.. _sec-triton: TRITON: A Multipurpose Transport, Depletion, And Sensitivity and Uncertainty Analysis Module ============================================================================================ **Code Responsible(s):** F. Bostelmann **Contributors:** F. Bostelmann, K. Bekar, D. Wiarda, M. A. Jessee, U. Mertyurek, W. A. Wieselquist, S. W. Hart, S. E. Skutnik, B. R. Langley ABSTRACT The TRITON computer code is a multipurpose SCALE control sequence for transport, depletion, and sensitivity and uncertainty analysis. TRITON automates the processing of cross sections, the neutron transport calculations for one-, two-, and three-dimensional (1D, 2D, and 3D) configurations, and the depletion calculations to estimate the neutron flux, mixture-wise powers, isotopic concentrations, source terms, decay heat and other quantities as well as few-group homogenized cross sections for nodal core calculations as a function of burnup. TRITON can be used in combination with any one of SCALE's neutron transport kernels. Deterministic multigroup transport calculations for 1D and 2D geometries are performed using XSDRNPM and NEWT, respectively. The application of the Monte Carlo codes KENO V.a, KENO-VI, and Shift enables depletion calculations of 3D geometries in either multigroup or in continuous-energy mode. In MG mode, TRITON automates the preparation of problem-dependent MG cross sections for use in MG neutron transport calculations using SCALE's cross section processing module XSProc. The depletion calculations are performed by the ORIGEN depletion module. The SAMS module is used to determine the sensitivity of the calculated value of responses to the nuclear data used in the calculation as a function of nuclide, reaction type, and energy. The uncertainty in the calculated value of the response, resulting from uncertainties in the basic nuclear data used in the calculation, is estimated using energy-dependent cross section covariance matrices. The implicit effects of the cross section processing calculations are also treated. Acknowledgments --------------- The U.S. Nuclear Regulatory Commission (NRC) has been instrumental in supporting TRITON development and enhancements in each SCALE release. For SCALE 6.3 support, we express gratitude to M. Aissa (former NRC), D. Algama (former NRC), R. Y. Lee (former NRC), D. Barto, L. Kyriazidis, and H. Esmaili. Previous releases of SCALE-TRITON were funded by other government agencies and developed by several current and former ORNL staff. The original development of TRITON was led by Mark DeHart (former ORNL) who authored several sections in the current manual. Introduction ------------ TRITON (Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion) is a multipurpose SCALE control sequence for transport and depletion analysis for reactor physics applications. By calling the appropriate SCALE modules, TRITON automates the processing of cross sections, the neutron transport calculations for one-, two-, and three-dimensional (1D, 2D, and 3D) configurations, and the depletion calculations to estimate the neutron flux, mixture-wise powers, isotopic concentrations, source terms, decay heat and other quantities as a function of burnup. An overview can be found in :cite:`TRI-dehart2011`. The choice of the neutron transport kernel determines whether TRITON is run in multi-group (MG) or in continuous-energy (CE) mode. TRITON can be used in combination with any one of SCALE's neutron transport kernels. Deterministic MG transport calculations for 1D and 2D geometries are performed using XSDRNPM and NEWT, respectively. The application of the Monte Carlo codes KENO V.a, KENO-VI, and Shift enables depletion calculations of 3D geometries in either MG or in CE mode. In MG mode, TRITON automates the preparation of problem-dependent MG cross sections for use by the MG neutron transport kernels (see :numref:`fig-triton-sequence-flowchart`). Nodal data for use in nodal core simulations can be generated with the TRITON sequence that uses the NEWT deterministic transport code and with the TRITON sequences using the Shift Monte Carlo code. .. _fig-triton-sequence-flowchart: .. figure:: figs/TRITON/triton-sequence-flowchart.* :width: 400 :align: center General flowchart of the TRITON reactor physics sequence. The SAMS module is used to determine the sensitivity of the calculated value of the response to the nuclear data used in the calculation as a function of nuclide, reaction type, and energy. The uncertainty in the calculated value of the response, resulting from uncertainties in the basic nuclear data used in the calculation, is estimated using energy-dependent cross section covariance matrices. The implicit effects of the cross section processing calculations are predicted using SENLIB and BONAMIST. As a SCALE control module, TRITON automates execution of SCALE functional modules and manages data transfer and input/output processes for multiple analysis sequences. Each of TRITON's eleven calculational sequences is provided in :numref:`tab-triton-sequences`, which lists the sequence name keyword, the sequence description, and the function modules invoked within each sequence. The method for cross section processing is selected using a separate "\ *parm=*\" keyword, which is described in more detail in the next section. .. tabularcolumns:: |p{4cm}|p{3cm}|p{3cm}|p{4cm}| .. table:: Overview of TRITON sequences. :widths: 25 25 25 25 :class: longtable :name: tab-triton-sequences +--------------------+-------------------+-----------------+-----------------------------------------+ | **Sequence** | **Primary SCALE** | **parm=** | **Sequence** | | **keyword** | **modules** | **options** | **function** | +====================+===================+=================+=========================================+ | **Cross section processing sequences** | +--------------------+-------------------+-----------------+-----------------------------------------+ | ``=T-XSEC`` | XSProc | bonami | Preparation of | | | | | multigroup (MG) | | | | centrm\ :sup:`a`| cross section | | | | | library. | | | | xslevel=1/2/3/4 | | +--------------------+-------------------+-----------------+-----------------------------------------+ | **Transport sequences** | +--------------------+-------------------+-----------------+-----------------------------------------+ | ``=T-XSDRN`` | XSProc, XSDRNPM | bonami | 1D MG | | | | | deterministic | | | | centrm\ :sup:`a`| transport | | | | | calculation. | | | | xslevel=1/2/3/4 | | | | | | | | | | weight\ :sup:`b`| | +--------------------+-------------------+-----------------+-----------------------------------------+ | ``=T-NEWT`` | XSProc, NEWT | | 2D MG | | | | | deterministic | | | | | transport | | | | | calculation. | +--------------------+-------------------+-----------------+-----------------------------------------+ | **Depletion sequences** | +--------------------+-------------------+-----------------+-----------------------------------------+ | ``=T-DEPL-1D`` | XSProc, | bonami | 1D MG | | | XSDRNPM, | | deterministic | | | ORIGEN, OPUS | centrm | transport, | | | | | coupled with | | | | xslevel=1/2/3\ | ORIGEN | | | | *a*/4 | depletion. | | | | | | | | | addnux=0/1/2\ | | | | | :sup:`a`/3/4 | | | | | | | | | | weight\ :sup:`b`| | +--------------------+-------------------+-----------------+-----------------------------------------+ | ``=T-DEPL`` | XSProc, NEWT, | | 2D MG | | | ORIGEN, OPUS | | deterministic | | | | | transport, | | | | | coupled with | | | | | ORIGEN | | | | | depletion. | +--------------------+-------------------+-----------------+-----------------------------------------+ | ``=T5-DEPL`` | XSProc\ :sup:`c` | | 3D, Monte Carlo | | | KENO-V.a, | | transport | | | ORIGEN, OPUS | | (KENO-V.a), | | | | | coupled with | | | | | ORIGEN | | | | | depletion. | +--------------------+-------------------+-----------------+-----------------------------------------+ | ``=T6-DEPL`` | XSProc\ :sup:`c` | | 3D, Monte Carlo | | | KENOVI, ORIGEN, | | transport | | | OPUS | | (KENO-VI), | | | | | coupled with | | | | | ORIGEN | | | | | depletion. | +--------------------+-------------------+-----------------+-----------------------------------------+ | ``=T5-DEPL-SHIFT`` | XSProc\ :sup:`c` | | 3D, Monte Carlo | | | Shift, | | transport | | | ORIGEN, OPUS | | (Shift, | | | | | coupled with | | | | | ORIGEN | | | | | depletion. | +--------------------+-------------------+-----------------+-----------------------------------------+ | ``=T6-DEPL-SHIFT`` | XSProc\ :sup:`c` | | 3D, Monte Carlo | | | Shift, ORIGEN, | | transport | | | OPUS | | (Shift), | | | | | coupled with | | | | | ORIGEN | | | | | depletion. | +--------------------+-------------------+-----------------+-----------------------------------------+ +--------------------+-------------------+-----------------+-----------------------------------------+ | :sup:`a`\ Default parm value. | | | | :sup:`b`\ parm=weight is used to generate a broad group cross section | | library. This parm option is only available for the T-DEPL sequence. | | | | :sup:`c`\ T5-DEPL and T6-DEPL are also available in CE-mode, which does not | | invoke XSProc for cross section processing. | +--------------------+-------------------+-----------------+-----------------------------------------+ Overview of TRITON Sequences ---------------------------- The TRITON control module supports eleven calculational sequences, each with its own design and applications. Each of these sequences is described in the following subsections. The first subsection covers the basic cross section processing sequence T-XSEC. The T-XSEC sequence prepares problem-dependent multigroup cross sections for subsequent transport analysis. The second subsection covers TRITON's transport analysis sequences, while the third subsection discusses TRITON's depletion analysis sequences. .. _3-1-2-1: Cross section processing sequence (T-XSEC) ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ The T-XSEC sequence provides the ability to prepare a problem-dependent multigroup cross section library using SCALE cross section processing modules to appropriately account for spatial and energy self-shielding effects. The problem-dependent cross section library contains microscopic cross sections for each nuclide for each material composition defined in the TRITON input. SCALE provides several unit cell types (e.g., a lattice of pins, an infinite medium, a multiregion problem, or a doubly heterogeneous cell) to correct the cross sections for spatial and energy self-shielding. Multiple cell calculations can be used in the same calculation. The calculation of multigroup cross sections is executed by :ref:`XSProc `). Transport sequences (T-XSDRN, T-NEWT) ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ The TRITON transport sequences build upon the cross section processing sequence by automating a transport calculation after cross section processing. Both 1D and 2D discrete-ordinates transport calculations can be performed using XSDRNPM and NEWT, respectively. The T-XSDRN sequence calls XSDRNPM for transport analysis in slab, sphere, or cylindrical geometries, while the T-NEWT sequence calls NEWT for analyses in 2D *xy-*\ geometries. In addition to the input necessary for cross section processing, an XSDRN or NEWT input model is also required. The XSDRN model input is discussed in Appendix A of TRITON; the NEWT model input requirements are described in the NEWT chapter. Similar capabilities and applications for KENO-V.a and KENO-VI are handled through the CSAS5 and CSAS6 sequences, respectively. Depletion sequences (T-DEPL, T-DEPL-1D, T5-DEPL, T6-DEPL, T5-DEPL-SHIFT, T6-DEPL-SHIFT) ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ The TRITON depletion sequences build upon the transport sequences by automating depletion/decay calculations after the transport calculations for each material designated for depletion. One or more materials in the model can be designated for depletion. Each designated material is depleted using region-averaged reaction rates, accounting for all regions in the model associated with a given depletion material. The TRITON depletion calculation procedure is described further in the next subsection. TRITON automates the various computational processes-cross section processing, transport, and depletion-over a series of depletion and decay intervals supplied by the user. The depletion procedure is discussed in :numref:`3-1-2-3-1`. The 2D TRITON depletion sequence (T-DEPL), which uses NEWT for the transport calculations and the 3D TRITON depletion sequences which use Shift in CE mode for the transport calculations (T5-DEPL-Shift, T6-DEPL-SHIFT) provide the capability to generate lattice-physics data for nodal core calculations. Within TRITON depletion calculations, TRITON invokes the ORIGEN depletion module for the time-dependent transmutation of each user-defined material. TRITON provides ORIGEN the neutron flux space-energy distribution, the multigroup cross sections, material concentrations, and material volumes. ORIGEN performs the flux normalization, cross section collapse, and multi-material depletion/decay operations to determine new isotopic concentrations for the next calculation. .. _3-1-2-3-1: Predictor-corrector depletion process ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ For all depletion sequences, TRITON automates cross section processing, transport, and depletion calculations over a series of depletion-decay intervals supplied by the user. A **depletion interval** represents a time interval in which the model power level is assumed constant. A depletion model that exhibits various power level changes will require multiple depletion intervals to accurately model the changes in power. Each depletion interval can be followed by a decay calculation over a user-specified **decay interval**. Within a given depletion interval (e.g., an LWR operating at constant power for a 12-month fuel cycle), the isotope concentrations of different depletion materials change, which induces changes in the problem-dependent multigroup cross sections (through spatial and energy self-shielding effects) as well as the neutron flux distribution, leading to different power distributions and transmutation rates in depletion materials. This requires TRITON to represent each depletion interval as a series of smaller time intervals in which cross section processing and transport solutions are recomputed to accurately model these time-dependent effects. A **depletion subinterval** represents a time interval in which TRITON performs cross section processing and transport calculations to determine cross sections and flux distributions used in the depletion calculations. All depletion subintervals for a given depletion interval have the same length-for example, one 12-month depletion interval can be represented as a series of 12 one-month depletion subintervals, or as 6 two-month depletion subintervals. Alternatively, the 12-month depletion interval can be modeled as two consecutive 6-month depletion intervals, each one having a different number of subintervals. Therefore the formulation of a **depletion scheme** in TRITON is highly flexible. A depletion scheme is the set of user-defined depletion and decay intervals with associated power levels and number of subintervals. .. Caution:: TRITON does not provide automated means to determine the appropriate depletion scheme for a given application. The user must determine the accurate depletion scheme specific to his or her application. TRITON uses a predictor-corrector approach to process the user-defined depletion scheme. The predictor-corrector approach performs cross section processing and transport calculations based on anticipated isotope concentrations at the *midpoint* of a depletion subinterval. Depletion calculations are then performed over the full subinterval using cross sections and flux distributions predicted at the midpoint. Depletion calculations are then extended to the midpoint of the next subinterval (possibly through a decay interval and into a new depletion interval), followed by cross section processing and transport calculations at the new midpoint. The iterative process is repeated until all depletion subintervals are processed. In order to start the calculation, a "bootstrap case" is required using initial isotope concentrations for the initial cross section processing and transport calculation. The bootstrap calculation is used to determine the anticipated isotope concentrations at the midpoint of the first depletion subinterval. The predictor-corrector approach is best explained by an example. :numref:`fig-triton-pred-corr-algorithm` illustrates the predictor-corrector process for a hypothetical depletion scheme with two depletion intervals. The first depletion interval contains two subintervals, followed by a decay interval. The second depletion interval contains one subinterval and is also followed by a decay interval. In :numref:`fig-triton-pred-corr-algorithm`, cross section processing and transport calculations are represented by the 'T' label, and depletion calculations are represented by the 'D' label. For this example, four sets of calculations would be necessary: one for each of the three depletion subintervals, and one for the initial "bootstrap case." These calculations are represented in the following eight steps. - Step 1 + T\ :sub:`0`: Cross section processing and transport calculation using initial (i.e., time-zero) isotope concentrations. - Step 2 + D\ :sub:`1`: Depletion calculation from time-zero to the midpoint of the first depletion subinterval. The dashed horizontal arrow denotes a "predictor" depletion step. - Step 3 + T\ :sub:`1`: Cross section processing and transport calculation at the midpoint of the first depletion subinterval. - Step 4 + D\ :sub:`1`: Depletion calculation for the first depletion subinterval. The solid horizontal arrow across the subinterval denotes a "corrector" depletion step. *Corrector steps use cross sections and flux distribution computed at the subinterval midpoint.* This is represented by a solid arrow from T\ :sub:`1` to D\ :sub:`1`. + D\ :sub:`2`: Predictor depletion calculation for the second depletion subinterval. *Predictor steps use cross sections and flux distribution computed at the*\ **previous**\ *subinterval midpoint.* This is represented as the dashed arrow from T\ :sub:`1` to D\ :sub:`2`. - Step 5 + T\ :sub:`2`: Cross section processing and transport calculation at the midpoint of the second depletion subinterval. - Step 6: + D\ :sub:`2`: Corrector depletion calculation for the second depletion subinterval, followed by the decay calculation at the end of the first depletion interval. + D\ :sub:`3`: Predictor depletion calculation for the third depletion subinterval. The third depletion subinterval is the first and only subinterval associated with the second depletion interval. - Step 7 + T\ :sub:`3`: Cross section processing and transport calculation at the midpoint of the third depletion subinterval. - Step 8 + D\ :sub:`3`: Corrector depletion calculation for the third depletion subinterval. This calculation is followed by a second decay calculation. .. _fig-triton-pred-corr-algorithm: .. figure:: figs/TRITON/fig1.png :align: center :width: 500 Predictor/corrector depletion algorithm used by TRITON. The depletion calculations are performed by ORIGEN and span either the first half of a subinterval (predictor step) or the full subinterval (corrector step). ORIGEN performs these depletion calculations and possible decay calculations over a series of smaller time intervals. The **ORIGEN time intervals** are automatically determined by TRITON depending on the length of the depletion subinterval and decay interval. Additionally, TRITON will automatically adjust the number of subintervals per depletion interval if the time length of the user-defined subinterval is large (i.e., >400 days). TRITON writes the utilized depletion scheme near the top of the output file. The depletion scheme output edit is further described in :numref:`3-1-5-4-1`. Lattice physics analysis ^^^^^^^^^^^^^^^^^^^^^^^^ The 2D depletion sequence (T-DEPL) may be used to generate lattice physics data for subsequent core analysis calculations using core simulator software. Core simulators typically employ few-group nodal diffusion theory for neutronic calculations, coupled with other calculation methods for thermal hydraulics, fuel performance, and plant operation (e.g., soluble boron letdown or control rod movement). Core simulation requires the use of pretabulated **lattice physics data** for the neutronic calculations-that is, few-group homogenized cross sections, with appropriate discontinuity factors, pin powers, and kinetic parameters, functionalized in terms of burnup and other system conditions such as fuel temperature and moderator density. To support lattice physics database preparation, the NEWT transport module contains flexible input options to define the few-group energy structure, spatial homogenization regions, and discontinuity factors. After the transport calculation at the midpoint of each depletion subinterval, NEWT computes the lattice physics data and stores this data on a temporary file. TRITON reads the temporary file and archives the lattice physics data onto a separate database file. In addition, the *T-DEPL* sequence supports branch calculations in which perturbations may be applied to certain system conditions such as fuel temperatures and moderator density. TRITON automates the cross section processing and transport calculations for each branch condition at the midpoint of the depletion subinterval. NEWT computes the lattice physics data for the branch calculations, and TRITON archives this data onto the lattice physics database file. The TRITON input options for branch calculations are described in :numref:`3-1-3-3-2`, and the file format of the lattice physics database is provided in the Appendix B of TRITON. .. note:: The TRITON input options for branch calculations are designed to be highly flexible to support a large range of core analyses; therefore, TRITON does not provide automated means to determine the branch calculations. The user must determine the necessary branch calculations for his or her core analysis and be knowledgeable of the capabilities and limitations of the cross section treatment of the core simulator. The TRITON Lattice Physics Primer has been developed to provide guidance on appropriate TRITON branch calculations for LWR core analysis (NUREG/CR-7041) and in "Cross Section Generation Guidelines for TRACE-PARCS" (NUREG/CR-7164). Input Description ----------------- TRITON input is free-form and keyword based, similar in form to many other modules in SCALE. With a few exceptions, the following formatting rules apply: - Data is limited to 255 columns but may wrap into as many lines as are needed. - Comment lines start with a tick mark (') in the first column of a line and may be placed anywhere in the input. - The keyword-based input is case insensitive. - TRITON input is organized into blocks of data. Each data block begins with *read blockname* and terminates with *end blockname*. - Blocks of data may appear in any order. Each block of data may