Computer codes for shielding calculations — 1969

NUCLEAR ENGINEERING AND DESIGN 10 (1969) 505-517. NORTH-HOLLAND PUBLISHING COMPANY, AMSTERDAM

COMPUTER

CODES FOR SHIELDING CALCULATIONS

- 1969"

D.K.TRUBEY and Betty F.MASKEWlTZ Radiation Shielding Information Center, Oak Ridge National Laboratory, Oak Ridge. Tennessee 37830, USA

Received 16 June 1969 An extensive library of computer codes useful for radiation transport or shielding calculations is available from the Radiation Shielding Information Center at Oak Ridge National Laboratory. In addition to the point kernel, Monte Carlo, and discrete ordinates codes used for neutron and gamma-ray transport calculations, the collection includes cross-section libraries and codes for processing cross sections, calculating fission product inventories, proton penetration of spacecraft, electron-photon transport, and analyzing neutron activation detector data to determine spectra. A list of the most current codes is given and essential information for each is included.

1. Introduction The Radiation Shielding Information Center [ 1] at Oak Ridge National Laboratory, as an integral part of its information processing activities, collects, makes operable, packages, and distributes computer code packages to nuclear scientists and engineers engaged in shielding research or design. The various codes are designed for calculations related to radiation from reactors, radioisotopes, weapons and accelerators and to radiation occurring in space. The Center uses the word "package" to mean all the items needed to utilize a code effectively. The package normally includes documentation describing the theory and code operation (contributor's report plus RSIC abstract and notes) and one or more reels of tape on which is written the source program, operating (binary or hex) program, input and output for a sample problem, data libraries, and auxiliary routines. A list of the first 60 codes packaged was published in 1966 [2]. Codes collected since that time and selected ones from the previous table are given in table 1. Each is identified by a CCC (Computer Code Collection) number and by name. * Research sponsored by the U.S. Atomic Energy Commission under contract with Union Carbide Corporation. The work is also supported by the National Aeronautics and Space Administration and the Defense Atomic Support Agency.

Most of the new codes are written in FORTRAN IV, which makes them nearly machine independent. Several of the packages actually represent coding systems. These are represented in the collection by prototypes, which are not necessarily useful in themselves, but which achieve generality in that they are designed to be easily changed. Such code systems are most useful to the research worker who will invest a great deal of effort in learning to use the system. The radiation treated by the majority of the codes is either neutron or gamma radiation, but some codes are for charged particles. The types of geometry treated vary widely, with a few codes allowing quite a general geometry. 2. Methods of solution for neutron and gamma-radiation transport problems The neutron and gamma-ray codes are used for nuclear reactor, shelter, radioisotope, and neutron generator shielding. The energy range considered is generally less than about 15 MeV. Most of these codes incorporate a certain way of solving the Boltzmann transport equation (or use results of such solutions). Some codes calculate dose rates while others calculate energy spectra and angular distributions as well. The most generally used methods are outlined below. For further information on these

506

D.K.TRUBEY and B.F.MASKEWITZ

methods, see the recently published Engineering Compendium on Radiation Shielding [3].

2.1. A ttenuation kernel integration If the dose rate at various distances from a point isotropic source is known from experiment or calculation, one can integrate over any source region to obtain the dose rate at given points [4]. That is

D(R) : f S(R')K(R',R)dR' v where V = volume containing the source D(R) = dose rate at R S(R') = source strength at R' K(R',R) = dose rate at R due to a source at R ' . Often the kernel K(R',R) is derived from infinite medium moments method results or AlbertoWelton [5] kernels based on experiment. Codes using this method are very fast (and thus are often used for parameter studies) and can give reasonably good results, especially for gamma-ray transport if the system is essentially homogeneous. In the case of neutrons, the system must also be hydrogenous. This method can account for inhomogeneities only very approximately. 2.2. Monte Carlo method The Monte Carlo [6] method is very popular, especially as a research tool, because any geometry or process can be accounted for in principle. This technique might be described as a theoretical experiment in which many particles are released from the source and their subsequent life histories traced. The main problem with this method is the difficulty of ascertaining that sufficient samples have been obtained. Often straight-forward (analog) methods will not yield sufficient sampling of histories important to the answer desired. Thus complicated importance sampling (biasing or variance reduction) techniques are introduced, but these do not necessarily succeed in improving results.

