0000 .9629 .8413 1918 .1015 ,6016 .4014 .3240 13 1.5600 1.5098 .0000 1.3511 1.2181 .0000 1.1206 1.0264 .9522 .0000 1002 .6938 .0000 .4815 .3193 14 1.5424 1.4614 1.4095 1.3106 1 2334 1.1103 1,0132 .9822 .9116 .8435 .7511 .6664 .5061 .4454 .3121 15 1.5313 1.4401 1.3609 1.2015 1.2045 1.1210 1.0469 .9663 .4400 .8122 1331 .6525 .5643 .4551 .3065 tor l Legend:
For each assembly in the top right quadrant of the core an array of 15 x 15 relative pin powers p(1,j) is given. The assembly is oriented as in Fig. 1. 5, pin (1,1) is in the top left corner, pin (1,15) in the- top right corner, pin (15,1) in the bottom left corner and pin (15,15) in the bottom right corner.
For each assembly in the top right quadrant of the core the following is given:
ASSEMBLY NUMBER # . . . . . . . . . . assembly number as in Fig. 1.5; one record.
Format: (' ASSEMBLY NUMBER ' ,I3)
Row number 1, pin powers p (i, j ) , j =1,15; fifteen records.
Format: (I3,15F7.4).
Fig.1.8 Content and format of the FILE 3.DAT. Beginning and end of file are shown 19 NUREG/CR-6453
I ASSEMBLY RELATIVE POWERS FOR BURNUP STEP 1( 147 mwd /MTV)
CORE AVERAGE ASSEMBLY POWER 1.000
.241 .324 .244
.705 .874 .853 .910 .855 .876 707 784 1.180 1.289 1.014 .883 1.016 1.292 1.183 786 781 1.054 1.190 1.116 1.266 .995 1.269 1.118 1.193 1.058 784 701 1.173 1.181 1.038 1.201 1.028 1.138 1.031 1.207 1.052 1.190 1.177 703
.668 1.280 1.108 1.196 1.113 1.258 1.086 1.264 1.120 1.207 1.114 1. 83 .867
.239 .846 1.008 1.260 1.024 1.257 1.171 1.371 1.177 1.263 1.029 1.263 1.007 .042 .235
.317 .897 .876 .991 1.134 1.082 1.365 1.133 1.374 1.086 1.136 .992 .876 .896 .317
.239 .844 1.006 1.261 1.027 1.259 1.171 1.367 1.1"4 1.262 1.027 1.261 1.006 .845 .239
.868 1.280 1.112 1.203
- 114 1.256 1.082 1.258 1.115 1.198 1.108 1.280 .867 702 1.175 1.187 1.047 1. A 98 1. 02 4 1.131 1. 02 4 1.198 1. 036 1.17 9 1.17 2 701 783 1.055 1.187 1.13' 1.257 .984 3.256 1.109 1.183 1.050 .780
.783 1.171 1.2v2 1.003 .857 1.002 1.279 1.171 700 703 .868 .845 .896 .843 .867 701
.241 .322 .241 ASSEMBLY RELATIVE POWERS FOR BURNUP STEP B kl0379 mwd /HTV)
CORE AVERAGE ASSEMBLY POWER 1.000
.382 .508 .386
.741 1.016 1.184 1.234 1.185 1.016 741
.847 1.121 1.215 1.050 .971 1.051 1.215 1.120 .847
.847 1.262 1.135 1.018 1.159 .96' 1.158 1.017 1.134 1.262 .847 741 1.121 1.132 .969 1.060 .957 1.257 .957 1.060 .975 1.133 1.120 741 1.017 1.215 1.016 1.059 .963 1.072 .958 1.073 .964 1.061 1.017 1.214 1.016
.337 1.188 1.053 1.159 .957 1.072 .950 1.070 .951 1.072 .957 1.159 1.051 1.184 .382
.514 1.237 .972 .962 1.257 .958 1.068 .869 1.071 .950 1.257 .962 .972 1.237 .514
.386 1.186 1.051 1.159 .957 1.073 .950 1.069 .951 1.074 .957 1.160 1.053 1.188 4387 1.017 1.214 1.018 1.062 .965 1.072 .959 1.074 .965 1.060 1.017 1.215 1.018
.741 1.121 1.134 .975 1.060 .957 1.258 .958 1.061 .970 1.132 1.121 742
.847 1.263 1.135 1.018 1.159 .959 1.159 1.019 1.135 1.263 .848 848 1.122 1.216 1.050 .956 1.049 1.216 1.122 .848 742 1.017 1.185 1.233 1.185 1.018 742
. 3 8 t, .504 .386 Eor Legend:
Assembly relative powers at eight core burnups.
Assembly relative powers are normalized to the average core-wise value of 1.00.
Format: free format Fig.1.9 Content and format of the FILE 4.DAT. Begianing and end of file are shown NUREG/CR-6453 20 i
CYCLE 9 AT 14100 mwd /T (MIM08UM-P CASE 9 MP Cycle 09 MAP 413, 0173 ppm, list MW, 00141 mwd /MTU)
A35tHBLY BUP#JPS IN 1000 mwd /MTU 29.214 22.141 29.143
.106 .131 .221 .136 .128 .311 .106
.111 1.106 6.970 18.806 23.040 14.000 6.947 1.105 .118
.117 .158 12.146 19.192 1.328 18.495 7.322 19.800 12.091 .158 .117
.305 1.115 12.202 21.411 12.075 22.990 .166 22.990 12.071 21.517 12.225 1.111 .105
.130 1.001 19.015 12.004 19.658 10.232 21.999 10.215 19.658 12.090 19.824 1.001 .130 29.272 .126 10.196 7.333 23.032 10.252 19.613 8.214 19.676 10.262 23.040 1.331 18.198 .125 29.143 21.141 .133 23.046 10.893 .266 22.014 8.373 21.264 6.329 22.080 .366 18.894 23.049 .133 21.149 29.215 .126 10.199 1.340 23.034 10.275 19.673 8.214 19.678 10.J59 23.025 7.336 18.000 .126 29.275
.330 1.040 19.835 12.005 19.655 10.216 21'.999 10.230 19.653 12.081 19.835 6.999 .130 %
.105 1.710 12.265 21.511 12.069 22.907 .165 22.999 12.014 21.034 12.217 1.111 .105
.111 -158 12,103 19.106 1.321 18.894 1.325 19.003 12.138 .351 .111
.111 1.104 6.941 14.906 23.C45 18.508 6.961 1.105 .117
.105 .130 .126 .133 .126 .130 .105 29.143 22.115 29.145 CYCLt 0 AT 10319.00 mwd /T (MICROEUM-P Cast 34 MP Cycle 09 MAP 455, 0040 ppe,1261 MW, 10379 mwd /HTU)
ASSEMBLY BUMVPS 1A 1000 mwd /MTU 32.491 26-411 32.406 1.316 9.149 10.704 11.323 10.127 9.757 7.380 8.430 19,295 19.616 29.438 32.739 29.451 19.604 19_299 8.435 9.423 12.299 23.936 30.586 19.817 29.462 19.319 30.593 23.891 12.310 0.428 1.369 19.245 23.945 31.964 23.464 33.192 13.041 33.199 23.415 31.145 23.999 19.291 7.364 1.143 19.632 30.515 23.446 30.080 21.956 32.380 21.910 30.110 23.503 30,606 19.620 9.717 l
32.528 10.717 29.431 38.817 33.211 21.980 30.281 20.417 30.314 22.01' 33.239 19.817 29.401 10.661 32.349 I 25.456 11.212 32.133 29.461 13.032 32.449 20.542 31.588 20.551 32.412 13.040 29.462 32.129 11.291 25.446 32.525 10.699 29.417 19.823 33.235 22.019 30.283 20.461 30.309 22.011 33.226 19.823 29.431 10.109 s2.521 5.731 19.660 30.621 23.491 30.091 21.941 32.311 21.972 30.092 23.463 30.590 19.624 9.141 1.369 19.296 24.046 31.740 23.446 33.170 13.017 33.108 23.461 31.916 23.954 19.203 7.361 8.431 12.313 23.888 30.560 19.175 29.412 19.117 30.515 23.905 12.284 8.419 0.433 19.289 19.571 29.390 32.542 29.383 19.586 19.279 8.422 1.311 9.131 10,666 11.240 10,659 9.125 1.366 32.384 27.050 32.385 EOF Legend: Assembly burnups at eight core burnup steps.
