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SINBAD ABSTRACT NEA-1517/71

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."         
  
 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 × 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 × (pi / pDORT) × (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 superceeded 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

    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 
      


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.

     The tables are in Excel format, and the figures in pdf format.


SINBAD Benchmark Generation Date: 6/2004
SINBAD Benchmark Last Update: 6/2004