EIR-Bericht Nr CH&A om. The NJOY Nuclear Data Processing System: The MICROR Module. Dezembef 1964

Transcription

1 CH&A om 0) CO If) I* i LU Q EIR-Bericht Nr. 539 Dezembef 1964 The NJOY Nuclear Data Processing System: The MICROR Module D. R. Mathews, J. Stepanek, S. Pelloni, C. E. Higgs A report on work done under the joint General Atomic Technologies Inc./Swiss Federal Institute for Reactor Research / Hochtemperatur- Reaktorbau Company Cooperative Program Published also as General Atomic Technologies Report, GA- A Eidgendssisches Instrtut fur Reaktorforschung Instrtut Federal de Recherche* en Matiere de Reacteurs Swiss Federal Institute for Reactor Research CH-5303 WUrenlingen Tel Telex eirch

2 GA-A17851 EIR-Bericht Nr The NJOY Nuclear Data Processing System: The MICROR Module by D.R. Mathews, J. Stepanek, S. Pelloni, C.E. Higgs A report on work done under the joint General Atomic Technologies Inc / Swiss Federal Institute for Reactor Research / Hochtemperatur-Reaktorbau Company Cooperative Program Published also as General Atomic Technologies Report December 1984

3 - 2 - Contents Page 1.0 INTRODUCTION NJOY SYSTEM MICROR MODULE Fast Neutron Cross Sections Thermal Neutron Cross Sections Pointwise Resonance Cross Sections General Comments INPUT INSTRUCTIONS General Comments Free-format Input NJOY-MICROR Input Specifications SAMPLE PROBLEMS Example Example Example Example 4 42 REFERENCES 43 ACKNOWLEDGEMENTS 46 APPENDIX A 47 APPENDIX B 56 APPENDIX C 62 APPENDIX D 64 APPENDIX E 78

4 - 3 - ABSTRACT The NJOY nuclear data processing system is a comprehensive computer code package for producing pointwise and multigroup neutron and photon cross sections and related nuclear parameters from ENDF/B-IV and V evaluated nuclear data. The MICROR overlay is a reformatting module that produces cross sections library files for the MICROX, MICROX-2 and MICROBURN postprocessor codes. Using the data on the pointwise and groupwise NJOY tapes, MICROR produces the tapes containing basic nuclear data, FDTAPE, GAR and G6TAPE used by two-region spectrum codes MICROX and MICROX-2 and by two-region spectrum burn-up code MICROBURN. ZUSAMMENFASSUNG Das NJOY System ist ein umfassendes Programm, dass Punkt-und Gruppen-Wirkungsquerschnitte fur Neutronen und Photonen, sowie andere Kernparameter ausrechnen kann. Die Berechnung basiert auf den gemessenen Werten, die in den ENDF/B-IV und V Basisbibliotheken enthalten sind. Der MICROR Modul formatiert dann die so erzeugten Datensatze derart, dass sie dann in den Rechenprogrammen MICROX, MICROX-2 und MICROBURN verwendet werder, konnen. Ausgehend von de.i NJOY-Punkt-und Gruppenbandern, kann MICROR die Bander erzeugen, die die notigen Grunddaten FDTAPE, GAR und GGTAPE fur die zwei- Zonen-Spektrum Rechenprogramme MICROX, MICROX-2 und MICROBURN enthalten.

5 - A INTRODUCTION The NJOY nuclear data processing system is a comprehensive computer code package for producing pointwise and multigroup cross sections from ENDF/B-IV and -V evaluated nuclear data (1). A concise deccription of the code system and references to the ancestors of NJOY are given in (2) to (6). NJOY is a modular system to generate pointwise ENDF and groupwise ENDF nuclear data libraries called PENDF and 6ENDF files, respectively. Each module is essentially free-standing and reflects a give physical task during the generation of the nuclear data libraries. The present modular structure of NJOY is shown in the scheme on page 4. The RECONR module reconstructs pointwise (energy- dependent) cross sections from ENDF/B resonance parameters and interpolation schemes. Resonance cross sections are calculated with an extended version of the methods of RESEND (7). BROADR Doppler-broadens and thins pointwise cross sections using the method of SIGMA modified for better behaviour at high temperatures and low energies (8). UNRESR computes effective self-shielded pointwise cross sections in the unresolved-resonance region using the methods of ETOX (9). 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 cross sections data. GAMINR calculates multigroup photon interaction cross sections ano kerma factors and group-to-group photon scattering matrices. ERRORR produces multigroup covariince matrices from ENDF/B uncertainties. COVR reads the output cf ERRORR and performs covariance ^lotting and output formatting operations. DTFR

6 formats multigroup data for transport codes, such as DTF-IV (10) and ANISN (11). CCCCR formats multigroup data for the CCCC standard (12) interface files ISOTXS, BR0K0XS, and DLAYXS. MATXSR formats multigroup data *or the MATXS cross- section interface file. The ACER module prepares libraries for the Los Alamos continuous-energy Monte Carlo code MCNP (13). POUR prepares libraries for the EPRI-CELL and EPRI-CPM codes. Finally, MODER changes ENOF/B "tapes" and other ENDF-like NJOY interface files back and forth between formatted (that is, BCD or ASCII) and blocked-binary modes. NJOY incorporates and improves upon the features of its direct ancestor, MINX (14). It also includes and extends the photon production capabilities of LAPHANO (15), the photon interaction capabilities of GAMJ.EG (16), the heating capabilities of MACK (17), the covariancc capabilities of PUFF (18), and the thermal capabilities of FLANGE-II (19) and HEXSCAT (20). NJOY ENDF/B PENDF n GENDF Main program Working Module MODER RECONR BROADR UNRESR HEATR THERMR GROUPR GAMINR ERRORR MOOER DTFR CCCCR MATXSR COVR ACER POWR MICROR Basic structure of the NJOY code (overlay configuration)

7 - 6 - In a joint cooperation between the Eidgenoessisches Institut fuer Reaktorf orschung in Uuerenl ingen, Switzerland (EIR), Genera!. Atomic Technologies Inc. in San Diego, USA (GA) and Hochtemperatur-Reaktorbau GmbH in Mannheim, Federal Republic of Germany (HRB) a new reformatting module MICROR was developed. This module prepares cross section data files for use in the two region spectrum codes MICROX (21) and MICROX-2 (22) and in the two region spectrum-burnup code MICROBURN (23). Uhereas the module POWER in connection with EPRI-CELL or EPRI-CPM and the module MATXSR in connection with the post processor codes TRANSX (24) or T RANSX-CTR (25) use just self-shielded multigroup cross sections based on the- Bondarenko flux-weighting model (26) to shield the resolved as well as unresolved resonances, the module MICROR in connection with MICROX, MICROX-2 and/or MICROBURN uses additionally pointwise (slowing down) resonance shielding.

8 NJOY SYSTEM The main program of the NJOY system is NJOY. It simply reads a module name in free format and calls in the requested module. The first card read by any module contains the unit numbers for the various input and output files. In this way, the output of one module can be assigned to be the input of another module, thereby linking the modules to perform the desired processing task. Using the subsequent calls for the NJOY modu'es RECONR and/or BROADR, UNRESR, HEATR and THERMR (see flow diagram on the next page) pointwise neutron cross sections are written onto a "point-endf" (PENDF) tape for future use. RECONR reads an ENOF-format tape with neutron data and produces a common energy grid for all neutron reactions (the union grid) such that all cross sections can be obtained to within a specified tolerance by linear interpolation and writes them onto the PENDF tape for future use. BROADR reads a PENDF tape and Doppler-broadens the data. After broadening, the summation cross sections are again reconstructed from their parts. The results are written out on a new PENDF tape for future use. UNRESR produces effective self-shielded pointwise cross sections, versus temperature and background cross sections, 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 or energy balance. The ENDF/B photon production files are used in both methods when available. The heating results are added to the PENDF tape. THERMR produces pointwise cross sections in the thermal range. The results for all the processes (incoherent, coherent scattering etc.) are added to the PENDF tape.

9 * UNRESR To reconstruct the photon pointwise ENDF/B cross sections just the module RECONR has to be used (see flow diagram below). In this case RECONR reads an ENDF/B library with photon data and produces a common energy grid for all photon reactions using the same method as for neutron reactions. The results are written onto a PENDF tape for photon data.

10 - 9 - Using the PENDF tapes' the "groupwise-endf" tapes can be generated using the modules GROUPR, for neutron reactions and photon production, and GAMINR, for photon reactions, respectively. In GROUPR the resolved and unresolved resonance ranges are shielded for a single isotope in a background scattering medium using the Bondarenko flux-weighting model. As an option, a pointwise flux solution can be generated for a heavy absorber in a light moderator. The results are saved onto neutron and photon GENDF tapes, respectively. The next two flow diagrams describe the generation of neutron and photon GENOF tapes.