Format: free format Fig.1.10 Content and format of the FILES.DAT. Beginning and end of file are shown 21 NUREG/CR-6453
71N MWLP.A l'OR 6'JP>lVP STTP 4 1{ 141 thlFH71/)
MSLMPLY NUMILR 2 AA8LH&LY AVtPMt FULL *PDD M*th DEF%t NOP7Att TAT 100 .326 1 .2141 .2823 .2904 .2915 .3021 .3069 .3046 .3001 .3019 .3016 .3039 .2960 .2930 .2040 .2115 2 .411% .4249 .4416 .4431 .4S01 .4601 .4166 .4134 .4514 .4622 .4519 .4461 .4446 .4292 .41$1 3 1142 .5316 .0000 .66)1 .5101 .0000 .$t29 .6100 .6030 .0000 .6132 .$662 .0000 .1419 .5189 4 .$991 .6179 .6443 .6540 .6104 .6166 .6146 .0000 .6114 .6114 .till .6$16 .6400 .6221 .6039
$ .6100 .5963 1291 1480 .0000 .1$16 1419 4494 4421 .t$31 .0000 .is29 1342 .
- 0$1 .6830 6 .1661 1920 .0000 .93t1 .431$ .6114 .9099 .8111 .0111 .9199 .4349 .4394 .0000 1694 1646 1 .6361 .0649 .6D44 .9200 .9041 .9920 .6006 .6136 .6020 .9944 .6016 .9261 .9130 .5116 .4441 6 .6234 .9430 . Stet .0000 1.0023 .6004 1.0026 .0000 1.0038 .9825 1.0059 .0000 .9953 .6503 .9309 6 1.0126 1.0424 1.0t28 1.1011 1.0966 1.0119 1.t019 1.0611 1.0031 1.0143 1.0907 1.1132 1.0990 1.0500 1.0199 10 1.1044 1.1506 .0000 1.1083 1.2016 1.1910 1.1680 1.1100 1.1100 1.1934 1.2059 1.2141 .0000 1.1566 1.1120 11 1.1946 1.2218 1.;161 1.3D66 .0000 1.3011 1.2002 1.3013 1.2904 1.3096 .0000 1.3129 1.2034 1 2391 4.2022 12 1.2006 1.3234 1.3139 1.3660 1.4200 1.4290 1.4281 .0000 1.4299 1.4311 1.4250 1.3955 1 3015 1.3314 1.2660 13 1.3913 1.4445 .0000 1.4964 1 $110 .0000 1.$159 1.1220 1.1311 .0000 1.5163 1.6025 . 0000 1. 4 t2 4 1. 3 60 9 14 1.1019 1.$341 1. $ 006 1.117 3 1. 5 909 1. 6k 22 1. 604 0 1. 5 00 3 1. 6012 1. 6250 1. $ t 52 1. $ 130 1. $ l 82 1. 542 9 1. $ 0ll I b 1. f 411 1. 6554 1. 6119 1. 6 9 34 1.106 4 1.116 0 1.1106 1.1186 1.1190 1.1195 1.1104 1. 6 9 91 1. 64 51 1. 66 36 1. 64 0 4 AsstNBLY NUMBER 86 AASLMBLY AVIPMC f7,'LL.kOD LOWER TEM $1 NW%L11AT]DN .$14 1 1.44.1 1.4069 1.3382 1.2654 1.1920 1.1195 1.0406 .6626 .Stil .9102 .1300 .6491 .5591 .4494 .3009 2 1.etti 1.430$ 1.1831 1.2630 1.2224 1.1616 1.0691 .6001 .9116 .4443 .it34 .6693 .1852 .4642 .3090 3 1.t001 1.4692 .0000 1.3392 1.2656 .0000 1.1169 1.0264 .tS42 .0000 1939 .6915 .0000 .4829 .3196 4 1.$104 1.4626 1.42$8 1.3532 1.2940 1.2156 1.1309 .0000 .6616 .8tl6 .0049 100D .6136 .4062 .3213 5 1.5194 1.4144 1.4315 1.3006 .0000 1.2014 1.1095 1.0393 .6503 .8841 .0000 1211 .6230 .4945 ,3342 6 1.5264 1 4tv3 .0000 1.3811 1.2916 1.1814 1.0640 1.01tt .9301 .4111 .0098 1335 .0000 .5064 ,3394 1 1 $246 1.4012 1.4673 1.3851 1.2139 1.1143 1.1031 1.0311 .9410 ,9635 1961 1353 .6300 .5011 .3416 6 1.5229 1.4612 1.4441 .0000 1.28ti 1.1151 1.1101 .0000 .9601 .66$3 .0006 .0000 .6341 .5011 .3435 6 1.t22$ 1.400) 1 4 513 1. 30 65 1.21$ $ 1.1165 1.106 4 1.04 02 .t*05 .4612 .3034 1392 .6411 .5010 .3443 10 1.5221 1.4914 .0000 1.3860 1.2910 1.1920 1.0996 1.0218 .6,49 . tit) .9164 1413 .0000 .$134 .3443 11 1.5134 1.4115 1.4311 1.3935 .0000 1.2146 1.1141 1 0400 .6606 .06S6 .0000 .ille .6331 .$041 .3410 12 1.5023 1.4tet 1 4262 1 3513 1.3014 1.2253 1.1426 .0000 .9811 .9032 .4196 1238 .6216 .4992 .3363 13 1.4900 1.4643 .0000 1.3442 1.2144 .0000 1.1313 1.0424 .9115 .0000 .0020 1116 .0000 .4600 .3303 14 1.4?21 1.4246 1.3839 1.2995 1.2325 i.litt 1.0850 .9990 .6314 .4453 7146 .6901 .6051 .4819 .3221 15 1.4630 1.3991 1.3310 1.2111 . 2031 1.1346 1.0593 .9836 .9089 .3343
- it h .61;8 .5836 .4108 .3110 .
tsr Legend!
For eight core burnup steps the assembly pin powers are qive.n for each assembly in the top right quadtant of the e.*re. For each assembly an array of 15 x 15 relative pin powers p(1,j) is given. The assembly 11 oriented as in Fig. 1.53 pin (1,1) is in the top left corner, pin n,15) in the top right corner, pin (15,1) in the bottom left corner and pin (15,15) in the bottom right corner, for each assembly in the top right quadrant of the core the following is givent ASSEMBLY NUMBER W.......... assembly number as in Fig. 1.53 ene record.
Formatl (' ASSEMBLY NUMBER ',I3)
ASSEMBL, AVERAGE FUEL-ROD POWER BEFOR*. FO'URLI ZAT ION . . . . assembly average fuel-rod power is equal to the relative asumbly power.
Formatt (' ASSEMBLY AVERAGE FUEL-ROD POWER BOTORE NORMALIZATION ' F5.3)
Row number 1, pin powers p (1, $ ) , j =1,15 3 fif teen records. Pin powert are normalized so that the average of the fuel-pin powers (204 per ass,embly) is 1.000.
Formatt (13,15F7.4).
Fig.1.11 Content and format of the FILE 6.DAT. Beginning and end of file are shown NUREG/CR-6453 22
\
_ - _ _ _ - _ _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~
C0147HWd/HT 1 .04662 .07495 .08855 .09196 .09584 .09618
.09072 .09612 .09394 .08788 .07724 .05199 2 .04607 .07432 .08827 .09209 .09602 .09636
.09891 .09631 .09413 .08805 .07739 .05208 156 .05030 .07704 .08063 .09508 .09722 .09021
.09386 .09596 .09400 .08686 .07278 .04404 157 .03839 .06957 .08568 -.09152 .09607 .09696
.09859 .09707 .09587 .09165 .08082 .05780 10379 mwd /MT 1 .07266 .09436 .09279 .08585 .08551 .08275
.08349 .08130 .08161 .08257 .08181 .07530 2 .07118 .09300 .09242 .08613 .08589 .08312
.08366 .08166 .08197 .08295 .0821A .07564
\
157 .05238 .08230 .090$$ .08942 .08940 .08700
.08669 .08578 .08371 .08479 .08520 .08278 EOF Legend For eight core burnup steps, the following is given.
Ass. # ........ assembly number as in Fig. 1.5.
Relative powers of the axial segments of the assembly.
