H.B. Robinson-2 Pressure Vessel Dosimetry Benchmark
1. Name of Experiment:
------------------
H.B. Robinson-2 In- and Ex-Vessel Neutron Dosimetry Experiment
2. Purpose and Phenomena Tested:
----------------------------
The in- and ex-vessel neutron dosimetry measurements were performed at the
H.B. Robinson-2 (HBR-2) nuclear power plant, which is a commercial pressurized
light-water reactor designed by Westinghouse. HBR-2 is a three loop reactor with
2300-MW (thermal) power. It was placed in operation in March of 1971, and is owned
by Carolina Power and Light Company.
The measurements served several purposes. By performing the measurements on both sides
of the pressure vessel, that is, in the surveillance capsule located in the downcomer
region between the thermal shield and the pressure vessel, and outside the pressure
vessel, in the cavity between the vessel and the biological shield, the consistency
between the (traditional) in-vessel dosimetry and the cavity dosimetry (which was new
at the time of these experiments) was tested. The cavity measurements were used to
experimentally verify the effectiveness of the low-leakage fuel loading pattern which
was introduced in the HBR-2 in fuel cycle 9 to reduce the pressure vessel irradiation.
The measurements also provided a means to test the ability of neutron transport
calculations to predict the through the wall attenuation.
The measurements were used to prepare the "H. B. Robinson-2 Pressure Vessel Benchmark"
(Ref. 1), which can be used for partial 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 1.190, "Calculational
and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."
This package was updated in regards to the benchmark calculation inputs of TORT-3.2
(court. of R. Orsi, ENEA Italy). The summary of the update is reported on “H.B. Robinson-2
Pressure Vessel Dosimetry Benchmark: Summary of ENEA-Bologna Three-Dimensional Deterministic
Analyses, was added to the material” (SICNUC-P9H6-005). The thorough analysis is described in
ENEA-Bologna SICNUC-P9H6-003.
The input files of MCNP5 were also added (court. of P. Ortego, SEA Spain).
3. The H. B. Robinson-2 Reactor
----------------------------
The data presented in this section were obtained from Refs. 2 and 3, and from personal
communications (Refs. 4 and 5).
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.
An octant of the horizontal cross-section of the reactor is shown schematically in
Fig. 1, which also shows the locations of the capsule and cavity dosimeters.
Axial geometry and dimensions are given in Fig. 2. The core baffle geometry is further
specified in Fig. 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. 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 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 2. The regions given in Table 2 correspond to the ones shown in Figs. 1 and 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 2.
The densities and chemical compositions of the other materials are given in Table 3.
The concrete of the biological 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. 2).
4. CORE POWER DISTRIBUTION AND POWER HISTORY
-----------------------------------------
The fuel assemblies in the core are numbered as shown in Fig. 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.
For each assembly in the core, the mass of uranium, burnup at the beginning of cycle
life (BOL) and end of cycle life (EOL), burnup increment in cycle 9, and cycle-average
relative power are listed in the data file FILE1.DAT. Part of the file is shown in
Fig. 6. These data were taken from the TOTE output, except for the cycle-average
assembly power, which 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 FILE2.DAT.
Part of FILE2.DAT is shown in Fig. 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 burnup distributions given in the TOTE output for each assembly.
The cycle-average assembly-pin-power distributions are given in FILE3.DAT.
The content of the file is illustrated in Fig. 8. Distributions are given
for the assemblies 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 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
corner of the assembly, the last value in row 1 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.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 burnup increment from
the midpoint of the (I-1)-th and I-th burnup step and I-th and (I+1)-th burnup step.
For cycle 9, a low-leakage core loading pattern 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 changed significantly. 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 the core neutron leakage and consequently the dosimeter
reaction rates. For this reason, the cycle-average core power distribution data,
described previously, are supplemented 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 (FILE4.DAT, see
Fig. 9), assembly burnups (FILE5.DAT, see Fig. 10), pin-power distributions for the
assemblies in the upper left quadrant of the core (FILE6.DAT, see Fig. 11),
and assembly-wise axial power distributions in 12 axial segments (FILE7.DAT,
see Fig. 12).
