.. |nbsp| unicode:: 0xA0 :trim: Acknowledgements ================ Development and testing of ORIGEN data resources, libraries, and methods have been sponsored by many organizations including the US Nuclear Regulatory Commission (NRC), the US Department of Energy (DOE), and nuclear power and research institutions. Version Information =================== Following is a description of the data resources available for use with ORIGEN in different SCALE versions. Methodologies and algorithms used in applying the data are described in the ORIGEN chapter. Version 6.3 (2021) ------------------ *Data lead(s):* A. Holcomb and W. Wieselquist Nuclear data in ORIGEN has had minor updates for SCALE 6.3. The old ENDF/B-VII.0 libraries have been removed. There are new libraries for 302-groups (for fast-spectrum systems) and 1597-groups as an very fine group option for any spectrum. Data for recoverable energy per capture for Gd-155 and Gd-157 has been updated from 5 MeV to 8.5360 and 7.9370 MeV, respectively. Version 6.2 (2016) ------------------ *Data lead(s):* I. C. Gauld, D. Wiarda, M. Pigni, and W. Wieselquist Nuclear data in ORIGEN are unchanged from SCALE 6.1.3 except for the modification of independent fission yields for thermal fission of :sup:`235`\ U and :sup:`241`\ Pu and fast fission of :sup:`238`\ U to provide greater compatibility between the direct and cumulative fission yields when using the updated decay data from ENDF/B-VII.1. Additionally, ORIGEN no longer has its own independent source of nuclide mass and abundance data and now relies on the SCALE Standard Composition library so that there is consistency in this data across SCALE. D. Mueller and W. Wieselquist are acknowledged for testing of the new yield data. W. Wieselquist and S. Hart are acknowledged for the revision of this chapter. Version 6.1.3 (2011) -------------------- *Data lead (s):* I. C. Gauld and D. Wiarda SCALE 6.1 represented a complete revision and update of the nuclear data available in ORIGEN. The following is a summary from the SCALE 6.1 manual. The ORIGEN data libraries include nuclear decay data, neutron reaction cross sections, neutron induced fission product yields, delayed gamma-ray emission data, and neutron emission data. The nuclear decay data libraries have been updated based on ENDF/B-VII evaluations and expanded to include 903 activation products and structural materials, 174 actinides, and 1149 fission products. The cross section libraries have been revised using evaluations from the JEFF-3.0/A neutron activation file, containing data for 774 target nuclides, more than 12,000 neutron-induced reactions, and more than 20 different reaction types below 20 MeV. The JEFF-3.0/A activation file is processed into several multigroup cross section libraries, from 44 groups to 238 groups, that can be used to determine the neutron reaction transition rates in ORIGEN. Energy-dependent ENDF/B-VII fission product yields are provided for 30 fissionable actinides. Photon yield data libraries have been updated based on the most recent ENSDF nuclear structure evaluations processed using the NuDat program. The photon libraries contain discrete photon line energy and intensity data for decay gamma and x-rays emission for 982 radionuclides, prompt and equilibrium continuum fission product spectra from spontaneous fission, :math:`\left( \alpha,n \right)` reactions in oxide fuel, and bremsstrahlung from decay beta (negatron and positron) particles slowing down in either UO\ :sub:`2` fuel or water matrix. Methods and data libraries used to calculate the neutron yields and energy spectra for spontaneous fission, :math:`\left( \alpha,n \right)` reactions in any matrix, and delayed neutron emission are adopted from the SOURCES code. The libraries used by ORIGEN can be coupled directly with detailed and problem-dependent physics calculations to obtain self-shielded problem-dependent cross sections based on the most recent evaluations of ENDF/B-VII. In addition, the library formats allow multiple sets of cross section data to be stored on a library to represent the changes in cross sections during irradiation. Introduction ============ ORIGEN data resources include nuclear decay data, multigroup neutron reaction cross sections, neutron-induced fission product yields, and decay emission data for photons, neutrons, alpha particles and beta particles. The available resources are summarized in :numref:`tab-origen-resources` and described in greater detail in the subsequent sections. The "Unit" column shows the corresponding unit number for use with FIDO input systems (e.g. with COUPLE). .. tabularcolumns:: |\Y{0.25}|\Y{0.1}|\Y{0.05}|\Y{0.15}|\Y{0.45}| .. table:: Available resources in ORIGEN :class: longtable :name: tab-origen-resources +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | **Description** | **Alias** | **Unit** | **Category** | **Location in SCALE data directory** | +================================+===========+==========+==============+===========================================+ | ENDF/B-VII.1 decay data | decay | 27 | Decay | origen_data/origen.rev03.decay.data | | | | | | | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | ENDF/B-VII.0-based | yields | 17 | Yield | origen_data/origen.rev05.yields.data | | fission yield data | | | | | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | JEFF-3.0/A - 56g | n56 | 75 | Reaction | origen.rev01.jeff56g | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | JEFF-3.0/A - 200g | n200 | 78 | Reaction | origen.rev03.jeff200g | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | JEFF-3.0/A - 252g | n252 | 74 | Reaction | origen.rev01.jeff252g | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | JEFF-3.0/A - 302g | n302 | | Reaction | origen.rev00.jeff302g | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | JEFF-3.0/A - 1597g | n1597 | | Reaction | origen.rev00.jeff1597g | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Energy per fission and capture | | | Energy | n/a | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Master photon (x-ray and | | | Emission | origen_data/origen.rev04.mpkkxgam.data | | gamma) emission data | | | | | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Spontaneous fission and | | | Emission | origen_data/origen.rev00.mpsfangm.data | | :math:`\left(\alpha,n \right)` | | | | | | reaction gamma rays | | | | | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Bremsstrahlung from beta | | | Emission | origen_data/origen.rev00.mpbrh2om.data | | particles slowing down in | | | | | | water | | | | | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Bremsstrahlung from positrons | | | Emission | origen_data/origen.rev00.mpbrh2op.data | | slowing down in water | | | | | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Bremsstrahlung from beta | | | Emission | origen_data/origen.rev00.mpbruo2m.data | | particles slowing down in | | | | igen_data/or | | UO\ :sub:`2` | | | | | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Bremsstrahlung from positrons | | | Emission | origen_data/origen.rev00.mpbruo2p.data | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Neutron source emission and | | | Emission | origen_data/origen.rev01.alphdec.data | | alpha decay data | | | | rigen_data/o | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Alpha particle stopping cross | | | Emission | origen_data/origen.rev00.stcoeff.data | | section expansion coeffcients | | | | | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Target | | | Emission | origen_data/origen.rev00.alphyld.data | | :math:`\left( \alpha,n\right)` | | | | | | product excited level | | | | | | branching data | | | | | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Target | | | Emission | origen_data/origen.rev00.alphaxs.data | | :math:`\left(\alpha,n \right)` | | | | | | cross section data | | | | | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ | Beta source emission data | | | Emission | origen_data/origen.rev00.ensdf95beta.data | +--------------------------------+-----------+----------+--------------+-------------------------------------------+ Decay Resource ============== The nuclear data stored on the decay resource is based on ENDF/B-VII.1 evaluations :cite:`ODR-chadwick_endfb-vii_2011`, including half-lives, decay modes and branching fractions, and recoverable energy per disintegration. Decay modes include beta (:math:`\beta^-`), positron (:math:`\beta^+`) and electron capture (EC), isomeric transition (IT), alpha (:math:`\alpha`), spontaneous fission (SF), delayed neutron (:math:`\beta^-\,n`) emission, neutron emission (n), double beta decay :math:`\left( \beta^- \beta^- \right)`, and decay by beta and alpha emission (:math:`\beta^- \alpha`). The decay resource also includes radiotoxicity factors based on the radioactivity concentration guides (RCGs) for air and water as defined in Part 10, Title 20, of the Code of Federal Regulations (10CFR20) :cite:`ODR-10cfr20`. RCGs specify the maximum permissible concentrations of an isotope in soluble and insoluble forms for both ingestion and inhalation and for occupational and unrestricted exposure. The radiotoxicity is calculated as the dilution volume of a nuclide for cases of direct ingestion or inhalation. The values are defined to be the smaller (i.e., more toxic) of the values for soluble and insoluble forms of the isotope. The maximum permissible RCGs for air and water are the public exposure limits for adult ingestion and inhalation dose coefficients of ICRP Publication 72 :cite:`ODR-icrp72`. External exposure dose coefficients for noble gases were obtained from the Environmental Protection Agency (EPA) Federal Guidance Report 12 :cite:`ODR-epa1993`. Recoverable energy includes the delayed energy from all electron-related radiations (e.g., :math:`\beta^-`, :math:`\beta^+`, Auger electrons), all gamma rays, x-rays, annihilation radiations, and the average energy of all heavy charged particles and delayed neutrons. The average alpha energy includes the energy of the recoil nucleus. A part of the recoverable energy per decay not included in the ENDF/B-VII.1 values is the additional contribution from spontaneous fission. This energy was calculated as the product of the spontaneous fission branching fraction and recoverable energy per fission using a value of 200 MeV per fission and then added to the ENDF/B-VII.1 recoverable Q energy. A value of 12.56 MeV gamma energy per fission was used in computing the fraction of recoverable spontaneous fission energy from gamma rays. External Bremsstrahlung radiation is **not** included in the Q-value since the Bremsstrahlung spectrum depends on electron interactions with the medium that contains the decay nuclide. The energy from capture gamma rays accompanying :math:`\left( \alpha,n \right)` reactions is not included either since it also depends on the medium. Appendix A describes the decay resource file format. It is important to note that the decay resource not only defines fundamental decay data, but also the complete ORIGEN nuclide set, including the "duplicates" of nuclides across sublibraries. For example, a version of :sup:`155`\ Gd is contained in both the light nuclide/activation product and fission product sublibraries. Appendix D includes the full list of the nuclides on the ORIGEN decay library "end7dec" created by COUPLE based on the current decay resource, including duplicates. Appendix E contains a list of the fundamental decay data only, without duplicates. To consider a different set of nuclides in an ORIGEN calculation, the current process is to alter the decay resource and then regenerate