appear only once in the input. - Input can be redirected from an auxiliary file by using the open angle bracket (<) and the name of the file-for example, *element->nuclide. For example, the input .. code-block:: scale nuclides u all u-235 end rates 1e-3 1e-4 1e-5 end is processed in the following order: 1. assign rate 1e-4 to *all* 2. assign rate 1e-3 to *u* 3. assign rate 1e-5 to *u-235* In this way, a rate of 1e-5 is assigned to U-235, a rate of 1e-3 is assigned to all uranium isotopes besides U-235, and a rate of 1e-4 is assigned to all nuclides besides all uranium isotopes. The order in which *all*, elements, and nuclides are specified is irrelevant. **Nuclide removal and feed interaction** If nuclides are transferred from mixture 1 into mixture 2---that is, they are **removed** from mixture 1 and **fed** into mixture 2---then users might observe slight inconsistencies in the removed compared to the fed nuclide amounts after a TRITON depletion/decay step. The removal of nuclides is handled exactly according to the expected behavior with the assigned removal rate. This means that in the absense of any other nuclide production or removal mechanism, the nuclide density in mixture 1 would follow: .. math:: \frac{dN_{i}}{\text{d}t} = - \lambda_{i,rem} N_{i}(t) In order to connect the removal from material 1 to the feed rate into material 2, the removal is integrated numerically over a substep and recast as a nuclide feed for this specific substep *s*: .. math:: S_{i,s} = \int_{t_{s-1}}^{t_s} \lambda_{i,rem}N_{i}(t) \text{d}t The approximation for *N*\ :sub:`i` over a substep uses a weighted essentially non-oscillatory (WENO)--like scheme that takes the nuclide amounts at the beginning and end of a substep and determines a combination of logarithmic and linear interpolation, preferring logarithmic if the amounts differ greatly and linear if they are similar. If there are particularly strong removals, then it may be necessary to introduce additional small time steps (i.e., additional depletion/decay steps in the *BURNDATA* block) to better handle the numerical integration on the substeps. **Decay of mixtures outside of the system** If nuclides are transferred from mixture 1 into mixture 2, but mixture 2 is not part of the system (e.g., contained in a loop outside the reactor), then the decay of the nuclides in this mixture can still be considered by specifying keyword *decayonly* in the *DEPLETION* block. These *decayonly* mixtures exist only on the ORIGEN side, but they are not relevant for the neutron transport calculation. Their volume is by default set to 1 cm\ :sup:`3`, and their mass is set according to the specified density for this volume. TRITON's mass normalization does not affect the volumes or masses of these mixtures. However, if a *continuous_feed* is applied to such a mixture, then the specified feed rate is still multiplied with the multiplier used for the mass normalization. The following is an example in which xenon is transfered from mixture 1 to mixture 2, and the decay of nuclides in mixture 2 is enabled in the *DEPLETION* block: .. code-block:: scale read depletion -1 decayonly 2 end depletion read timetable flow from 1 to 2 type fractional_removal units pers nuclides xe end rates 2e-2 end time 0.0 end multiplier 1.0 end end flow end timetable OPUS block ^^^^^^^^^^ The OPUS module of SCALE is fully documented in the OPUS chapter of the SCALE manual. OPUS provides the ability to extract specific data from ORIGEN output libraries, perform unit conversions, and generate plot data for post-calculation analysis. In essence, OPUS is an ORIGEN post-processor that provides data in the desired form for a desired subset of nuclides. TRITON by default calls OPUS to extract nuclide concentrations for selected nuclides for all depletion materials and for the most important nuclides. TRITON provides the capability to specify the full set of OPUS commands to tailor OPUS calculations to obtain specific information. TRITON allows a stacked set of OPUS calculations in order to retrieve selected data for selected nuclides. The content of the *OPUS* block is based on standard OPUS input parameters, as described in the OPUS chapter; the details of OPUS control and use are not repeated here. However, additional input is necessary to support TRITON operations with OPUS, because TRITON enables additional capabilities beyond those provided for in standard OPUS input. For example, OPUS is designed to process the output file from a single ORIGEN calculation; because ORIGEN is a point depletion solver, the output represents data from a single material. TRITON is typically used to perform multiple depletion calculations at each depletion step-one calculation for each material being depleted. Hence, multiple OPUS calculations are needed to obtain results from multiple materials. The OPUS calculations are performed automatically by TRITON but require the user to specify the materials for which OPUS processing is desired. Additionally, TRITON supports stacked OPUS cases within the *READ OPUS* data block; hence, keywords are introduced to separate stacked cases. There are two alternatives available to SCALE users that are complimentary to the OPUS block within TRITON. First, standalone OPUS case(s) can be used to post process the ORIGEN binary concentration file (.f71 extension). This file is automatically saved in the output directory with the file name ``${OUTBASENAME}.f71``. (e.g. if the input file is reactor.inp, the concentration file is saved in the output directory as reactor.f71) Second, the user may also open the concentration file within Fulcrum to enable similar post-processing capabilities. Selection of materials for OPUS processing '''''''''''''''''''''''''''''''''''''''''' Beyond standard OPUS input keywords (see OPUS chapter), TRITON reads a *matl=* keyword to allow specification of ID number(s) for the material(s) in the problem for which outputs are desired. The *matl=…end* input keyword accepts one or more materials from the *DEPLETION* data block for which OPUS processing is desired. If omitted, OPUS processing will be performed for all materials in the *DEPLETION* block. For example, consider the following *DEPLETION* and *OPUS* data blocks: .. code-block:: scale READ depletion 1 2 3 4 5 6 END depletion READ opus units=gram symnuc=u-234 u-235 u-236 u-238 pu-238 pu-239 pu-240 pu-241 pu-242 pu-243 np-237 end time=year END opus In this example, OPUS processing will be performed for all depletion materials, 1--6. Adding a subset of materials using the *matl=* keyword, for example. .. code-block:: scale READ depletion 1 2 3 4 5 6 END depletion READ opus units=gram symnuc=u-234 u-235 u-236 u-238 pu-238 pu-239 pu-240 pu-241 pu-242 pu-243 np-237 end time=year matl=1 2 3 end END opus will result in OPUS calculations for materials 1, 2, and 3 only. Although ORIGEN calculations are performed only for individual materials, TRITON provides the capability of combining the results of all or a subset of all depletion materials to get a multimaterial average set of ORIGEN responses. TRITON provides three special ID numbers for combining material results: - material ID 0 returns system-averaged results for the entire set of depletion materials (i.e., all mixtures included in the *depletion* block), - material ID -1 returns the average of only those materials with ID > 0 present in OPUS *matl=* list, - material ID -2 returns averaged results for all fuel materials in the system (i.e., all mixtures with a non-negligible heavy metal mass) Again, this is best illustrated by example. Specification of the data blocks .. code-block:: scale READ depletion 1 2 3 4 5 6 END depletion READ opus units=gram symnuc=u-234 u-235 u-236 u-238 pu-238 pu-239 pu-240 pu-241 pu-242 pu-243 np-237 end time=year matl=1 2 3 0 -1 end END opus will result in five OPUS calculations and five sets of results-one for each of materials 1, 2, and 3, one for the average of materials 1--6 (due to input of material ID 0), and one for the average of materials 1--3 (due to input of material ID –1). Specification of stacked OPUS cases ''''''''''''''''''''''''''''''''''' In a given calculation, multiple output units may be desired (e.g., grams, curies, and watts), or multiple time scales (e.g., seconds and years), or a combination of these or other parameters. TRITON provides the ability to stack inputs such that multiple cases may be run within a single TRITON calculation. In order to stack cases, the keywords *new case* are entered in the input stream. Any parameters following these keywords are used to define a new OPUS case. There is no limit on the number of stacked cases that may be input; however, the *matl=* specification may be used only in the first case. OPUS calculations are run for each of the materials in this list, for all cases. Consider a depletion calculation where gadolinium pins are present in the assembly design. One may wish to determine the quantities of gadolinium nuclides from the initial poison rods (tracked as a light element by ORIGEN within TRITON) and from fission (tracked as a fission product by ORIGEN). One may also need masses of selected actinides as well as the total (:math:`\alpha`,n) reaction rate. :numref:`fig-triton-opus-block-example` shows how the *new case* keyword set is used to define unique OPUS calculations. In this example, the *new case* keywords are shown in upper case and on a line by themselves, but this has been done only for readability. The text may be entered in lower case and on the same line as other keywords. Note, however, that the *matl=* specification is given only in the first case. All OPUS calculations will be performed for materials 1, 2, and 3 and for the average of these three materials. .. _fig-triton-opus-block-example: .. figure:: figs/TRITON/fig15.svg :align: center :width: 500 Example OPUS block input. .. _sec-triton.fgxs: FGXS block ^^^^^^^^^^ The 3D TRITON depletion sequences which use Shift in CE mode for the transport calculations (T5-DEPL-Shift, T6-DEPL-SHIFT) provide the capability to generate lattice-physics data for nodal core calculations based on complex 3D geometries. The use of CE data avoids any approximations due to MG cross section processing and it allows the generation of nodal data in arbitrary group structures. The generation of nodal data in a TRITON-Shift input is requested through the *FGXS* block. The general input format of this block is as follows: .. code-block:: scale read fgxs energy id= ... end id= [options] shape id= [dimensions] [ mesh id= [definition] ] end fgxs - **id**: Integer identifier of the requested nodal data set. Multiple sets of data can be requested, and the id allows the association of parameters to one specific set. - **energy**: Energy group boundaries [eV] in increasing order, including the uppermost and lowermost boundaries. For an N-group structure, N+1 boundaries need to be provided. - ****: The type of requested tally. The only supported type is currently *tallyset t16* to request the full set of nodal data. - **shape**: Either the name of one of the permitted cell shapes [*cuboid*, *rhexprism*] or keyword *global* to request data on a mesh over the entire geometry. The use of *global* requires a *mesh* record. All dimensions are provided in cm. - **mesh**: Gobal mesh definition in case of a *global* shape. See details below. **Global energy and tallytype definition** If multiple sets of nodal are requested within one *FGXS* block, and if the same type of tally in the same energy group structure is desired for *all* tallies, a global *energy* definition and the *tallytype* can be defined. The energy and tallytype id can be set to 0 (*id=0*) and is then valid for all other defined tallies. The energy and tallytype definition can then be skipped for the individual tallies (see Example 5 below). **Shapes** The supported shapes follow the following input syntax: 1. **cuboid**: .. code-block:: scale shape cuboid id= 2. **rhexprism**: .. code-block:: scale shape rhexprism id= hpitch= origin x= y= 3. **global**: .. code-block:: scale shape global id= with the following parameters: - *, , , , , *: minimum and maximum x-, y-, or z-coordinates [cm] - *hpitch=