11 The structure of the PENDF and GENDF tapts is similar to the structure of the ENDF/B files. The ENDF/B evaluated nuclear data files are well documented elsewhere (1) but for the convenience of the reader, some features of the format will be described here. ENDF-format "tapes" consist of a series of "record" with special structures. Tapes are originally distributed in formatted mode (EBCDIC, ASCII, BCD, etc.) in which the records are composed of one or more "card images)" (80 column FORTRAN records). NJOY also uses a special binary format where the "ecords are divided into one or more "blocks". Each tape starts with a "tape identification" record. ENDF/B "tapes" are subdivided internally into " materials" (MAT), "files" (MF), and "sections" (MT). A MAT contains all data for a particular evaluation for an element or isotope (for example, MAT=1276 is an evaluation for ). A " file" contains a particular type of data for that MAT: MF=3 is cross-section versus energy data; MF=15 contains secondary photon energy distributions, for example. A " section" refers to a particular reaction

12 (for example, MT=2 is elastic scattering and MT=107 is the (n,>» ) reaction). Every record contains the current MAT, MF, and MT values. Two materials are separated by a record with MAT=0 (the material-end or MEND record). Two files are separated by a record with MF=0 (the file- end or FEND record). Two sections are separated by a record with MT=0 (the section-end or SEND record). Finally, the tape is terminated with a record with MAT=-1 (tape-end or TEND record). - PENDF tapes The structure of the PENDF tapes is summarized in Appendix A. It is quite similar to that of ENDF except for tne following differences. File 1 contains only a short description and the standard ENDF "dictionary", which is really a directory to the files and sections given for the material. File 2 is a short version which gives only the effective scattering radius. All other resonance parameters (if any) have been removed because their effects have been included in File 3. File 3 uses only linear interpolation for cross sections and the energy grid is the same for all sections. Redundant cross sections except for total <MT=1) have been removed. There is no File I*. There is no File 5. A specially-defined File 6 may be included for thermal scattering data. All higher files are omitted. CENDF tapes The GENDF tape structure uses standard ENDF records, but it is quite different in other ways (see Appendix B). It begins with a File " that contains summary information including group structure and dilution values. File 3 contains cross sections by group, Legendre order, and dilution. File 5 contains delayed neutron spectra and time constants. File 6 contains neutron or photon scattering matrices by incident group, secondary group,

13 Legendre order, and sometimes dilution. File 16 contains photon production matrices by neutron group, photon group, Legendre order, and dilution (not usually used). No other files are used. Weighting fluxes by group Legendre order, and dilution are also given in each section of each file. One important difference between GENPF and PENDF is that GENOF allows sections of File 3 and 6 to be intermixed in any order if desired. vthe order is determined by the order of requesting reactions in GROUPR). Therefore only File 1 has a file end record (FEJ.O). As a consequence, GENOF tapes cannot be searched with utility routines such as FINDF in NJOY that assume sections are in correct order.

14 MICROR MODULE Using the data on the PENDF and GENOF tapes, MICROR produces three tapes containing basic nuclear data, FDTAPE, GAR and GGTAPE, used by MICROX, MICROX-2 and MICROBURN. E.QIAPE tape contains fine group dilution- and temperature-dependent cross sections for the fast energy range for use in MI CROX-2. The GGTAPE consists of two files (sections) which contain infinite dilution cross sections for the fast and thermal energy ranges, respectively,- for use in the MICROX and MICROBURN codes. The MICROX-2 cede uses only the thermal section of the GGTAPE. The GAR tap.e, contains pointwise Doppler- broadened resonance cross sections in the resolved resonance range. The number of energy points as well as the energy range is arbitrary. Typical are energy points between ev and about 3 kev to 8 kev. 3.1 Fast Neutron Cross Sections Fast neutron cross sections for the MICROX codes are prepared from fine group (GENDF) data prepared by the NJOY GR0UPR module. An expanded version of the NJOY POUR module is used to prepare GAM-II format files for specific temperatures and dilutions from the GENDF data. The GAM-II format file is used by adaptations of the FDTAPE and GGTAPE codes to prepare fast neutron data for the MICROX-2 (FD> and MICROX (GG) codes, respectively. The GAM-II file is not presently saved, although it would be suitable for use directly in the GGC-5 code (28) in combination with the GAROL and GANDY resolved and unresolved resonance ca Iculational options, respectively, after processing onto GAM-II library tapes with the MAKE code (MAKE is available at EIR but was not converted to CDC).

15 The absorption cross sections are calculated sections and total scattering. from the total cross The edit cross sections as well as scattering matrices are calculated either using the given cross sections or their linear combinations from the GENOF file. For example the scattering matrix includes elastic (MT = 2) and all inelastic <M." = 51 to 91), <n,2n), (n,3n), (n,4n), etc. partial cross sections. The total (n,2n) cross section is calculated by summing all the particular excited states (MT = 6,7,8 and 9) and the direct d.,2n) cross section (MT = 16). The scattering matrix includes all available reactions from the ENDF/B files,such elastic, inelastic, all available (n,xn), and other.reactions. Fission Spectra The prompt fission source into energy group g is S g = 11 f g-g' V ' g' where <J f g^gi is the total fission matr i x f or MF=6 and MT = 18in the GENDF file and $$ is the isotropic neutron flux in the group g. For some important materials the fission process is divided into direct fission (n,f), MT=19; second-chance fission (n,n'f), MT=21; third-chance fission (n,2n)f; MT=20 and fourth-chance fission (n,3n)f, MT=38. This procedure makes possible a more accurate representation of high- energy portion of fission spectrum

16 when fission is induced by neutrons with energies above 5 to 6 MeV. This is important for the systems with very hard neutron spectrum, such as fusion-fission hybrid blankets, for example. In such cases MICROR adds all fission matrices to get the total prompt fission matrix: f g+g Z HT f g*-g MT for MT = 19, 20, 21 and 38. The above fission source includes the prompt part of fission only. The steady-state (SS). fission is obtained adding the delayed portion of fission process:.ss g' f g-g'vj' + x g g' V'WJ 1 where v is the number of neutrons per fission <MF=3 and MT=455), X_ is the normalized spectrum for delayed neutrons summed over all time groups, and fa is tnetotal fission cross section in MF=3 and MT=18. However, most existing transport and diffusion codes do not use the above described matrix representation of the fission source because full upscatter is expensive to handle. Instead assuming no or weak dependence on incident neutron spectrum, the steady-state fission source can be expressed as S = x ss g g g' ss V a fg'*0g' ' where event SS and is the number of steady-state neutrons per fission,ss is the average steady-state fission spectrum de-

17 fined by SS a* l = 'fg f g<-g» + V D and SS Z f g-g-v + X?Z v g' a fg'v gj_ 3L Y, Z «ro'v + Z v g' fg'v q' g' where X g is fission spectrum of the delayed neutrons (MF=5 and MT=471) for incident energy 2 MeV. X is calculated using the energy distributions ov delayed neug -j_ trons 3 in six delayed groups: i=l -i B g which are in MF=5 and MT=455. SS The X is exact only for the incident neutron spectrum *gg but normally varies slowly with changes in incident neutron spectrum. - Transport cross section The transport correction is calculated in the fast as well as in the thermal energy range, for each fine energy group, using the

18 diagonal transport approximation: tr tl si g g^g where o, and 0, are the P< weighted total and self-scattering cross sections E E 'tl - 1 a t (E)<J> 1 (E)dE / j * x (E)dE,dE' 'g+i 'g+i and E g E g 0 sl 9 = I I OgiE'+B^CEJdE de' / I t ± (E)dE, E g+1 Vl Vi where the P^ neutron flux is : (^(E) = C(E) (o t (E)+0 0 ) and C(E) is the weighting spectrum.

19 Thermal Neutron Cross Sections Thermal neutron cross sections (the thermal portion of the G6 tape used by the MICROX codes) are prepared from fine group (GENDF) data prepared by the NJOY THERMR and GROUPR nodules. Since, the MICROX code expects ppintwise (per unit energy) as opposed to groupwise thermal neutron data, the fine group data is divided by the width (in energy) of the incident groups and, in the case of the scattering transfer matrices, by the width of the final groups. The thermal neutron energies supplied to the thermal sections of the MICROX codes are averages of the group boundary energies and can be made very close to the standard GATH- ER-II values used ai GA by appropriate choice of group boundaries in GROUPR. By This way, the General Atomic fast GAM-II (99 Groups) and thermal GATHER-II (101 Points) library results in an unified 193 group library with energy group boundaries given in AP PENDIX C. The use of "pseudo" pointwise thermal neutron data obtained from groupwise data allows a reasonable representation of the reaction rates for nuclides such as Hf-177 (important to HRB), PA-233, U-233, ect. which have relatively narrow resonances in the upper portion of the thermal neutron energy range without requiring an excessive number of thermal neutron energies (groups). All incoherent inelastic scattering matrices for free or bound scatterers, coherent elastic scattering matrices for hexagonal materials, and incoherent elastic scattering matrices available on the GENDF tape are considered. If there is inelastic as well as elastic scattering, the total scattering matrix is produced as the sum of the both parts. The user can specify the type of thermal scattering. If he specifies the bounded scattering, MICROR searches for incoherent inelastic and coherent elastic scattering matrices. If they are not available, the free gas, or if it is not available, the elastic scattering matrix is used.

20 REACTION NUMBERS FOR THERMAL SCATTERING ENDF/B-V ENOF/B-IV CONTENTS free gas H 2 in H inelastic H in CH elastic H in CH inelastic H in ZrH elastic H in ZrH benzene , 0 in D^O inelastic graphite elastic graphite inelastic Be elastic Be inelastic BeO elastic BeO inelastic Zr in ZrH elastic Zr in ZrH If the user specifies the free gas scattering, MICROR searches first for the free gas scattering matrix. If it is not available, MICROX uses just elastic scattering. rtlcror gives a message if the required scattering was not found and tells what type of scattering was used. The available MT numbers of thermal scattering are listed above. The effective temperature T.. for calculation of the thermal sources for graphite and for other moderator nuclides are usually somewhat larger than the corresponding Maxwellian temperature T. For the convenience of the U3er, the values of J.. for the common moderators and temperature T are calculated by MICROR using tabulation of values from Ref.27 (see the following table).