The core power history for cycle 9 is given in the FILE8.DAT as is illustrated
in Fig. 13.
Descriptions of the contents and formats of the files are given at the end of each
file and are shown in Figs. 6-13.
5. Dosimetry Measurements
----------------------
During cycle 9, comprehensive sets of dosimeters were irradiated in the surveillance
capsule position and in several locations in the reactor cavity (Ref. 3). 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 at the 20° azimuthal angle location in the downcomer. The region that
usually contains metallurgical specimens was filled with carbon steel, and the
dosimeters were installed in the holes drilled in the steel. Specific activities
given in Table 4 are for the core-midplane set. (Dosimetry sets were installed
in the capsule at the core midplane and approximately 28 cm (11 in.) above and
below the midplane. The measured activities showed axial variations of only ~ 3%,
which is not considered important, and therefore only the results for the midplane
set are given. Radially, the dosimeters were installed at the capsule centerline
at the radius of 191.15 cm (see Fig. 4). The specially built capsule was
irradiated during cycle 9 only.
Specific activities of the cavity dosimeters irradiated at 0° azimuth, on the core
midplane, are also given in Table 4. In the present benchmark, only the midplane
measurements are considered. However, at the 0° azimuth multiple dosimeter sets were
irradiated at the midplane and at 213 cm ( 7 ft) and 107 cm (3.5 ft) above and below
the midplane; and activities of the gradient wire [54Fe(n,p)54Mn and 58Ni(n,p)58Co
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.
The dosimeters were 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 perturbations 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 insulation and the biological
shield at a radius of 238.02 cm (93.71 in.). The sketch of the 0° azimuth cavity
dosimetry axial locations is given in Fig. 14.
Specific activities listed in Table 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. 2) are given in the footnotes to Table 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.).
6. Neutron Transport Calculations:
------------------------------
The transport calculations were performed using the DORT computer code (Ref. 6) 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) 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 transport calculations. One 2-D calculation models
the horizontal cross section of the reactor in the r - θ 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 - z geometry, in which a core is modeled as a finite-height cylinder. The third
calculation is made for the 1-D (r) cylindrical model of the reactor. The r - z 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 analysis (Refs. 2, 7). The r - θ model covered one octant of the
horizontal cross section with 74 azimuthal (θ) 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 - z 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 - z 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 - z model.
For the transport calculations, the cross sections of the macroscopic mixtures were
prepared by the GIP code (Ref. 8), using the homogenized zone compositions given in
Table 2 of this report. The P3 approximation to the angular dependence of the
anisotropic scattering cross sections (i.e., the P0 to P3 Legendre components) were
taken into account, and a symmetric S8 "directional quadrature set" (i.e., a set of
discrete directions and angular quadratures) were used for all transport calculations.
The benchmark was analyzed with three cross-section libraries based on ENDF/B-VI:
BUGLE-93 (Ref. 9), SAILOR-95 (Ref. 10). and BUGLE-96 (Ref. 11), which have 47 neutron
and 20 gamma energy groups.
The neutron sources for the r - θ, r - z and one-dimensional calculation were
prepared by the DOTSOR code (Ref. 12). For the r - θ 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 - θ geometry
mesh. The power-to-neutron-source conversion factor was based on the average burnup 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 235U and 239Pu to the fission neutron
source. The power-to-neutron-source conversion factor of 8.175E+16 neutrons s-1MW-1
was calculated by the DOTSOR code for the fuel burnup of 28596 MWd/MTU, which
corresponds to the cycle-average burnup of the fuel assemblies number 71, 86, and 101.
The source energy spectrum was taken as the average of 235U and 239Pu fission spectra.
The ENDF/B-VI fission spectra for 235U and 239Pu were used.