*: half-pitch [cm] in case of the hexagon - *origin x= y=*: coordinates in the global unit coordinate system for the center of the defined shape [cm] **Meshes** Global meshes are superimposed over the entire geometry. The half pitches of the mesh needs to be provided, the origin of the mesh (the origin of the central mesh cell), and the axial discretization. At this point, only square lattices and **rotated** hexagonal lattices are supported: 1. **square**: .. code-block:: scale mesh square id= hpitch=

origin x= y= dz ... end 2. **hexagonal** .. code-block:: scale mesh hexagonal id= hpitch= origin x= y= dz ... end with the following parameters: - *hpitch=

*: half-pitch of the lattice [cm] - *origin x= y=*: and are coordinates in the global unit coordinate system in which to place the center of a lattice element [cm]. Lattice elements are repeated in the negative-x, negative-y, positive-x, and positive-y directions to fill the global unit. - *dz ... *: definition of the relative axial mesh widths. Each ** value must be positive, and the sum of all ** values must be 1.0. M entries in the dz-list will create M axial zones. The axial zone boundaries is determined by the relative mesh widths and the bottom and top axial boundaries of the global unit. **Output** On overview of the requested nodal data is provided in the output file: .. code-block:: none ============================================================================================= Nodal FGXS Tally Summary ============================================================================================= Tally ID=30 tally spatial type: MESH tally boundary: RHEXPRISM energy grid: 2e+07,0.625,1e-05 volume mesh inscribed radius: 0.500000 volume mesh center: -0.288675,-0.5 volume mesh planes z: -1e-06,0.9,1 volume tallies: absorption,fission,flux,kappa_sigma,nu_delayed_fission,nu_fission,transfer_1n,transfer_2n,transfer_3n,transfer_4n volume tallies on isotopes (nuclide id/reaction mt): 1001/18, ... , 96245/1018 fine energy grid: 2e+07,... , 1e-05 fine energy grid volume tallies: flux,absorption,fission,transfer_1n,transfer_2n,transfer_3n,transfer_4n volume tally name (shift): mesh_nodal_tally_t16_30 num tally energy groups: 2 num tally mesh rings: 2 num tally fine energy groups: 1000 Tally ID=40 ... The t16 output file names are composed of the ${BASENAME} of the input file, the tally id, the mesh ids in case of global meshes, and they have the ending t16. **Example 1:** The following block requests nodal data in a cuboid within the interval *x=[2, -5]*, *y=[8, -5]*, *z=[2.5, 0]*. .. code-block:: scale read fgxs energy id=1 1e-5 0.625 20E6 end tallyset t16 id=1 shape cuboid id=1 2 -5 8 -5 2.5 0 end fgxs One output file with name *${BASENAME}.id1.t16* will be generated. **Example 2:** The following block requests nodal data in a rotated hexagon with half pitch of 1.25 cm, in the axial zone *z=[2,-2]*. The center of the hexagon is located at *(x,y,z) = (10.0, 0.5, 0.0)*. .. code-block:: scale read fgxs energy id=2 1e-5 0.625 20E6 end tallyset t16 id=2 shape rhexprism id=2 1.25 2.0 -2.0 origin x=10 y=0.5 end fgxs One output file with name *${BASENAME}.id2.t16* will be generated. **Example 3:** The following block requests nodal data in a square lattice that stretches over the entire geometry. The half pitch of the square lattice is 0.8 cm and the origin is at (x,y) = (0,0). Nodal data is requested for 2 axial zones of equal height. .. code-block:: scale read fgxs energy id=3 1e-10 0.625 20e6 end tallyset t16 id=3 shape global id=3 mesh square id=3 hpitch=0.8 origin x=0 y=0 dz 0.5 0.5 end end fgxs Multiple files named *${BASENAME}.id3---.t16* will be generated. **Example 4:** The following block requests nodal data in a rotated hexagonal lattice that stretches over the entire geometry. The half pitch of the hexagonal lattice is 0.5 cm and the origin is at (x,y) = (-0.288675,-0.5). Nodal data is requested in two axial zones, the lower zone stretching over 90% of the geometry, the upper zone covering the upper 10% of the geometry. .. code-block:: scale read fgxs energy id=4 1E-5 0.625 20E6 end tallyset t16 id=4 shape global id=4 mesh hexagonal id=4 hpitch=0.5 origin x=-0.288675 y=-0.5 dz 0.9 0.1 end end fgxs Multiple files named *${BASENAME}.id3---.t16* will be generated. The output files contains an overview of all file names with the corresponding (x,y)-origin of the cells and the axial range. A visual representation of the mesh superimposed on the geometry indicates the centrl cell and the cells for which nodal data was generated. .. code-block:: none T16 file summary for hex mesh tally id=4 ========================================= Apothem (inner radius): 0.5 cm Number of rings: 2 Grid size: 5 basename.id4---.t16 x y zmin zmax --------------------------------------------------------------------------------------------------------- basename.id30-02-04-01.t16 -1.15470e+00 0.00000e+00 -1.00000e-06 9.00000e-01 basename.id30-02-04-02.t16 -1.15470e+00 0.00000e+00 9.00000e-01 1.00000e+00 basename.id30-03-03-01.t16 -2.88675e-01 -5.00000e-01 -1.00000e-06 9.00000e-01 basename.id30-03-03-02.t16 -2.88675e-01 -5.00000e-01 9.00000e-01 1.00000e+00 basename.id30-03-04-01.t16 -2.88675e-01 5.00000e-01 -1.00000e-06 9.00000e-01 basename.id30-03-04-02.t16 -2.88675e-01 5.00000e-01 9.00000e-01 1.00000e+00 basename.id30-04-02-01.t16 5.77350e-01 -1.00000e+00 -1.00000e-06 9.00000e-01 basename.id30-04-02-02.t16 5.77350e-01 -1.00000e+00 9.00000e-01 1.00000e+00 basename.id30-04-03-01.t16 5.77350e-01 0.00000e+00 -1.00000e-06 9.00000e-01 basename.id30-04-03-02.t16 5.77350e-01 0.00000e+00 9.00000e-01 1.00000e+00 basename.id30-04-04-01.t16 5.77350e-01 1.00000e+00 -1.00000e-06 9.00000e-01 basename.id30-04-04-02.t16 5.77350e-01 1.00000e+00 9.00000e-01 1.00000e+00 _____ / \ _____/ 05 \ / \ 05 / _____/ 04 \_____/ / \ 05 / \ _____/ 03 \_____/ 05 \ / \ 05 / \ 04 / _____/ 02 \_____/*04****\_____/ / \ 05 / \*04****/ \ / 01 \_____/*03****\_____/ 05 \ \ 05 / \*04****/ \ 03 / \_____/*02****\_____/*04****\_____/ / \*04****/ \*03****/ \ / 01 \_____/#03####\_____/ 05 \ \ 04 / \#03####/ \ 02 / \_____/ 02 \_____/*04****\_____/ / \ 03 / \*02****/ \ / 01 \_____/ 03 \_____/ 05 \ \ 03 / \ 02 / \ 01 / \_____/ 02 \_____/ 04 \_____/ / \ 02 / \ 01 / / 01 \_____/ 03 \_____/ \ 02 / \ 01 / \_____/ 02 \_____/ / \ 01 / / 01 \_____/ \ 01 / \_____/ #### mesh origin cell **** tally cells cell column/x-index cell row/y-index **Example 5:** The following block requests three sets of nodal data for different cuboids. The energy group definition and the tallyset are only defined once and apply to all defined tallies. .. code-block:: scale read fgxs energy id=0 1e-5 0.625 20E6 end tallyset t16 id=0 shape cuboid id=10 1.0 -1.0 5.0 -5.0 1.0 0.1 shape cuboid id=20 2.0 -2.0 6.0 -6.0 2.5 1.2 shape cuboid id=30 3.0 -3.0 7.0 -7.0 5.0 2.0 end fgxs TALLIES and DEFINITIONS block ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ Since SCALE 6.3, a *TALLIES* input block and a *DEFINITIONS* input block can be used with CSAS-Shift and TRITON-Shift for flexible definition and output control of mesh tallies. The mesh responses for neutron flux, the neutrons produced from fission, and the fission rate (new since 6.3) can be requested on different spatial and energy grids. The syntax is very similar to the corresponding blocks used in the MAVRIC sequence. The input parameters previously used to request these responses (i.e., *gfx*, *cds*, and *fis*) are still permitted in SCALE 6.3, but it is recommended to use the new input. The new input further permits the mesh tally responses based on different spatial grids and energy grids within one calculation, whereas previously only one spatial and energy grid was permitted per calculation. **DEFINITIONS BLOCK** The new definitions input block allows multiple spatial grids to be defined using the *gridGeometry* keyword and multiple energy grids to be defined using the *energyBounds* keyword. The syntax for defining a *gridGeometry* inside a definitions block is the same as defining a standalone grid at the root level of input (see :numref:`8-1-2-14`). .. note:: For TRITON-Shift sequences, the grid boundaries must be inside the specified geometry, while TRITON-KENO permits the grid boundaries to go beyond the geometry. The syntax for defining *energyBounds* is already used for defining energy grids in the MAVRIC sequence (see :numref:`tab-module.monaco.ce_keywords`). The energy grid definition permits specification of individual energy boundaries, equal-width energy bins and equal-width lethargy bins within a specified energy, and SCALE energy group structures (such as the 56-group structure). The SCALE energy group structure can be any group structures that is used by a SCALE multigroup library that is available in the DATA directory. The syntax is *n* for neutron libraries and *p* for photon (gamma) libraries. A combination of the different options is also supported. In continuous-energy mode, a special *default* keyword allows modification of the default energy group structure previously defined with the *NGP* parameter and/or the standalone energy block. For example, TRITON-Shift can use these *energyBounds* to tally flux and cross sections for depletion calculations. The following examples demonstrate how the *DEFINITIONS* block in TRITON is used. .. code-block:: scale read definitions read grid 1 xlinear 30 -10 70 ylinear 10 -20 60 zlinear 50 -30 40 end grid read grid 2 numxcells=10 xmin=-18.5 xmax=+68.5 numycells=25 ymin=-28.5 ymax=+58.5 numzcells=10 zmin=-38.5 zmax=+48.5 end grid ' global energy grid, default is 252g in CE mode, but can be modified here energyBounds default bounds 2e7 0.625 1e-5 end end energyBounds ' user specified energy grid energyBounds 1 bounds 2e7 0.625 1e-5 end end energyBounds ' user specified energy grid using equal-energy bins energyBounds 2 linear=10 1e-5 2e7 end energyBounds ' user specified energy grid using equal-lethargy bins energyBounds 3 logarithmic=10 1e-5 2e7 end energyBounds 'SCALE 56-group neutron structure with additional energy points energyBounds 10 56n bounds 1.1 0.11 0.011 0.0011 end energyBounds end definitions **TALLIES BLOCK** The new tallies input block allows mesh responses to be requested using any energy grid and/or spatial grid from the definitions block. TRITON-Shift also allows specifying a time input array to enable or disable these mesh tally responses for specific depletion steps. The last depletion step may be conveniently requested using the special *LAST* keyword, and the special *ALL* keyword may be used to request the tally for every depletion step. +--------------------------------+------------------------------+------------------------------+ | **Description** | **SCALE 6.2 parameter name** | **SCALE 6.3 parameter name** | +================================+=======+======================+==============================+ | neutron flux | GFX | flux | +--------------------------------+------------------------------+------------------------------+ | fission rate | FIS | fission_density | +--------------------------------+------------------------------+------------------------------+ | neutrons produced from fission | CDS | fission_source | +--------------------------------+------------------------------+------------------------------+ The following examples demonstrate how the *TALLIES* block in TRITON is used. .. code-block:: scale read tallies read mesh 1 response = FLUX grid = 1 time 0 LAST end energy = 1 end mesh read mesh 2 response = FISSION_DENSITY grid = 2 time 0 1 2 end energy = 2 end mesh read mesh 3 response = FISSION_SOURCE grid = 3 time ALL end energy = default end mesh end tallies .. _3-1-3-5: *ALIAS* block ~~~~~~~~~~~~~ The optional *ALIAS* block may be used to simplify model development within TRITON by defining a set of material numbers that will be inserted in place of the alias when that alias is used in subsequent data blocks. Aliases function as variables for which a user-defined set of materials are inserted; they are identified by a dollar character ($) preceding a single-word alphanumeric label. The *ALIAS* block is used to preprocess an input, creating a new, modified input deck with all alias variable substitutions included. TRITON then processes the modified input deck before proceeding with the calculation. The use of an alias variable is best illustrated by a brief example. Assume that the alias ``$fuel`` is defined as materials 1, 2, and 3, and ``$mod`` as materials 4, 5, and 6. (The input format for defining aliases is described below.) The user wishes to create three identical sets of materials and use them in three identical pin cell specifications. In the *COMPOSITION* data block, specifications could be written in the following form .. code-block:: scale h2o $mod den=0.6616 1.0 595 end wtpt-boron $mod 0.6616 1 5000 100 655e-6 595 end TRITON would create a modified input with the alias expanded as follows: .. code-block:: scale uo2 1 den=10.29 0.9322 920 92235 3.0 92238 97.0 end uo2 2 den=10.29 0.9322 920 92235 3.0 92238 97.0 end uo2 3 den=10.29 0.9322 920 92235 3.0 92238 97.0 end h2o 4 den=0.6616 1.0 595 end h2o 5 den=0.6616 1.0 595 end h2o 6 den=0.6616 1.0 595 end wtpt-boron 4 0.6616 1 5000 100 655e-6 595 end wtpt-boron 5 0.6616 1 5000 100 655e-6 595 end wtpt-boron 6 0.6616 1 5000 100 655e-6 595 end Similarly, if the alias were used in the *CELLDATA* block as .. code-block:: scale latticecell squarepitch pitch=1.26 $mod fuelr=0.4095 $fuel end then TRITON would expand the aliases to .. code-block:: scale latticecell squarepitch pitch=1.26 4 fuelr=0.4095 1 end latticecell squarepitch pitch=1.26 5 fuelr=0.4095 2 end latticecell squarepitch pitch=1.26 6 fuelr=0.4095 3 end In a depletion calculation, one may wish to deplete a large number of fuel rods independently because of different geometric locations in a fuel assembly. Even though each fuel rod may have the same initial composition, each must be specified as a unique material composition in order to be depleted independently. Furthermore, multiple cell specifications must all use unique material identifiers for each cell component. Thus, if one desired to deplete 25 fuel materials in a fuel/clad/moderator pin cell, one would need to set up material composition definitions for 25 fuels, 25 moderators, and 25 clads. Then one would need to provide 25 pin cell specifications. By using aliases, one need only specify the material identifiers corresponding to each alias and then provide only one material composition specification for each alias type, and then one pin cell specification. TRITON will automatically expand the aliases and create a revised input with all materials and cell specifications explicitly defined. .. note:: Note that although this will simplify the pin cell input in the CELLDATA, 25 pin cell calculations would still be required. The number of pin cell calculations can be reduced by using the ASSIGN function described in :numref:`3-1-3-3-4-3`. The purpose of the *ALIAS* block is to define a set of alias variables to be used in subsequent data blocks. The *ALIAS* block is optional, but aliases may not be used in other blocks if an *ALIAS* block is not present to define the aliases. An *ALIAS* block may contain as many aliases as desired. Each alias specification consists of three parts: the alias name, consisting of a dollar sign followed by up to 11 alphanumeric characters with no embedded spaces; the material number or numbers; and an *end* keyword. Material numbers may be entered in any order and may be separated by spaces or commas (or both). Material numbers may also be separated by a dash (-), but this represents an inclusive list. In other words, a material specification of 1-3 (or 1 - 1) indicates materials 1, 2, and 3. The example *ALIAS* block below illustrates the various means for assigning a set of materials for an alias definition. .. code-block:: scale read alias $fueltype1 1 2 3 end $fueltype2 4,5,6, 31-33 end $clad1 21,22,23 end $clad2 24 25 26 34-36 end $mod1 11 - 13 end $mod2 14-16, 37-39 end end alias The *ALIAS* block simply serves to assign material identifiers to specific variables, and the variables are used in subsequent data blocks. The same material identifier can be used in more than one alias if desired. As indicated earlier, TRITON will preprocess any input deck containing an *ALIAS* block and replace instances of alias variables with the appropriate material identifiers. The following subsections describe how aliases are implemented in TRITON's various input blocks, as the form of alias variable substitution is block dependent. Aliases are processed only in these input blocks; aliases used in other blocks will result in an error. *COMPOSITION* block aliases ^^^^^^^^^^^^^^^^^^^^^^^^^^^ The *COMPOSITION* block uses aliases to create multiple copies of each standard composition specification, replacing the alias variable with each material identifier associated with the alias definition. For example, consider the following alias definition in an *ALIAS* block: .. code-block:: scale read alias $fuel 1 2 10 end end alias and the standard composition specification: .. code-block:: scale uo2 $fuel den=10.045 1 800 92235 2.5 92238 97.5 end A modified TRITON input would be created with the standard composition specification replaced by .. code-block:: scale uo2 1 den=10.045 1 800 92235 2.5 92238 97.5 end uo2 2 den=10.045 1 800 92235 2.5 92238 97.5 end uo2 10 den=10.045 1 800 92235 2.5 92238 97.5 end *CELLDATA* block aliases ^^^^^^^^^^^^^^^^^^^^^^^^ *CELLDATA* block *latticecell* specifications typically contain more than one material; therefore, multiple aliases are permitted in each cell specification. However, this constrains the set of aliases used in the cell specification to have the same number of material identifiers associated with it. Consider the *ALIAS* block: .. code-block:: scale read alias $fuel 1-3 10 end $clad 4,5,6,11 end $mod 7 8-9 12 end end alias All three aliases contain four materials each. One could then create a single cell specification that uses one or more of these alias variables, such as .. code-block:: scale latticecell squarepitch pitch=1.26 $mod fuelr=0.41 $fuel cladr=0.50 $clad end This would result in the following alias expansion by TRITON: .. code-block:: scale latticecell squarepitch pitch=1.26 7 fuelr=0.41 1 cladr=0.50 4 end latticecell squarepitch pitch=1.26 8 fuelr=0.41 2 cladr=0.50 5 end latticecell squarepitch pitch=1.26 9 fuelr=0.41 3 cladr=0.50 6 end latticecell squarepitch pitch=1.26 12 fuelr=0.41 10 cladr=0.50 11 end Material identifiers are substituted according to their position in the alias definition (i.e., the first substitution will use the first material associated with each alias, and the second expansion will use the second material associated with each alias, etc.) Material numbers should not be entered manually in a cell specification; for example, .. code-block:: scale latticecell triangpitch pitch=1.26 $mod fuelr=0.4095 1 end TRITON would allow this to occur and would create a set of cell specifications as follows: .. code-block:: scale latticecell triangpitch pitch=1.26 2 fuelr=0.4095 1 end latticecell triangpitch pitch=1.26 3 fuelr=0.4095 1 end where $mod was defined as materials 2 and 3. However, SCALE does not allow the same material identifier to occur in two different cell specifications, and the fact that material 1 occurs in two different cell specifications would result in TRITON ending with an error. *Note that alias expansions for*\ **multiregion**\ *and*\ **doublehet**\ *cell specifications are not supported. Also note that TRITON will not copy*\ **centrmdata**\ *and*\ **moredata**\ *specifications that follow a cell specification that uses an alias variable.* *DEPLETION* block aliases ^^^^^^^^^^^^^^^^^^^^^^^^^ Aliases in the TRITON *DEPLETION* are simply replaced by the set of materials associated with the alias. For example, the *ALIAS* block .. code-block:: scale read alias $fuel 1 2 10 end end alias and DEPLETION block .. code-block:: scale read depletion 7 8 9 $fuel end depletion would be expanded to .. code-block:: scale read depletion 7 8 9 1 2 10 end depletion Aliases may be mixed with actual material numbers in the depletion block, along with the flux and assign keywords. *However, the negative sign-used to define the basis for power normalization-cannot precede an alias definition.* *TIMETABLE* block aliases ^^^^^^^^^^^^^^^^^^^^^^^^^ *TIMETABLE* block alias expansion is similar to that of the *COMPOSITION* block: TRITON will create a new timetable entry for each material associated with the alias used in the *TIMETABLE* definition. For the *TIMETABLE* block below, using the alias *$allmod*, unique timetables will be created for each material identifier associated with this alias variable. .. note:: Note that alias expansion of **density** timetable entries is not yet supported. .. code-block:: scale read timetable temperature $allmod 0.0 615 121.0 615 121.01 685 322.5 685 352.5 610 738.75 610 end end timetable *BRANCH* block aliases ^^^^^^^^^^^^^^^^^^^^^^ Aliases may be used within the *define* keyword definitions of the *BRANCH* block. Aliases are simply replaced by the list of materials associated with the alias, as is done for the *DEPLETIO*\ N block. Hence, .. code-block:: scale read alias $fuel 1 2 10 end end alias used with .. code-block:: scale read branch define fuel $fuel end md=0.75 tm=559 tf=880 sb=0.0 cr=0 end tf=1600 end end branch would be expanded to .. code-block:: scale read branch define fuel 1 2 10 end md=0.75 tm=559 tf=880 sb=0.0 cr=0 end tf=1600 end end branch NEWT *MATERIAL* block aliases ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ The *MATERIAL* block within the NEWT model section of a TRITON input can also use aliases. As with *COMPOSITION* and *TIMETABLE* entries, TRITON will create a new material specification for each material represented by an alias. For the sample material block below, using the alias *$fuel*, unique material block entries will be created for each material associated with the alias variable. .. code-block:: scale read materials mix=$fuel pn=1 com="3.25 wo uo2 fuel" end mix=21 pn=1 com="zirc cladding" end mix=31 pn=1 com="water" end end materials If an alias were defined as .. code-block:: scale $fuel 10 11 12 end then the *MATERIAL* block would be expanded to .. code-block:: scale read materials mix=10 pn=1 com="3.25 wo uo2 fuel" end mix=11 pn=1 com="3.25 wo uo2 fuel" end mix=12 pn=1 com="3.25 wo uo2 fuel" end mix=21 pn=1 com="zirc cladding" end mix=31 pn=1 com="water" end end materials .. _3-1-3-6: *KEEP_OUTPUT* block ^^^^^^^^^^^^^^^^^^^ When performing depletion calculations for a number of different materials, TRITON output can