The attitude that large amounts of computer time are needed for Monte Carlo is not generally valid since the costs of most calculations are largely programming costs rather than machine costs. It is clear that Monte Carlo becomes more attractive with each generation of computers. For 3-dimensional problems, this is the only rigorous method. 2.3. Discrete ordinates method Most codes in this class are derived from the work of Carlson and associates at Los Alamos and are often called Carlson S n codes [7]. These codes were originally developed to calculate the reactivity of nuclear chain reactors, but are now being routinely applied to neutron and gamma-ray [8] shielding problems. In the general case, tire transport equation involves tile variables of position R, direction ~ , energy E, and time t, and, dependent on these, the neutron flux qS. The discrete ordinates method, which basically is a quad-rature method over ~ , is applied to the transport equation as one of several steps in the solution of that equation. More explicitly, it furnishes a means for replacing 4~by a set of functions q~m,m = 1,2 . . . . . Mn, not dependent on ~ . The ~bm ' s are the ordinates in the quadrature and n specifies the order of precision. The effect of applying the method is that an integrodifferential equation in the angular flux ¢ is transformed into a set of much simpler differential equations in tire ordinate fluxes ~bm . The energy dependence is handled by a "maltigroup" treatment. That is, cross sections averaged over an energy range, or group, must be supplied and the output directional fluxes are averaged over each group. For 1-dimensional problems, this is usually the recommended method since computer time is short even for relatively many groups and high order angular quadrature. For 2-dimensional problems, computer time is much greater and the only clear advantage over Monte Carlo is that statistical variation is avoided. Several 2dimensional codes are now available. 2.4. Spinney method The removal-diffusion method pioneered by the late K.T.Spinney [3] has been further developed along different lines by several groups. In this tech-

COMPUTER CODES nique the transport of neutrons is divided into two parts - an exponential attenuation (removal) phase for the fast neutrons and a multigroup diffusion phase for the more isotropic "diffusion" neutrons. The subsequent secondary gamma-ray transport is usually treated by a kernel method. The method has been extended to two dimensions [9]. The removal part, being a kernel method, can be easily extended to three dimensions and so the method is limited in dimensions to the diffusion theory part. 2.5. Others Other methods are presently being used successfully, especially for 1-dimensional problems, such as invariant imbedding and spherical harmonics, but these are not yet represented in the RSIC collection. The moments method, from which most buildup factors were derived, is represented in the collection but not, at present, by a modern code. There are also codes in the RSIC collection which solve the single scattering problem, integrate emerging radiation from a surface, determine fission product inventories, optimize shield weight, solve design problems of shipping casks for radioactive material, determine radioactive fallout protection factors in buildings, determine neutron spectra from activation measurements, or solve other problems connected with shielding. Codes which process nuclear cross sections for the transport codes are also being acquired and distributed by RSIC.

3. Methods of solution for charged particle transport problems The charged particle transport codes were designed to solve shielding problems connected with spacecraft design and charged particle radiation transport in simple geometries. The most powerfulmethod is Monte Carlo because it can treat secondary radiation rigorously, provided the cross sections are known. However, simpler treatments are often utilized and can be much faster, especially for complicated geometries. These generaUy involve ray-tracing and evaluate dose by using a kernel or slowing down theory (stopping power, dE/dx). The proton dose can be adequately computed by assuming that the primary particles proceed through matter straight ahead

507

[ 10-12]. Secondary particles, including cascade protons, cascade neutrons, and evaporation neutrons are treated in an approximate way. That is, the highenergy "cascade" particles from non-elastic collisions are emitted in the same direction as the incident particle (straight ahead) and lower-energy secondary particles are produced isotropically. Some of the codes useful for space shielding studies determine the primary radiation incident on earth-orbiting vehicles by taking into account such things as the trajectory, radiation belts, and the geomagnetic storm associated with a solar flare.

4. General The Center attempts to maintain the most up-todate version of each code in the collection. If errors are corrected or improvements made by the contributor, the changes are incorporated in the code and announced in the Center's monthly newsletter. In general each program is written for a certain purpose and with certain assumptions. Hence there is seldom an attempt at intercomparison by running all codes on a given problem, but rather the Center will recommend particular codes best suited to a requester's purposes. When comparisons are made by the Center, the emphasis is on method comparison. In this procedure, one or two of the most suitable codes are selected that represent each particular method, and the results are compared. RSICcooperates with other information analysis and computer code centers, such as the Argonne Code Center at Argonne National Laboratory, Illinois, and the ENEA Computer Programme Library, at Ispra, Italy. Shielding codes collected at the latter library are also available at RSIC, and European requesters obtain RSIC codes through ENEA-CPL. More detailed information on the codes listed here is available in ref. [13].

