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PSR-0355 NJOY-94.

NJOY-94, General ENDF/B Processing System for Reactor Design Problems

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1. NAME OR DESIGNATION OF PROGRAM:  NJOY94: Code System for
Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.
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2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Only liaison officers are authorised to submit online requests. Rules for requesters are available here.
Program name Package id Status Status date
NJOY-94.61 PSR-0355/02 Tested 28-APR-1997

Machines used:

Package ID Orig. computer Test computer
PSR-0355/02 UNIX W.S. DEC ALPHA W.S.
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3. DESCRIPTION OF PROGRAM OR FUNCTION

The NJOY nuclear data processing system is a comprehensive computer code system for producing pointwise and multigroup cross sections and related quantities from ENDF/B evaluated nuclear data in the ENDF format, including the latest US library, ENDF/B-VI. The NJOY code works with neutrons, photons, and charged particles and produces libraries for  a wide variety of particle transport and reactor analysis codes.

It is capable of processing data in ENDF/B-4, ENDF/B-5, and ENDF/B-6 formats for evaluated data (to the extent that the latter have been frozen at the time of this release).



Short descriptions of the different modules follow:

RECONR  Reconstructs  pointwise (energy-dependent)  cross  sections
        from ENDF/B resonance parameters and interpolation schemes.


BROADR  Doppler broadens and thins pointwise cross sections.



UNRESR  Computes  effective  self-shielded pointwise cross sections
        in the unresolved-resonance region.



HEATR   Generates  pointwise heat production cross sections  (KERMA
        factors) and radiation-damage-energy production cross

        sections.



THERMR  Produces incoherent inelastic energy-to-energy matrices for
        free  or bound scatterers, coherent elastic cross  sections
        for  hexagonal materials, and incoherent elastic cross

        sections.



GROUPR  Generates  self-shielded multigroup  cross sections, group-
        to-group neutron scattering matrices, and photon production
        matrices from pointwise input.



GAMINR  Calculates multigroup photon interaction cross sections and
        KERMA factors and group-to-group photon scattering

        matrices.



ERRORR  Produces  multigroup covariance matrices from ENDF/B

        uncertainties.



COVR    Reads the output of ERRORR and performs covariance plotting
        and output-formatting operations.



DTFR    Formats multigroup  data for transport codes such as DTF-IV
     (5) and ANISN (6).



CCCCR   Formats multigroup  data  for  the CCCC standard  interface
        files ISOTXS, BRKOXS, and DLAYXS.



MATXSR  Formats multigroup data for the  MATXS cross section

        interface file.



ACER    Prepares  libraries  for  the  Los Alamos continuous-energy
        Monte Carlo code MCNP (8).



POWR    Prepares libraries for the EPRI-CELL and EPRI-CPM codes



MODER   Changes  ENDF/B "tapes" and other ENDF-like  NJOY interface
        files back and forth between formatted (i.e., BCD or ASCII)
        and blocked-binary modes.



PLOTR   Plots 2-d graphs of ENDF, PENDF, or GENDF data.
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4. METHOD OF SOLUTION

NJOY94.61 consists of a set of modules, each performing a well-defined processing task. Each of these modules is essentially a separate computer program linked together by input and output files and a few common constants. The methods and instructions on how to use them are documented in the LA-12740-M report on NJOY91 and in the README file. No published document is planned for NJOY94, which is an interim release on the way to NJOY95. NJOY94.0 included all the changes in NJOY91.118, the up119 patch, the PostScript plotting package, the new LEAPR module for computing thermal scattering laws, and a major update to PURR incorporating a new version of the unresolved resonance probability table method including temperature correlations. The November 1995 update to RSIC's package was to replace NJOY94.5 which was packaged in June 1995 with NJOY94.10, which was used to process all of the materials from ENDF/B-VI Release 3 into PENDF format, ACE format, and a 69-group library. In the course of this work, a few additional problems were found and corrected by updates 6-10, which involve changes to the therm, error, acer and njoy modules. Details on the August 1996 NJOY94.35 can be found in the README35 file. In December 1996 the package was upgraded to NJOY94.61.

RECONR:
Reads an ENDF/B tape and produces a common energy grid for all reactions (the union grid) such that all cross sections can be obtained to within a specified tolerance by linear interpolation. Resonance cross sections are calculated with the methods of RESEND (17) but a new method of choosing the energy grid is used which incorporates control of the number of significant figures generated and a resonance-integral criterion to reduce the number of grid points generated for some materials. Summation cross sections (e.g. total, inelastic) are reconstructed from their parts. The resulting pointwise cross sections are written onto a ?point-ENDF" (PENDF) tape for future use.