become quite voluminous. Often, much of that output is not needed for calculations that seek only eigenvalues, sources, or concentrations as a function of irradiation history. TRITON provides the ability to trim output to only those portions for which output is desired. Output produced directly by the TRITON module is always provided and cannot be disabled, but output from any other code in the sequence can be automatically removed from the output listing. Retaining certain output is accomplished using the *KEEP_OUTPUT* data block. The *KEEP_OUTPUT* data block provides the ability to preserve only selected outputs. The format of this data block is .. code-block:: scale read keep_output module_1 module_1 ... module_i ... module_N end keep_output where ``module_i`` represents any valid module name from the list of modules invoked by TRITON, as listed here: xsproc xsdrn newt kenova kenovi couple origen Without the *KEEP_OUTPUT* data block, the output from the neutron transport kernel (xsdrn, newt, kenova, kenovi) is retained and the output of all other modules (xsproc, couple, origen) is suppressed. Note that the output of SAMS and OPUS is not controlled by this block; the output of these modules is always retained. If a *KEEP_OUTPUT* data block is included, then only the output of the specified modules is retained. If the output of the neutron transport kernel is desired, then the corresponding module has to be listed since the above mentioned default is no longer applicable. A *KEEP_OUTPUT* data block without any module name can be specified to suppress the output of all modules. When using the TRITON-Shift sequence, the generation of Shift's HDF5 output files can be controlled through a time array in the *KEEP_OUTPUT* output. The special keywords "ALL" for all depletion steps and "LAST" for the last depletion step are supported. **Examples:** 1. Only the output of the neutron transport kernel is retained: *The input does not contain a KEEP_OUTPUT data block.* 2. Only the XSProc output is retained; the output of the neutron transport kernel is suppressed: .. code-block:: scale read keep_output xsproc end keep_output 3. The output of both XSProc and the neutron transport kernel KENO-VI is retained: .. code-block:: scale read keep_output xsproc kenovi end keep_output 4. The output of all modules is suppressed: .. code-block:: scale read keep_output end keep_output 5. Shift's HDF5 output files are requested for the first, second and last depletion step: .. code-block:: scale read keep_output shift 0 1 LAST end end keep_output .. _3-1-3-7: TRITON control parameters ~~~~~~~~~~~~~~~~~~~~~~~~~ TRITON supports the following of control parameter options: parm= CHECK, CENTRM, 2REGION, XSLEVEL=N, WEIGHT, WEIGHT=N, ADDNUX=N, INFDCUTOFF=X, CXM=N, MAXDAYS=N If an invalid control parameter option is specified, including misspelled keywords, an error message will be generated and execution terminated. TRITON also provides the ability to nest several control parameter keywords together; to combine keywords (where appropriate), a list may be entered, enclosed in parentheses, and separated by commas. For example, to specify CHECK, 2REGION, and ADDNUX=1 at the same time, input would begin with .. code-block:: scale =t-depl parm=(check, 2region,addnux=1) The following subsections provide more detail on each of the control parameters listed above. Check mode: *parm=check* ^^^^^^^^^^^^^^^^^^^^^^^^ Specification of *parm=check* will request that TRITON read all input and ensure that no input errors are present, without running additional calculations. In this mode, all input is set up as if a full calculation will be run, but the sequence exits without any functional module execution. The check mode is useful for debugging or obtaining processed standard composition data, without actually running a calculation. It can also be used to generate plot files for embedded NEWT and KENO inputs for additional review and checking of input specifications. Of course, some errors may be uncovered only by dynamically executing the functional modules; hence, there are rare occasions where a *parm=check* run will complete with no errors but will fail when run outside of check mode as the problem begins to run. .. _3-1-3-7-2: Multigroup cross section processing options ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ The most common use of *parm=* sequence control is in the selection of an alternate multigroup cross section processing mode. By default, XSProc enables both the BONAMI and CENTRM modules for cross section processing. BONAMI-only XSProc calculations can be performed using the control parameter *parm=bonami*. TRITON also supports the control parameter *parm=(xslevel=N)*. The *xslevel* parameter option initializes various CENTRM options for the XSProc calculations. The *xslevel* option is equivalent to initializing all unit cell calculations with the following *centrmdata* specifications: .. code-block:: scale parm=(xslevel=1): centrmdata npxs=5 nfst=0 nthr=3 nmf6=-1 alump=0.3 demin=0.125 pmc_omit=1 pmc_dilute=5.0e5 end centrmdata parm=(xslevel=2): centrmdata npxs=5 nfst=0 nthr=3 nmf6=-1 end centrmdata parm=(xslevel=3): centrmdata alump=0.3 demin=0.125 pmc_omit=1 pmc_dilute=5.0e5 end centrmdata parm=(xslevel=4): [no centrmdata statement] The option *parm=(xslevel=4)* is equivalent to *parm=centrm*. The option *parm=(xslevel=3)* is the default for depletion sequences and is equivalent to *parm=centrm* but with some minor approximations to decrease run time. The option *parm=(xslevel=2)* is equivalent to *parm=2region* for all sequences. Note that the *xslevel=1* and *xslevel=3* options have additional specifications for keywords *alump*, *demin*, *pmc_omit*, and *pmc_dilute*. These keywords are further discussed in the XSProc chapter. The additional keyword specifications are used to decrease run-time for the CENTRM and PMC calculations. Internal investigations have shown that the approximations introduced by the additional keyword specifications have minimal impact on solution accuracy for a wide range of calculations. Therefore the additional keyword specifications are used by default for depletion calculations, where several CENTRM and PMC calculations are necessary. The additional keyword values are not used by default for nondepletion calculations to be consistent with the SCALE CSAS5 and CSAS6 criticality sequences. .. _3-1-3-7-3: Creating a broad group library: *parm=weight, parm=(weight=N)* ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ Used in tandem with the TRITON T-NEWT sequence, the specification *parm=weight* extends the sequence by setting up and executing the MALOCS2 module to generate a weighted broad-group cross-library (AMPX master format). The spectrum generated in the NEWT transport calculation is used as the weighting function for the collapse. Additionally, the broad-group library energy structure is defined by the NEWT *COLLASPE* block. The *parm=weight* option uses the problem-averaged flux spectrum for the weighting function in the collapse. The problem may be a simple pin cell or a full assembly. However, there may be cases where the flux in a specific region or material is most appropriate for the spectral collapse. TRITON allows identification of a specific material from which the collapsing spectrum should be used. When specified in the form *parm=(weight=N)*, the average flux determined for material N is used in place of the total domain spectrum to perform the collapse. TRITON sample problem 1 (:numref:`3-1-6-1`) provides an example of the use of T-NEWT to produce a new broad-group library. Note that the broad-group library produced in this calculation will reside in the SCALE temporary working directory with the name *newxnlib* at the end of the calculation. If the library will be needed for future calculations, the user should use a shell script to copy the library back to a more permanent location, and perhaps give it a more meaningful name. In sample problem 1, the SCALE 252-group master library is collapsed to 56 energy groups. The process for creating a broad-group master library is also supported in the 2D depletion sequence T-DEPL. When *parm=weight* or *parm=(weight=N)* is specified in a depletion calculation, the input cross section library must be one of the SCALE 238-group or 252-group libraries, which will automatically be collapsed to the SCALE 49-group or 56-group structure, respectively. An initial fine group calculation is performed for the input configuration, and the flux from the solution is used to create the broad group library. The initial calculation is then repeated with the new broad group library, followed by the remainder of the depletion calculation. *Note that for lattice physics calculations, the NEWT* COLLAPSE *block will be based on the 49-group (or 56-group) energy structure, not the fine group structure.* It is important to note that the 252-group library contains intermediate resonance parameters and other data that cannot be accurately collapsed into 56-group data with the collapsing procedures available in MALOCS2. These parameters are important for bonami-only cross section processing calculations, i.e., *parm=bonami*. Therefore, the *parm=centrm* option is recommended for follow-on application of the collapsed 56-group collapsed library. The 238-group and 49-group libraries do not contain intermediate resonance parameter data, and bonami-only processing is available, provided that this cross section processing option and group structure is suitable for the intended application. Inclusion of additional nuclides for depletion: *parm=(addnux=N)* ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ For depletion calculations, it is important to add trace quantities (1 :math:`\times` 10\ :sup:`-20` at/b-cm) of certain nuclides to the inventories of depletion materials in order to accurately track the nuclides' impact on cross section processing and transport calculations as a function of burnup. By default, TRITON automatically adds to all fuel materials trace quantities of a set of nuclides that have been determined to be important in the characterization of spent fuel. TRITON recognizes fuel materials as any material containing quantities of heavy metals (Z > 89) in the standard composition specification. TRITON provides user control of the set of nuclides added to a fuel material through the *parm=(addnux=N)* control parameter, where N is an integer value. For N = 0, no nuclides are added, which is generally a very poor approximation and should only be used when the ramifications are fully understood. For N = 1, a bare minimum set of 15 nuclides (actinides) are added; this will generate improved number density estimates for actinides in low-burnup fuels but will not update cross sections for fission products of primary importance. Again, use of this option is discouraged unless it addresses special modeling needs. For N = 2, the default setting for the TRITON depletion sequences, 95 nuclides are added. N = 3 and N = 4 add 231 and 388 nuclides, respectively. Note that in previous versions of TRITON, N = 2 would add 64 nuclides. The set of 64 nuclides is still supported by specifiying *parm=(addnux=-2)* in the input. The default in the SCALE 6.1 release remains *parm=(addnux=2).