Acknowledgement The authors gratefully acknowledge the dedicated service of Hemma Francis, Henrietta Hendrickson, Juanita Brown, Vivian Jacobs and Marie Anthony of the RSIC code center staff who make this service possible.

ORNL-N

ORNL-N

7-B/NTC

17-A/05R

LASL-T

42-A/DTF-IV

48-B/QAD

48-A/QAD

46-C/OGRE 46-D/OGRE

46-B/OGRE

46-A/OGRE

LASL-N, NASA-LE, BE BE,ORNL-N, BC-H

ORNL-N

ANLb LASL-T & AFWL

LG

28/FPIC

42-B/DTF-IV 42-C/DTF-IV 42-D/DTF-IV

AFWL ORNL-N GT

17-C/05R 17-D/05R 17-E/05R 17-F/05R

17-B/05R

Contributor

CCC No./ Code name

Fortran

Fortran II & FAP Fortran 63 & Codap Fortran IV Fortran IV

Fortran IV

Fortran

Fortran II & Fap Fortran 63 &CODAP Fortran IV

Fortran II

Program language

IBM 7090 7094 IBM 360/75

IBM 360/75 IBM 7090

CDC 1604

IBM 7090

IBM 7030 IBM 360[75 CDC 6600

IBM 7090

IBM 7090 7094

IBM 7090 CDC 6600 IBM 360]75 UNIVAC 1108

CDC 1604

IBM 7090

IBM 7090

Computer

Method

%n

7,n

7

n,7

3'

n

Kernel integr.

Monte Carlo

Discrete ordinates

Fission-product inventory

Monte Carlo

Nucleon Monte Carlo

Rad. type

Complex

Multilayer slab or complex quadric surfaces

Multilayer plane, cylinder sphere

Plane & cylindrical

Complex quadric surfaces

Complex

Geometry

Table 1 Computer code packages available from RSIC.

10,12

8,9

7

6

5

1-4

Ref.

Uses single buildup factor or base neutron kernel; neutron kernel is moments method result or Alber t-Welton kernel; many versions available.

RSIC package contains two prototypes of system, one of these uses 05R geometry routine.

Package contains GAMLEG code for "r-ray cross sections.

Fission product "r-ray source strength in energy groups is calculated based on reactor operation and shutdown time.

Uses very detailed cross section system; writes collision parameter tape for later analysis; RSIC package contains several prototypes of system.

High-energy ( ~ 400 MeV) nucleons are treated but below 50 MeV only neutrons are treated. Version B incorporates revised CCC- 17/05 R.

Comments

N

t:h

~0

ORNL-C

56-A/MYRA

PPC

NASA-MSC

AI

NASA-LE

NASA-LE

EURATOMa

EURATOM a

NASA-LA RAC

AI

AI

61[CEP

62/K009

63/OPEX

64/LPSC

65/TDSN

66-A/BIGGI-3P

66-B/BIGGI-4T

67/STORM

68/TYCHE IIl

69/CURIE-DOSE THUNDERHEAD

56-B/MYRA

Contributor

CCC No./ Code Name

Fortran II & FAP

Fortran II & FAP

Fortran IV

Fortran IV

Fortran II

Fortran IV & MAP

Fortran IV

Fortran II

Fortran II

MAP

Fortran

Program language

IBM 7090 7094

IBM 7090 7094

IBM 7090 7094

IBM 7090

IBM 7090

IBM 7090 7094

IBM 7090 7094

IBM 7090

IBM 7090

IBM 7040

IBM 7090 CDC 1604

Computer

Ray tracing dE/dx

Monte Carlo

Kernel integr.

Method

7

n

Solar p

7

n

p,sec.n

Fission product inventory, kernel integr.

Monte Carlo

St6rmer cone approx, for permitted proton trajectories plus other factors

Discrete ordinates

Discrete ordinates

Straightahead dE/dx

Weight optimization by steepest descent

p,tx

n

7

Rad. type

Table 1 (cont.)

Homogeneous medium

Infinite homogeneous medium

Multilayer slabs, spherical shells

2-dimen.~

Multilayer slab

Multilayer slab

Spherical shell solid angle increments

Cylinder

Cylinder

Geometry

25

23,24

22

19-21

18

17

16

15

14

13

Ref.

External and internal radiation dose from a radioactive cloud.

Calculates 2nd, 4th, 6th moments of slowing down distribution

Solar flare flux and dose to earth orbiting vehicles, includes geomagnetic effects.

Angular flux exponential interpolation, exponential transform, flexible angular mesh.

PI or transport corrected Po scattering.