the "end7dec" decay library with COUPLE. By default, all subsequent libraries created from COUPLE using problem-dependent reaction transitions are based on the "end7dec" decay library and will therefore include the modified nuclide set. Neutron Reaction Resource ========================= The neutron cross sections defining the nuclear reaction transmutation rates use a comprehensive collection of nuclear data evaluations compiled from the JEFF-3.0/A neutron activation files :cite:`ODR-jeffdoc-982`. The JEFF-3.0/A files contain continuous energy neutron data for 774 target nuclei, including ground and metastable excited states, and 12,617 neutron-induced reactions below 20 MeV. The JEFF-3.0/A cross section data are developed directly from the European Activation File (EAF-2003) :cite:`ODR-fus-486` formatted as standard ENDF-6 format data. JEFF-3.0/A cross sections are stored using File 3, multiplicities on File 10, and isomeric branching to different metastable levels using File 9. The evaluations include many reactions that may be important for modeling fast fission and other high-energy systems. Neutron reactions are available for 23 reaction types, including :math:`\left(n,n' \right)`, :math:`\left(n,2n \right)`, :math:`\left(n,3n \right)`, :math:`\left(n,f \right)`, :math:`\left(n,n' \alpha \right)`, :math:`\left(n,2n\alpha\right)`, :math:`\left(n,3n\alpha \right)`, :math:`\left(n,n'p \right)`, :math:`\left(n,n2\alpha \right)`, :math:`\left(n,n'd \right)`, :math:`\left(n,n't \right)`, :math:`\left(n,n'{}^3 He \right)`, :math:`\left(n,4n \right)`, :math:`\left(n,2np \right)`, :math:`\left(n,\gamma \right)`, :math:`\left(n,p \right)`, :math:`\left(n,d \right)`, :math:`\left(n,t \right)`, :math:`\left(n,{}^3\ He \right)`, :math:`\left(n,\alpha \right)`, :math:`\left(n,2\alpha \right)`, :math:`\left(n,2p \right)`, and :math:`\left(n,p\alpha \right)`. The JEFF-3.0/A evaluations also include extensive compilations of energy-dependent branching fractions that define neutron reaction transitions to ground and metastable energy states. Energy-dependent branching is fully implemented in the ORIGEN cross section libraries. Implementation of the JEFF-3.0/A cross sections as ORIGEN multigroup data was accomplished by processing and collapsing the JEFF-3.0/A pointwise cross sections into a standard multigroup AMPX format using ENDF data-processing modules of the AMPX :cite:`ODR-dugr2002` cross section processing code system. The collapse is performed using a thermal Maxwellian-1/E-fission-1/E weighting spectrum (see :numref:`fig-origen-collapse-flux`) to provide infinite dilution multigroup cross sections. .. _fig-origen-collapse-flux: .. figure:: ../figs/ORIGEN/collapse-flux.png Pointwise flux spectrum used to generate collapsed cross section libraries. Neutron reactions with transitions to multiple states of the daughter product are represented using separate cross sections to the ground and metastable states. A special reaction identifier (MT') is defined for this implementation of metastable transitions as .. math:: \text{MT}' = \text{MT}*10000 + 100*\text{LP} + \text{LT} :label: eq-MT-special where MT is the reaction identifier, LP is the product metastable state, and LT is the target metastable state. Using the :sup:`187`\ W(n,3n)\ :sup:`185`\ W cross section (MT=17) as an example, the reaction identifier 170000 defines the partial cross section to the ground state of :sup:`185`\ W, and 170100 defines the cross section to metastable :sup:`185m`\ W. Cross section data from the JEFF-3.0/A neutron activation file are first converted to point-wise cross section data, are Doppler broadened to 900K, and then they are collapsed to different group structures. The following group strucures are available in SCALE: * 238-group neutron (thermal applications), * 252-group neutron (thermal applications), * 56-group neutron (thermal applications), * 200-group neutron (fast applications and shielding), * 47-group neutron (applications using the BUGLE shielding transport library), * 49-group neutron (collapsed version of 238 groups), * 44-group neutron (collapsed version of 238 groups), and * 999-group neutron (multipurpose). Several minor modifications were made to the JEFF-3.0/A data: * The :sup:`239`\ Np radiative neutron capture cross section was replaced with data from ENDF/B-VII.0. Neutron capture using JEFF-3.0/A cross sections was significantly larger than ENDF/B-VII.0 due to differences in the resonance cross section region. Although experimental resonance parameters are not available for :sup:`239`\ Np, comparisons of :sup:`240`\ Pu production during irradiation :cite:`ODR-gump1954` obtained using the two evaluations showed that better agreement with the experiment was obtained using the ENDF/B-VII.0 evaluation. * The :math:`{}^{241} Am (n,\gamma)` branching fraction to the :sup:`242`\ Am ground and metastable states was replaced by the evaluation from ENDF/B-VII.0 to yield better agreement with the results of destructive radiochemical assay measurements of irradiated fuels. The branching fraction of :sup:`241`\ Am to :sup:`242m`\ Am for thermal neutron capture changed from 8.2% in JEFF-3.0/A to 10.0% in ENDF/B-VII.0. The cross section library header record information and a complete list of nuclides in JEFF-3.0/A libraries developed for ORIGEN are provided in Appendix E. Because JEFF-3.0/A-based libraries are formatted as standard AMPX working libraries, they can be accessed and/or manipulated using standard AMPX utility modules in SCALE. For example, multigroup cross sections may be listed using the PALEALE module. Additionally, the data may be visualized using the Fulcrum user interface. Cross section plots of the 238-group JEFF-3.0/A library are illustrated in :numref:`fig-origen-jeff-w187` for :math:`(n,\gamma)`, :math:`(n,\alpha)`, (n,2n), and (n,3n) cross sections to the ground and metastable states. Before the cross sections in ORIGEN can be used, they must be collapsed with a user-defined multigroup flux to a one-group cross section and added to the ORIGEN binary library (see the COUPLE input description). .. _fig-origen-jeff-w187: .. figure:: ../figs/ORIGENdata/JEFF-w-187-plot.