There are 12 segments per assembly and each segment is 30.48 cm long (1 ft). The first segment (after Ass. #) is at the top of the active fuel and the last segment is at the bottom of the fuel. Normalization is to the sum of the .c segment powers equal to 1.00 in any assembly.
Format: (I 3, 6 F10,5/ 3X , 6 F10,5)
Fig.1.12 Content and format of the FILE 7.DAT. Beginning and end of file are shown 23 NUREG/CR T6453 g r
__ _ .-.._._._.-.-_ _ _ _ _ ___ __,_ _ ~___ _______ _ . _ - . _ _ . _ .
i, 1
J YYMMDD BU LF DTG Temp.
mwd /MTU % mwd 100'F 820001 .0 .0 .0 5.250 820802 .0 .0 .0 .000 820003 .0 .0 .0 .000 820804 .0 .0 .0 4.000 820805 .0 .0 .0 5.300
- 820806 .0 .0 .0 3.440 f
840126 10636.8 34.0 782.2 5.354
! 840127 10636.8 .0 .0 5.030 840128 10636.8 .0 .0 .000 840129 10636.8 .0 .0 .000 4
840130 10636.8 .0 .0 1.2E0 840131 10636.8 .0 .0 1.170 EOF Legendt YYMMDD... year, month, day.
BU....... core average cycle burnup in mwd /MTU.
LF....... load factor in percents; LF-100* (daily average power /2300MW).
DTG...... daily thermal generat an in mwd.
Temp..... core average coolant camperature in 100'F.(e.g., 5.354 is 535.4*F) .
Format: (X,3I2, F8.1, F7.1, F8.1, F6. 3)
Fig.1.13 Content and fonnat of the FILE 8.DAT. Beginning and end of file are shown NUREG/CR-6453 24
i f
, .p. .; . . .
t o ll
$l. ' . ' . .. ,
- l f. ,.,.,. +213 cm (7 ft)
'" ~, -
e +107 cm (3.511)
~M
/ '
~~ ~
- ;-,, - - CORE MIDPLANE
?..
,.'.,.j.--.
-107 cm (3.S ft) h[,% .-
- - -213 cm (7 ft) n g i
.....1
..i l .
i a
- I ,.
i ',*,'..'.
I I R = 238.023 cm (93.71 in.)
Fig.1.14 Schematic drawing of the axial positions of the cavity dosimeten. At each of the marked locations a multiple dosimeter set was irradiated. Data in Table 1.4 are for the set in the core midplane 4
25 NUREG/CR-6453 i
o
2 BENCHMARK ANALYSIS 2.1 METHODOLOGY 1 This section describes the analysis of the HBR-2 benchmark. The transpon calculations were ;
performed using the DORT computer code (Ref.1) and the flux synthesis method.' The synthesis .
method, described in more detail in Ref. 2, relies on two and one dimensional (2 D and 1.D) i transport calculations to obtain an estimation of the neutron fluxes in the three-dimensional (3 D) :
geometries. When the method is used to analyze a neutron field in a region outside the core of a pressurized water reactor, it calls for three transpon calculations. One 2 D calculation models the horizontal cross section of the reactor in the r - 6 geometry. It is used to compute the variations of :
the neutron field in the radial direction (which is the main direction of the neutron transport from the - !
core toward the pressure vessel and beyond) and in the azimuthal direction. The second calculation ;
is a 2 D calculation in cylindrical r - geometry, in which a core is modeled as a finite-height cylinder. l The third calculation is nude for the 1 D (r) cylindrical model of the reactor, The r- and 1 D r calculations are combined to obtain the axial variations of the neutron field. ,
Geometry models used in this analysis were almost identical to those used in the previous HBR 2 i analysis (Refs. 2, 3). The r - 6 model covered one octant of the horizontal cross section with 74 :
azimuthal (6) intervals. In the radial direction-which extended from the core axis to the pressure !
vessel, the reactor cavity and inside the concrete shield (from a radius of 0 to 345 cm)-the number of radial intervals was varied with azimuthal interval (variable mesh option) and ranged from 93 to 116 intervals. The surveillance capsule was included in the model. The r - model used 75 axial (z-axis) intervals (57 intervals covered the active fuel height of 365.76 cm) and 93 radial intervals (from the axis of the core to the radius of 335 cm). The r - mesh outside the core described the geometry at the azimuth of 0', since the benchmark cavity dosimetry is at the 0* azimuth. The one- .
dimensional calculation used the same radial mesh as the r - model.
For the transpon calculations, the cross sections of the macroscopic mixtures were prepared by the
- GIP code (Ref. 4), using the homogenized zone compositions given in Table 1.2 of this repon. The P3 approximation to the angular dependence of the anisotropic scattering cross sections (i.e., the P.
- to P3 Legendre components) were taken into account, and a symmetric S " directional quadrature set" (i.e., a set of discrete directions and angular quadratures) were used for all transpon calculations.
The benchmark was analyzed with three cross-section libraries based on ENDF/B-VI: BUGLE 93 (Ref. 5), SAILOR 95,t and BUGLE-% (Ref. 6), which have 47 neutron and 20 gamma energy groups. ;
9
'DORT version 2.12.14, dated December 14,1994, was used.
i Regardmg SAILOR 95, see M. L. Williams, M. Asgari, and 11 Manohara," letter Report on Generating SAILOR 95
- Library," personal communication to F. B. K. Kam,' ORNL, February 1995.
The neutron sources for the r - 8, r - : and one-dimensional calculation were prepared by the DOTSOR code.8 For the r - 6 source, the cycle averaged pin power distributions in x y geometry and cycle average assembly powers were input into the DOTSOR, which transformed the power distribution into the r 6 geometry mesh. The power to neutron source conversion factor was based on the average bumup of the peripheral assemblies (i.e., assemblies 71, 86, and 101) at the middle of cycle 9, in order to account for the contributions of 2"U and2 "Pu to the fission neutron source."
The source energy spectrum was taken as the average of 2"U and 8"Pu fission spectra."
The source for the r - calculation was generated by averaging the cycle average pin powers of the
- top halves of the fuel elements 79 to 86 over they axis (see Fig.1.1; the y axis is perpendicular to the O' radial direction) and multiplying the average pin power values by the cycle average axial power distribution of the corresponding fuel assembly. The x -: power distribution obtained was then transformed into r : mesh by the DOTSOR code, which also prepared the source for 1 D r calculation by integrating the r - source over the : axis. The same source energy spectrum as in the
- r - 6 calculation was used for the r
- and r calculations.
I From the three transport calculations, the neutron fluxes in the core midplane, in the surveillance l capsule at the azimuth of 20', and in the cavity at the azimuth of 0' were synthesized. Reaction rates were calculated with the CROSS 95 dosimetry library (Ref. 7). The CROSS 95 cross sections were collapsed from the 640 to 47 energy groups using the FLXPRO code from the LSL-M2 code package (Ref. 8) and the reference spectra as calculated in the capsule and cavity location. The reaction rates are listed in Table 2.1.
To calculate the specific activities at the end ofirradiation, which are the measured quantities provided fo. comparison with the calculations, it is necessary to take into account the reactor power changes during irradiation and other changes that may affect the reaction rates. As a result of fuel burnup the power distribution in the core changes gradually throughout the fuel cycle, causing changes in neutronleakage from the core and consequent changes in reaction rates at the dosimetry -
locations. Since the reaction rates were calculated for one power distribution only (i.e., the cycle-average power distribution) approximations are necessary to account for these gradual changes.
t 1
Regarding DOTSOR, see M. L Willisma, DOTSOR: A Afodule in she LEPRICON Computer Code System for Repnsenting the Neutron Sourre Distribution in LWR Cons, EPRI Research Project 1399 1 Intenm Report (December 1985),RSIC Peripheral Shielding Routine Collection PSR 277.
"The power to-neutron-source conversion factor of 8.175 x 10" neutrons sM%M was calculated by the DOTSOR code for the fuel bumup of 28596 M%WMTU, which corresponds to the cycle-average burnup of the fuel assemblies number 71,86, and 101, tt The ENDF/B VI fission spectra for '"U and '"Pu were used.