BROADR:
Reads a PENDF tape and Doppler broadens the data using the method of SIGMA1 (18), modified for better behaviour at high temperatures and low energies. The union grid allows all resonance reactions to be broadened simultaneously resulting in great savings of processing time. After broadening, the summation cross sections are again reconstructed from their parts. The results are written out on a PENDF tape for future use.

UNRESR:
Uses the methods of ETOX (19) to produce effective self-shielded pointwise cross sections, versus temperature and background cross section, in the unresolved-resonance region. The results are added to the PENDF tape in a special format.

HEATR:
Computes both heating and radiation-damage-energy production using momentum balance (for capture) or energy balance (for all other reactions). The ENDF/B photon production files are used in both methods, when available. The heating results are added to the PENDF tape using ENDF/B reaction numbers in the 300 series, and the radiation damage results are the special identifier 444.

THERMR:
Produces cross sections in the thermal range. Bragg edges in coherent scattering are produced using the method of HEXSCAT (16) with an improved treatment at high energies.
Energy-to-energy incoherent scattering matrices can be computed for free scattering or for bound scattering using a precomputed form factor S(alpha,beta) in ENDF format. The secondary angle and energy grids are determined adaptively so as to represent the function to a desired precision by linear interpolation; the angular representation is converted to one based on equally-probable angles. Elastic incoherent scattering is represented using equally-probable angles computed analytically. The results for all the processes are added to the PENDF tape using special formats and reaction numbers.

GROUPR:
Processes the pointwise cross sections produced by the modules described above into multigroup form using the Bondarenko flux weighting model (20). As an option, a pointwise flux solution can be generated for a heavy absorber in a light moderator. Self-shielded cross sections, scattering matrices, and photon production matrices are all averaged in a unified way, the only difference being in the function which describes the feed" into secondary group g' with Legendre order l from initial energy E. The feed for two-body scattering is computed using a centre-of-mass Gaussian integration scheme which provides high accuracy even for small Legendre components of the scattering matrix. Special features are included for delayed neutrons, the coupled angle and energy dependence of the thermal scattering matrix, and the discrete scattering angles arising for thermal coherent reactions. Prompt fission is treated with a group-to-group matrix. The results are written in a special "groupwise-ENDF" format (GENDF) for later use by the output formatting modules.

GAMINR:
Uses a specialized version of GROUPR. Coherent and incoherent form factors (21) are processed in order to extend the useful range of the results to lower energies. Photon heat production cross sections are also generated. The results are saved on a GENDF tape.

ERRORR:
Can either produce its own multigroup cross sections using the methods of GROUPR or start from a pre-computed set. The cross sections and ENDF covariance data are combined in a way which includes the effects of deriving one cross section from several others. Special features are included to process covariances for data given as resonance parameters or ratios (e.g. fission nu-bar).

COVR:
Uses the widely available DISSPLA (22) plotting software to make publication-quality plots (23) of covariance data; it also provides a site for user-supplied routines to prepare covariance libraries for various sensitivity systems.

DTFR:
Is a simplereformatting code which produces cross section tables acceptable to most discrete-ordinates codes. It also converts the GROUPR fission matrix to chi (fission spectrum) and nu-sigma-fission and prepares a photon production matrix if desired. The user can define edit cross sections which are any linear combination of the cross sections on the GENDF tape. This makes complex edits such as gas production possible. DTFR also contains system-dependent plotting routines for the cross sections and P0 scattering and photon production matrices.

CCCCR:
CCCCR is also a straightforward reformatting code. All of the CCCC-IV (7) options are supported. In the cross section file (ISOTXS), the user can choose either isotope chi matrices or isotope chi vectors collapsed during any specified flux. The BRKOXS file includes self-shielding factors for elastic removal. It should be noted that some of the cross sections producible with NJOY are not defined in the CCCC-IV files.

MATXSR:
Reformats GENDF data into the MATXS file format, which is suitable for input to the TRANSX post-processor program. The MATXS format uses flexible naming conventions which allow it to store all NJOY data types except delayed neutron and delayed photon spectra.