* 238-group JEFF-3.0/A cross sections for :sup:`187`\ W. Fission Yield Resource ====================== The fission-yield resource contains the energy-dependent direct yields of each fission product for 30 fissionable actinides, including :sup:`227,228,232`\ Th, :sup:`231`\ Pa, :sup:`232--238`\ U, :sup:`238-242`\ Pu, :sup:`241,242m,243`\ Am, :sup:`237,238`\ Np, :sup:`242-246,248`\ Cm, :sup:`249,252`\ Cf, and :sup:`254`\ Es. Independent (direct) fission product yields are stored as atom percent per fission, and except for :sup:`235`\ U(thermal), :sup:`238`\ U(fast), and :sup:`241`\ Pu(thermal), they are obtained from ENDF/B-VII.0 :cite:`ODR-chadwick_endfb-vii_2006` File 8 and MT=454. Revised independent fission product yields for :sup:`235`\ U(thermal), :sup:`238`\ U(fast), and :sup:`241`\ Pu(thermal) were adopted to address inconsistencies between the direct and cumulative fission yields in ENDF/B-VII.0 caused by the use of updated nuclear decay schemes in the decay sublibrary :cite:`ODR-pifrga2015,ODR-frwepi2015`. Namely, recent changes in the decay data, particularly the delayed neutron branching fractions, result in calculated fission product concentrations that do not agree with the cumulative fission yields in the ENDF/B-VII.0 library. These issues were particularly evident for the three cited isotopes because their fissioning systems result in a preferential formation of fragments that are sensitive to the changes in the decay data. For example, a study on :sup:`239`\ Pu(thermal) showed negligible differences between cumulative yields calculated (using the recent decay data sublibrary) and the cumulative yields in ENDF/B-VII.0. Energy-dependent product yields are available for thermal, fast, and high-energy incident neutron energies. For fast fission, the value of the energy of incident neutron was modified from the value of 500 keV tabulated in ENDF/B-VII.0 to more accurately represent the relationship between the energy distribution of the neutrons causing fission and the and the fission neutron spectrum energy. For this implementation of the yield data, the effective incident neutron energy for fast fission was adjusted from 500 keV to 2.0 MeV to better reflect the average fission energy of most nuclides. The neutron energies for thermal fission (0.0253 eV) and high energy fission (14 MeV) are unchanged. The fission product yields also include cumulative ternary yields from the JEF-2.2 fission yield library :cite:`ODR-jef-2.2` for :sup:`3`\ H and :sup:`4`\ He. The nuclide :sup:`3`\ He was also added to the fission product library since it is a decay product of tritium. Note that inclusion of fission yields for each actinide in an ORIGEN library can be controlled by the user through COUPLE. Actinides not assigned with explicit yields do not produce fission products during fission. .. table:: Fissionable isotopes having explicit fission yields :class: longtable :name: tab-origen-fiss-isotopes +----------------------+-------------------------------------------+ | **Nuclide** | **Neutron-induced fission energies** | | | [#fiss-energy]_ | +======================+======================+======+=============+ | :sup:`227`\ Th | Thermal | | | +----------------------+----------------------+------+-------------+ | :sup:`229`\ Th | Thermal | | | +----------------------+----------------------+------+-------------+ | :sup:`232`\ Th | | Fast | High energy | +----------------------+----------------------+------+-------------+ | :sup:`231`\ Pa | | Fast | | +----------------------+----------------------+------+-------------+ | :sup:`232`\ U | Thermal | | | +----------------------+----------------------+------+-------------+ | :sup:`233`\ U | Thermal | | | +----------------------+----------------------+------+-------------+ | :sup:`234`\ U | | Fast | High energy | +----------------------+----------------------+------+-------------+ | :sup:`235`\ U | Thermal | Fast | High energy | +----------------------+----------------------+------+-------------+ | :sup:`236`\ U | | Fast | High energy | +----------------------+----------------------+------+-------------+ | :sup:`237`\ U | | Fast | | +----------------------+----------------------+------+-------------+ | :sup:`238`\ U | | Fast | High energy | +----------------------+----------------------+------+-------------+ | :sup:`237`\ Np | Thermal | Fast | High energy | +----------------------+----------------------+------+-------------+ | :sup:`238`\ Np | | Fast | | +----------------------+----------------------+------+-------------+ | :sup:`238`\ Pu | | Fast | | +----------------------+----------------------+------+-------------+ | :sup:`239`\ Pu | Thermal | Fast | High energy | +----------------------+----------------------+------+-------------+ | :sup:`240`\ Pu | Thermal | Fast | High energy | +----------------------+----------------------+------+-------------+ | :sup:`241`\ Pu | Thermal | Fast | | +----------------------+----------------------+------+-------------+ | :sup:`242`\ Pu | Thermal | Fast | High energy | +----------------------+----------------------+------+-------------+ | :sup:`241`\ Am | Thermal | Fast | High energy | +----------------------+----------------------+------+-------------+ | :sup:`242\ m`\ Am | Thermal | | | +----------------------+----------------------+------+-------------+ | :sup:`243`\ Am | | Fast | | +----------------------+----------------------+------+-------------+ | :sup:`242`\ Cm | | Fast | | +----------------------+----------------------+------+-------------+ | :sup:`243`\ Cm | Thermal | Fast | | +----------------------+----------------------+------+-------------+ | :sup:`244`\ Cm | | Fast | | +----------------------+----------------------+------+-------------+ | :sup:`245`\ Cm | Thermal | | | +----------------------+----------------------+------+-------------+ | :sup:`246`\ Cm | | Fast | | +----------------------+----------------------+------+-------------+ | :sup:`248`\ Cm | | Fast | | +----------------------+----------------------+------+-------------+ | :sup:`249`\ Cf | Thermal | | | +----------------------+----------------------+------+-------------+ | :sup:`251`\ Cf | Thermal | | | +----------------------+----------------------+------+-------------+ | :sup:`254`\ Es | Thermal | | | +----------------------+----------------------+------+-------------+ .. [#fiss-energy] Neutron energy causing fission Energy Resource =============== The energy resource is a set of data defined internally to ORIGEN to compute the total power during irradiation if the flux is known, or the total flux if the power is known. The data include the energy contributed by fission and capture. The recoverable energy values taken primarily from ENDF/B evaluations are listed in :numref:`tab-origen-recoverable-en-fc` and :numref:`tab-origen-recoverable-en-cap`. The recoverable energy for fission and neutron capture for nuclides not listed in the tables are assumed to be 200 MeV and 5.0 MeV, respectively. .. table:: Recoverable energy (MeV) values for actinides :class: longtable :name: tab-origen-recoverable-en-fc =============== =========== =========== **Nuclide** **Fission** **Capture** :sup:`230`\ Th 190.00 5.010 :sup:`232`\ Th 189.21 4.786 :sup:`233`\ Th 190.00 6.080 :sup:`231`\ Pa 190.00 5.660 :sup:`233`\ Pa 189.10 5.197 :sup:`232`\ U 200.00 5.930 :sup:`233`\ U 191.29 6.841 :sup:`234`\ U 190.30 5.297 :sup:`235`\ U 194.02 6.545 :sup:`236`\ U 192.80 5.124 :sup:`238`\ U 198.12 4.804 :sup:`237`\ Np 195.10 5.490 :sup:`239`\ Np 200.00 4.970 :sup:`238`\ Pu 197.80 5.550 :sup:`239`\ Pu 200.05 6.533 :sup:`240`\ Pu 199.79 5.241 :sup:`241`\ Pu 202.22 6.301 :sup:`242`\ Pu 200.62 5.071 :sup:`243`\ Pu 200.00 6.020 :sup:`241`\ Am 202.30 5.529 :sup:`242m`\ Am 202.29 6.426 :sup:`243`\ Am 202.10 5.363 :sup:`244`\ Cm 200.00 6.451 :sup:`245`\ Cm 200.00 6.110 =============== =========== =========== .. table:: Recoverable energy (MeV) values for activation and fission products :class: longtable :name: tab-origen-recoverable-en-cap =============== =========== **Nuclide** **Capture** :sup:`1`\ H 2.225 :sup:`10`\ B 2.790 :sup:`16`\ O 4.143 :sup:`56`\ Fe 7.600 :sup:`58`\ Ni 9.020 :sup:`90`\ Zr 7.203 :sup:`91`\ Zr 8.635 :sup:`92`\ Zr 6.758 :sup:`96`\ Zr 5.571 :sup:`95`\ Mo 9.154 :sup:`95`\ Tc 7.710 :sup:`101`\ Ru 9.216 :sup:`103`\ Rh 6.999 :sup:`105`\ Rh 7.094 :sup:`109`\ Ag 6.825 :sup:`131`\ Xe 8.936 :sup:`135`\ Xe 7.880 :sup:`133`\ Cs 6.704 :sup:`134`\ Cs 6.550 :sup:`143`\ Nd 7.817 :sup:`145`\ Nd 7.565 :sup:`147`\ Pm 5.900 :sup:`148`\ Pm 7.266 :sup:`148m`\ Pm 7.266 :sup:`147`\ Sm 8.140 :sup:`149`\ Sm 7.982 :sup:`150`\ Sm 5.596 :sup:`151`\ Sm 8.258 :sup:`152`\ Sm 5.867 :sup:`153`\ Eu 6.444 :sup:`154`\ Eu 8.167 :sup:`155`\ Eu 6.490 =============== =========== Emission Resources ================== The two main groups for emission resources are the photon (gamma) resource, which includes beta particle emission data, and the neutron resource, which includes alpha emission data. Gamma Emission -------------- The resources for gamma emission are stored as separate files (see :numref:`tab-origen-photon-files`) containing the photon data associated with different modes of decay or photon production. The photon data sets include decay gamma and x-ray line-energy data, gamma rays accompanying spontaneous fission, gamma rays accompanying :math:`\left( \alpha,n \right)` reactions in oxide fuels, and Bremsstrahlung spectra from decay electrons/positrons slowing down in UO\ :sub:`2` and water. The photon energy spectra can be generated in any energy group structure for all activation products, actinides, and fission product nuclides with photon yield data. .. table:: Photon data files :class: longtable :name: tab-origen-photon-files ============== ================================================================== **File name** **Description** ============== ================================================================== MPDKXGAM x-ray and gamma emissions line data MPSFANGM spontaneous fission and :math:`\left( \alpha,n \right)` reactions MPBRH2OM bremsstrahlung from beta particles slowing down in water MPBRH2OP bremsstrahlung from positrons slowing down in water MPBRUO2M bremsstrahlung from beta particles slowing down in UO\ :sub:`2` MPBRUO2P bremsstrahlung from positrons slowing down in UO\ :sub:`2` ============== ================================================================== All photon data sets are constructed with the same format (see Appendix C). The majority of the photon emissions are discrete energy lines. Photon continuum data, used to represent Bremstrahlung and some other gamma-ray emission spectra, are stored at discrete energies and approximately expanded to a continuum, as needed. Gamma and x-ray yields are directly from ENDF/B-VII.1 decay files containing spectral data for decay transitions of 1,132 nuclides. A separate file contains emission spectra for gamma rays accompanying spontaneous fission and for gamma rays accompanying :math:`\left( \alpha, n \right)` reactions in oxide fuels :cite:`ODR-crhagr1979`. The spontaneous fission spectra combine prompt and equilibrium fission product gamma-ray components. The prompt spectrum is similar to that of :sup:`235`\ U, and the delayed fission product gamma intensity at equilibrium is about 0.75 of that from the prompt fission gamma rays. Based on measured prompt fission gamma spectra from :sup:`235`\ U, spontaneous fission spectra are computed from the following approximation: .. math:: N\left(E \right) \cong \begin{cases} 11.5 & 0.1 \leq\ E \leq 0.6\ MeV \\ 35.4 e^{-1.78 E} & 0.6 \leq E < 1.5\ MeV \\ 12.6 e^{-1.09 E} & 1.5 \leq E \leq 10.5\ MeV \\ 0 & \text{otherwise.