Reaction rates at the dosimetry location are affected mostly by the closest fuel assemblies. Therefore, for the cavity-dosimetry location, the following appioximation was used. The cycle was divided into e'ght time intervals, based on the burnup steps at which the power distributions were provided. That is, the first interval was taken from the beginning of cycle to the core burnup halfway between the first and wond power distribution provided, the second interval from the end of the first interval to halfway b, ween the second and third power distribution, etc. The average relative powerp, of the three fuel clenients on the core flat edge (i.e., assemblies 71, 86, and 101) was calculated and assumed constant during the corresponding interval. The average relative powers (p,) were normalized so that, when integrated over the cycle, they provide the correct total energy produced (i e., the average energy produced in the three fuel elements, as given in FILEl.DAT). Using the daily power history, the reaction rate was then approximated as R, = R, x (p,Ipw7) x (P,I P,), (2.1) where R, = reaction rate at cavity location duringJ th day, R, = reaction rate obtained from transport (DORT) calculations, for nominal core power, p, = normalized average relative power of the fuel elements 71,86, and 101 during /-th time interval.
Per = average relative power of the fuel elements 71,86, and 101 used in the transport calculation (DORT),
P, = daily average reactor core power during dayJ. (Dayj is in the time interval ().
P, = nominal core power (2300 MW).
The same procedure was used for the calculation of activities of the dosimeters in the surveillance capsule; however, the fuel assemblics considered were the ones closest to the capsule location- that is, assemblies 43,56, and 71.
Different approaches can be used to account for the changes of reaction rates during the cycle; for example, one can (1) simply neglect the effects of redistribution and account only for the core power
! variations or (2) use the adjoint scaling techniques described in Ref. 2. The impact of different approaches on the calculated specific activities is further discussed in Appendix A.
2.2 RESULTS AND DISCUSSION The reaction rates calculated as described in the previous subsection, for the cycle average power distribution, are given in Table 2.1. The reaction rates obtained from the transport calculations with the BUGLE 93, SAILOR 95, and BUGLE-% are practically identical in the surveillance capsule, for all the reactions considered, the maximum differences are less than 1%. In the cavity the reaction rates obtained by BUGLE 96 and SAILOR 95 agree to better than 1%. The reaction rates obtained by BUGLE 93 for the Cu(n.n) and Ti(n.p) reactions are practically identical to those obtained by the other two libraries, while BUGLE 93 reaction rates for "Fe(n.p), "Ni(n.p), "'U(nf), cnd "'Np(nf) are 1%,2%,4% and 10% lower, respectively, than reaction rates calculated with the other two libraries. These observations are consistent with the results of the Pool Critical Assembly Pressure Vessel Facility Benchmark analysis, where good agreement of the reaction rates obtained by all three libraries was found for the dosimeters located inside the pressure vessel, while in the void box behind the pressure vessel (simulating the reactor cavity), the BUGLE-93 predicted lower reaction rates than the other two libraries, for all the dosimeters except "'Np, for which the BUGLE 93 predicted a higher reaction rate (Ref. 9).
Table 2.1 Reaction rates calculated for the cycle-average power distributloa and core power of 2300 MW (100% of nominal power), with different cross-section libraries for transport calculations Reaction Rate (s atom *')
3 g ,' "'Np(nf) "'U(nf) "Ni(n.p) "Fe(n,p) "Ti(n.p) Cu(n, a)
Library Capsule BUGLE-93 1.05E-13 1.54E- 14 4.74E- 15 3.50E- 15 5.62E-16 3.57E- 17 SAILOR 95 1.06E- 13 1.55E- 14 4.77E- 15 3.52E- 15 5.64E- 16 3.58E- 17 BUGLE.96 1.06E- 13 1.54E- 14 4.74E- 15 3.51E- 15 5,62E- 16 3.57E- 17 Cavity BUGLE-93 3.72E- 15 2.04E- 16 4.72E- 17 3.20E- 17 5.16E-18 3.63E- 19 SAILOR-95 4.14E- 15 2.12E- 16 4.82E- 17 3.24E- 17 5.18E-18 3.64E- 19 BUGLE 96 4.13E- 15 2.1 lE- 16 4.79E- 17 3.23 E- 17 5.16E-18 3.63E- 19 29 NUREG/CR-6453
With the reaction rates from Table 2.1 the specific activities were calculated as described in subsection 2.1. The calculated specific activities are given in Teble 2.2. Conversion from reaction rates to specific activities does not affect the differences between results obtained by different cross-section libraries; therefore, for the compadson of specific activities the comments given above for the l reaction rates apply. !
Table 2.2 Calculated specific activities Specific activity (Bq/mg)
Section "'Np(nf) "'U(nf) "Ni(n,p) "Fe(np) "Ti(n.p) Cu(n, a)
Library '"Cs 8"Cs "Co "Mn Sc **Co T,/
i 30 years 30 years 71 days 313 days 84 days 5.3 years Capsule BUGLE-93 3.28E+2 4.52E+1 1.70E+4 8.68E+2 2.96E+2 2.39E+1 SAILOR-95 3.31 E+2 4.56E+1 1.71E+4 8.73E+2 2.98E+2 2.40E+1 BUGLE-96 3.30E+2 4.54E+1 1.71E+4 8.69E+2 2.96E+2 2.39E+1 Cavity BUGLE-93 1.17E+1 6.06E- 1 1.88E+2 8.27 2.99 2.47E- 1 SAILOR-95 1.30E+1 6.30E- 1 1,91E+2 8.36 3.00 2.47E- 1 BUGLE.96 1.30E+1 6.28E- 1 1.91E+2 8.32 2.99 2.47E- 1
- Reaction product half life.
Table 2.3 lists the ratios of the calculated and measured specific activities. Calculated specific activities are taken from Table 2.2. Measured specific activities are taken from Table 1.4. The average C/M ratios in the capsule, for BUGLE-93, SAILOR 95, and BUGLE-96, are 0.90
- 0.04,0.90
- 0.04, and 0.90
- 0.04, respectively. If the corrections, discussed in notes to Table 1.4, are applied to the measured activities of the "'Np, "'U, andCu dosimeters, the C/M ratios increase by ~3%, 6%,
and 3% in the capsule, respectively, and by -6%,11%, and 3% in the cavity, respectively. The C/M ratios for the corrected measured actisities are listed in Table 2.3 in parentheses.
Table 2.3 Ratios of calculated-to measured (C/M) specific activiiles*
Cross.
)'
f Section Library 2"Np(nf) 2nU(nf)
"'Cs "'Cs "Ni(n.p)
"Co "Fe(",p)
"Mn "Ti(n p) Cu(n, a)
"Sc "Co Average' 8
Tv2 30 years 30 years 71 days 313 days 84 days 5.3 years Capsule BUGLE 93 0.89 0.85 0.95 0.93 0.85 0.90 0.90
- 0.04 (0.92) (0.89) (0.93) (0.91
- 0.04)
SAILOR 95 0.90 0.85 0.96 0.93 0.85 0.91 0.90
- 0.04 (0.92) (0.90) (0.93) (0.92 0.04) 13UGLE 96 0.90 0.85 0.96 0.93 0.85 0.90 0.90
- 0.04 (0.92) (0.89) (0.93) (0.91 0.04)
Cavity BUGLE-93 0.52 0.71 0.96 0.95 0.90 0.93 0.89 i 0.10 (0.55) (0.79) (0.96) (0.91
- 0.07)
SAILOR-95 0.58 0.74 0.98 0.96 0.91 0.94 0.91
- 0.10 (0.61) (0.82) (0.96) (0.93
- 0.06)
BUGLE-96 0.58 0.74 0.97 0.96 0.90 0.93 0.90
- 0.09 (0.61) (0.82) , (0.96) (0.92
- 0.06) llatios CM are given for the as-mee.sured specific activities. The ratios given in parentheses are calculated with corrections, specified in Table 1.4, applied 2to "Np(nf)"'Cs,"'U(n))"'Cs, and "Cu(n.a)"Co measured reaction rates, i Average C/M ratio and standard deviatic,n. For the cavity location averages are calculated without D'Np(nf)'"Cs reaction. The averaFes with "'Np(nJ)"'Cs reaction are 0.83
- 0.18 (0.85
- 0.16),0.85
- 0.16 (0.87
- 0.14), and 0.85
- 0.16 (0=87
- 0.14), for BUGLE 93, SAILOR 95, and BUGLE 96 libraries, respectively. Values in parentheses are calculated with cortections applied to "'Np,8"U, and "Cu dosimeters, as discussed in the footnote above.
Reaction product half-hfe.