ACER:
Prepares a data library tape for MCN, the LASL continuous energy Monte-Carlo code. Reaction cross sections are written out on the grid of the total cross section from the input PENDF tape (assumed to be linearized and unionized). Redundant reactions are removed. Angular distributions (MF4) and tabulated energy distributions (MF5, LF1 or 5) are converted into equal probability bins. The incident energy grids of both are thinned for linear interpolation using the specified tolerance. Analytic secondary energy distributions are copied. All photon production cross sections are combined on the cross section energy grid and written as MF13,MT3. Multigroup photon production cross sections are obtainedfrom the NGEND input tape.

The photon distributions are summed and converted into a set of equally probable mean energies and written as MF15,MT3LF3 (a specially defined law). MF14 is made isotropic. Unresolved region probability tables are generated by a least square fit to the self-shielded cross sections on MT152 (see UNRESR) and written out at MT153 in a special format. The dictionary on the output tape is corrected to reflect the changes.

POWR:
Produces input for the EPRI-CELL codes GAMTAP (fast) and LIBRAR (thermal), and the EPRI-CPM code CLIB.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM:  None noted.
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6. TYPICAL RUNNING TIME

All eight test cases ran in less than ten minutes on an IBM RS/6000 Model 590.
PSR-0355/02
The NEA Data Bank executed the 8 test cases included in this package on a DEC 3000 Model 300x Alpha station. The following execution times (in seconds) were found:
     Sample           User time             System time
      in1               89.08                  10.77
      in2               79.23                  10.38
      in3               34.45                   4.99
      in4               10.18                   3.97
      in5              151.05                  85.28
      in6               29.21                   7.99
      in7               45.65                  19.60
      in8              102.53                  32.71
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7. UNUSUAL FEATURES OF THE PROGRAM:
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8. RELATED AND AUXILIARY PROGRAMS:  UPD 1.3: Portable Update Emulator.
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9. STATUS
Package ID Status date Status
PSR-0355/02 28-APR-1997 Tested at NEADB
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10. REFERENCES:
PSR-0355/02, included references:
- R.E. MacFarlane:
  "README: NJOY 94.10"
  (August 24, 1995)
- R.E. MacFarlane & D.W. Muir:
  The NJOY Nuclear Data Processing System, Version 91
  LA-12740-M (October 1994)
- R.E. MacFarlane:
  New Thermal Neutron Scattering Files for ENDF/B-VI, Release 2
  LA-12639-MS (ENDF 356) (March 1994)
- R.E. MacFarlane & D.C. George:
  UPD: A Portable Version-Control Program
  LA-12057-MS (April 1991)
- R.E. MacFarlane:
  "How to NJOY ENDF-6", The International Workshop on NJOY,
  Saclay, France (April 1992)
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11. MACHINE REQUIREMENTS

NJOY94 runs on Cray computers. Machine-specific updates are included for Cray, VAX, Sun, and IBM RS/6000 and DEC Alpha workstations.
PSR-0355/02
NEA-DB installed the program on a DEC 3000 Model 300x Alpha Station.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
PSR-0355/02 FORTRAN-77
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED

A Fortran 77 compiler is required. This system runs on Cray under either UNICOS or CTSS, on the VAX under VMS, on the Sun under either Sun OS4 or OS5, IBM RS/6000 under AIX and on DEC Alpha under OSF/1. The CFT77 compiler was used under UNICOS. RSIC tested this release on a
SunSparc 5 running SUN OS 5.3 using the Fortran 2.0 compiler and on  the IBM RS/6000 Model 590 under AIX 3.2.5 using the XLF 3.2.2.3 compiler. NJOY 94 produces graphs directly in Postscript and no longer requires the proprietary DISSPLA software.
PSR-0355/02
The test compilation and executions were done at the NEA Data Bank under OSF/1 using the standard Fortran compiler.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS:
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15. NAME AND ESTABLISHMENT OF AUTHORS

Contributed by: Radiation Safety Information Computational Center
                Oak Ridge National Laboratory
                Oak Ridge, Tennessee, U. S. A.