Includes doses from primary protons and secondary radiation consisting of cascade protons, cascade neutrons, and evaporation neutrons.

Minimizes shield weight subject to the constraint of a constant dose rate at some selected detection point.

Calculates point dose inside solid angle-defined spacecraft.

First flight escape probability.

Shipping cask design, shipping procedures, schedules and costs.

Comments

0

O

MD-M

PPC

UKAEA AEEW a

LRL-B

OCD

LASL-N NASA-LE & AGN

NASA-LA BC & AFWL

NASA-LA BC & AFWL

7 l/MIST

72/COMPRASH

73/ASTROS

74/CAPS-2

75/G33

76/BPPC

77/BEBC

• Contributor

70/CHARGE

CCC No./ Code Name

Fortran IV

Fortran IV Fortran II

Fortran Il & FAP Fortran IV

Fortran 63

Fortran IV & MAP

Fortran I1 & FAP

Fortran II & Fortran IV

Fortran IV

Program language

IBM 7090 7094

IBM 7090 7094

IBM 7090 7094

CDC 1604

IBM 7090 7094

IBM 7090

IBM 7090

IBM 7090 7094

Computer

brems.

s e c . -y

p sec. n

7

7

p

n sec. 7

n

p, e, brems., sec. n.

Rad. type

Kernel integr.

Straightahead dE/dx

Kernel integr.

Engineering Manual method

dE/dx

Straightahead

Spinney

Discrete ordinates

Straight-ahead dE/dx

Method

Table 1 (cont.)

Multilayer slab

Multiple slab

Complex structure

Spheres

Multilayer slab

Multilayer slab

Multilayer slabs or spherical shells

Geometry

Analysis of structures for fallout radiation shielding Single scatter and optional multiple scattering using buildup factor with first scatter as source, QUAD geometry and data base. Space vehicle analysis. Tissue dose calculation from primary protons, cascade protons and neutrons, and gamma rays. Space vehicle analysis. Dose from bremsstrahlung produced at outer surface or in the volume of a shield.

35

36,37

36,37

Calculates depth dose in spheres. Primary and cascade protons and protons elastically scattered by hydrogen 32-34

30,31

Removal cross section is usual transport cross section, age-diffusion theory limitation on slowing down.

Double Sn formulation, Po scattering group-togroup, P2 within group, converges rapidly.

27

28,29

Doses from primary protons, cascade protons and neutrons; also, primary electrons and bremsstrahlung.

Comments

26

Ref.

,..] N

BN-PNL

BN-PNL

BN-PNL

UI

UNC, BRL & NDL

79-A/ISOSHLD

79-B/ISOSHLD II

79-C/ISOSHLD

80/GASS

81-A/UNC-SAM

CTC UNC CTC AFWL

GD

GD

NRDL

UNC & NASA-LE

82-A/ANISN 82-B/ANISN 82-C/ANISN 82-D/ANISN

83/RAID

84/SHADRAC

85/MOMGENMOMDIS

86/HANGER

81-B/UNC-SAM II

NASA-LA BC & AFWL

Contributor

78/BED

CCC No./ Code Name

IV 63 IV IV

Fortran IV

Fortan II IV

Fortran IV

Fortran IV

Fortran Fortran Fortran Fortran

Fortran 63 & CODAP

Fortran IV & MAP

Fortran IV

Fortran IV

Fortran IV

Fortran IV

Program language

IBM 7090

IBM 704 7090

IBM 7090

IBM 7090

IBM 7090 CDC 1604 IBM 360/75 CDC 6600

CDC 1604

IBM 7090 7094

UNIVAC 1108

IBM 360

IBM 7090

IBM 7090 7094

Computer

n

3"

n,3"

n,3"

n,3,

n,7

3"

3' brems.

3"

e

Rad type

Monte Carlo

Moments

Kernel integr.

Monte Carlo

Multigroup discrete ordinates

Monte Carlo

Monte Carlo

Kernel integr.

Straightahead dE/dx

Method

Table 1 (cont.)

44

43

41,42

40

39

38

36,37

Ref.

Cylindrial sym- 47 metry

lnfff,ite homo- 46 geneous medium plane, monodirectional source

Rectangular 45 pyramids and coaxial cylinders

Cylindrical (multiaxis)

Slab, cylin&ical, spherical

Complex

Cylindrial spherical

Simple 3dimens.

Simple 3dimens.

Multilayer spherical shell

Geometry

Neutron transport in cylindrically symmetric NERVA configuration, with emphasis on neutron heating in liquid hydrogen. (SAM typeC).