In the cavity the CA1 ratio for the2"Np dosimeter is significantly lower than cat ratios for other dosimeters, regardless of the cross section libraty used.' Therefore, the average CAi values in the cavity, given in Table 2.3 in the last column on the right, were calculated without the Np dosimeter.
- This well known problem of the HBR 2 cycle 9 cavity dostmetry measurements was addressed in several analyses, but has not been completely explained Currently the most probable explanation appears to be an incorrect measured value.
The average C/hi values in the cavity for BUGLE-93, SAILOR 95, and BUGLE-96 are 0.89
- 0.10, 0.91
- 0.10, and 0.90
- 0.09, respectively.' The C/hi ratios given in parentheses are for the measured activities of "Np,2"U, and "Cu dosimeters, corrected as discussed in notes to Table 1.4. The average C/hi ratios in the cavity are practically identical to those in the capsule; therefore, no de:rease in the C/h1 ratios with increasing distance from the core and increasing thickness of steel penetrhted is observed Such decrease was typical for the pre-ENDF/B-VI cross section libraries and is illustrated in Appendix A.
The variations of the C/M values for different dosimeters at the same location are small: the standard deviation of the average C/hi ratios is ~0.04 in the capsule and ~0.10 in the cavity.8 These values suggest that the shapes of the calculated spectra, in the energy range to which the measured dosimeters are sensitive, are adequate. To further assess the differences between the three libraries the calculated multigroup neutron spectra are tabulated and compared in Appendix B. The tabulated spectra were used to determine the reaction rates given in Table 2.1. In the capsule the multigroup fluxes calculated with the BUGLE 93, SAILOR-95, and BUGLE-96 libraries agree to within ~2%,
except at thermal energies where differences are bigger: below -0. lev SAILOR 95 and BUGLE-93 predicted, respectively, ~2 times lower and 2.7 times higher flux than BUGLE-96 (see Fig. B.2).
These differences at the low energies are not important for predicting radiation damage in the steel specimens and reaction rates of the threshold neutron dosimeters in the capsule.
In the cavity, the group fluxes calculated with the SAILOR 95 and BUGLE-96 libraries agree to better than 1% over the entire energy range while the BUGLE-93 fluxes differ considerably (see Fig.
B.4). BUGLE 93 predicted up to two times higher fluxes below lev, and, more importantly, lower fluxes at higher energies, except between ~10 kev and 70 kev. Between ~0.lhieV and 2hieV, BUGLE 93 predicted at least 5% lower fluxes than BUGLE-96, with the maximum difference about 18% at ~0.7 hieV. This comparison, combined with the observation that the calculations underpredicted the reaction rates, suggests that neutron flux and spectmm in the cavity are more accurately predicted by the BUGLE-% library than by the BUGLE-93 library. Some support for this suggestion can also be found from the comparison of the calculated and measured specific activities (see Table 2.3). In the cavity, the BUGLE-93 library gave slightly lower C/hi ratios than the other 2
two libraries for the "Ni dosimeter and in particular for the "U and 2"Np dosimeters, which have lower reaction energy thresholds and are sensitive to the neutrons below ~2hieV. Similar difTerences, as observed here between the multigroup fluxes calculated by the BUGLE-93 and BUGLE-96 libraries, were also found in the analysis of the Pool Critical Assembly Pressure Vessel Facility (Ref. 9).
'If the "Np 8
dosimeter in the cavity is taken into account, the average CM values are 0.83
- 0.18,0 85
- 0.16, and 0 85
- 0.16, for the BUGLE-93, SAILOR 95, and BUGLE-% library, respectively.
8 1f the 2"Np dosimeter in the cavity is taken tnto account, the standard deviation of the avange C/M is -0.16.
1 2J REFERENCES
- 1. - W. A. Rhoades et al., " TORT DORT Two and Three Dimensional Discrete Ondinates Transport, Version 2.8.14," CCC 543 Radiation Shielding Infonnation Center, Oak Ridge National Laboratory,1994.
- 2. R. E. Maerker, *LEPRICON Analysis of the Pressure Vessel Surveillance Dosimetry inserted into H. B. Robinson 2 During Cycle 9," Nuc. Scl. Eng.,9&263 (1987). t
- 3. M. L. Williams, M. Asgari, F. B. K. Kam, Impact ofENDF/B VI Cross-Section Data on H. l B. Robinson Cycle 9 Dosimetry Calculations, NUREG/CR-6071 (ORNlJTM 12406),
October 1993.
- 4. W. A. Rhoades,"The GIP Program for Preparation of Group-Organized Cross-Section Libraries," inforraal notes, November 1975, RSIC Peripheral Shielding Routine Collection PSR 75.
- 5. D. T. Ingersoll et al.,' Bugle-93: Coupled 47 Neutron,20 Gamma Ray Group Cross Section Library Derived from ENDF/B VI for LWR Shielding and Pressure Vessel Dosimetry Applications," RSIC Data L!brary Collection, DLC-175, February 1994.
- 6. J. E. White et al.," BUGLE 96: Coupled 47 Neutron,20 Gamma Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," RSIC Data Library Collection, DLC-185, March 1996.
- 7. 1, Remec and F. B. K. Kam, An Update of the Dosimetry Cross-Section Data Basefor the Adjustment Code LSL-M2, ORNL/NRC/LTR 95/20, June 1995.
- 8. F. W. Stallmann, LSL-M2: A Comp' uter Programfor Least-Squares Logarithmic Adjustment i
ofNeutron Spectra, NUREG/CR-4349 (ORNL/TM 9933), March 1986.
- 9. 1. Remec and F. B. K. Kam, Pool Critical Assembly Pressure l'rssel Facility Benchmark, NUREG/CR-6454 (ORNL/TM 13:05), July 1997.
I l
- l. 33 NUREG/CR-6453 1
1
3 CONCLUSIONS Section 1 of this report describes the HBR 2 pressure vessel dosimetry benchmark and provides all the dimensions, material compositions and neutron source data necessary for the analysis. The neutron source data are provided on the floppy disk accompanying this report, in Section 2, the analysis of the HBR 2 benchmark is presented. The transport calculations with the computer code DORT, based on the discrete-ordinates method, were performed with three a ENDF/B.VI based multigroup libraries: BUGLE.93, SAILOR-95, and BUGLE.96. Excellent agreement of the calculated specific activities with the rnessurements was obtained. For the '
dosimeters in the surveillance capsule, the average C/M ratios for BUGLE 93, SAILOR 95, and y BUGLE-%, are 0.90
- 0.04,0.90
- 0.04, and 0.90
- 0.04, tespectively. For the dosimeters irradiated !
in the cavity, the average C/M ratios (excluding "'Np dosimeter) for BUGLE 93, SAILOR 95, and l HUGLE-06, are 0.89
- 0,10,0.91
- 0.10, and 0.90
- 0.09, tespectively. The C/M ratios given above are for the as measured specific activities (e.g., no corrections were applied to the "'Np, "U, and <
Cu dosimeters). No systematic decrease in the C/M ratios with increasing distance from the core was observed for any of the libraries used. ,
it is expected that the agreements of the calculations with the measurements, similar to those shown -
In this report, should typically be obtained when the discrete-ordinates method and the ENDF/B-VI cross-section libraries are used for the HBR 2 benchmark analysis.
1 i
I T
-t NUREG/CR-6453 _
34
APPENDIX A COMPARISON OF APPROXIMATIONS FOR MODELING TIIE REACTION RATE
, VARIATIONS DUE TO CORE POWER REDISTRIBUTION AND COMPARISON OF RESULTS OHTAINED WITil ENDF/B IV AND ENDF/B VI CROSS SECTIONS i
l l
In the steady state neutron field the activities of the dosimeters during irradiation gradually approach the saturated activities, which are proportional to the reaction rates. For a given dosimeter and reaction rate, the activity of the dosimeter depends only on the time ofirradiation. The transformation of the measured specif: activit) into the reaction rate is simple; it does not involve approximations and it does not introduce uncertainties other than those rela'ad to the characteristics of the dosimeter and reaction product. Therefore, the reaction rates deduced from activities measured in a steady state neutron field are usually referred to as " measured" reaction rates.