Developed by:   Los Alamos National Laboratory
                Los Alamos, New Mexico, USA
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16. MATERIAL AVAILABLE
PSR-0355/02
File name File description Records
PSR0355_02.001 Information file 164
PSR0355_02.002 README61 - Author's information file 1061
PSR0355_02.003 src - NJOY-94.61 upd source file 69749
PSR0355_02.004 up61 - upd directives input file for 94.61 3152
PSR0355_02.005 upvms - updates for VAX/VMS 71
PSR0355_02.006 upsun - updates for Sun workstations 39
PSR0355_02.007 upuni-la - updates for Los Alamos UNICOS 33
PSR0355_02.008 upuni-sd - updates for SDSCC UNICOS 33
PSR0355_02.009 upnersc - updates for NERSC UNICOS 28
PSR0355_02.010 uprs6k - updates for IBM RS/6000 machines 50
PSR0355_02.011 updecau - updates for DEC alpha Unix WS. 42
PSR0355_02.012 upuser - upd file to extract user instr. 171
PSR0355_02.013 userin - user input instructions 2520
PSR0355_02.014 makef.uni- UNICOS maintenance make file 241
PSR0355_02.015 makef.sun- Sun maintenance make file 236
PSR0355_02.016 makef.alp - DEC alpha maintenance make file 236
PSR0355_02.017 makef.ris - IBM RISC maintenance make file 236
PSR0355_02.018 Install - Unix installation script 24
PSR0355_02.019 gam23 - dlc-7 photon interaction lib, part 1 9930
PSR0355_02.020 gam27 - dlc-7 photon interaction lib, part 2 5222
PSR0355_02.021 t322 - scattering law data from ENDF/B-III 15284
PSR0355_02.022 t404 - U234 data from ENDF/B-IV 17490
PSR0355_02.023 t511 - materials from ENDF/B-V 21769
PSR0355_02.024 eni61 - Ni61 data from ENDF/B-VI 6984
PSR0355_02.025 upd.f - version control code for NJOY update 1643
PSR0355_02.026 in1 - test problem 1 input data 85
PSR0355_02.027 out1 - test problem 1 printed output 1995
PSR0355_02.028 in2 - Test problem 2 input data 83
PSR0355_02.029 out2 - test problem 2 printed output 7720
PSR0355_02.030 in3 - test problem 3 input data 56
PSR0355_02.031 out3 - test problem 3 printed output 1388
PSR0355_02.032 in4 - test problem 4 input data 41
PSR0355_02.033 out4 - test problem 4 printed output 584
PSR0355_02.034 in5 - test problem 5 input data 30
PSR0355_02.035 out5 - test problem 5 printed output 63072
PSR0355_02.036 in6 - test problem 6 input data 104
PSR0355_02.037 in7 - test problem 7 input data 48
PSR0355_02.038 out7 - test problem 7 printed output 18786
PSR0355_02.039 in8 - test problem 8 input data 51
PSR0355_02.040 out8 - test problem 8 printed output 23832
PSR0355_02.041 acer.ps - PostScript documentation 3336
PSR0355_02.042 broadr.ps - PostScript documentation 4147
PSR0355_02.043 ccccr.ps - PostScript documentation 4030
PSR0355_02.044 covr.ps - PostScript documentation 1882
PSR0355_02.045 dtfr.ps - PostScript documentation 5561
PSR0355_02.046 errorr.ps - PostScript documentation 7807
PSR0355_02.047 front.ps - PostScript documentation 1163
PSR0355_02.048 gaminr.ps - PostScript documentation 3181
PSR0355_02.049 groupr.ps - PostScript documentation 12988
PSR0355_02.050 heatr.ps - PostScript documentation 9908
PSR0355_02.051 index.ps - PostScript documentation 1003
PSR0355_02.052 intro.ps - PostScript documentation 2022
PSR0355_02.053 matxsr.ps - PostScript documentation 3234
PSR0355_02.054 mixr.ps - PostScript documentation 840
PSR0355_02.055 moder.ps - PostScript documentation 731
PSR0355_02.056 njoy.ps - PostScript documentation 2343
PSR0355_02.057 plotr.ps - PostScript documentation 7566
PSR0355_02.058 powr.ps - PostScript documentation 735
PSR0355_02.059 purr.ps - PostScript documentation 740
PSR0355_02.060 reconr.ps - PostScript documentation 4092
PSR0355_02.061 resxsr.ps - PostScript documentation 822
PSR0355_02.062 thermr.ps - PostScript documentation 4542
PSR0355_02.063 unresr.ps - PostScript documentation 2821
PSR0355_02.064 wimsr.ps - PostScript documentation 2065
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17. CATEGORIES
  • B. Spectrum Calculations, Generation of Group Constants and Cell Problems
  • C. Static Design Studies
  • J. Gamma Heating and Shield Design
  • K. Reactor Systems Analysis

Keywords: Doppler broadening, Monte Carlo method, charged particles, covariance matrices, data uncertainties, discrete ordinate method, gamma radiation, kerma factors, kernels, neutrons, photon interaction, resonance integrals, scattering, self-shielding, thermal neutrons, thermal scattering, transport theory, unresolved region.