Constructs energy-angular flux distributions from moments. Requires input scattering moments.

Revision of CCC-5/C- 17. Compatible with CDC 6600.

Designed for duct calculations; revision of CCC9/L-05. Compatible with CDC 6600.

One-dimensional Sn transport with general anisotropic scattering.

L~

0

O Ex tension of CCC- 13/ ADONIS. SAM-II has time dependence, other extensions (SAM typeC).

Designed to compute leakage spectra from encapsulated source.

Isotope shielding analysis, RIBD code computers fission product inventory. Includes bremsstrahlung from # sources.

Space vehicle analysis. Dose from electrons penetrating shield

Comments

JAERI & HCRL a

SRL

CTC AI CTC AFWL

ORNL-N

GGA

WANL

NASA-LE

WANL & NASA-MSFC

WANL & NASA-MSFC

88/RADOS

89-A/DOT 89-B/DOT-II 89-C/DOT-II 89-D/DOT-II

90/AMC

91/NEFIRS

92/SAP-N&G

93/MCFLARE

94/KAP-V

95/TAPAT

Contributor

87/LG-H

CCC No./ Code Name

Fortran IV

Fortran IV

Fortran IV

Fortran IV

Fortran IV & MAP

Fortran IV & MAP

Fortran IV

Fortran IV

Fortran IV

Program language

IBM 7090 7094

IBM 7090 7094

IBM 7090 7094

IBM 7090 7094

IBM 360

IBM 360

IBM 7090 IBM 360/75 IBM 360/75,91 CDC 6600

IBM 360

IBM 7090 7040, 7044

Computer

Spinney

Monte Carlo

Multigroup discrete ordinates

Kernel integr.

Kernel integr. (ray analysis)

Method

n,7

n,7

Solar flare p

Multigroup discrete ordinates

Kernel integr.

Monte Carlo

Reflected Numerical n,3" integr.

n

n sec.-r

n,'r

7

n,7

Rad. type

Table 1 (cont.)

49

Homogeneous medium (line

Multigroup neutron flux and reaction rate calculated; removal cross section is usual transport cross section.

51

52

Rectangular ducts Planes, cylinders, spheres

Uses empirical fit of moments data and AIbert-Welton kernel (WANL seriesd). Includes POINT (data generator and cross section library) (WANL seriesd).

54

55 (Vol. 1,4) 56 (Vol. 7) 55 (Vol. 2,3) 56 (Vol. 1,2,3)

1-dimens.

Complex

1-dimens.

Total dose in shielded crew compartment during interplanetary voyage, base dose and trajectory parameters are input; uses transport data of CCC-64/LPSC.

53 Reflection from planes or cylindrical surface

Uses empirical differential albedo formulas to calculate dose at a point, uses input incident flux.

Designed to compute transport in large ducts albedo data required.

50

Anisotropic scattering, "point scaling" accelerated convergence.

Dose from radioactive cloud.

Gamma and neutron streaming through straight cylindrial ducts using ray analysis method.

Comments

2-dimens.

sources)

48

Ref.

Straight cylindrical ducts

Geometry

N

h~

Fortran IV

Fortran IV

WANL

BC-H ART & AC NASA-MSFC

ORNL-0

NAS A-MSC

IIT & NASAMSFC

GEN

NAS A-LE

NASA-MSFC

g&A/FASTER

SgB/FASTER 9&C/F ASTER g&D/FASTER

99/PLUME

lOO/KOl9

lOI/NAP

1OZ/SURF

103/OPEX-II

104/~lJIyA

IBM 7090 7094 IBM 7090 7094

Fortran IV

IBM 360175

IBM 7090 7094

IBM 360/75

IBM 360/75

IBM 7090 7094 IBM 360/75 IBM 360/75 UNIVAC 1108

IBM 7090 7094

IBM 7090 7094

Computer

Fortran IV

Fortran IV

Fortran IV

Fortran IV

Floco Fortran IV MAP

WANL & NASA-MSFC

97/ODDK

Fortran IV

Program language

WANL & NASA-MSFC

Contributor

96/TIC-TOC-TOE

CCC No./ Code Name

e

1,2,4,5,6)

Recoded and extended CCC-63/0PEX. Requires dose-thickness data. Integrates, empirical kernel deduced from Monte Carlo data.

61

62,63

Kernel integr.

Slab

Spherical shell

Computes uncollided and singly-scattered compo nents of flux in the presence of a thin scattering structure.

60

Conical shells, multilayer slabs Numerical integr.

by

Includes (n,r), (n,p), (n&r) and (n,2n) reactions; incident flux can be time-dependent; decay chain followed.