Ilowever, in a power reactor the neutron field and consequently the reaction rates vary with time because (1) the power distribution in the core gradually changes with fuel burnup (" power-redistribution") and (2) the changes of the reactor power. The reaction rates are often calculated for a given core condition only (e g., nominal thermal power, at certain core burnup), and approximations are necessary in the calculation of specific activitics. To approximate the effect of the changes of the core power it is usually assumed that the reaction rates are proportional to the core power (at all locations of the dosimeters ). The changes in reaction rates caused by the power redistribution may vary from negligible to ~30 to 40%. These changes depend primarily on the fuel loading pattern (and, therefore, vary from cycle to cycle) and may be different for different dosimetty locations. Since the treatment of the power redistribution effect is less standardized, the effect of a few different approximations is illustrated on the example ofilBR 2 cycle 9 dosimetry analysis. The following approaches were considered:
(a) Changes due to redistribution were approximately accounted for as described insubsection 2.1 [e.g., the reaction rates were taken proportional to the core power (daily-averaged) and average relative power of the fuel assemblies closest to the location of the dosimeters).
(b) The redistribution effect was neglected, and reaction rates were taken proportional to the core power (daily averaged).
(c) Reaction rates from the present analys'.s were converted into the specific activities by the conversion factors determined from Ref.1. In Ref. I the adjoint scaling technique was used to determine the reaction rates for eight core power distributions during the cycle and then the specific actisities were calculated by superimposing the power history. This method should be more accurate than the two approaches described above. However, in Ref. I the core power distributions from an older analysis were used, which may affect the reaction rate to-activity conversion factors and consequently the comparison with the results from steps (a) and (b).
Using the reaction rates obtained from the transport calculation with the BUGLE-95 library, the specific activities were calculated according to the three approximations described above. The C/M ratios for the capsule and cavity dosimeters are listed in Table A.I. In the capsule the three approximations give very similar average C/M ratios and corresponding standard deviations. This similarity exists because the changes of the power of the peripheral fuel assemblies closest to the capsule are relatively small; the average power of the elements number 43, 56, and 71 increased only
-20% from the beginning to the end of cycle. However, the average power of the assemblies on the 37 NUREG/CR-6453
flat edge (i e., assemblies 71,86 and 101) increased over 60% from the beginning to the end of the cycle, and this power increase affects the comparisons in the cavity. Approximations (a), (b), and (c) gave the average C/M values in the cavity of 0.90 i 0.09,0.83
- 0.08, and 0.84
- 0.05, respectively.
. The advantage of approximation (a) over (b) is clearly shown. The largest differences it; the CMi ratios are observed for the reactions with short lived products, "Ni(n.p)"Co and "Ti(n.p)"Sc; the C/M for the fission dosimeters, for which the activity oflong lived "Cs is measured, are almost unaffected. The approximations (b) and (c) give very similar results: average C/M and its standerd deviation in the cavity are 0.83
- 0.08, and 0.84
- 0.08, respectively, and in the capsule are 0.88
- 0.04, and 0.89
- 0.05, respectively. Therefore, in this case it appears that using the adjoint scaling technique [i.e., approximation (c)) gives little advantage over the simpler approximation (b),which accounts for core power variations only. Ilowever, the application of conversion factors, calculated from results obtained by adjoint scaling in Ref.1, to the reaction rates calculated in the present analysis, is approximate, as described above.
The IIBR 2 cycle 9 dosimetry has been analyzed before; see for example Refs. I and 2. In Ret 1, ELXSIR cross sections (Ref 3) based on ENDF/B.IV were used. In Ret 2, SAILOR cross sections (Ref 4) based on ENDF/B lV with iron, oxygen, and hydrogen cross sections from the ENDF/BN1 library and ENDF/BNI dosimetry cross sections were used. To assess the impact of the ENDF/BNI-based cross-section library for transport calculations, the analysis was repeated with exactly the same neutron source (i. e , spatial power distribution in the core, and source energy spectrum between "'U and "Tu ENDF/B V fission spectra) and modeling approximations that were used in Refs. I and 2.
For consistency (with Refs. I and 2), the time-dependent variations of reaction rates were approximated by using the mid cycle reaction rates to the end-of-cycle activities conversion factors from Rcf 1, and the measured reaction rates were corrected as described in the r,ote to i s1.4.
The SAILOR 95 (ENDF/BNI based) cross sections for transport calculations were usu, and dosimetry cross sections were taken from CROSS 95. This analysis will be referred to in the following discussion as the "new" analysis. The Cai ratios for the capsule and cavity dosimeters from the new analysis are compared with tb values from Refs. I and 2 in Table A.2.
In the cap.;ule, the new analysis gave thu average C/M of 0.93 t 0.05, slightly lower than the average of 0.96
- 0.05 from Ref. 2. This lower value is present probably because in the new analysis and in Rc0 2 the reaction rates inside the capsule were determined at slightly different radial locations. Both Ret 2 and the new analysis gave significantly improved CAi values over the values from Ref.1: the increase in the average C/M in the capsule is ~12% for the new analysis and ~16% for the Ret 2.
In the cavity location, the new analysis and Ref. 2 gave practically identical results, with the average CAi of 0.88
- 0.14, while the C81 average for the Ref.1 is 0.66
- 0.04. Therefore, the increase in 2
the average C/M ratio is ~33%. For the "Np(nf)'"Cs reaction the C/M ratio in the new analysis and in Ref. 2 is about 0.61 and differs significantly from the C/M ratios for the other dosimeters, as can be seen frcir 'able A.2. The average C/M for the cavity location calculated without the 2"Np(nf)'"Cs reaction is 0.93 for the new analysis and Ret 2, and 0.67 for Ref.1; therefore, an improvement of 39% was obtained.
The ENDF/B VI based cross sections for transport calculations resulted in improved agreement of calculatians and measurements, both in the capsule and in the cavity. The average C/M in the capsule, for the six dosimeters used, is 4.93
- 0.05; the ENDF/B l% based library gave 0.83
- 0.03. In the 2
cavity, the average (excluding "Np dosimeter) is 0.93
- 0.06, and 0.67
- 0.03 for the ENDF/B VI-and ENDF/B I% based libraries, respectively. Therefore, the ENDF/B VI based cross sections eliminated the decrease of the C/M ratios with increr. sing distance from the core and increasing thickness of the steel penetrated by neutrons.
i 39 NUREG/CR-6453
_-- _u
I Table A.1 Ratios of calculated to measured (C/M) specific activities obtained with different approximations for the time-dependent variations of reaction rates *
- U(nf) "Ni(n.p) "Fe(n.p) "Ti(n.p) Cu(n, a) 2nNfnf)
" Cs "'Cs "Co "Sc "Co Average t
_"Mn Tn8 30 years 30 years 71 days 313 days 84 days 5.3 years Capsule Approx. 0.90 0.85 0.96 0.93 0.85 0.90 0.90
- 0.04 (a)" (0.92) (0.89) (0.93) (0.91
- 0.04)
Approx. 0.90 0.85 0.90 0.91 0.80 0.90 0.88
- 0.04 >
(b)" (0.92) (0.89) (0.92) (0.89
- 0.05)
Approx. 0.86 0.82 0.98 0.91 0.87 0.87 0.89
- 0.05 (c)" (0.88) (0.86) (0.89) (0.90
- 0.04)
Cavity Approx. 0.58 0.74 0.97 0.96 0.90 0.93 0.90
- 0,09 (a)" (0.61) (0.82) (0.96) (0,92
- 0.06)
Approx. 0.57 0.73 0.83 0.90 0.78 0.92 0.83
- 0.08 (b)" (0.60) (0.80) (0.94) (0.85
- 0.07)
Approx. 0.55 0.70 0.90 0.90 0.84 0.88 0.84
- 0.08 (c)" (0.57) (0.77) (0.91) (0.86
- 0.06)
- Ratios CN are given for the as-measured specific activities. The ratios given in parentheses are calculated with conectims, speciGed in Table i A, applied to "'Np(n/)"'Cs, '"U(n/)"'Cs, and"Cu(n. a)"Co measured reaction rates.
' Average CM ratio and standard deviation. For the cavity location, averages are calculated without "'Np(nf)"'Cs reaction. 'the averages with "'Np(nf)"'Cs reaction are 0 05
- 0.16 (0.87
- 0.14),0.79
- 0.I3 (0.81
- 0.12), and 0.80
- 0.14 (0.82
- 0. I 3), for methods (a), (b), and (c), respectively. Values in parentheses are calculated with correctioas applied to "'Np, '"U, and "Cu dosimeters, as discussed in the footnote above.