21 T( K) H ( H 20) MAT = 1002 D(D 2 0; MAT = 1004 BeO Graphite HAT= MAT = Zr(ZrH) MAT = 1096 H(ZrH) MAT = 1097 H(CH 2 > MAT = C Since the Cadilhac spectrum is no longer used in MICROX, MICROX-2 and MICROBURN, the record 10 arrays of GGTAPE were filled with zeroes. Also the mean lethargy gain per collision was filled with zero. PQ and Pi sources into the thermal group g are calculated using the weighting neutron flux spectrum in the epithermal energy range for infinite dilution:

22 G 2 0 V, S 0 g'=l and G 2 Q g = \' 9«-g' ** I=Z s/ *1 I g'=i where C? is the lowest epithermal energy group. 3.3 Pointwise Resonance Cross Sections Pointwise resonance cross section (GAR data) are prepared by optmized interpolation from pointwise (PENDF) data prepared by the NJOY, RECONR and BRCADR modules. Equality of the infinite dilution NJOY resonance integrals calculated using the originai energy poi.it distribution on the PENDF tape and the equidistant (velocity or lethargy) energy point distribution in MICROX is enforced. The energy point distribution is obtained in NJOY ( RECONR and BROADR modules) using the method of cross section linearization. For a required accuracy of the cross section reconstruction and of the resonance integrals, this method leads to the smallest number of energy points and to the most efficient point positions. But the distribution of the energy points is irregular over the given energy range. There are more points with a higher

23 density of distribution in the regions with sharp cross section shapes and fewer points in the parts with more smoother cross sect ion shapes. In such a case, the smallest number of energy points enables a quick calculation of the group integrals if just one absorber in a homogeneous scattering medium and the Bondarenko flux-weighting model are considered. If the neutron flux is calculated solving the "slowing down" equation, the method is no longer as efficient as even for just one absorber in a homogeneous medium. Greater problems occurs if one considers resonance overlap in mixture of absorbers when solving the "slowing down" equation. First, the mixture of energy point grid,s for more aosorbers could lead to an enormous number of energy points. Secondly, the irregular distribution of energy points would lead to an inefficient calculation of "stowing down" scattering integrals. Therefore, an equidistant lethargy or velocity energy point distribution is used in MICROX, MICROX-2 and in MICROBURN. This leads to a reasonable number of energy points in the resonance region even for mixtures of resonance absorbers. Whereas the equidistant lethargy energy point distribution is suitable for well thermalized systems, the equidistant velocity energy point distribution is used to produce the library for fast reactor systems. As mentioned above, equality of the infinite diluted resonance integrals calculated using the original energy point distribution on the PENDF tape and using equidistant (velocity or lethargy) energy point distribution on GAR tape is enforced. The "effective" cross section (E^) at energy Ei on the GAR tape is defined as:

24 (E i ) = E i+v2 { "(E)f E i-v2 E >i+v2 I ^E 1 'i-v2 INT, l,(e i + V2 /E i-v2 ) where Ei+v2 and E i-v2 are the energies in the middle between E i+1' E i and E i» E i-1 }» respectively. Resonance integral INT is calculated considering the linear energy dependence of the cross section between the energy points Ej and E., on the PENDF tape: 3-1 P P P < E - E j-i> o(e) = c F (E..) + (a*(e.)-<r (E..)) L -^ D ' i 3 3_i (E.-E..) 3 D-l where index P denotes the cross sections on the PENDF tape and E. is the energy at the j-th energy point on the PENDF tape. The resonance integral is calculated considering three possibilities: Between the energies E. ^2 and E._^2, there is no energy point on the PENDF tape: INI. '''W^-^WW,.,^,..,,^. (E i+v2 " E i-v2 )

25 Betyeen the energies E i+^2 and E ^ ^. there is just energy p int E on the PENOF tape: INT = (q(e._ V2 )E. - a(e.)e.. V2 ) ^ n( ) + (E j - E i-v2 } E i-v2 + (g(e i )E i+v2 " (E i±v2hl ) E - tot-itvi, + a(e._)-a(e. 'E. i+v2' i- (E i + V2 ' E j> 1 where E ^ < E. < E.^. Between the energies E i+v2 and E i-i/2» there is one energy point on the PENDF tape: more INT. = (a(e i-v2 i-i= 3 )E j ' a(e i 3 ^_Vi_,E i-v2 ) E j n( J ) + (E j ' E i-v2 ) E i-v2 + ^<y fc -i)v fc - g <y k >v fc -i> n(! i±!l _ ) + k=l (E jh-k " E j+k-l ) j+k-1 + ( <V E i+v2 -»W»1+H ln(!lii2 + 0 (E i+v2 " E j+n ) j+n i+v2~ a i-v2 ' where E d _ v2 * E j+k * E i+v2 ks ' N

26 I I I I I- I i I I Using the above described method, the desired number of energy dependent cross sections can be reconstructed from the PENOF tape in the resolved resonance region to obtain the required accuracy of the resonance screening. On the other hand, in the unresolved resonance region, the PENDF tape does not include enough energy points to enable the pointwise resonance screening to obtain a desirable accuracy, since the Bondarenko method of resonance scree ing is considered in NJOY in this region and therefore the module UNRESR produces resonance shielded dilution dependent tables. Therefore, the libraries produced by MICROR could lead to inaccurate unresolved resonance screening in mixtures of resonance absorbers with different boundaries between the resolved and unresolved resonance regions. For example, this boundary lies at about 82 ev for U-235 but at 8000 ev for U-238. It would not be simple to overcome this problem in MICROX and MI- CPOSURN (includes MICROX) without additional changes in MICROX. Because MICROX should be replaced by MICR0X-2,only the existing library produced by the GA cross section generation system should oe used in MICROX in the meantime. MICROR should be used only to produce the cross section libraries for additional non- resonant

27 or weakly resonant isotopes for which the previously mentioned limitation leads to no decrease of accuracy. In MICROX-2 the boundary between the pointwise and Bondarenko resonance screening can be chosen as an input parameter by the user. Then, if the user sets this boundary to be lower or equal to the lowest boundary between the reso'ved and unresolved resonance region in the problem defined resonant absorbers, MICROX-2 will screen the resonances properly with the caution that a large part of the resonances will be screened using Bondarer.ko formalism. However, MICROX-2 will be generalized in the future to enable mixing of both Bondarenko and pointwise, screening of resonances of different isotopes for the same energy range. In the NJOY module 6R0UPR, there exists the possibility to calculate the resonance shielded tables in the resolved resonance region not only using the narrow resonance flux approximation but also using an accurate slowing down calculation in which no narrow resonance approach is considered. The narrow resonance approach has to be used in MICROX-2 only to select the background cross section to interpolate in these tables. Therefore, in such a way the use of shielded tables leads to good accuracy even in the resolved resonance range. 3.4 General Comments As noted in the separate report by C. Higgs covering the conversion from UNIVAC to CDC of the MICROX-2, FDTAPE, TAPER, GGTAPE and related codes, we have changed the definitions of the fast, resonance and thermal neutron data files used by the MICROX codes in two significant respects.

28 All A6 character data was replaced by A4 character data. Single A6 words became A4, A2; 2A6 became 3A4 and 12A6 became 18A4. Modifications to all of the above listed codes were required. The FD (fast neutron data for MICROX-2) and the GAR pointwise resonance data tapes (files) were changed to use standard unformated FORTRAN instead of the UNIVAC NTRAN format previously used. These two changes will greatly facilitate the conversion of the MICROX-2 and related codes and data files to other computers such as VAX and IBM. The FD and GAR data tapes prepared by the MICROR module are not directly usable at GA and probably HRB until the corresponding changes are made in the MICROX codes involved. The EIR version of MICROX will also have to be changed to make use of data f-'les prepared by NJCY.

29 INPUT INSTRUCTIONS 4.1 General Comments All NJOY input is in free form. A subroutine FREE (see description below) has been included among the NJOY utilities to provide this capability. Fields on the input cards are limited by any character not used for another purpose (.-,numeral,e,h,*,r,/). For exponent fields, the E must be present, and spaces are not allowed before the E. O.ecimal points are not required after numbers. Hollerith entries may use nhstring or *STRING*. The form nr causes the following number to be repeated n times. The (/' terminates the input for one call to FREE (it may involve more than one card) leaving any unread variables unchanged. This feature is often used to default variables from the right. The part of the input line to the right of the (/) can be used for comments if desired. As an example of when the (/) is useful, in several NJOY routines a record of Hollerith information is constructed from user input. This is accomplished by calling FREE with NZA=17 (the number of Hollerith words required to fill the 66 columns available for an ENOF/B "comment"). The array is preset to 17 blank words, so that the user needs not blank-fill the line explicitly. Instead, he can write *MESSAGE*/ where the (/) terminates the process of replacing the default blanks with actual input ("MESSAGE" in this example). The user should be cautioned that if the (/) is omitted from an input data block that is incomplete, as in the Hollerith example above, FREE will go on reading successive input data cards until the expected NZA words are found, usually resulting in an error condition. For this reason, if the user is uncertain whether he has supplied enough input parameters to "satisfy" a particular

30 call to FREE, it is good practice to use a (/) at the end of the input data for that data block. Some input examples follow. legal *U235* 5R0 1.2E1 4HU235 2R 1.2E2 it. legal E1 4RU235 (does not mean right-justify) The input to the NJOY-MICROR sample problems. Section 5, provides additional illustrations of the syntax of free-form input. In the next input examples the typical flow of the job/ input cards is shown: JOB CARD EOR. 0/1 <I0P T = Input/output option) 4/5 (IVERF = ENDF/B version IV/V) * MICROR* input for MICROR STOP*.EOF. NJOY provides for a special blocked-binary mode for the ENDF/B files. Such files are indicated with negative unit numbers. The MODER module can be used to convert back and forth between formatted (that is, BCD or ASCII) and blocked-binary modes. The user may assign unit numbers from 20 through 99 for input and/or output MICRO? tapes. 0 is accepted as a un-'t number, which means that the corresponding file is not used. Unit numbers from 10-19

31 are reserved for scratch files, and units 5-7 are used for system input and output files. 4.2 Free-format Input For a card-input program, free-form input is convenient, but in a time-sharing environment, it is almost essential. Therefore, a subroutine FREE has been included among the NJOY utilities to provide a simple free-format input capability. This routine contains a machine- dependent subroutine PACK, that may have to be adapted to local conditions. FREE(NIN,Z,NZA,NCU) NIN input logical unit containing free-format card images Z(I) dimensioned variable containing numbers decoded from input cards. NZA on call, number of words desired NCU number of of Hollerith characters to be loaded in each word, blank fill to right All numbers read from the input cards are returned as real in Z. The calling program can convert selected numbers to integer mode as required. Hollerith variables are returned in integer from using the internal N-bit code of the machine. If NCW is larger than the number of characters per word, successive locations of Z will be used.