59

58

Complex

l-dimens.

Computes equivalent thickness of base material for solid angle increments

57

Considerable importance sampling (WANL setiesd).

Extension of DDK, PO in-group scattering, Pr group to group, includes POINT (WANL seriesd).

Designed to compute heating rate in hydrogen; uses empirical Monte Carlo kernels (WANL seriesd).

Comments

Simple 3dimens.

fGo1. 9) 56 (Vol. 8)

;toL

(VoL 2,6,7,8)

55

;;01. 5) 56 (Vol. 1,9)

Ref.

Dose from radioactive cloud Effects of ground surface and thermal inversion layer included.

Complex

2dimens.

Conical & cylindrical

Geometry

Multigroup disCrete ordinates

Kernel integr.

Monte Carlo

Multigroup

Multigroup discrete ordinates

Kernel integr.

Method

Weight optimization steepest descent

n,7

n,7 activation source

P

7

n,7

n,7

n,7

Rad. type

Table 1 (cont.)

;

8

8 2 cl E

CDC 1604 UNIVAC 1108 UNIVAC 1108 IBM 7090 UNIVAC 1108 IBM 360/75 CDC 6600 IBM 7090 7094 CDC 6600

Fortran 63 Fortran IV Fortran IV

Fortran IV

Fortran IV

Fortran IV

110-A/AIRTRANS UNC

110-B/AIRTRANS LMSC

MD-M

BN-PNL, AI

AFWL, & TRW

UNC & NASA-LE

MAGI, BRL, & NDL

111/FLORA

112-A/SAND

ll2-B/SAND 112-C/SAND

113/ATHENA

114/SAM-C

CDC 1604 CDC 3600

UNIVAC 1108 IBM 360/75 IBM 360/75

CDC 3600

7094 IBM 360/75

IBM 7090

Computer

IBM 360/75

AI

109/SOSUM

Fortran 63 Fortran

Fortran IV

Fortran 64

Fortran IV

Program language

Fortran IV

SL

108-A/SPECTRA 108-B/SPECTRA

107-B/ETRAN-16 107-C/ETRAN-16B

107-A/ETRAN-15 NBS

106/PF-COMP

RTI & OCD

EURATOMa

105-A/RDMM

105-B/RDMM

Contributor

CCC No./ Code Name Method

Monte Carlo

Engineering manual method

Kernel integr.

Monte Carlo

n,3"

n,3"

Monte Carlo

Monte Carlo

Neutron spectra by activation detector analysis

Fluorescence X-ray

n,3" sec. 3'

13,3"source from fission products

Neutron spectra by activation detector analysis

e,3"

3'

Neutron spectra by activation detector analysis

Rad. type

Table 1 (cont.)

76

77

Complex 3-dimens. (combinatorial)

Extension of CCC-81/ SAM-II permits combination of geometric figures, has time-dependence (SAM typeC).

Emphasis on 3' heating in reactor core (SAM typeC).

Fully automated iteration procedure. 74,75

Complex 3-dimens.

Calculates fluorescence source and transmission. 73

Slab

Computes fission product source spectra from CCC-69/CURIE data.

A trial spectrum is used to introduce knowledge of spectral shape.

Calculates electron-photon cascade, electron tracking in steps each representing multiple scattering.

Analysis oi" structures for fallout radiation shielding.

Relative Deviation Minimization Method used to determine polynomial spectra.

Comments

Computes time- and energy-dependent flux at point or ring detectors (SAM typeC).

70

69

67,68

66,33,34

64,65

Ref.

Variable density 71,72 air-over ground

Slab

Complex structure

Geometry

,.-] N

ga,

r~

7~

RI, NRDL

II5/GADJET

FortranIV

Program language CDC 6600

Computer 3,

Rad. type

1. ORNL-3610 (1964). 2. ORNL-TM-196 (1961). 3. Phys. Rev. 131 (1961) 1801. 4. ORNL-TM-1866 (1969). 5. ORNL-3622 (1965). 6. ER-6906 (1964). 7. LA-3373 (1965). 8. ORNL-3805 (1965). 9. ORNL-TM-1212 (1966). 10. LA-3573 (1967). 11. NASA TM X-1397 (1967). 12. ORNL-4181 (1968). 13. ORNL-3648 (1964). 14. IDO-17075 (1965). 15. NASA MSC Project 3208P (1966). 16. NAA-SR-TDR-11516 (1965). 17. NASA TM X-52166 (1966). 18. NASA TN D-3573 (1966). 19. Atomkernenergie 12 (1967) 81. 20. EUR 2488e (1965). 21. EUR 3555e (1967). 22. RAC-1395-1,2,3,4 (1964,65). 23. NAA-SR-7357 (1962). 24. NAA-SR-memo-9069 (1963). 25. NAA-SR-8884 (1965). 26. SM-46335 (1965).