8 Reaction product half life.
" See text for the explanation of the approximations (a),(b), and (c).
Table A.2 Comparison of the C/M ration of specific activities from the present analysis with the values from the previous analyses (Refs. I and 2)
C/M ratios
"'Np(nf) "'U(nJ) "Ni(np) "Fe(n.p) *&ri(n.p) "Cu(n, a) Average C/M
'"Cs 8"Cs "Co "Mn "Sc "Co io Capsule 0.91 0.89 1.01 0.95 0.89 0.91 0.93
- 0.05 Ana y is.
Analysis from 0.85 0.80 0.87 0.83 0.81 0.83 0.83
- 0.03 Ref. l' Analysis from 0.94 0.93 1.05 0.98 0.92 0.95 0.96
- 0.05 Ref. 28 Cavity
- ,. 0.62 0.84 0.98 0.97 0.89 0.95 (0 93
- 0 )**
ro 1 0.61 0.65 0.66 0.68 0.66 0.72 **
06 *0 Ref.1 Analysis 088*014
- from 0.61 0.86 0.97 0.97 0.90 0.96 Ref 28 (0.93
- 0.05)**
' New analysis, using SAILOR 95 and CROSS 95 cross sections.
T Results from Ref.1. using ELXSIR cross sectiov.s. based on ENDF/B-IV.
8 Results fran Ref. 2, using SAILOR cross sections (based on ENDF/B IV) with iron, oxygen, and hydrogen cross r,ecti ons from ENDF/B.VI library and ENDF/B VI dosimeuy cross-sections.
C/M fa rettunium omitted from the average.
APPENDIX A REFERENCES
- 1. R E. Mawker,"LEPRICON Analysis of the Pressure Vessel Surveillance Dosimetry Inserted into 11. H. Robinson 2 During Cycle 9," Nuc. Scl. Eng.,96:263 (1987).
- 2. M. L. Williams, M. Asgari, F. B. K. Kam, impact ofENDF/B 17 Cross-Section Data on H. B.
Robinson Cycle 9 Dosimetry Calculations, NUREG/CR 6071 (ORN1/rM -12406), October 1993.
3 .' M. L. Williams et al., The ELA51R Cross-Section Libraryfor L WR Pressure l'esselIrradiation Studies: Part of the LEPRICON Computer Code System, EPRI NY 3654, E%ctric Power Research Institute, Palo Alto, Calif.,1984.
- 4. O. L. Simmons and R. W. Roussin, *S AILOR. Coupled, Self Shielded,47 Neutron, 20 Gamma Ray, P3, Cross-Section Library for Light Water Reactors," DLC-76, Radiation Shielding Information Center, Oak Ridge National Laboratory,1985.
i=
i i
a
- NUREG/CR-6453 42 I
t
. ~ -
APPENDIX B CALCULATED NEUTRON SPECTRA AT TIIE DOSIMETRY LOCATIONS NUREG/CR-6453
Table H.1 Calculated multigroup neutron fluses in the surveillance capsule (20' azimuth, core midplane, at the radius of 191.15 cm from core vertical axis)
Group Group upper Neutron flux j number energylimit BUGLE 93 SAlLOR-95 BUGLE 96 i eV cm-2s* I em-2 ,- t em-2 ,-l 1 1.733E+07 8,870E+06 . 8.870E+06 8.870E+06 2 1.419E+07 2.808E407 2,808E+07 2.808E+07 3 1.221E+07 1.207E+08 1.207E+08 1.207E+08 ;
4 1.000E+07 2.379E+08 2.379E408 2.379E+08 5 8.607E+06 4.113E408 4.117E+08 4.112E+0B 6 7.408E+06 9.984E+08 1.001E+09 9.984E+08 7 6.065E+06 1.485E+09 1.493E409 1.485E+09 8 4.966E406 2.865E+09 2.891E+09 2.866E+09 9 3.679E+06 2.219E+09 2.229E+09 2.220E+09 10 3.012E+06 1.710E+09 1.719E+09 1.713E+09 11 2.72SE+06 1.998E+09 2.005E+09 2.001E+09 12 2,466E+06 9.928E+08 9.989E+08 9.963E+08 13 2.365E+06 2.768E+08 2.786E+08 2.779E408 14 2.346E+06 1.3R3E+09 1.392E+09 1.388E+09 15 2.231E+06 3.767E+09 3.790E+09 3.780E+09 16 1.920E+06 4.382E+09 4.427E+09 4.411 E+09 17 1.653E+06 6.576E+09 6,664E+09 6.633E+09 18 1.353E+06 1.190E+10 1.207E+10 1.203E+10 19 1.003E+06 8.127E+09 8.258E+09 8.228E+09 20 8.208E+05 4.085E+09 4.147E+09 4.136E+09 21 7.427E+05 1.150E+10 1.180E+10 1.176E+10 22 6.081E+05 9.276E+09 9.333E+09 9.306E+09 23 4.979E+05 1.001E+10 1.040Ei10 1.035E+10 24 3.688E+05 9.367E+09 9.430E+09 9.409E+09 25 2.972E+05 1.284E+10 1.319E+10 1.316E+10 26 1.832E+05 1.177E+10 1.185E+10 1.181E+10 ..
27 1.111E405 8.987E+09 9.102E+09 9.076E+09 28 6.738E+04 7.444E+09 7.496E+09 7.473E+09 29 4.087E+04 2.752E+09 2.796E+09 2.790E+09 30 3183E+04 1.316E+09 1.432E+09 1.429E+09 45 NUREG/CR-6453
Table B.1 (continued)
Group Group upper Neutron flux number energylimit BUGLE 93 SAILOR 95 BUGLE 96 eV cm- 2 ,- i em-2, l em- 2 ,- 1 31 2.606E+04 2.557E409 2.578E+09 2.571E+09 32 2.418E+04 1.539E+09 1.552E409 1.546E+09 33 2.188E+04 3.961 E+09 4.038E409 4.027E+09 34 1.503E+04 7.440E+09 7.589E+09 7.5$0E+09 35 7.102E+03 8.556E+09 8.686E+09 8.663E+09 36 3.355E+03 7.928E+09 8.047E+09 8.024E+09 37 1.585E+03 1.302E+10 1.332E+10 1.328E+10 38 4.540E+02 7.223E+09 7.381E+09 7.362E+09 39 2,144E+02 7.900E+09 8.050E+09 8.031E+09 40 1.013E+02 1.037E+10 1.056E+10 1.053E+10 41 3.727E+01 1.265E+10 1.288E+10 1.284E+10 42 1.068E+01 7.259E409 7.383E+09 7.365E+09 43 5.043E+00 9.579E+09 9.360E +00 9.384E+09 44 1.855E+00 7.103E+09 6.297E+09 6.357E+09 45 8.764E-01 6.081E409 4.893E+09 4.933E+09 46 4.140E 01 1.268E+10 7.068E+09 7.074E+09 47 1.000E 01 2.709E+10 5.339E+09 9.820E+09 1.000E-05'
- Low-energy boundary of the last group.-
?