32 FREE contains several parameters that may have to be changed yhen converting between different machines: NBPC is the number of bits per character for Hollerith data (6 on CDC, 8 on IBM), MACHUD is the number of Hollerith characters in a machine word (10 on CDC. 4 on IBM), and RNDOFF is a constant that should be approximately equivalent to one bit in the last place for the target machine. The rest of the machine dependence is incorporated into FUNCTION PACK, which inserts characters into words. Two versions are supplied: one is based on masking for CDC machines, and the other uses one-byte variables and equivalencing for IBM systems. 4.3 NJ0Y-MICR0R Input Specifications Detailed input instructions are included as comment cards at the start of MICROR module. The current set of comment- card and input card instructions for main overlay-njoy module and for MI CROR module is reproduced below. In the following some details are given in connection with the input for NJOY and MICROR modules. A. NJOY Module IOPT... yhen operating in a time/sharing environment (I0PT=1), the code routes input prompts and shares output messages to the therminal (TTY). The regular long output is still available for the system printer.

33 B. MICROR Module * * PRODUCE OPTIONALLY MICROX,MICROX-2 OR MICROBURN INPUT DATA * * files FROM NJOY INTERMEDIATE CROSS SECTION LIBRARY TAPES * * PENDF AND NEUTRON AND PHOTON GENDF. * * (1) FD AND/OR FAST PART OF GG DATA * * (2) GAR DATA * * (3) THERMAL PART OF GG DATA * * * * * * * INPUT SPECIFICATIONS (FREE FORMAT) * * # * CARD 1 UNITS * * NPENDF INPUT UNIT FOR PENDF TAPE * * NGEN1 INPUT DATA FOR UNIT FROM GROUPR * * NGEN2 INPUT UNIT FOR DATA FROM GAMINR * * NFDIN INPUT UNIT FOR OLD FD DATA * * NGARIN INPUT UNIT FOR OLD GAR DATA * * NGGIN INPUT UNIT FOR OLD GG DATA *» NFD OUTPUT UNIT FOR NEW FD DATA * * NGAR OUTPUT UNIT FOR NEW GAR DATA * * NGG OUTPUT UNIT FOR NEW GG DATA * * * * WARNING: IF GAR DATA IS DESIRED (NGAR.NE.O) ONE HAS TO DEFINE * * THE GENDF FILE AS WELL (NGEN1.NE.O) * * * * ALLOWED UNIT NUMBERS ARE RESTRICTED TO * * * * CARD 2 USER IDENTIFICATION * * LGPRNT 0/1 MEANS SUMMARY PRINT/FULL PRINT» * LFPRNT 0/1/2/3 MEANS DIFFERENT PRINT OPTIONS * * LTPRNT 0/1/2/3 MEANS DIFFERENT PRINT OPTIONS * * 0: HEADING * * 1: HEADING,FLUXES AND GROUP BOUNDARIES *» 2: HEADING,FLUXES,GROUP BOUNDARIES AND RESPONSE * * FUNCTIONS * 3: FULL PRINT * * * * * LIB LIBRARY OPTION (O-EDIT OR DELETE GIVEN DATA ON * * NEW UNITS NGAR,NFD,OR NGG. 1«FD,2-GG,3-BOTH, * 4-FAST GG) * * IOPT -0 PROCESS FOR EACH OF THE NMATN MATERIALS GIVEN * * NTEMP TEMPERATURES AND NSIGZ DILUTIONS * * (CARDS 4-5. FIRST TEMPERATURE AND FIRST * * DILUTION,FIRST TEMPERATURE AND SECOND DILUTION * * ARE FIRST PROCESSED * * -1 PROCESS FOR EACH OF THE NMATN MATERIALS EVERY * * TEMPERATURE AND DILUTION ON THE GENDF FILE» * #

34 * NTEMP NUMBER OF TEMPERATURES TO PROCESS *» (MEANINGFUL ONLY FOR IOPT.EQ.O) * * NSIGZ NUMBER OF SIGMA ZEROES TO PROCESS * * (MEANINGFUL ONLY FOR IOPT.EQ.O) * * NMATN NUMBER OF MATERIAL NAMES TO BE READ * * IGN NEUTRON GROUP STRUCTURE OPTION * * IWT WEIGHT FUNCTION OPTION *» NOT YET USED * IVERS MICROX FILE VERSION NUMBER (DEFAULT=0) * HUSE USER ID (12 CHARACTERS, DELIMITED BY *,ENDED BY /)* (DEFAULT=BLANK) * * AT THE PRESENT MOMENT * * IGN. EQ.O NUMBER OF FAST GROUPS IS COMPUTED * * IN THE CODE * *.NE.O FORCE TO HAVE IGN FAST GROUPS * * * * NOTES * * NO MEANING OF VARIABLES ON CARD 2 FOR LIB.EQ.O * * FOR CREATING NEW GAR DATA THE PARAMETER LIB HAS NO MEANING BUT * LIB HAS NOT TO BE EQUAL TO 0 (1.LE.LIB.LE.4) * IT IS NOT ALLOWED TO EDIT AND TO DELETE GAR DATA AT THE SAME * TIME * FOR NFDIN. NE.O. AND. 1*FD. NE.O. AND. LI B.EQ.O. AND. NSUBFD.EQ.O *.AND.KSPECFD.EQ.O THE FD TAPE ON UNIT NFDIN IS COPIED ONTO * * UNIT FD AND THEN EDITED * *» SUPPLY CARDS 3-5 ONLY IF LIB.NE.O * * * * SUPPLY CARD 3 NMATN TIMES (MATERIAL DATA) * * CARD 3 * * HMATR HOLLERITH MATERIAL IDENTIFIER (6 CHARACTERS * * DELIMITED BY *) * * MATR INTEGER MATERIAL IDENTIFIER * * NTHER INELASTIC THERMAL SCATTERING MT NUMBER * * (DEFAULT=0 MEANS ELASTIC SCATTERING) * * SUPPLY CARDS 4-5 ONLY IF IOPT.EQ.O * * CARD 4 * * (T(l),1 = 1,NTEMP) * * TEMPERATURES IN K * * CARD 5 * * (S(I),I-1,NSIGZ) * * SIGMA ZEROES VALUES * INPUT-SPEC IFICATIONS-FCR-GAR -TAPE» * * SUPPLY CARDS 1A-5A IF NGAR.GT.O» * * SUPPLY CARD 1A ONLY IF LIB.NE.O * * CARD 1A * * NPT NUMBER OF RESONANCE CROSS SECTION POINTS TO BE * * RECONSTRUCTED ON GAR FILE * * EO TOP ENERGY OF RESONANCE RANGE (EV) EL LOWEST ENERGY OF RESONANCE RANGE (EV) * NPRK ILOG NUMBER 0/1-EQUIDISTANT OF POINTS/RECORD VELOCITY/LETHARGY (DEFAULT-10) SPACING» *