The source for the r - z calculation was generated by averaging the cycle-average pin
powers of the top halves of the fuel elements 79 to 86 over the y axis (see Fig. 1;
the y axis is perpendicular to the 0° radial direction) and multiplying the average
pin-power values by the cycle-average axial power distribution of the corresponding fuel
assembly. The x - z power distribution obtained was then transformed into r - z mesh
by the DOTSOR code, which also prepared the source for 1-D r calculation by integrating
the r - z source over the z axis. The same source energy spectrum as in the r - θ
calculation was used for the r - z and r calculations.
The input for the GIP code is HBR2-GIP.INP; the input for the DORT 1-D r- calculation
is HBR2-1D.INP, the input for the DORT 2-D r - θ calculation is HBR2-RT.INP; and the
input for the DORT 2-D r - z calculation is HBR2-HBR2-RZ.INP.
From the three transport calculations, the neutron fluxes in the core midplane, in
the surveillance capsule at the azimuth of 20°, and in the cavity at the azimuth of 0°
were synthesized. For this purpose the code DOTSYN was used (Ref. 13). The DOTSYN inputs,
used to synthesize the fluxes at the capsule and cavity location are SYN-CAPSULE.INP,
and SYN-CAVITY.INP.
Reaction rates were calculated with the CROSS-95 dosimetry library (Ref. 14). 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. 15) and the reference spectra as
calculated in the capsule and cavity location. The 47-group cross sections which were
used to calculate the reaction rates are given in are given in CAPSULE-XS.DAT and
CAVITY-XS.DAT for the dosimeters located in the capsule and in the cavity, respectively.
To calculate the specific activities at the end of irradiation, which are the measured
quantities provided for 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 neutron leakage 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.
Reaction rates at the dosimetry location are affected mostly by the closest fuel
assemblies. Therefore, for the cavity-dosimetry location, the following approximation
was used. The cycle was divided into eight 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 second power
distribution provided, the second interval from the end of the first interval to halfway
between the second and third power distribution, etc. The average relative power (pi) of
the three fuel elements 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 (pi) 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 FILE1.DAT). Using the daily power history, the reaction rate
was then approximated as
Rj = Rc x (pi / pDORT) x (Pj / Po),
Where:
Rj = reaction rate at cavity location during j-th day,
Rc = reaction rate obtained from transport (DORT) calculations,
for nominal core power,
pi = normalized average relative power of the fuel elements 71, 86, and 101
during i-th time interval,
pDORT = average relative power of the fuel elements 71, 86, and 101
used in the transport calculation (DORT),
Pj = daily-average reactor core power during day j. (Day j is in the
time interval i).
Po = nominal core power (2300 MW).
The specific activities are easily obtained by numerically integrating the daily
reaction (Rj) rates over the reactor power history, and accounting for the decay
of the measured isotope to the end of cyle 9.
The same procedure was used for the calculation of activities of the dosimeters in
the surveillance capsule; however, the fuel assemblies 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 activities
is further discussed in Appendix A of "H. B. Robinson-2 Pressure Vessel Benchmark".
7. Results and Discussion:
----------------------
The reaction rates calculated as described in the previous subsection, for the
cycle-average power distribution, are given in Table 5. The reaction rates
obtained from the transport calculations with the BUGLE-93, SAILOR-95 and BUGLE-96
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 63Cu(n,a) and 46Ti(n,p) reactions are practically
identical to those obtained by the other two libraries, while BUGLE-93 reaction rates
for 54Fe(n,p), 58Ni(n,p), 238U(n,f), and 237Np(n,f) are 1%, 2%, 4% and 10% lower,
respectively, than reaction rates calculated with the other two libraries.
The specific activities, given in Table 6 were calculated as described in
subsection 5. Conversion from reaction rates to specific activities does not affect
the differences between results obtained by different cross-section libraries;
therefore, for the comparison of specific activities the comments given above for the
reaction rates apply.