} \end{cases} :label: eq-origen-sfn-spec where N(E) = number of photons per unit energy per fission (photons/MeV per fission) at energy E, where E is the photon energy (MeV). For medical and industrial spontaneous fission source applications, a more accurate simulation of the source may be desirable. Work has been performed on :sup:`252`\ Cf source modeling to explicity represent the fission product generation from fission and the delayed gamma emission. In this application, the equilibrium spontaneous fission gamma spectrum was replaced with an evaluation of the :sup:`252`\ Cf prompt gamma spectrum, and the delayed fission product gamma rays was modeled explicitly in ORIGEN by generating the time-dependent fission products using :sup:`252`\ Cf spontaneous fission product yields from ENDF/B-VII.0 :cite:`ODR-fogawa2011`. This was performed by adding decay transitions to the ORIGEN library from the actinides to the fission products. The spectrum of gamma rays accompanying :math:`\left( \alpha,n \right)` reactions is based on reaction data for alpha interactions on :sup:`18`\ O and from studies for :sup:`238`\ PuO\ :sub:`2` systems. The spectrum is computed from the approximation: .. math:: N\left( E \right) \cong 2.13 \cdot 10^{-8}\ e^{- 1.38E} where N(E) = number of photons per unit energy per alpha decay (photons/MeV per disintegration) at energy E (MeV). The photon yields in this data set are continuum spectra represented by discrete lines with an energy width of 500 keV and range from 250 keV to 10.25 MeV. Two photon data sets contain bremsstrahlung spectra from decay electrons and positrons slowing down in a UO\ :sub:`2` fuel matrix. The yields are in the form of continuum spectra represented in the data sets as discrete lines using up to 70 quasi-logarithmic spaced energy points over the energy range between 0 and 13.5 MeV. Two libraries contain bremsstrahlung spectra from decay electrons and positrons slowing down in water. Bremsstrahlung spectra were calculated using a computer program developed by Dillman *et al.* :cite:`ODR-disnfo1973` using beta spectra derived from **Evaluated Nuclear Structure Data Files (**\ ENSDF) decay data with a computer program written by Gove and Martin :cite:`ODR-goma1971`. Neutron Emission ---------------- There are four neutron emission resources used by ORIGEN to calculate the neutron intensities and spectra: (1) neutron decay data, (2) an \alpha-particle stopping power, (3) a target :math:`\left( \alpha,n \right)` cross section, and (4) a target :math:`\left( \alpha,n \right)` product level branching. All of the neutron data are stored in a text format with names and descriptions given in :numref:`tab-origen-neutron-libs`. The neutron decay data contain basic decay information for decay processes that lead to direct and indirect emission of neutrons, including spontaneous fission branching fractions, alpha decay branching fractions, delayed neutron branching fractions, alpha-particle decay energies, Watt fission spectrum parameters, and delayed neutron spectra. The stopping cross sections, :math:`\left( \alpha,n \right)` target cross sections, and product-level branching data are used in calculating the neutron yield and spectra from :math:`\left( \alpha,n \right)` reactions. The neutron data were obtained directly from the updated SOURCES-4B code package. The sources of the neutron data are described by Shores :cite:`ODR-sh2000`. An update was made to correct an error in the :sup:`250`\ Cf spontaneous fission neutron branching fraction in the neutron source decay data distributed with the SOURCES code. The :sup:`250`\ Cf branching fraction was incorrectly assigned the value from :sup:`252`\ Cf of :math:`3.092 \cdot 10^{-2}`. A corrected value of :math:`7.700 \cdot 10^{-4}` from ENDF/B-VII.1 is used. .. table:: Neutron source data libraries :class: longtable :name: tab-origen-neutron-libs ============= ==================================================================== **File name** **Description** ALPHDEC Neutron source decay data STCOEFF Stopping cross section expansion coefficients ALPHYLD Target :math:`\left( \alpha,n \right)` product level branching ALPHAXS Target :math:`\left( \alpha,n \right)` cross section ============= ==================================================================== The neutron source decay contains spontaneous fission data for the 49 actinides listed in :numref:`origen-nucl-sfn`. These data include the spontaneous fission branching fraction, the number of neutrons per fission (:math:`nu`), and the watt spectrum parameters for spontaneous fission. The spontaneous fission neutron energy spectrum is approximated using spectral parameters A and B, such that .. math:: N\left( E \right) \cong \text{C\ }e^{ - \frac{E}{A}}\sinh \sqrt{\text{BE}} :label: eq-watt-spec where *E* is the neutron energy and *C* is a normalization constant. .. table:: Nuclides with spontaneous fission data and spectral parameters :class: longtable :name: origen-nucl-sfn ================ ================ ================ ================ ================ :sup:`230`\ Th :sup:`239`\ U :sup:`240`\ Pu :sup:`244`\ Am :sup:`250`\ Cm :sup:`232`\ Th :sup:`236`\ Np :sup:`241`\ Pu :sup:`244m`\ Am :sup:`249`\ Bk :sup:`231`\ Pa :sup:`236m`\ Np :sup:`242`\ Pu :sup:`240`\ Cm :sup:`248`\ Cf :sup:`232`\ U :sup:`237`\ Np :sup:`243`\ Pu :sup:`241`\ Cm :sup:`250`\ Cf :sup:`233`\ U :sup:`238`\ Np :sup:`244`\ Pu :sup:`242`\ Cm :sup:`252`\ Cf :sup:`234`\ U :sup:`239`\ Np :sup:`240`\ Am :sup:`243`\ Cm :sup:`254`\ Cf :sup:`235`\ U :sup:`236`\ Pu :sup:`241`\ Am :sup:`244`\ Cm :sup:`253`\ Es :sup:`236`\ U :sup:`237`\ Pu :sup:`242`\ Am :sup:`245`\ Cm :sup:`254m`\ Es :sup:`237`\ U :sup:`238`\ Pu :sup:`242m`\ Am :sup:`246`\ Cm :sup:`255`\ Es :sup:`238`\ U :sup:`239`\ Pu :sup:`243`\ Am :sup:`248`\ Cm ================ ================ ================ ================ ================ Delayed neutron branching fractions and neutron spectra for 105 fission products are listed in :numref:`tab-origen-nucl-dn`. The delayed neutron spectra are tabulated in discrete 10 keV bins from 50 keV to about 2 MeV. .. table:: Nuclides with delayed neutron emission spectral data :class: longtable :name: tab-origen-nucl-dn ============== ============== =============== =============== =============== :sup:`79`\ Zn :sup:`89`\ Br :sup:`97`\ Y :sup:`128`\ In :sup:`41`\ I :sup:`79`\ Ga :sup:`90`\ Br :sup:`97m`\ Y :sup:`129`\ In :sup:`42`\ I :sup:`80`\ Ga :sup:`91`\ Br :sup:`98`\ Y :sup:`129m`\ In :sup:`43`\ I :sup:`81`\ Ga :sup:`92`\ Br :sup:`98m`\ Y :sup:`130`\ In :sup:`141`\ Xe :sup:`82`\ Ga :sup:`93`\ Br :sup:`99`\ Y :sup:`131`\ In :sup:`142`\ Xe :sup:`83`\ Ga :sup:`92`\ Kr :sup:`100`\ Y :sup:`132`\ In :sup:`143`\ Xe :sup:`83`\ Ge :sup:`93`\ Kr :sup:`104`\ Zr :sup:`133`\ Sn :sup:`144`\ Xe :sup:`84`\ Ge :sup:`94`\ Kr :sup:`105`\ Zr :sup:`134`\ Sn :sup:`141`\ Cs :sup:`85`\ Ge :sup:`95`\ Kr :sup:`103`\ Nb :sup:`135`\ Sn :sup:`142`\ Cs :sup:`86`\ Ge :sup:`92`\ Rb :sup:`104`\ Nb :sup:`134m`\ Sb :sup:`143`\ Cs :sup:`84`\ As :sup:`93`\ Rb :sup:`105`\ Nb :sup:`135`\ Sb :sup:`144`\ Cs :sup:`85`\ As :sup:`94`\ Rb :sup:`106`\ Nb :sup:`136`\ Sb :sup:`145`\ Cs :sup:`86`\ As :sup:`95`\ Rb :sup:`109`\ Mo :sup:`137`\ Sb :sup:`146`\ Cs :sup:`87`\ As :sup:`96`\ Rb :sup:`110`\ Mo :sup:`136`\ Te :sup:`147`\ Cs :sup:`87`\ Se :sup:`97`\ Rb :sup:`109`\ Tc :sup:`137`\ Te :sup:`147`\ Ba :sup:`88`\ Se :sup:`98`\ Rb :sup:`110`\ Tc :sup:`138`\ Te :sup:`148`\ Ba :sup:`89`\ Se :sup:`99`\ Rb :sup:`122`\ Ag :sup:`139`\ Te :sup:`149`\ Ba :sup:`90`\ Se :sup:`97`\ Sr :sup:`123`\ Ag :sup:`137`\ I :sup:`150`\ Ba :sup:`91`\ Se :sup:`98`\ Sr :sup:`128`\ Cd :sup:`138`\ I :sup:`147`\ La :sup:`87`\ Br :sup:`99`\ Sr :sup:`127`\ In :sup:`139`\ I :sup:`149`\ La :sup:`88`\ Br :sup:`100`\ Sr :sup:`127m`\ In :sup:`140`\ I :sup:`150`\ La ============== ============== =============== =============== =============== Neutron yields from \alpha-particle interaction are available for 19 :math:`\left( \alpha,n \right)` target nuclides: :sup:`7`\ Li, :sup:`9`\ Be, :sup:`10`\ B, :sup:`11`\ B, :sup:`13`\ C, :sup:`14`\ N, :sup:`17`\ O, :sup:`18`\ O, :sup:`19`\ F, :sup:`21`\ Ne, :sup:`22`\ Ne, :sup:`23`\ Na, :sup:`25`\ Mg, :sup:`26`\ Mg, :sup:`27`\ Al, :sup:`29`\ Si, :sup:`30`\ Si, :sup:`31`\ P, and :sup:`37`\ Cl. The neutron decay data contain discrete alpha-particle energies and branching fractions for 89 actinides and 7 fission products listed in :numref:`tab-origen-nucl-alpha`. The sources of the level branching fraction data and the :math:`\left( \alpha,n \right)` cross section data are listed in :numref:`tab-origen-an-cxs`. The stopping cross sections and :math:`\left( \alpha,n \right)` target cross section and product level branching libraries are used in calculating the neutron yield and spectra from Ziegler :cite:`ODR-zieg1977` for all elements with Z < 93, and from Wilson :cite:`ODR-wp1983` for all elements \geq 93. .. table:: Nuclides with :math:`\alpha`-particle emission data for neutron yield calculations :class: longtable :name: tab-origen-nucl-alpha ================ ================ ================ ================= ================= :sup:`142`\ Ce :sup:`216`\ Po :sup:`226`\ Ac :sup:`237`\ Np :sup:`245`\ Cm :sup:`144`\ Nd :sup:`218`\ Po :sup:`227`\ Ac :sup:`235`\ Pu :sup:`246`\ Cm :sup:`146`\ Sm :sup:`215`\ At :sup:`226`\ Th :sup:`236`\ Pu :sup:`247`\ Cm :sup:`147`\ Sm :sup:`217`\ At :sup:`227`\ Th :sup:`237`\ Pu :sup:`248`\ Cm :sup:`148`\ Sm :sup:`218`\ At :sup:`228`\ Th :sup:`238`\ Pu :sup:`249`\ Bk :sup:`149`\ Sm :sup:`219`\ At :sup:`229`\ Th :sup:`239`\ Pu :sup:`248`\ Cf :sup:`152`\ Gd :sup:`217`\ Rn :sup:`230`\ Th :sup:`240`\ Pu :sup:`249`\ Cf :sup:`210`\ Pb :sup:`218`\ Rn :sup:`232`\ Th :sup:`241`\ Pu :sup:`250`\ Cf :sup:`210`\ Bi :sup:`219`\ Rn :sup:`230`\ Pa :sup:`242`\ Pu :sup:`251`\ Cf :sup:`211`\ Bi :sup:`220`\ Rn :sup:`231`\ Pa :sup:`244`\ Pu :sup:`252`\ Cf :sup:`212`\ Bi :sup:`222`\ Rn :sup:`230`\ U :sup:`240`\ Am :sup:`253`\ Cf :sup:`213`\ Bi :sup:`221`\ Fr :sup:`231`\ U :sup:`241`\ Am :sup:`254`\ Cf :sup:`214`\ Bi :sup:`222`\ Fr :sup:`232`\ U :sup:`242m`\ Am :sup:`253`\ Es :sup:`210`\ Po :sup:`223`\ Fr :sup:`233`\ U :sup:`243`\ Am :sup:`254`\ Es :sup:`211`\ Po :sup:`222`\ Ra :sup:`234`\ U :sup:`240`\ Cm :sup:`254m`\ Es :sup:`212`\ Po :sup:`223`\ Ra :sup:`235`\ U :sup:`241`\ Cm :sup:`255`\ Es :sup:`213`\ Po :sup:`224`\ Ra :sup:`236`\ U :sup:`242`\ Cm :sup:`254`\ Fm :sup:`214`\ Po :sup:`226`\ Ra :sup:`238`\ U :sup:`243`\ Cm :sup:`255`\ Fm :sup:`215`\ Po :sup:`225`\ Ac :sup:`235`\ Np :sup:`244`\ Cm :sup:`256`\ Fm |nbsp| |nbsp| |nbsp| |nbsp| :sup:`257`\ Fm ================ ================ ================ ================= ================= .. tabularcolumns:: |\Y{0.15}|\Y{0.15}|\Y{0.35}|\Y{0.35}| .. table:: Target :math:`(\alpha,n)` cross section and branching level isotopes and sources :class: longtable :name: tab-origen-an-cxs +------------------+----------+--------------------------+--------------------------+ | **Isotope** | **ZAID** | **Level branching** | **Cross section data** | | | | **fraction source data** | | +==================+==========+==========================+==========================+ | :sup:`7`\ Li | 30070 | GNASH | Gibbons and Macklin | | | | | :cite:`ODR-gima1959` | +------------------+----------+--------------------------+--------------------------+ | :sup:`9`\ Be | 40090 | Geiger and Van der | Geiger and Van der | | | | Zwain | Zwain | | | | :cite:`ODR-gezw1975` | :cite:`ODR-gezw1975` | +------------------+----------+--------------------------+--------------------------+ | :sup:`10`\ B | 50010 | GNASH | Bair *et al.