35 * SUPPLY CARDS 2A-5A IF LGPRNT.GT.O.OR.(LIB.EC.0.AND.NGARIN.EQ.0)* * CARD 2A * * NGRP MAXIMUM NUMBER OF ENERGIES FJ* WHICH GAR DATA IS * * TO BE EDITED * * ' ET NGRP.LB.0 TO EDIT ONLY TABLE OF CONTENTS * * NNUC NUMBER OF MATERIALS FOR WHICH GAR DATA IS * * TO BE EDITED (NOT NEEDED IF NGRP.LE.0)(NNUC.LE.40)* * * IF NGRP.LC.O,THIS IS END OF T ' JT OF PRINTING GAR FILE * * CARD 3A * * (MID(I), 1 = 1,KNUC) * * GAR DATA SET MATERIAL ID NUMBERS * * REPEAT CARD(S) 4A AND 5A FOR EACH OF THE NNUC MATERIALS FOR * * WHICH GAR DATA IS TO BE EDITED (1-1,NNUC) * * CARD 4A * * J NUMEER OF PRINT RANGES (J.LE.6) * * CARD 5A * * (IFF(K,I),IPL(K,I),K = 1,J) * * BEGINNING (IPF) AND ENDING GAP POINT NUMBERS FOR * * EACH PRINT RANGE * * (lpf(k,l).le.ngrp.and.ipl(k,l).le.ngrp) * * * * INPUT-SPECIFICATIONS-TO-DELETE-ON-OLD-GAR-TAPE * * * * SUPPLY CARDS 6A-7A ONLY IF * * LIB.EQ.O.AND.NGARIN.NE.O.AND.NGAR.NE.O * * CARD 6A * * NSUBG NUMBER OF NUCLIDES TO DELETE FROM OLD GAR TAPE * * CARD 7A (NSUBG.GT.O) * * XSUBG IDENTIFICATION NUMBERS OF NUCLIDES TO BE DELETED * * FROM OLD GAR TAPE (NSUBG VALUES) * * * * INPUT-SPEC IFICATIONS-FOR-FD-T APE * * # * SUPPLY FOLLOWING CARDS ONLY IF NFD.GT.0.OR.NGG.GT.0 * * INPUT AND OUTPUT UNIT NUMBERS HAVE TO BE DIFFERENT * * FOR LIB.EQ.1.0R.LIB.EQ.3-(FD TAPE OUTPUT) * * (NFD. GT.O) * * * * CARD1B * * NCARDS NUMBER OF 72 CHARACTER CARD IMAGES TO USE TO * * DESCRIBE NEW FD TAPE * * ICSPEC NUMBER OF FISSION SPECTRA TO INPUT *» FROM CARDS 6B AND 7B * * AT THE PRESENT ONLY ICSPEC.EQ.O IS ALLOWED * * IFSWTH 0 ' BOUND AND NORMALIZE NEW FISSION SPECTRA * * 1 - NORMALIZE ONLY * * 2 = NEITHER * * AT THE PRESENT ONLY IFSWTH.EQ.O IS ALLOWED *» ISQUEZ 0 * COMPRESS TRANSFER MATRICES * * 1 - DO NOT COMPRESS TRANSFER MATRICES * * CARD 2B TO BE REPEATED NCARTS TIMES * * AT THE PRESENT ONLY ISQUEZ.EQ.O IS ALLOWED * * IREC2 NEW FD TAPE DESCRIPTION (72 CHARACTERS PER CARD * * IMAGE * * REPEAT CARDS 3B-5B NMATN TIMES * * CARD 3B * * ITSPEC NUMBER OF NEW FISSION SPECTRA TO PREPARE FROM * * THE GROUPR DATA FOR THIS MATERIAL * * (ITSPEC DIFFERENT INCIDENT NEUTRON SP3CTRA) *

36 * AT ""HE PRESENT MOMENT ONLY ONE SPECTRUM CAN BE PREPARED * * ITSPEC = -1 USE GROUPR WEIGHTING FLUX * * TO CALCULATE THE SPECTRUM * * 0 DO NOT PREPARE ANY FISSION SPECTRUM * * = 1 READ IN WEIGHTING FLUX FROM CARD 5B * * TO COMPUTE THE SPECTRUM * * * * REPEAT CARDS 4B TO 5B, IABS(lTSPEC) TIMES * * FOR ITSPEC.LT.O REPEAT ONLY CARD 4B * * CARD 4B * * IDUM 72-CHARACTER (18A4) FISSION SPECTRUM TITLE * * CARD 5B * * ENFIS INCIDENT NEUTRON ENERGY SPECTRUM FOR THIS * * FISSION SPECTRUM (HIGH TO LOW IN ENERGY) * * REPEAT CARDS 6B AND 7B FOR EACH OF THE ICSPEC FISSION SPECTRA * * TO BE READ FROM CARDS (ICSPEC.GT.0) * * CARD 6B * * IDUM 72-CHARACTER (18A4) FISSION SPECTRUM TITLE * * CARD 7B * * DUM FINE GROUP FISSION SPECThUM (HIGH-TO-LOW IN * * ENERGY) (NUMBER OF VALUES SAME AS NUMBER OF * * FAST NEUTRON GROUPS IN GROUPR FILE) * * END OF LIB = 1 (FD TAPE) INPUT * * * * INPUT-SPEC IF ICATIONS-FOR-FAST-GG-FILE * * * FOR LIB.GE.2 (FAST GG TAPE OUTPUT) (NGG.GT.O) * * * * NOTE THAT,FOR LIB.EQ.3 (BOTH FD AND GG),CONFLICTS BETWEEN THE * * FOLLOWING GG DATA INPUT AND THE PREVIOUSLY READ FD DATA INPUT * * ARE RESOLVED IN FAVOR OF THE FD INPUT * * # * CARD 1C * * ICSPEC NUMBER OF FISSION SPECTRA TO INPUT * * (FROM CARDS 3C AND 4C ) * * AT THE PRESENT, ONLY ICSFEC.EQ.O IS ALLOWED * * ISWTH 0 = BOUND AND NORMALIZE NEW FISSION SPECTRA * * 1 - NORMALIZE ONLY * * AT THE PRESENT, ONLY ISWTH.EQ.O IS ALLOWED * * IFLUX 0 = USE BUILT-IN REFERENCE NEUTRON FLUX SPECTRUM * * 1 * READ REFERENCE NEUTRON FLUX SPECTRUM * * ON CARD(S) 7C * * ICUR 0 = USE BUILT-IN REFERENCE NEUTRON CURRENT *» SPECTRUM * * 1 - READ REFERENCE NEUTRON CURRENT SPECTRUM * * ON CARDS 8C * *» * REPEAT CARDS 2C-4C NMATN TIMES *» CARD 2C *» L^YPE DATA SET (NUCLIDE) TYPE (-1/0 - SECONDARY/PRIMARY)* * (SECONDARY DATA SETS MAY NOT BE USED WITH GAR DATA* * AS FOR EXAMPLE MINOR FISSION PRODUCTS) * * AT THE PRESENT, ONLY LTYPE.EQ.O IS ALLOWED * * ITSPEC NUMBER OF FISSION SPECTRA TO PREPARE FROM THE * * GROUPR DATA FOR THIS MATERIAL * * (ITSPEC DIFFERENT INCIDENT NEUTRON SPECTRA) *

37 * AT THE PRESENT MOMENT ONLY ONE SPECTRUM CAN BE PREPARED * * ITSPEC - -1 USE GROUPR WEIGHTING FLUX * * TO CALCULATE THE SPECTRUM * * 0 DO NOT PREPARE ANY FISSION SPECTRUM * * = 1 READ IN WEIGHTING FLUX FROM CARD 4C * * TO COMPUTE THE SPECTRUM * * REPEAT CARDS 3C TO 4C, IABS(lTSPEC) TIMES * * FOR ITSPEC.LT.O REPEAT ONLY CARD 3C * * * * CARD 3C * * ASPEC 72-CHARACTER FISSION SPECTRUM TITLE * * CARD 4C * * ENFIS INCIDENT NEUTRON ENERGY SPECTRUM FOR THIS FISSION * * SPECTRUM (HIGH TO LOW IN ENERGY) * * * * REPEAT CARDS 5C AND 6C FOR EACH OF THE ICSPEC FISSION SPECTRA * * TO BE READ FROM CARDS (ICSPEC.GT.0 ) * * * * CARD 5C. * * ASPEC 72-CHARACTERS FISSION SPECTRUM TITLE * * CARD 6C * * SPEC FINE GROUP FISSION SPECTRUM (HIGH TO LOW IN ENERGY)* * (NUMBER OF VALUES SAME AS NUMBER OF FAST NEUTRON * * GROUPS IN GROUPR FILE) * * CARD 7C (IFLUX.EQ.1 ) * * PHI REFERENCE NEUTRON FLUX ENERGY SPECTRUM (HIGH TO LOW* * IN ENERGY)(NUMBER OF VALUES SAME AS NUMBER OF FAST * * NEUTRON GROUPS IN GROUPR FILE) * * CARD 8C (ICUR.E0.1) * * CUR REFERENCE NEUTRON CURRENT ENERGY SPECTRUM (HIGH TO * * LOW IN ENERGY)(NUMBER OF VALUES SAME AS NUMBER OF * * FAST NEUTRON GROUPS IN GROUPR FILE) * * * * * * END-0F-LIB=2--(FAST-GG)--INPUT * * *» * * * *» * SUPPLY CARDS 1D-11D ONLY IF LIB.EQ.O * * * * EDITING-PARAMETERS - --* * * NOTE: IT IS NOT ALLOWED TO EDIT AND DELETE USIFG THE SAME JOB * «#» * * SUPPLY CARDS 1D-4D IF NFD.GT.O * * # * * * INPUT-SPEC IF ICATIONS-TO-EDIT-OLD-FD-T APE * * * * CARD 1D * * IFPRFD 0- SUMMARY EDIT * * N.GT.O SUMMARY EDIT+SELECTIVE EDIT FOR N NUCLIDES * * N.LT.O EDIT EVERYTHING * * NSUBFD NUMBER OF NUCLIDES TO BE DELETED FROM THE OLD * * FD TAPE * * NSUBFD.LT.O: SUPPLY NUCLIDE ID NUMBER ON CARD 2D * * NSUBFD.GT.O: SUPPLY NUCLIDE SEQUENCE NUMBER ON *