The comparison of calculations and measurements is given in Table 7 which lists
the ratios of the calculated and measured specific activities. The calculated specific
activities are taken from Table 6, and the measured specific activities are taken from
Table 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 4, are applied to the measured activities of the 237Np, 238U,
and 63Cu 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 activities are listed in Table 7 in parentheses.
In the cavity the C/M ratio for the 237Np dosimeter is significantly lower than C/M
ratios for other dosimeters, regardless of the cross-section library used. This
well-known problem of the HBR-2 cycle 9 cavity dosimetry measurements was addressed
in several analyses, but has not been completely explained. The most probable explanation
appears to be an incorrect measured value. Therefore, the average C/M values in the
cavity, given in Table 7 in the last column on the right, were calculated without the
Np dosimeter.
The average C/M 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. (If the 237Np dosimeter
in the cavity is taken into account, the average C/M values are 0.83 +/- 0.18,
0.85 +/- 0.16, and 0.85 +/- 0.16, for the BUGLE-93, SAILOR-95, and BUGLE-96 library,
respectively.) The C/M ratios given in parentheses are for the measured activities
of 237Np, 238U, and 63Cu dosimeters, corrected as discussed in notes to Table 4. The
average C/M ratios in the cavity are practically identical to those in the capsule;
therefore, no decrease in the C/M ratios with increasing distance from the core and
increasing thickness of steel penetrated is observed. Such decrease was typical for
the pre-ENDF/B-VI cross-section libraries and is illustrated in Appendix A of
"H. B. Robinson-2 Pressure Vessel Benchmark".
The variations of the C/M values for different dosimeters at the same location are
small: the standard deviation of the average C/M ratios is ~0.04 in the capsule and
~0.10 in the cavity. These values suggest that the shapes of the calculated spectra,
in the energy range to which the measured dosimeters are sensitive, are adequate.
Further assessment of the differences between the three libraries can be found in
the Appendix B of "H. B. Robinson-2 Pressure Vessel Benchmark". Since the publication
of the Ref. 1 the BUGLE-93 library was superseded by the BUGLE-96 library. The
BUGLE-96 library is at present the broad-group library recommended by the American
National Standards Institute and American Nuclear Society for reactor pressure
vessel neutron flux calculations (Ref. 16).
8. Special Features:
----------------
1. Experimental reference data obtained by simultaneous in-vessel and
ex-vessel dosimetry experiment.
2. Detailed reactor and neutron source description.
3. Neutron transport calculations by discrete ordinates method
and intercomparison of results for different cross-section libraries.
4. Comparison between calculated and experimental reference data.
9. Author/Organizer:
----------------
The dosimetry experiment at the HBR-2 was performed as a cooperative venture
between Carolina Power and Light Company and the United Stated Nuclear Regulatory
Commission sponsored Light Water Reactor Pressure Vessel Surveillance Dosimetry
Improvement Program (LWR-PV-SDIP). The in-vessel experiment was in part funded
by the LWR-PV-SDIP with additional support from Electric Power Research Institute.
The "H. B. Robinson-2 Pressure Vessel Benchmark" was prepared by the Oak Ridge
National Laboratory and was sponsored by U.S. Nuclear Regulatory Commission, Office
of Nuclear Regulatory Research, Divison of Engineering Technology.
Compiler of data for Sinbad:
I. Remec
Oak Ridge National Laboratory
P.O. Box 2008
Oak Ridge, TN 37831-6172, USA
Reviewer of compiled data:
I. Kodeli
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France
10. Availability:
------------
Unrestricted
11. References:
----------
[ 1] I. Remec and F. B. K. Kam, "H. B. Robinson-2 Pressure Vessel Benchmark,
"NUREG/CR-6453, ORNL/TM-13204, October 1997.
[ 2] R. E. Maerker, "LEPRICON Analysis of the Pressure Vessel Surveillance
Dosimetry Inserted into H. B. Robinson-2 During Cycle 9," Nuc. Sci.
Eng., 96:263 (1987).