* | | | | | :cite:`ODR-baca1979` | +------------------+----------+--------------------------+--------------------------+ | :sup:`11`\ B | 50110 | GNASH | Bair *et al.* | | | | | :cite:`ODR-baca1979` | +------------------+----------+--------------------------+--------------------------+ | :sup:`13`\ C | 60130 | GNASH\ *a* | Bair and Haas | | | | | :cite:`ODR-baha1973` | +------------------+----------+--------------------------+--------------------------+ | :sup:`14`\ N | 70140 | N/A | GNASH | +------------------+----------+--------------------------+--------------------------+ | :sup:`17`\ O | 80170 | Lessor and Schenter | Perry and Wilson | | | | :cite:`ODR-lesc1971` | :cite:`ODR-pewi1981` | +------------------+----------+--------------------------+--------------------------+ | :sup:`18`\ O | 80180 | Lesser and Schenter | Perry and Wilson | | | | :cite:`ODR-lesc1971` | :cite:`ODR-pewi1981` | +------------------+----------+--------------------------+--------------------------+ | :sup:`19`\ F | 90190 | Lesser and Schenter | Balakrishnan *et al.* | | | | :cite:`ODR-lesc1971` | :cite:`ODR-bakame1978` | +------------------+----------+--------------------------+--------------------------+ | :sup:`21`\ Ne | 100210 | N/A | GNASH | +------------------+----------+--------------------------+--------------------------+ | :sup:`22`\ Ne | 100220 | N/A | GNASH | +------------------+----------+--------------------------+--------------------------+ | :sup:`23`\ Na | 110230 | GNASH | GNASH\ *a* | +------------------+----------+--------------------------+--------------------------+ | :sup:`25`\ Mg | 120250 | GNASH | GNASH | +------------------+----------+--------------------------+--------------------------+ | :sup:`26`\ Mg | 120260 | GNASH | GNASH | +------------------+----------+--------------------------+--------------------------+ | :sup:`27`\ Al | 130270 | GNASH | GNASH\ *a* | +------------------+----------+--------------------------+--------------------------+ | :sup:`29`\ Si | 140290 | GNASH | GNASH\ *a* | +------------------+----------+--------------------------+--------------------------+ | :sup:`30`\ Si | 140300 | GNASH | GNASH\ *a* | +------------------+----------+--------------------------+--------------------------+ | :sup:`31`\ P | 150310 | GNASH | GNASH | +------------------+----------+--------------------------+--------------------------+ | :sup:`37`\ Cl | 170370 | GNASH | Woosley et. al. | | | | | :cite:`ODR-wofohozi1975` | +------------------+----------+--------------------------+--------------------------+ .. [#gnash] GNASH-calculated data and measured data are available for these nuclides in the library. By default, the GNASH values are used. To use the measured data, the user must reverse the order of th GNASH and measured data in the library since the code uses the first set encountered in the library (GNASH set). Beta Emission ------------- Beta emission rates and energy spectra are calculated using an analytic expression for the kinetic energy of the emitted :math:`\beta^-` particles :cite:`ODR-goma1971`: .. math:: N\left( Z,W \right) = \frac{g^{2}}{{2\pi}^{3}} F\left( Z,W \right) \rho W \left( W_{0}- W \right)^{2}S_{n}\left( W \right) dW where :math:`Z =` atomic number of the daugher nucleus :math:`g =` weak interaction coupling constant :math:`W =` kinetic energy of beta particle (in :math:`m_{e}c^{2}` units) :math:`F\left( Z,W \right) =` Fermi function :math:`W_{0} =` endpoint beta energy :math:`\rho = \sqrt{W^{2} - 1}` = electron momentum :math:`S_{n}\left( W \right) =` spectral shape factor based on transition type :math:`n =` classification of the transition type **Internal conversion electron emission is not considered.** The calculation requires nuclear data on the fraction of the beta transition to each exicited state of the daughter nucleus, the maximum endpoint energy of the transition (W\ 0), and a classification of the beta transition (n) defined by the spin and parity change of the transition which defines the spectral shape factor. The transition classification uses n\ =0 for allowed and forbidden non-unique transitions, n\ =1 for first forbidden unique transitions, n\ =2 for second forbidden unique transitions, and n\ =3 for third forbidden unique transitions. These data are not stored in the decay data resource but are included in a separate beta decay resource used only for the beta calculation. The beta decay data are stored in the formatted file origen.rev00.ensdf95beta.data. The data are derived from ENSDF as compiled in 1995. The file includes beta decay information for 715 beta decay nuclides and has 8486 beta transition branches. Alpha Emission -------------- Calculation of the alpha emission intensity and spectrum requires detailed information that is not available on the decay resource. The calculation requires the alpha particle energy and branching fraction for each transition branch. Unlike the beta spectrum, the alpha particles are emitted with discrete energies, and the source spectrum may be generated by straightforward binning into the user-defined group structure. Alpha particle emission data are also used in the :math:`\left( \alpha,n \right)` neutron source calculation. Therefore, the alpha emission spectra are calculated using the same alpha decay library in the neutron emission resource: origen.rev01.alphdec.data. .. only:: html .. rubric:: References .. bibliography:: ../zSCALE.bib :cited: :keyprefix: ODR- :labelprefix: Origen-Data-Resources