38 * CARD 3D * * KSPECFD NUMBER OF FISSION SPECTRA TO DELETE FROM THE OLD * * FD TAPE» * (.LT.O MEANS DELETE ALL) * * CARD 2D (NSUBFD.LT.O) * * XSUBFD IDENTIFICATION NUMBERS OF NUCLIDES TO BE DELETED * * FROM OLD FD TAPE * * CARD 3D (NSUBFD.GT.O) * * JSUBFD SEQUENCE NUMBERS OF DATA SETS TO BE DELETED FROM * * OLD FD TAPE (NSUBFD VALUES) * * CARD 4D (KSPECFD.GT.O) * * IRSPECD *» SEQUENCE NUMBERS OF FISSION SPECTRA TO BE DELETED * * (KSPECFD VALUES) * * CARD 5D (IFPRFD.GT.O) * * JPFD SEQUENCE NUMBERS OF FD DATA SETS TO BE EDITED * * (IFFRFD VALUES) * * * * SUPPLY CARDS 6D-9D IF NGG.GT.O *» * INPUT-SPEC IF IC AT IONS-TO-EDIT-OLD-FAST-GG-FILE * * * * CARD 6D * * IFPRNF 0= NO EFFECT * * N= EDIT FAST GG DATA FOR THE N DATA SETS GIVEN * * ON CARDS» * -1= EDIT ALL FAST GG DATA * * NSUBF NUMBER OF NUCLIDES TO BE DELETED FROM OLD FAST * * GG FILE * * KSPECF NUMBER OF FISSION SPECTRA TO DELETE FROM THE OLD * * FAST GG FILE (AT THE PRESENT MOMENT USE 0) * * IFLAG 0= NO TREATMENT OF THERMAL GG FILE *» 1= EDIT OR DELETE ON THERMAL GG FILE AS WELL *» * NOTE: FOR DELETING ONE HAS TO USE IFLAG.EQ.O IF THERE IS NO * * THERMAL DATA ON UNIT NGGIN.IF THERMAL DATA IS PRESENT USE * * IFLAG.EQ.1 * * * * CARD 7D (NSUBF.GT.O) * * XSUBF IDENTIFICATION NUMBERS OF NUCLIDES TO BE DELETED * * FROM OLD FAST GG FILE (NSUBF.LT.51 VALUES) * * CARD 8D (KSPECF.GT.O) * * IRSPECF SEQUENCE NUMBERS OF FISSION SPECTRA TO B DELETED * * FROM OLD FAST GG FILE (NOT YET USED) * * CARD 9D (IFPRNF.GT.O) *» JPFG SEQUENCE NUMBERS OF FAST GG FILE DATA oets TO BE * * EDITED (IFPRNF VALUES) * * * SUPPLY CARDS 10D-12D IF NGG.GT.0.AND.IFLAG.EQ.1» * * * INPUT-SPEC IFICATIONS-TO-EDIT-OLD-THERMAL-GG-TAPE * * «* CARD 10D * * JPRINT» 0: NO EFFECT *» N: EDIT THERMAL GG DATA FOR THE N DATA SETS GIVEN *» ON CARDS * * -1: EDIT ALL THERMAL GG DATA * * NSUBT NUMBER OF NUCLIDES TO BE DELETED FROM OLD THERMAL * * GG FILE *

39 * IF(NSUBT.GT.O) SUPPLY CARD 11D * * CARD 11D * * (XSUBT(l),I=1,NSUBT) * * IDENTIFICATION NUMBERS OF NUCLIDES ON OLD THERMAL * * GG TAPE NOT WANTED ON NEW THERMAL GG TAPE * * IF(JPRINT.GT.O) SUPPLY CARD 12D * * CARD 12D * * (JI'T(I).I-I, JPRINT) * * MATERIAL NUMBERS OF NUCLIDES ON THERMAL GG TAPE FOR WHICH DATA * * PRINTING IS DESIRED * * * END-OF--INPUT * * *» * OPTIONS-FOR-INPUT-VARIABLES * * * * IGN MEANING *» * * 1 ARBITRARY.STRUCTURE (READ IN) * * 2 CSEWG 239 GROUP STRUCTURE * * 3 LANL 30 GROUP STRUCTURE * * 4 ANL 27 GROUP STRUCTURE * * 5 RRD 50 GROUP STRUCTURE * * * 6 7 GAM-1 68 GROUP STRUCTURE GAM GROUP STRUCTURE * * * 8 LASER-THERMOS 35 GROUP STRUCTURE * * 9 EPRI-CPM 69 GROUP STRUCTURE * * 10 LANL 187-GROUP STRUCTURE * * 11 LANL 70-GROUP STRUCTURE * * 12 SAND GROUP STRUCTURE * * 13 LANL 80-GROUP STRUCTURE *» 14 GA 193-GROUP STRUCTURE * * * * IWT MEANING * *» * 1 READ IN SMOOTH WEIGHT FUNCTION * * 2 CONSTANT *» 3 1/E * * 4 1/E+FISSION SPECTRUM+THERMAL MAXWELLIAN * * 5 EPRI-CELL LWR * * 6 (THERMAL) -- (l/e) -- (FISSION + FUSION) * * 7 FAST REACTOR * * 8 THERMAL--1/E--FAST REACTOR FISSION + FUSION» * # * *

40 SAMPLE PROBLEMS The sample problems were designed to demonstrate the major options of MICROR module. The complete sample inputs and outputs are reproduced in APPENDIX C. 5.1 Example 1 FD. GAR and GG files are created on units 23, 24 and 25 respectively for the isotope U-235 at 300 K and at infinite dilution. Thus LIB=3 (both FD and GG files are constructed), I0PT=0, NTEMP=1 NSIGZ=1 (only the specified temperature and dilution are processed). One fission spectrum is prepared from the GENDF file on units 22 (ITSPEC=-1). GAR data is generated from the PENDF file on units 21. In each case minimal printing is specified. 0 4 MICROR* *PELL0NI*/ *U-235 * 1261/ E E E NEW FD TAPE FROM MICROR CONTAINING U235*/ *AT 300 K AND AT INFINITE DILUTION * -1 FISSION SPECTRUM COMPUTED USING B0NDARENKO WEIGHTING*/ / 0-1 FISSION SPECTRUM COMPUTED USING 80NDARENK0 WEIGHTING*/ STOP*

41 Example 2 S imilar to example 1. Add to the old FD. GAR and GG data on units 23, 24, 25 more recent data for isotope U-235 at 300 K and at infinite dilution to create the new FD. GAR and GG files on units 26, 27 and 28 respectively. Since the identification of new data corresponds to the old one a new version number has to be specified (IVERS=2). 0 4 MICROR* *PELL0NI*/ *U-235 * 1261/ E E E *NEW FD TAPE FROM MICROR CONTAINING U235*/ IN TWO VERSIONS*/ -1 FISSION SPECTRUM COMPUTED USING BONDARENKO WEIGHTING*/ / 0-1 FISSION SPECTRUM COMPUTED USING BONDARENKO WEIGHTING*/ STOP*

42 Example 3 Delete from the FO, GAR and GG data on vinits , 28 the data pertaining to the first isotode to create new FD, GAR and GG tapes on units 23, 24 and 25 respectively. Thus IIB=0 and the second input card has no meaning. The isotope to delete is U-235 at 300 K and infinite dilution, with identification number in he GAR case, in the FD and fast GG cases and in the thermal case. The first fission spectrum is deleted from the old FD file (KSPECFD=1 on card 1D, IRSPECD (1)=1 on card 40) and che identification number is specified (NSUBDF=-1 on card 10). The theimal part of the GG file is treated as well <IFLAG=1). 0 4 *MICROR* *PELL0NI*/ STOP*

43 Example 4 Edit table of contents of FD, GAR and GG files on units 2?. 24 and 25 respectively. Thus LIB=0 and the second input card has no meaning. Card 1A is not supplied since LIB=0. Card 2A is specified with NGRP=0 and NNUC=0 (summary editing of GAR tape only). Cards 6A-7A are not supplied since NGARIN=0 (no deleting of GAR data). ON card 1D IFPRFD=0 is defined (summary FD tape print). On card 6D IFPRNF=0 is specified (summary fast GG tape print) and on card 10D JPRINT=0 has to be used (summary thermal GG tape print). 0 4 MICROR* *PELL0NI*/ STOP*

44 S R. Kinsey. Ed., "ENDF-102, Data formats and Procedures for the Evaluated Nuclear Data File, ENDF," Brookhaven National Laboratory report BNL-NCS (ENDF-102) 2nd Edition (ENDF/B-V) (October 1979). R.E. MacFarlane, D.U. Muir and R.M. Boicourt, "The NJOY Nuclear Data Processing System, Volume I User's Manual", LA-9303-M (ENDF-324) (1982) R.E. MacFarlane, Q.W. Muir and R.M. Boicourt, "The NJOY Nuclear Data Processing System, Volume II : The NJOY, RECONR, BROADR, HEATR and THERMR Modules", LA-9303-M (ENDF-324) (1982) R.E. MacFarlane and R.M. Boicourt, "NJOY : A Neutron and Photon Cross Section Processing System," Trans. Am. Nucl. Soc. 22, 720 (1975). R.E. MacFarlane, D.U. Muir, and R.J. Barrett, "Advanced Nuclear Data Processing Methods for the Fusion Power Program," Trans. Am. Nucl. Soc. 23, 16 (1976). R.E. MacFarlane, R.J. Barrett, D.U. Muir, and R.M. Boicourt, "NJOY: A Comprehensive ENDF/B Processing System," in "A Review of Multigroup Nuclear Cross-Section," Proceedings of a Seminar-Uorkshop, Oak Ridge, Tennessee, March 14-16, 1978, Oak Ridge National Laboratory report ORNL/RSIC-41, p. 25 (October 1978) 0. Ozer, "RESEND: A Program to preprocess ENDF/B Materials with Resonance Files into a Pointwise Form," Brookhaven National Laboratory report BNL ( 1972).