[ 3] E. P. Lippincott et al., Evaluation of Surveillance Capsule and Reactor
Cavity Dosimetry from H. B. Robinson Unit 2, Cycle 9, NUREG/CR-4576
(WCAP-11104), Westinghouse Corp., Pittsburgh, Pa., February 1987.
[ 4] S. L. Anderson, Westinghouse Electric Corporation, personal communication
to I. Remec, Oak Ridge National Laboratory, 1996.
[ 5] R. M. Kirch, H. B. Robinson Steam Electric Plant, Unit No. 2, response
to request for information regarding operating cycle 9, personal communication to J. V. Pace,
Oak Ridge National Laboratory, Oct. 1, 1996.
[ 6] W. A. Rhoades et al., "TORT-DORT Two- and Three-Dimensional Discrete
Ordinates Transport, Version 2.8.14," CCC-543, Radiation Shielding
Information Center, Oak Ridge National Laboratory, 1994.
[ 7] M. L. Williams, M. Asgari, F. B. K. Kam, Impact of ENDF/B-VI Cross-Section
Data on H. B. Robinson Cycle 9 Dosimetry Calculations, NUREG/CR-6071
(ORNL/TM-12406), October 1993.
[ 8] W. A. Rhoades,"The GIP Program for Preparation of Group-Organized
Cross-Section Libraries," informal notes, November 1975, RSIC Peripheral
Shielding Routine Collection PSR-75.
[ 9] 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 Library Collection, DLC-175, February 1994.
[10] M. L. Williams, M. Asgari, and H. Manohara, "Letter Report on Generating SAILOR-95 Library,"
personal communication to F. B. K. Kam, ORNL, February 1995.
[11] 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.
[12] M. L. Williams, "DOTSOR: A Module in the LEPRICON Computer Code System for Representing the
Neutron Source Distribution in LWR Cores, EPRI Research Project 1399-1 Interim Report (December 1985),
RSIC Peripheral Shielding Routine Collection PSR-277.
[13] M. L. Williams, P. Chowdhury, B. L. Broadhead, "DOTSYN: A Module for
Synthesizing Three- Dimensional Fluxes in the LEPRICON Computer Code
System," EPRI Research Project 1399-1 Interim Report (Dec. 1985); RSIC Peripheral Shielding Routine
Collection PSR-277.
[14] I. Remec and F. B. K. Kam, An Update of the Dosimetry Cross-Section Data Base
for the Adjustment Code LSL-M2, ORNL/NRC/LTR-95/20, June 1995.
[15] F. W. Stallmann, LSL-M2: A Computer Program for Least-Squares Logarithmic Adjustment of Neutron Spectra,
NUREG/CR-4349 (ORNL/TM-9933), March 1986.
[16] ANSI/ANS-6.1.2-1999: Neutron and Gamma-Ray Cross Sections for Nuclear Radiation Protection
Calculations for Nuclear Power Plants.
[17] The ENEA-Bologna technical report SICNUC-P9H6-003:
H.B. Robinson-2 Pressure Vessel Dosimetry Benchmark - Deterministic Analysis in Both Cartesian
(X,Y,Z) and Cylindrical (R,θ,Z) Geometries Using the TORT-3.2 3D Transport Code,
the BUGJEFF311.BOLIB, BUGENDF70.BOLIB, BUGLE-B7 and the BUGLE-96 Cross Section Libraries.
[18] The ENEA-Bologna technical report SICNUC-P9H6-005:
H.B. Robinson-2 Pressure Vessel Dosimetry Benchmark: Summary of ENEA-Bologna Three-Dimensional Deterministic Analyses
[19] R. Orsi, H.B. Robinson-2 Pressure Vessel Dosimetry Benchmark:
Deterministic three-dimensional analysis with the TORT transport code, Nuclear Engineering and Technology,