Adjoint Monte Carlo

Method

27. IDO-16856 (1963). 28. AEEW-R-361 (1964). 29. AEEW-M-648 (1966). 30. UCRL-10980 (1964). 31. UCRL-16154 (1965). 32. OCD-JDI (1966). 33. TR-20 (1964). 34. PM-100-1 (1965). 35. EAD-119 (1964). 36. WL-TDR-64-71 (1964). 37. D2-90684-1 (1965). 38. BNWL-236 (1966). 39. BNWL-236 (Sup. 1) (1967) 40. UI-NRSS-3 (1966). 41. UNC-5093 (1964). 42. UNC-5157 (1966). 43. K-1693 (1967). 44. NARF-DC-Memo-1.115 (1966). 45. NARD-DC-Memo-1.097 (1966). 46. USNRDL-TR-67-9 (1967). 47. UNC-5043 (1962). 48. Informal memo (1966). 49. DP-1098 (1967). 50. K-1694 ($967). 51. ORNL-3964 (1967). 52. GA-8069 (1967).

Table references

a Via ENEA-CPL. b Via Argonne Code Center. c SAM type - similar codes based on UNC routines and cross section systems. d WANL series - group of related codes of shield design system.

& OCD

Contributor

CCC No./ Code Name

Table 1 (cont.)

78,79

Ref. Computes exposure in open or covered basement from fallout source (SAM typeC).

Comments

53. WANL-TME-1273 (Rev. A) (1966). 54. NASA TN D-4311 (1968). 55. WANL-PR-(LL)-010 (1967). 56. WANL-PR-(LL)-014 (1967). 57. ORNL-4086 (1968). 58. NASA-MSC-3066 (1965). 59. IITRI-A6088-21,-22 (1966). 60. GEMP-582 (1968). 61. NASA TM X-1769 (1969). 62. IN-SSL-N-68-13 (1968). 63. NASA-SP-169 (p. 529) (1968). 64. EUR 2985e (1966). 65. Nucl. Sci. Eng. 23 (1965) 344. 66. OCD-RTI-1 (1965). 67. NBS-9836 (1968). 68. NBS-9837 (1968). 69. SC-RR-67-746 (1967). 70. AI-AEC-Memo-12693 (1968). 71. UNC-5179 (1967). 72. LMSC-5234 (1968). 73. DAC-60654 (1967). 74. AFWL-TR-67-41 (VoL 1,2) (1967). 75. BNWL-855 (1968). 76. UNC-5148 (NASA CR-54905) (1966). 77. MAGI-6701 (1967). 78. NRDL-TRC-68-25 (1968). 79. NRDL-TRC-68-27 (1968).

Complex 3-dimen~ (combinatorial)

Geometry

o rn

-8

o

U.S. Air Force Weapons Lab., Kirtland AFB, N.M. Aerojet General Corp., Sacramento, Calif. Aerojet General Nucleonics, San Ramon, Calif. Atomics International, Canoga Park, Calif. ART Research Corp., Los Angeles, Calif. The Boeing Co., Aerospace Div., Nuclear and Space Physics, Seattle, Wash. BC-H The Boeing Co., Huntsville, Ala. BE Brown Engineering Co., Huntsville, Ala. BN-PNL Battelle-Northwest, Pacific Northwest Labs., Richland, Wash. BRL U.S. Army Ballistics Research Lab., Aberdeen Proving Ground Md. CTC Computing Technology Center, Union Carbide Corp., Oak Ridge, Tenn. EURATOM EURATOM Joint Nuclear Research Center, Ispra (Varese), Italy GD General Dynamics Corp., USAF Nuclear Aerospace Research Facility, Fort Worth, Tex. GE-N General Electric Co., Nuclear Systems Programs, Cincinnati.0. GGA Gulf General Atomic, San Diego, Calif. GT Georgia Institute of Technology, Atlanta, Ga. HCRL Hitachi Central Research Lab., Hitachi Ltd., Ozenji, Kawasakishi, Kanagawa-ken, Japan. liT IIT Research Institute, Chicago, Ill. JAERI Japan Atomic Energy Research Institute, Tokai Research Establishment, Tokai-Mura, Ibaraki-ken, Japan. LASL-N Los Alamos Scientific Lab., N. Div., Los Alamos, N.M. LASL-T Los Alamos Scientific Lab., T. Div., Los Alamos, N.M. LG Lockheed-Georgia Co., Nuclear Analysis Dept., Marietta, Ga. LMSC Lockheed Missiles and Space Co., Sunnyvale, Calif. LRL-B Lawrence Radiation Lab., University of California. Berkeley, Calif. MAGI Mathematical Application Group, Inc., White Plains, N.Y.