45 D.E. Cullen and C.R. Ueisbin, "Exact Doppler Broadening of Tabulated Cross Sections," Nucl. Sci. Eng. 60, 199 (1976). R.E. Schenter, J.L. Baker, and R.B. Kidman, "ETOX. A Code to Calculate Group Constants for Nuclear Reactor Calculations," Battelle Northwest Laboratory report BNUL-1002 (1962). K.D. Lathrop, "DTF-IV, A FORTRAN Program for Solving the Multigroup Transport Equation with Anisotropic Scattering," Los Alamos Scientific Laboratory report LA-3373 ( November 19*5). W.U. Engle, Jr., "A Users Manual for ANISN, A One-Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering" Oak Ridge Gaseous Diffusion Plant Computing Technology Center report K-1693 (1967). R.D. O'Dell, "Standard Interface File?; and Procedures for Reactor Physics Codes, Version IV," Los Alamos Scientific Laboratory report LA-6941-MS ( September 1977). "MCNP - A General Monte Carlo Code for Neutron and Photon Transport," Los Alamos Scientific Laboratory report LA-7396-M, Revised (November 1979). C.R. Weisbin, P.O. Soran, R.E. MacFarlane, D.R. Harris, R.J. LaBauve, J.S. Hendricks, J.E. White and R.B. Kidman, "MINX, A Multigroup Interpretation of Nuclear x- Sections from ENDF/B," Los Alamos Scientific Laboratory report LA-6486-MS (ENDF-237) September 1976). D.J. Dudziak, R.E. Seamon, and D.V. Susco, "LAPHANO: A Multigroup Photon/Production Matrix and Source Code for ENDF," Los Alamos Scientific Laboratory report LA MS (ENDF-156) (January 1972).

46 K.D. Lathrop, "GAMLEG - A FORTRAN Code to Produce Multigroup Cross Sections for Photon Transport Calculations," Los Alamos Scientific Laboratory report LA-3267 (April 1965). M.A. Abdou, C.U. Maynard, and R.Q. Wright, "MACK: A Computer Program to Calculate Neutron Energy Release Parameters (Fluence-to-Kerma Factors) and Multigroup Neutron Reaction Cross Sections from Nuclear Data in ENDF Format." Oak Ridge National Laboratory report ORNL-TM-3994 ( 1973). C.R. Weisbin, E.M. Ob low, J. Ching, J.E. White, R.Q. Wright, and J. Drischler, "Cross Section and Method Uncertainties: The Application of Sensitivity Analysis to Study Their Relationship in Radiation Transport Benchmark Problems," Oak Ridge National Laboratory report ORNL-TM (ENDF-218) (1975). H.C. Honek and O.R. Finch, "FLANGE-II (Version 71-1), A Code to Process Thermal Neutron Data from an ENDF/B Tape," E.I. DuPont de Nemours and Co. Savannah R'ver Laboratory report DP-1278 (1971). Y.D. Naliboff and J.U. Koppel, "HEXSCAT, A Coherent Elastic Scattering of Neutrons by Hexagonal Lattices," General Atomic report GA-6026 (1964). P. Waelti and P. Koch, "MICROX, A two-region Flux Spectrum Code for the Efficient Calculation of Group Cross Sections", Gulf-6A-A10827 (1972). D. Mathews and P. Koch, "MICROX-2, An Improved Two/Region Flux Spectrum Code for the Efficient Calculation of Group Cross Sections", GA-A15009, vol. 1, UC-77 (1979).

47 R.H. Brogli und P. Koch, "MICROBURN, A "wo-region Spectrum-Burnup Code to Calculate Sel f-shie.ding Factors", Gulf-GA-A12673 (1973). 24. R.E. MacFarlane, "TRANSX. A Code for Interfacing MATXS Cross-Section Libraries to Nuclear Transport Codes for Fast Reactor System Analysis", LA-Report (in preparation) 25. R.E. MacFarlane, "TRANSX-CTR, A Code for Interfacing MATXS Cross-Section Libraries to Nuclear Transport Codes for Fusion Systems Analysis', LA-Report (in preparation) 26. I.I. Bond&renko, Ed., "Group Constants for Nuclear Reactor Calculations (consultants Bureau, New York, (1964). 27. J.U. Koppel and O.H. Houston, "Reference Manual for ENDF Thermal Neutron Scattering Data", General Atomic report GA-8774 revised and reissued as ENDF-269 by the National Nuclear Data Center, Brookhaven National Laboratory (1978). 28. D.R. Mathews, et al, "GGC-5, A Computer Program for Calculating Neutron Spectra and Group Constants, "General Atomic Co. Report GA-8871, (1971). Acknowledgements The authors wish to thank Dr. U. Giesser/HRB for many valuable discussions and in particular his suggestions in setting up the strategy for the MICROR module; Dr. R.E. MacFarlane for many suggv JDS, carefull reading the manuscript and for his major participation in describing the structure of PENDF and GENDF files; and to Dr. S. Brandes/HRB, Dr. R. Brogli/EIR and Dr. K. Schulz/GA for their continuing support during the realization of the work.

48 APPENDIX A THE PENDF FILE General Binary Format Records consist of one or more "blocks" of the form MAT, MF, MT. NB. NW. (A(I). I=1,NU) where NB gives the number of words remaining in the record after this block. NB=0 for the last block in a record. The text of the record uses floating-point format for both real and integer fields, and appropriate type conversion must be performed when necessary. General Coded Format (SCO, EBCDIC, etc.) Each record consists of one or more "card images" (*".e., 80-column FORTRAN records). Each cara image contain-., MAT, MF, MT, and a sequence number in columns using 14, 12, 13, 15. The text of the record (called "A" above) is given using six 11-column fields on each card. Some are integer and some are real. The description of the binary file is given below. THE PENDF FILE BINARY STRUCTURE Record 1 (Tape identification) NMAT number of different MAT numbers on PENDF file not used 17 number of hollerith variables to identify the PENDF file (HA(I), I = 1, 17) Identification set for PENDF file, arranged according to 16A4, A2. Repeat records 2-12 for each MAT number and for each temperature. Temperature is first vari^j Record 2 (important summarizing parameters) MAT ENDF material identification number

49 MF number 451 MT number 0 not used 6 Total number of remaining variables on record 2 (not used) ZA Isotope identification number (1000.*Z+A) AUR Isotope atom mass ratio to neutron not used NLIB library type (0 for ENDF and PENDF) AM total number of (MF, MT) pairs for this (MAT, TEMP) pair. Record 3 (important summarizing parameters) MAT ENDF material identification number 1 MF number 451 MT number 0 not used NU total number of remaining vaiables on record 3 TEMP material temperature ( K) ERR reconstruction tolerance (from REC0NR module) Q. 0. not used ANTEXT number of hollerith variables to describe the material on this record (HA(I),I=1,IFIX(ANTEXT) Material hollerith identification set, arranged as 16A4, A2. Record 4 (dictionary) There are IFIX(AM) records 4 MAT ENDF material identification number 1 MF number for dictionary 451 MT number for dictionary 0 not used 6 number of remaining variables on record 4 (not used) not used AMF MF number on PENDF file for this (MAT, TEMP) pair AMT MT number on PENDF file for this (MAT, TEMP) pair

50 ANREC 0. Number of records for data pertaining to this (AMF, AMT) pair on the PENDF file not used Record 5 (end of section MT=451) MAT Record ENDF material identification number MF number not used number of remaining variables on record 5 (not used) not used (end of file MF=1) MAT ENDF material identification number not used number of remaining variables on record 6 not used Repeat records 7-11 var ied. for each (MF, MT) pair. MT is first Record 7 ((MF, MT) pair description) MAT MF MT 0 6 ENDF material identification number file identification number section identification number not used number of remaining variables on record 7 (not used) ZA material identification number (1000.*Z+A) AUR atom mass ratio to neutron not used

51 Record 8 (point cross sections) there are 3 kinds of record MF=3 (record 8.1) 2. MF=2 (record 8.2) 3. MF=6 (MT=201) : free gas (record 8.3) Record 8.1 MAT MF MT NB NW ENDF material identification number file identification number section identification number number of remaining variables pertaining to this MT number, which are not on record 8.1. number of remaining variables on record 8.1. TEMP ANEP Temperatur ( K) number of energy points to thi s MT number. pertaining (E(I), SIGMA(I), 1 = 1,..)energy (ev)-cross section pairs, from lowest to highest energy Record 3.2 (Resonance Data, for MT = OR. MT = 1 52) Record (MT=151) MAT ZA ENDF material identification number 0 NW as usual material identification number (1000.*Z+A) AUR atom mass ratio to neutron not used ZAI equal to ZA above ABN always not used EL,EH always 1.E-5, 2.E not used SPI spin, always given as 0. AP effective scattering radius not used (((E(I), SIGMA(SIGZ(K),IX), IX = 1,NX), K = 1,NSIG0),1 = 1,NUNR ) Energy (ev)-cross section sets, from the lowest to the highest energy

52 Note: Meaning of IX Index IX refers to a given material number IX MT Record (MT=152) MAT ENDF material identification number 2 MF number 152 MT number NB number of remaining variables pertaining to this MT number, NW which are not on record number of remaining variables on record TEMP Temperature < K) 0. not used NX number of reaction types (5) NSIGO number of dilution cross sections NP NUNR number of words in table (NUNR*(1+NX*NSIG0)> number of unresolved energy points on file Record 8.3 MAT ENDF material identification number 6 MF number MT material identification number NB number of remaining variables pertaining to this MT number which are not on record 8.3 NW number of remaining variables on record 8.3 ZA isotope identification number (1000.*Z+A) AWR isotope atom mass ratio to neutron 0 not used LTT (=5/6/7) record type index 0 not used 0 not used MAT ENDF material identification number 6 MF number MT material identification number 0 0 not used TEMP Temperature ( K) 0 0 not used NNR number of different interpolation energy ranges NNE number of incident energy points (E(I),1 = 1,NNE) Incident energy points (in ev) from low to high energy (S(I),1=1,NNE) Subsections for each of the NNE values of incident energy