TABLE B.2 Calcula'ed multigroup neutron fluses at the location of cavity dosimeters (O' azimuth, core midplane, at the radius of 238.02 cm from core vertical axes)
Group Group upper Neutron flux number energy limit BUGLE 93 SAILOR 95 BUGLE 96 eV cm-2 , l em 2, t em 2 ,- i 1 1.733E+07 1.385E+05 1.386E405 1.386E405 2 1.419E+07 3.917E+05 3.915E+05 3.918E+05 /
3 1.221E+07 1.544E+06 1.544E406 1.545E+06 4 1.000E+07 2.828E406 2.827E+06 2.827E+06 5 8.607E406 4.247E406 4.251E+06 4.248E406 6 7.408E+06 8.486E+06 8.507E+06 8.487E+06 6.065E+06 1,185E+07 7
4.966E406 1.190E407 1.185E+07
\
8 2.262E+07 2.284E+07 2.265E+07 9 3.679E+06 1.910E+07 1.929E407 1.918E+07 10 3.012E+06 1.572E+07 1.594E+07 1.585E+07 11 2.725E+06 1.992E407 2.020EM7 2.012E+07 12 2.466E406 1.051E+07 1.075E+07 1.071E+07 13 2.365E406 3.397E406 3.494E+06 3.480E406 14- 2.346E+06 1.736E+07 1.788E+07 1.782E+07 15 2.231E+06 4.876E407 5.0llE+07 4.996E+07 16 1.920E+06 7.343E+07 7.712E+07 7.683E+07 17 1.653E+06 1.291E+08 1.379E+08 1.373E+08 18 1.353E+06 3.391E+08 3.705E+08 3.694E+08
( 19 1.003E+06 3.610E+08 3.967E+08 3.952E+08 20 8.208E+05 1.607E+08 1.745E+08 1,740E+08 l 21 7.427E+05 8.500E+08 1.038E+09 1.034E+09 22 6.081E+05 8.319E+08 8.919E+08 8.884E+08 23 4.979E+05 8,257E'08 9.839E+08 9.820E+08 24 3.688E+05 1.350E+09 1.609E+09 1.604E+09 25 2.972E+05 1.530E+09 1,692E409 1.694E+09 26 1.832E+05 1.726E+09 1.870E409 1.862E+09 27 1.lllE+05 1.120E+09 1.150E+09 1.147E+09 28 6.738E+04 8.178E+08 7.908E+08 7.871E+08 29 4.087E+04 2.591E+08 2.583E+08 2.574E+08 L 30 3.183E+04 1.546E+08 1.613E+08 1.607E+08 47 NUREG/CR-6453
l Table B.2 (continued)
Oroup Group upper Neutron flux number energy limit HUGLE-93 SAll.OR 95 BUGLE %
eV cm- 2 ,- l em 3 ,- l em 2, l 31 2.606E404 5.410E+08 5.322E408 5.319E+08 t 32 2.418EM4 3.489E+08 3.299E+08 3.285E+08 j 33 2.188E404 5.325E+08 5.17$E+08 5.163E+08 1 34 1.503E+04 6.552E+08 6.657E+08 6.612E+08 j
35 7.102E+03 6.656E+08 6.775E+08 6.739E+08 36 3.355E+03 5.392E+08 5.508E+08 5.480E+08 37 1.585E+03 7.867E+08 8.201E+08 8.160E+08 38 4.540E+02 3.945E+08 4.130E+08 4.ll2E+08 39 2.144E+02 3.872E+08 4.057E+08 4.041E+08 40 1.013E+02 4.742E+08 4.975E+08 4.957E+08 41 3.727E+01 5.282E+08 5.547E+08 5.529E+08 42 1.068E+01 2.813E+08 2.95$E+08 2.946E+08 43 5.043E400 3.373E+08 3.392E+08. 3.385E+08 44 1.855E+00 2.306E%8 2.133E+08 - 2.135E+08 45 8.764E-01 1.836E+08 1.692E+08 1.692E+08 46 4.140E-01 3.610E+08 2.124E+08 2.123E+08 47 1.000E-01 8.827E+08 4.383E+08 4.397E+08 1.000E-05'
- Low-energy to:ndary of the last group.
NUREG/CR-6453 48 y e-- - ,.- . . - . - . . r- ,,-e w..-_r- awc x--- =-,vv - . . . , - - - e= -- _ - - - - - . , , e , . . - - .r. a -
e+.
1E12 -
S
- 1E10 - . . ~ . . . . . ~ . .
~
1E8 -. . . ~ . . ~ . - .
. E 1E6 - .~ . . . ~ . . . . . . . ~ . . . .
y1E4 - . . ~ . . . . . . .
E
-- . . ~ . . . . . ~ . . ~
h1E2-1EO : : . : : :
1 E-6 1 E-4 1E 2 1EC 1E2 1E4 1E6 1EB Energy [eV)
Fig. H.1 Multigroup neutron spectrum, calculated with HUGLE % library, in the surveillance espsule (20' azimuth, core midplane, at the radius of 191.15 cm from core vertical axis) 3
?.5 - .. . .
- Q-(g;g g g;g .
O =
$ SAILOR 95/ BUGLE 96 m 2 - . . . .. . . . . . . . . . ..
2 -
a g 1,5 -- . . ~ .
o
-. .. ~ .
1 -
l 0.5 . . : : : :
l 12-6 1 E-4 1 E-2 1E0 1E2 1E4 1E6 1E8 Energy (eV)
Fig. H.2 Comparison of multigroup neutron spectra, calculated with different cross-section libraries,in the surveillance capsule (20' azimuth, core midplane, at the radius of 191.15 cm from core veitical axis) 49 NUREG/CR-6453
- . . . __ __ _ ~ . _ - - _ _ . _ _ ._. __-..___ __
l' 1E10 ;
9
- 1E8- - - - ~ - - . . - - -- - -
N h1E6
-. - .~- ~.- - -- . . . . - -
E 1E4 ~ ~ . . - -- - ~ ~.
V> 1E2 .- ..- . . ~ . . ... ..
,1EO - ~ . - - . ~ ~ - . ... ~.. .
\ ,
1 E : : : : :
1 E-6 1E-4 1E 2 1EO 1E2 1E4 1E6 1E8 Energy [nV]
Fig. B.3 Multigroup neutron spectrum, calculated with BUGI E-96 library, at the position of cavity dosimeters (0' azimuth, core midplane, at the radius of 238.02 cm from core vertical axis) 2.2 2 - - . - .-
BUGLE-93/BUGLF. 96
.g1.8- - - . - - - -
=- -
y -
SAILOR-95/ BUGLE 96 u 1.6 - -- - - ~ - . - . - -
.2 u.
gj,4 . . . . . . -. . - . . . ...
e -
0 1.2 - - - - . - - - - --
1 -
. - A,"
0.8 : : : : . !
1 E-6 1 E-4 1 E-2 1E0 1E2 1E4 1E6 1E8 Energy [eV)
Fig. B.4 Comparison of multigroup neutron spectra, calculated with different cross-section libraries, at the position of cavity dosimeterr (0* azimuth, core midplane, at the radius of 238.02 cm from core vertical axis)
te#C romu 336 U.S. kJcLI AR LIGULAf oR Y COMR818 MON 1. RtPoRT fvuutt R 0.En am n52 815UOGRAPHIC DATA SHEET m". ."."*.".'s.",m,1%** *" ~
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- 2. TITLE AV %stilitt H. B. Robinson-2 Preesure Vessel Benchmark 3 DATE REPomT PugLl&HED
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February 1998
- 4. FIN oR ORANT NUMDS A W6164 6.AufHont&t 6. TYPE of AtPomi
- 1. Remec F. B. K. Ram
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Oak Ridge National Laboratory Oak Ridge, TN 37831-6363 es asonerme. seweise nac onsen, pesen se nsamma. u.a ansee .- Communea
- 9. aas 5Po*#So,MtNO ausse ammusu oRGANilATioN - NAMt AND ADDALES ### 4ac. mee 'asam m esm Division of Engineering Technology Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
- 10. SUPPLetetNT AR Y NOTE S C. J. Fairbanks, NRC Project Manager ..
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The HBR-2 benchmark is specified and analyzed in this report. Analysis of the HBR-2 benchmark can be used as partial fulfillment of the requirements for the qualification of the methodology for calculating n-utron fluence in pressure vessels, as required by the U.S. Nuclear Regulatory Commissa .: Regulatory Guide DG-1053, " Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."
Section 1 of this report provides all the dimencions, material compositions,.and neutron source data necessary for the analysis. The measured quantities, to be compared with the calculated values, are the specific activities of the neutron donimeters, on both sides of the pressure vessel: in the surveillance capsule attached to the thermal chield and in the reactor cavity.
Section 2 describes the analysis of the HBR-2 benchmark with the computer code DORT and three ENDF/B-VI based multigroup libraries. The average ratio of the calculated-to-measured specific activities (C/H) for the six dosimeters in the surveillance capsule was 0.90 1 0.04 for all three libraries. The average C/Ms for the cavity dosimeters (without neptunium dosimeter) were 0.89 0.10, 0.91 1 0.10, and 0.90 1 0.09 for the BUGLE-93. SAILOR-95, and BUGLE-96 libraries, respectively, is. aty woRos/otscR:ef ors eu. --me===== - ==-.- - u. avawu r m a n e Neutron dosimetry Unlimited
$5 cv"' " "^"'"'a Neutron transport calculations Qualification of the calculational methodology "aa a'~
Reactor pressure vessel Unclassified ENDF/B-V1 ""a=""
Benchmark Unclassified HBR-2 i&. avaast A o' '^666 16 PRIC6
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