https://doi.org/10.1016/j.net.2019.07.025
12. Data and Format:
----------------
FILE NAME bytes Content
---- ----------- ------ ----------------------------------
1 HBR-2.htm 37,927 This file.
2 FILE1.DAT 8,773 Cycle 9, Assembly power and burnup.
3 FILE2.DAT 19,085 Cycle average assembly axail-segment-powers.
4 FILE3.DAT 79,368 Cycle-average assembly pin powers.
5 FILE4.DAT 10,483 Core power distributions at eight burnups.
6 FILE5.DAT 12,240 Assembly burnups at eight core burnups.
7 FILE6.DAT 654,947 Assembly pin-powers at eight core burnups.
8 FILE7.DAT 163,933 Assembly axial power distributions at eight burnup steps.
9 FILE8.DAT 21,352 Reactor daily power history for Cycle 9.
10 HBR2-GIP.INP 5,929 Input for cross-section preparation (GIP code).
11 HBR2-1D.INP 4,900 Input for the DORT 1-D r- calculation.
12 HBR2-RT.INP 95,504 Input for the DORT 2-D r - θ calculation.
13 HBR2-RZ.INP 29,421 Input for the DORT 2-D r - z calculation.
14 SYN-CAPSULE.INP 45,606 Input for DOTSYN to synthesize the fluxes at the capsule location.
15 SYN-CAVITY.INP 45,604 Input for DOTSYN to synthesize the fluxes at the cavity location.
16 CAPSULE-XS.DAT 4,077 Reaction cross-sections for capsule location.
17 CAVITY-XS.DAT 4,077 Reaction cross-sections for cavity location.
18 Table1.xls 20,992 Selected general data and dimensions of the HBR-2.
19 Table2.xls 18,342 Materials of the components and regions.
20 Table3.xls 15,360 Densities and chemical compositions of materials.
21 Table4.xls 15,872 Measured specific activities of the dosimeters.
22 Table5.xls 14,336 Calculated reaction rates.
23 Table6.xls 14,848 Calculated Specific Activities.
24 Table7.xls 16,896 Ratios of calculated-to-measured (C/M) specific activities.
25 Fig. 1 48,355 Horizontal cross section of the HBR-2 reactor.
26 Fig. 2 33,021 Schematic cketch of the axail geometry.
27 Fig. 3 12,584 Core baffle geometry.
28 Fig. 4 16,663 Sketch of the surveillance capsule mounting on the thermal shield.
29 Fig. 5 9,491 The numbering of the fuel elements in the HBR-2 core.
30 Fig. 6 8,357 Content and format of the FILE1.DAT.
31 Fig. 7 32,597 Content and format of the FILE2.DAT.
32 Fig. 8 9,722 Content and format of the FILE3.DAT.
33 Fig. 9 8,863 Content and format of the FILE4.DAT.
34 Fig. 10 9,082 Content and format of the FILE5.DAT.
35 Fig. 11 10,110 Content and format of the FILE6.DAT.
36 Fig. 12 8,616 Content and format of the FILE7.DAT.
37 Fig. 13 43,419 Content and format of the FILE8.DAT.
38 Fig. 14 26,286 Schematic drawing of the axail positions of the cavity dosimeters.
39 HBR2.pdf 664,954 H. B. Robinson-2 Pressure Vessel Benchmark, NUREG/CR-6453.
40 ENEA_Bologna_HBR2.tar.gz 48,743,030 Input data for TORT-3.2.
41 ENEA_Bologna_HBR2_file_list.txt 2,549 Contents of ENEA_Bologna_HBR2.tar.gz.
42 SICNUC-P9H6-003.pdf 15,180,687 Reference 17.
43 SICNUC-P9H6-005.pdf 2,183,188 Reference 18.
44 R.Orsi_NET2019.pdf 2,212,710 Reference 19.
45 HBRHE74 44,731 Input data for MCNP5.
46 HBRHE81 39,458 Input data for MCNP5.
47 INPUT_EXPLANATION.pdf 6,741 Memorandum of the MCNP5 input files
The tables are in Excel format, and the figures in pdf format.
SINBAD Benchmark Generation Date: 6/2004
SINBAD Benchmark Last Update: 11/2019