53 MAT not used ENDF material identification number MF number not used number of remaining variables on record 8.3 not used The structure of SCI) for LTT=5 is MAT 6 MT E 0 0 NP NL2 ENDF material identification number MF number material identification number not used incident energy in ev not used Number of words in table (=NL2 *NEP) number of discrete angles +2 E 1* * 1 KL '^2 ' *"' lj NL sets consisting of energies and E ' 2. f 2» Pi 'Vo >»I J NL discrete angles for the scattered E'NEP» f NEP >n. V2... IfcLPOintS (SIGMA(Ej, Eft), 1 = 1,NEP) cross section sets from the lowest to the highest incident and scattered energies E 2,E ji. Note: NL is the number of discrete angles NEP is the number of secondary energies and f is the normalized scattering function, i.e. JJf<E,E')dEdE'=1 Record 9.1 (remaining cross sections) Not present if NB=0 on record 8.1, i.e. if there are less than 306 data values on record 8.1. MAT MF MT NB ENDF material identification number file identification number section identification number Number of variables pertaining to the MT number, which are not on record 9.1. NU Number of remaining variables on record 9.1. (Ed), SIGMA(I),I = 1, )Remaining energy (ev) - cross section pairs pertaining to this MT number. Continuation of the data on record 8.1, using the same strategy,

54 Note There can be many records 9.1. The last one is characterized by NB=0. Record 9.2 (referring to records and 8.2.2) Not present if NB=0 on record or 8.2.2, i.e. if there are less than 306 data values on these. This is constructed using the same strategy as before. For LTT=6 is NEP=1, thus E'=E and f = 1. The format for LTT=7 is just provided to hold a position in File 6 because all the necessary information is implicit in MF=3. Record 9.3 Not present if NB=0 on record 8.3, i.e. if there are less than 306 data values on record 8.3. This is constructed in the same way as for record 9.1. Record 10 (end of section MT) Present only at the end of each section, i.e. when MT changes. MAT MF 0, ENDF isotope identification number EN0F file number. not used Number of remaining variables on record 10 (not used) not used

55 Record 11 (end of Vile MF) Present only at the end of each file, i.e. when MF changes. MAT ENOF isotope identification number not used number of remaining variables on record 11 (not used) not used Record 12 (end of material MAT) C. r<ot used number of remaining variables on record 12 (not used) not used Record 13 (end of PENDF file) not used not used

56 PENDF file loop structure Record Record Record Record Record Record Record A (tape identification) (summarizing parameters) (summarizing parameters) (diet ionary) (end of section MT=451) (end of file MF=1) ((MF, MT) pair description)) Record 8 (cross sections) Record 9 (Remaining cross sections) Record Record Record (End of section MT) (End of file MF) (End of material MAT) Record 13 (End of PENOF file)

57 APPENDIX B THE GENDF FILE BINARY STRUCTURE Record 1 (Tape identification) 1 not used (HA(I), 1=1, 17) not used number of Hollerith variables to identify the GENDF file Identification set for GENDF file srranged according to 16A4.A2 Repeat Records 2-9 for each MAT number and for each temperature. Temperature is first varied. Record 2 (Important summarizing parameters) MAT ZA AUR 0. ENDF material identification number MF number MT number not used Total number of remaining variables on record 2 (not used) Isotope identification number (1000.*Z+A) Isotope atom mass ratio to neutron Not used AMSIGZ Maximum number of dilution cross sections on the file NLIB Library type (-1 for GENDF) ANTEXT Number of hollerith variables on the next record 3 to describe material Record 3 (Important summarizing para»eters) MAT ENDF material number MF number MT number not used identification

58 NW Total, number of remaining variables on record 3 TEMP Temperature (K) 0. not used ANGN Number of neutron groups ANGG Number of gamma groups ANVAR Total number of remaining variables on record 3 0. not used (HA(I), 1=1, IFIX(ANTEXT) ) Material Hollerith identification Set (SIGZ(I), 1=1. IFIX(AMSIGZ)) Dilution cross section values in barns (from high to low) (BEN(I), 1=1, IFIX(ANGN)+1) Neutron energy boundaries in ev (from low to high) (GEN(I), 1=1, IFIX(ANGG)+1) Gamma energy boundaries in ev (from low to h igh) Record 4 (End of file 1) MAT ENOF material identification number not used Number of remaining variables on record 4 not used Repeat Records 5-8 (MF, MT) pair. MT is first varied. Record 5 ((MF, MT) pair description) MAT MF MT 0 6 ZA AWR ANLEG ANSIGZ ALRFLG ANGHA ENDF isotope identification number ENDF file number ENDF section number not used number of remaining variables on record 5 Isotope identification number Isotope atom mass Number of Legendre expansions Number of dilution cross sections flag to indicate that subsequent part icles are emitted after inelastic scattering Number of neutron groups

59 Record 6 (Cross sections) MF = 3 MF = 5 MF = 6 reaction cross sections or auxilliary data delayed neutron spectra by time group scattering or production matrices MAT MF MT NB NW TEMP 0. ANSI AGLA ANVAR to ENDF isotope identification number ENDF file number ENDF section number Number of variables pertaining to the (MF, MT) pair and referring to incident AG-th energy group, which are not on record 6. Number of remaining variables on record 6. Temperature (K) not used Number of scattering terms + 1. Lowest energetic group on record 6, i.e. AGL:= (ANGN-AGLA+1.)-th energy group Total number of cross sections pertaining incident group AG for this (MF, MT) pair AGA Incident neutron group, i.e. (AG := (ANGN-AGA+1.)-th energy group, corrected in order of decreasing energy ((PHI(ID.IL), IL = 1, IFIX(ANLEG)), 10 = 1. I F I X(ANSIGZ)) Neutron fluxes in the AG-th energy group (from the lowest to the highest expansion and from the highest to the lowest dilution) (((SIGMA(AG) (I,ID,ID IL = 1, IF IX(ANLEG)). 10=1, IFIX(ANSIGZ)), 1=1, IFIX(AS1-1.)) Cross sections pertaining to the (MF, MT) pair in the AG-th incident energy group Note For MF=3, MT=1 ANS1=2, AGLA=1, ANVAR=1 The P and P weighted total cross sections and fluxes are given For MF=3, MT=1 ANS1=1, AGLA = 1, ANVAR = 1 only P 0 weighted cross sections and P Q flux are given

60 For MF = 5. MT=455 ANS1=6 Delayed neutron spectra by time group are given using the Legendre index for time. Time constants are given in the flux positions. For MF=6 Scattering matrices are stored by Legendre order. Excess zeroes are removed by saving only a band of consecutive non-zero values. AGLA is used to determine the group index of the first element of the band. For MF=16, Production matrices Record 7 (Continuation of cross sections) Not present if NB=0 on record 6, 306 data values on record 6. if there are less than MAT MF MT NB NU SIGMA (I, ID, ID ENDF isotope identification number ENDF file number ENDF section number Number of variables pertaining to the (MF,MT) pair and referring to incident AG-th energy group, defined in previous record 6, which are not on record 7. Number of remaining variables on record 7. Remaining cross sections pertaining to the (MF, MT) pair in the AG-th energy group. Continuation of the data on record 6, using the same strategy. Note There can be many records 7. The last one is characterized by NB=0.

61 Record 8 (end of section MT> Present only at the end of each section, changes. i.e. when MT MT MF 0, ENDF isotope identification number ENDF file number not used total number of remaining variables on record 8 not used Note There is no end of file record for MF=3 or 6 on a GENOF file Record 9 (end of material MAT) not used not used Record 10 (end of GENOF file) not used not used

62 GENDF- file- loop-structure Record 1 (tape identification) * Record 3 (Summarizing parameters) * Record 4 (End of file 1) * repeat records 2-9 for each material on GENOF file. Each material consists of a pair (MAT, TEMP). TEMP is first varied. ****** Record 5 (<MF, MT) pair description) * Record 6 (Cross Sections) * Record 7 (Remaining Cross Sections) * Record 8 (End of section MT) * repeat records 5-8 for each (MF, MT) ***** pa i r. * Record 8 is present only at the end of each section. ***** Record 9 (End of material (MAT, TEMP)) Record 10 (End of GENOF file)

63 APPENDIX C GENERAL ATOMIC NEUTRON GROUP STRUCTURE 193-ENERGY GROUPS GROUP ENERGY BOUNDARIES (EV) OUP LOWER UPPFR GROUP LOWER UPPER E E E+07 1.OOOOOE E E E E E E E E E E E E E E E E+0C E E E E E E E E E E E E+O E E E E+O E E E E E E E E E E E E E E E E E E+07-1.O0OO0E E E E E E E E E E E E E E E E E E E E E ^E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E OE+OO OE+OO OE+O0 2.05O0OE+O OE E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E+00

64 GROUP ENERGY GROUP LOWER UPPER BOUNDARIES (EV) GROUP LOWER UPPER E E E E E+0Ö ÖÖOE O0OE + 0O - l.5500üe E+0O E OE+0O E OE+0O E E E E E E E+00 i _ OE+0O 14000E OE+0O E E E+O E E E E E E+00 i.05500e+00-1.o65ooe E E E E E E E E E E E E E E E E E E E-01-9 OOOOOE OE-O E E E E E E E E E E E E E E E OOE-O E E E E E E E E E E E E E E E E E E OE O0E-O E E E OOE-O E E E E-O E E O0E E E O0E E E O0E E O0E E E E E E E E E E E E E E E E E E E E E E E E E E E-02 9 OOOOOE E OOOOOE E E E E E C00E E E E E E E E E OO0E-O E E E E E E E E E E E E E E E E E E E E E E E E E E-02 OOOOOE E-03 OOOOOE-03 50O00E-O3 OOOOOE E-03