AFWL AG AGN AI ART BC McDonnell Douglas, Missile and Space Systems Division, Santa Monica, Calif. NASA-LA NASA Langley Research Center, Hampton, Va. NASA-LE NASA Lewis Research Center, Cleveland, O. NASAMSC NASA, Manned Spacecraft Center, Houston, Tex. NASAMSFC NASA, Geo. C. Marshall Space Flight Center, Huntsville, Aim NBS National Bureau of Standards, Washington, D.C. NDL U.S. Army Nuclear Defense Lab., Edgewood Arsenal, Md. NRDL U.S. Naval Radiological Defense Lab.,~San Francisco, Calif. OCD Department of the Army, Office of Civil Defense, Washington, D.C. ORNL-C Oak Ridge National Lab., Chemical Technology Div., Oak Ridge, Tenn. ORNL-N Oak Ridge National Lab., Neutron Physics Div., Oak Ridge, Tenn. ORNL-G Oak Ridge National Lab., Operations Div., Oak Ridge, Tenn. PPC Philips Petroleum Co., Atomic Energy Division, Idaho Falls, Id. RAC Republic Aviation Corp., Farmingdale, N.Y. RI Radioptics, Inc., Hainville, N.Y. RTI Research Triangle Institute, Raleigh, N.C. SL Sandia Laboratories, Albuquerque, N.M. SRL Savannah River Lab., E.l.du Pont deNenours and Co., Aiken, S.C. TRW TRW Inc., Redondo Beach, Calif. UI University of Illinois, Dept. of Civil Engineering and the Nuclear Engineering Program, Urbana, Ill. UKAEAAEEW United Kingdom Atomic Energy Authority, Atomic Energy Establishment, Winfrith, AERE, HarweU, England UNC United Nuclear Corp., Elmsford, N.Y. WANL Westinghouse Astronuclear Lab., Pittsburgh, Pa.

MD-M

Table 1 (cont.) Code contributors

N

COMPUTER CODES

References [1] D.K.Trubey, The Radiation Shielding Information Center A Technical Information Service for Nuclear Engineers, Nucl. Eng. and Design 9 (1969) 392-407. [2] D.K.Trubey, B.F.Maskewitz and S.K.Penny, Computer Codes for Shielding Calculations, Nucleonics 24(8), 112-115 (Aug. 1966). [3] R.G.Jaeger et al., editors, Engineering Compendium on Radiation Shielding, Vol. I, Shielding Fundamentals and Methods, Springer Verlag (1968). [4] E.P.Blizard et al., Reactor Handbook, 2nd edition, vol. 3, Part B, Shielding, p. 130 (Interscience Publishers, New York, 1962). [5] E.P.Blizard, op. tit., p. 72. [6] H.Kahn, Random Sampling (Monte Carlo) Techniques in Neutron Attenuation Problems, I and II, Nucleonics 6(5), 27 (1950) and 6(6) 60 (1950). [7] R.G.Jaeger, op. cir., p. 161.

517

[8] D.K.Tmbey and B.F.Maskewitz, editors, A Review of the Discrete Ordinates Sn Method For Radiation Transport Calculations, ORNL-RSIC- 19 (1968). [9] W.D.Collier and G.C.Curtis, ATTOW, A Two-Dimensional Shielding Program, UKAEA TRG Report 1466(R) (1967). [ 10] W.Wayne Scott and R.G.AlsmiUer Jr., Comparisons of Results Obtained with Several Proton Penetration Codes ORNL-RSIC-17 (1967). [11] W.Wayne Scott, Estimates of Primary and Secondary Particle Doses Behind Aluminum and Polyethylene Slabs Due to Incident Solar-Flare and Van Alien Belt Protons, ORNL-RSIC-18 (1967). [ 12] W.Wayne Scott and R.G.Alsmiller Jr., Comparisons of Results Obtained With Several Proton Penetration Codes, Part II, ORNL-RSIC-22 (1968). [ 13] B.F.Maskewitz, Abstracts of the Digital Computer Codes Assembled by the Radiation Shielding Information Center, ORNL-RSIC-13 (1966-1969).