* :numref:`tab-triton-addnux-1` through :numref:`tab-triton-addnux-4` list the set of nuclides added in trace quantities for each value of *addnux*. .. table:: Additional nuclides added in trace quantities for `parm=(addnux=1)` :name: tab-triton-addnux-1 :class: longtable +----------------------+----------------+----------------+----------------+ | | :sup:`234`\ U | :sup:`235`\ U | :sup:`236`\ U | +----------------------+----------------+----------------+----------------+ | :sup:`238`\ U | :sup:`237`\ Np | :sup:`238`\ Pu | :sup:`239`\ Pu | +----------------------+----------------+----------------+----------------+ | :sup:`240`\ Pu | :sup:`241`\ Pu | :sup:`242`\ Pu | :sup:`241`\ Am | +----------------------+----------------+----------------+----------------+ | :sup:`242`\ Am | :sup:`243`\ Am | :sup:`242`\ Cm | :sup:`243`\ Cm | +----------------------+----------------+----------------+----------------+ | \*15 nuclides total. | +----------------------+----------------+----------------+----------------+ .. table:: Additional nuclides added in trace quantities for `parm=(addnux= -2)` :name: tab-triton-addnux-m2 :class: longtable +-----------------+-----------------+-----------------+-----------------+ | :sup:`1`\ H | :sup:`10`\ B | :sup:`11`\ B | | +-----------------+-----------------+-----------------+-----------------+ | :sup:`14`\ N | :sup:`16`\ O | :sup:`83`\ Kr | :sup:`93`\ Nb | +-----------------+-----------------+-----------------+-----------------+ | :sup:`94`\ Zr | :sup:`95`\ Mo | :sup:`99`\ Tc | :sup:`103`\ Rh | +-----------------+-----------------+-----------------+-----------------+ | :sup:`105`\ Rh | :sup:`106`\ Ru | :sup:`109`\ Ag | :sup:`126`\ Sn | +-----------------+-----------------+-----------------+-----------------+ | :sup:`135`\ I | :sup:`131`\ Xe | :sup:`135`\ Xe | :sup:`133`\ Cs | +-----------------+-----------------+-----------------+-----------------+ | :sup:`134`\ Cs | :sup:`135`\ Cs | :sup:`137`\ Cs | :sup:`143`\ Pr | +-----------------+-----------------+-----------------+-----------------+ | :sup:`144`\ Ce | :sup:`143`\ Nd | :sup:`145`\ Nd | :sup:`146`\ Nd | +-----------------+-----------------+-----------------+-----------------+ | :sup:`147`\ Nd | :sup:`147`\ Pm | :sup:`148`\ Pm | :sup:`149`\ Pm | +-----------------+-----------------+-----------------+-----------------+ | :sup:`148`\ Nd | :sup:`147`\ Sm | :sup:`149`\ Sm | :sup:`150`\ Sm | +-----------------+-----------------+-----------------+-----------------+ | :sup:`151`\ Sm | :sup:`152`\ Sm | :sup:`151`\ Eu | :sup:`153`\ Eu | +-----------------+-----------------+-----------------+-----------------+ | :sup:`154`\ Eu | :sup:`155`\ Eu | :sup:`152`\ Gd | :sup:`154`\ Gd | +-----------------+-----------------+-----------------+-----------------+ | :sup:`155`\ Gd | :sup:`156`\ Gd | :sup:`157`\ Gd | :sup:`158`\ Gd | +-----------------+-----------------+-----------------+-----------------+ | :sup:`160`\ Gd | :sup:`244`\ Cm | | | +-----------------+-----------------+-----------------+-----------------+ | \*49 additional nuclides in addition to the 15 nuclides added in | | addnux=1, for a total of 64. | +-----------------+-----------------+-----------------+-----------------+ .. table:: Additional nuclides added in trace quantities for `parm=(addnux=2)` :name: tab-triton-addnux-2 :class: longtable +-----------------+-----------------+-----------------+-----------------+ | :sup:`91`\ Zr | :sup:`93`\ Zr | :sup:`95`\ Zr | :sup:`96`\ Zr | +-----------------+-----------------+-----------------+-----------------+ | :sup:`95`\ Nb | :sup:`97`\ Mo | :sup:`98`\ Mo | :sup:`99`\ Mo | +-----------------+-----------------+-----------------+-----------------+ | :sup:`100`\ Mo | :sup:`101`\ Ru | :sup:`102`\ Ru | :sup:`103`\ Ru | +-----------------+-----------------+-----------------+-----------------+ | :sup:`104`\ Ru | :sup:`105`\ Pd | :sup:`107`\ Pd | :sup:`108`\ Pd | +-----------------+-----------------+-----------------+-----------------+ | :sup:`113`\ Cd | :sup:`115`\ In | :sup:`127`\ I | :sup:`129`\ I | +-----------------+-----------------+-----------------+-----------------+ | :sup:`133`\ Xe | :sup:`139`\ La | :sup:`140`\ Ba | :sup:`141`\ Ce | +-----------------+-----------------+-----------------+-----------------+ | :sup:`142`\ Ce | :sup:`143`\ Ce | :sup:`141`\ Pr | :sup:`144`\ Nd | +-----------------+-----------------+-----------------+-----------------+ | :sup:`153`\ Sm | :sup:`156`\ Eu | :sup:`242m`\ Am | | +-----------------+-----------------+-----------------+-----------------+ | \*31 additional nuclides in addition to the 15 nuclides in | | :numref:`tab-triton-addnux-1` | | and 49 nuclides in :numref:`tab-triton-addnux-m2`, for a total of 95. | +-----------------------------------+-----------------+-----------------+ .. table:: Additional nuclides added in trace quantities for `parm=(addnux=3)` :name: tab-triton-addnux-3 :class: longtable +-----------------+-----------------+-----------------+-----------------+ | :sup:`72`\ Ge | :sup:`73`\ Ge | :sup:`74`\ Ge | :sup:`76`\ Ge | +-----------------+-----------------+-----------------+-----------------+ | :sup:`75`\ As | :sup:`79`\ Br | :sup:`76`\ Se | :sup:`77`\ Se | +-----------------+-----------------+-----------------+-----------------+ | :sup:`78`\ Se | :sup:`80`\ Se | :sup:`82`\ Se | :sup:`81`\ Br | +-----------------+-----------------+-----------------+-----------------+ | :sup:`80`\ Kr | :sup:`82`\ Kr | :sup:`84`\ Kr | :sup:`85`\ Kr | +-----------------+-----------------+-----------------+-----------------+ | :sup:`86`\ Kr | :sup:`85`\ Rb | :sup:`86`\ Rb | :sup:`87`\ Rb | +-----------------+-----------------+-----------------+-----------------+ | :sup:`84`\ Sr | :sup:`86`\ Sr | :sup:`87`\ Sr | :sup:`88`\ Sr | +-----------------+-----------------+-----------------+-----------------+ | :sup:`89`\ Sr | :sup:`90`\ Sr | :sup:`89`\ Y | :sup:`90`\ Y | +-----------------+-----------------+-----------------+-----------------+ | :sup:`91`\ Y | :sup:`90`\ Zr | :sup:`92`\ Zr | :sup:`92`\ Mo | +-----------------+-----------------+-----------------+-----------------+ | :sup:`94`\ Mo | :sup:`96`\ Mo | :sup:`94`\ Nb | :sup:`96`\ Ru | +-----------------+-----------------+-----------------+-----------------+ | :sup:`98`\ Ru | :sup:`99`\ Ru | :sup:`100`\ Ru | :sup:`105`\ Ru | +-----------------+-----------------+-----------------+-----------------+ | :sup:`102`\ Pd | :sup:`104`\ Pd | :sup:`106`\ Pd | :sup:`110`\ Pd | +-----------------+-----------------+-----------------+-----------------+ | :sup:`107`\ Ag | :sup:`111`\ Ag | :sup:`106`\ Cd | :sup:`108`\ Cd | +-----------------+-----------------+-----------------+-----------------+ | :sup:`110`\ Cd | :sup:`111`\ Cd | :sup:`112`\ Cd | :sup:`114`\ Cd | +-----------------+-----------------+-----------------+-----------------+ | :sup:`115m`\ Cd | :sup:`116`\ Cd | :sup:`140`\ Ce | :sup:`113`\ In | +-----------------+-----------------+-----------------+-----------------+ | :sup:`140`\ La | :sup:`112`\ Sn | :sup:`114`\ Sn | :sup:`115`\ Sn | +-----------------+-----------------+-----------------+-----------------+ | :sup:`116`\ Sn | :sup:`117`\ Sn | :sup:`118`\ Sn | :sup:`119`\ Sn | +-----------------+-----------------+-----------------+-----------------+ | :sup:`120`\ Sn | :sup:`122`\ Sn | :sup:`123`\ Sn | :sup:`124`\ Sn | +-----------------+-----------------+-----------------+-----------------+ | :sup:`125`\ Sn | :sup:`121`\ Sb | :sup:`123`\ Sb | :sup:`124`\ Sb | +-----------------+-----------------+-----------------+-----------------+ | :sup:`125`\ Sb | :sup:`126`\ Sb | :sup:`120`\ Te | :sup:`122`\ Te | +-----------------+-----------------+-----------------+-----------------+ | :sup:`123`\ Te | :sup:`124`\ Te | :sup:`125`\ Te | :sup:`126`\ Te | +-----------------+-----------------+-----------------+-----------------+ | :sup:`127m`\ Te | :sup:`128`\ Te | :sup:`129m`\ Te | :sup:`130`\ Te | +-----------------+-----------------+-----------------+-----------------+ | :sup:`132`\ Te | :sup:`130`\ I | :sup:`131`\ I | :sup:`124`\ Xe | +-----------------+-----------------+-----------------+-----------------+ | :sup:`126`\ Xe | :sup:`128`\ Xe | :sup:`129`\ Xe | :sup:`130`\ Xe | +-----------------+-----------------+-----------------+-----------------+ | :sup:`132`\ Xe | :sup:`134`\ Xe | :sup:`136`\ Xe | :sup:`134`\ Ba | +-----------------+-----------------+-----------------+-----------------+ | :sup:`135`\ Ba | :sup:`136`\ Ba | :sup:`137`\ Ba | :sup:`138`\ Ba | +-----------------+-----------------+-----------------+-----------------+ | :sup:`136`\ Cs | :sup:`142`\ Pr | :sup:`142`\ Nd | :sup:`150`\ Nd | +-----------------+-----------------+-----------------+-----------------+ | :sup:`151`\ Pm | :sup:`144`\ Sm | :sup:`148`\ Sm | :sup:`154`\ Sm | +-----------------+-----------------+-----------------+-----------------+ | :sup:`152`\ Eu | :sup:`157`\ Eu | :sup:`232`\ U | :sup:`233`\ U | +-----------------+-----------------+-----------------+-----------------+ | :sup:`159`\ Tb | :sup:`160`\ Tb | :sup:`160`\ Dy | :sup:`161`\ Dy | +-----------------+-----------------+-----------------+-----------------+ | :sup:`162`\ Dy | :sup:`163`\ Dy | :sup:`164`\ Dy | :sup:`165`\ Ho | +-----------------+-----------------+-----------------+-----------------+ | :sup:`166`\ Er | :sup:`167`\ Er | :sup:`175`\ Lu | :sup:`176`\ Lu | +-----------------+-----------------+-----------------+-----------------+ | :sup:`181`\ Ta | :sup:`182`\ W | :sup:`183`\ W | :sup:`184`\ W | +-----------------+-----------------+-----------------+-----------------+ | :sup:`186`\ W | :sup:`185`\ Re | :sup:`187`\ Re | :sup:`197`\ Au | +-----------------+-----------------+-----------------+-----------------+ | :sup:`231`\ Pa | :sup:`233`\ Pa | :sup:`230`\ Th | :sup:`232`\ Th | +-----------------+-----------------+-----------------+-----------------+ | \*136 additional nuclides in addition to the 15 nuclides in | | :numref:`tab-triton-addnux-1`, 49 nuclides in | | :numref:`tab-triton-addnux-m2`, and 31 nuclides in | | :numref:`tab-triton-addnux-2`, for a total of 231. | +-----------------------------------+-----------------+-----------------+ .. table:: Additional nuclides added in trace quantities for `parm=(addnux=4)` :name: tab-triton-addnux-4 :class: longtable +-----------------+-----------------+-----------------+-----------------+ | :sup:`2`\ H | :sup:`3`\ H | :sup:`3`\ He | :sup:`4`\ He | +-----------------+-----------------+-----------------+-----------------+ | :sup:`6`\ Li | :sup:`7`\ Li | :sup:`7`\ Be | :sup:`9`\ Be | +-----------------+-----------------+-----------------+-----------------+ | :sup:`15`\ N | :sup:`17`\ O | :sup:`19`\ F | :sup:`23`\ Na | +-----------------+-----------------+-----------------+-----------------+ | :sup:`24`\ Mg | :sup:`25`\ Mg | :sup:`26`\ Mg | :sup:`27`\ Al | +-----------------+-----------------+-----------------+-----------------+ | :sup:`28`\ Si | :sup:`29`\ Si | :sup:`30`\ Si | :sup:`31`\ P | +-----------------+-----------------+-----------------+-----------------+ | :sup:`32`\ S | :sup:`33`\ S | :sup:`34`\ S | :sup:`36`\ S | +-----------------+-----------------+-----------------+-----------------+ | :sup:`35`\ Cl | :sup:`37`\ Cl | :sup:`36`\ Ar | :sup:`38`\ Ar | +-----------------+-----------------+-----------------+-----------------+ | :sup:`40`\ Ar | :sup:`39`\ K | :sup:`40`\ K | :sup:`41`\ K | +-----------------+-----------------+-----------------+-----------------+ | :sup:`40`\ Ca | :sup:`42`\ Ca | :sup:`43`\ Ca | :sup:`44`\ Ca | +-----------------+-----------------+-----------------+-----------------+ | :sup:`46`\ Ca | :sup:`48`\ Ca | :sup:`45`\ Sc | :sup:`46`\ Ti | +-----------------+-----------------+-----------------+-----------------+ | :sup:`47`\ Ti | :sup:`48`\ Ti | :sup:`49`\ Ti | :sup:`50`\ Ti | +-----------------+-----------------+-----------------+-----------------+ | :sup:`50`\ Cr | :sup:`52`\ Cr | :sup:`53`\ Cr | :sup:`54`\ Cr | +-----------------+-----------------+-----------------+-----------------+ | :sup:`55`\ Mn | :sup:`54`\ Fe | :sup:`56`\ Fe | :sup:`57`\ Fe | +-----------------+-----------------+-----------------+-----------------+ | :sup:`58`\ Fe | :sup:`58`\ Co | :sup:`58m`\ Co | :sup:`59`\ Co | +-----------------+-----------------+-----------------+-----------------+ | :sup:`58`\ Ni | :sup:`59`\ Ni | :sup:`60`\ Ni | :sup:`61`\ Ni | +-----------------+-----------------+-----------------+-----------------+ | :sup:`62`\ Ni | :sup:`64`\ Ni | :sup:`63`\ Cu | :sup:`65`\ Cu | +-----------------+-----------------+-----------------+-----------------+ | :sup:`70`\ Ge | :sup:`69`\ Ga | :sup:`71`\ Ga | :sup:`74`\ As | +-----------------+-----------------+-----------------+-----------------+ | :sup:`74`\ Se | :sup:`79`\ Se | :sup:`78`\ Kr | :sup:`110m`\ Ag | +-----------------+-----------------+-----------------+-----------------+ | :sup:`113`\ Sn | :sup:`123`\ Xe | :sup:`130`\ Ba | :sup:`132`\ Ba | +-----------------+-----------------+-----------------+-----------------+ | :sup:`133`\ Ba | :sup:`136`\ Ce | :sup:`138`\ Ce | :sup:`139`\ Ce | +-----------------+-----------------+-----------------+-----------------+ | :sup:`138`\ La | :sup:`148m`\ Pm | :sup:`153`\ Gd | :sup:`156`\ Dy | +-----------------+-----------------+-----------------+-----------------+ | :sup:`158`\ Dy | :sup:`166m`\ Ho | :sup:`162`\ Er | :sup:`164`\ Er | +-----------------+-----------------+-----------------+-----------------+ | :sup:`168`\ Er | :sup:`170`\ Er | :sup:`174`\ Hf | :sup:`176`\ Hf | +-----------------+-----------------+-----------------+-----------------+ | :sup:`177`\ Hf | :sup:`178`\ Hf | :sup:`179`\ Hf | :sup:`180`\ Hf | +-----------------+-----------------+-----------------+-----------------+ | :sup:`182`\ Ta | :sup:`191`\ Ir | :sup:`193`\ Ir | :sup:`196`\ Hg | +-----------------+-----------------+-----------------+-----------------+ | :sup:`198`\ Hg | :sup:`199`\ Hg | :sup:`200`\ Hg | :sup:`201`\ Hg | +-----------------+-----------------+-----------------+-----------------+ | :sup:`202`\ Hg | :sup:`204`\ Hg | :sup:`204`\ Pb | :sup:`206`\ Pb | +-----------------+-----------------+-----------------+-----------------+ | :sup:`207`\ Pb | :sup:`208`\ Pb | :sup:`209`\ Bi | :sup:`223`\ Ra | +-----------------+-----------------+-----------------+-----------------+ | :sup:`224`\ Ra | :sup:`225`\ Ra | :sup:`225`\ Ac | :sup:`226`\ Ac | +-----------------+-----------------+-----------------+-----------------+ | :sup:`227`\ Ac | :sup:`226`\ Ra | :sup:`227`\ Th | :sup:`228`\ Th | +-----------------+-----------------+-----------------+-----------------+ | :sup:`229`\ Th | :sup:`233`\ Th | :sup:`234`\ Th | :sup:`232`\ Pa | +-----------------+-----------------+-----------------+-----------------+ | :sup:`235`\ Np | :sup:`236`\ Np | :sup:`238`\ Np | :sup:`239`\ Np | +-----------------+-----------------+-----------------+-----------------+ | :sup:`237`\ U | :sup:`239`\ U | :sup:`240`\ U | :sup:`241`\ U | +-----------------+-----------------+-----------------+-----------------+ | :sup:`236`\ Pu | :sup:`237`\ Pu | :sup:`243`\ Pu | :sup:`244`\ Pu | +-----------------+-----------------+-----------------+-----------------+ | :sup:`246`\ Pu | :sup:`244`\ Am | :sup:`244m`\ Am | :sup:`241`\ Cm | +-----------------+-----------------+-----------------+-----------------+ | :sup:`245`\ Cm | :sup:`246`\ Cm | :sup:`247`\ Cm | :sup:`248`\ Cm | +-----------------+-----------------+-----------------+-----------------+ | :sup:`249`\ Cm | :sup:`250`\ Cm | :sup:`249`\ Bk | :sup:`250`\ Bk | +-----------------+-----------------+-----------------+-----------------+ | :sup:`249`\ Cf | :sup:`250`\ Cf | :sup:`251`\ Cf | :sup:`252`\ Cf | +-----------------+-----------------+-----------------+-----------------+ | :sup:`253`\ Cf | :sup:`254`\ Cf | :sup:`253`\ Es | :sup:`254`\ Es | +-----------------+-----------------+-----------------+-----------------+ | :sup:`255`\ Es | | | | +-----------------+-----------------+-----------------+-----------------+ | \*158 additional nuclides in addition to the 15 nuclides in | | :numref:`tab-triton-addnux-1`, | | 49 nuclides in :numref:`tab-triton-addnux-m2`, 30 nuclides in | | :numref:`tab-triton-addnux-2`, and 136 nuclides in | | :numref:`tab-triton-addnux-3`, for a total of 388. | +-----------------------------------------------------------------------+ Few-group reaction cross section calculation control for continuous energy depletion: *parm=(cxm=N)* ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ In continuous energy depletion calculations, few group reaction cross sections are computed by KENO directly rather than using a post-processing approach that TRITON uses for multigroup mode. In addition to these region averaged multigroup reaction cross sections, KENO also provides problem-dependent region-averaged multigroup fluxes to TRITON that will be used by COUPLE to generate one-group cross section library for each depletion material. Option *parm=(cxm=N)* is used to setup continuous-energy depletion calculation with different modes of calculation, which tells KENO the details of the tallying process for the reaction cross sections and mixture fluxes. Available calculations modes and their descriptions are presented in :numref:`tab-triton-cxm`. .. tabularcolumns:: |p{1.5cm}|p{1.5cm}|p{1.5cm}|p{1.5cm}|p{8cm}| .. table:: `cxm` values and their descriptions. :name: tab-triton-cxm :class: longtable +------------+------------------------------------------+--------------------------+------------------------------------------------------------+ | **cxm** | **cross sections** | **flux** | **description** | | +----------------+-------------------------+--------------------------+ | | | reactions | number of energy groups | number of energy groups | | +------------+----------------+-------------------------+--------------------------+------------------------------------------------------------+ | 1 | All | NGP | NGP | KENO uses default NGP-group energy group boundaries | | | | | | to generate region-averaged reaction cross sections | | | | | | for all available reactions of the nuclides in each | | | | | | depletion mixture. KENO also computes region-averaged | | | | | | multigroup fluxes using the default NGP-group energy bins. | | | | | | multigroup fluxes using the default NGP-group energy bins. | +------------+----------------+-------------------------+--------------------------+------------------------------------------------------------+ | 2 | Transmutation | NGP | NGP | KENO uses default NGP-group energy group boundaries | | | (MT=16-18, | | | to generate region-averaged reaction cross sections | | | 102-125) | | | for only transmutation reactions of the nuclides in each | | | | | | depletion mixture. KENO also computes region-averaged | | | | | | multigroup fluxes using the default NGP-group energy bins. | | | | | | multigroup fluxes using the default NGP-group energy bins. | +------------+----------------+-------------------------+--------------------------+------------------------------------------------------------+ | 3 | All | 1 | NGP | KENO uses 1-group energy group boundaries | | | | | | to generate region-averaged reaction cross sections | | | | | | for all available reactions of the nuclides in each | | | | | | depletion mixture. KENO also computes region-averaged | | | | | | multigroup fluxes using the default NGP-group energy bins. | | | | | | multigroup fluxes using the default NGP-group energy bins. | +------------+----------------+-------------------------+--------------------------+------------------------------------------------------------+ | 4 | Transmutation | 1 | NGP | KENO uses 1-group energy group boundaries | | (default) | (MT=16-18, | | | to generate region-averaged reaction cross sections | | | 102-125) | | | for only transmutation reactions of the nuclides in each | | | | | | depletion mixture. KENO also computes region-averaged | | | | | | multigroup fluxes using the default NGP-group energy bins. | | | | | | multigroup fluxes using the default NGP-group energy bins. | +------------+----------------+-------------------------+--------------------------+------------------------------------------------------------+ .. note:: The energy group structure in KENO and associated number of energy groups, NGP, should be consistent with those from the ORIGEN library used in the problem. Infinite dilution cutoff control: *parm=(infdcutoff=X)* ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ The addition of nuclides to depletion materials as described in the previous section can lead to increased run-times for CENTRM-based XSProc calculations. However, many nuclides (e.g., low-density nuclides) are effectively infinitely dilute and can be treated as such and omitted from the expensive point-wise cross section collapse operation. For the option *parm=(infdcutoff=sigma0)* sequence option, XSProc will compute an effective background microscopic cross section for each nuclide. If the computed background cross section is greater than the cutoff value *sigma0*, recommended as 5 :math:`\times` 10\ :sup:`5` \ barns, then the nuclide is considered infinitely dilute and the infinitely dilute multigroup cross section is utilized from the cross section library. In general, a *sigma0* cutoff value of 5 :math:`\times` 10\ :sup:`5` barns will be acceptable for most applications. However, TRITON and the centrmdata card in the *CELLDATA* block provide a means for the user to control the cutoff value. The cutoff value may be assigned in either of two ways. A single global value may be assigned to all cells using the TRITON *parm=* specifier with the keyword *infdcutoff*, for example, *parm=(infdcutoff=1e10)*. Addition of the specifier with a value of 1 :math:`\times` 10\ :sup:`10` will set the cutoff value to 1 :math:`\times` 10\ :sup:`10` for all cells in the problem, which is generally appropriate for most calculations. However, a provision is made to specify a unique cutoff value to each cell using the *pmc_dilute* keyword in a *centrmdata* specification. An example of this is shown in the description of *parm=xslevel* in :numref:`3-1-3-7-2`. The default value of *sigma0* depends on the sequence and cross section processing option. For nondepletion sequences that use *parm=centrm*, the default is 0. The default value of 0 instructs PMC to include all nuclides for PMC processing. For depletion sequences that use *parm=centrm* or for any sequence that uses *parm=2region*, the default value is 5 :math:`\times` 10\ :sup:`5` barns. Override of the maximum number of days per depletion subinterval: *PARM=(MAXDAYS=N)* ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ TRITON is set to limit ORIGEN time intervals to no more than 40 days to avoid potential numerical error that would be introduced if depletion were performed over a long time interval. For depletion subintervals of more than 400 days (10 time intervals of 40 days), TRITON will automatically increase the number of depletion subintervals in a depletion interval. The depletion subinterval is based on a rule of thumb for ORIGEN depletion. However, the rule breaks down when burning at very low powers for extended time intervals. Thus, TRITON allows the user to override the default behavior by specifying a new value for the maximum number of days per ORIGEN time interval. A 100-day limit per ORIGEN time interval may be set using *parm=(maxdays=100)*. In overriding the default behavior, the user must be aware of any potential errors introduced in the approximation. Output Files Created by TRITON ------------------------------- TRITON produces a variety of output files that may be of use in related calculations. Of those files, only certain files are copied back to the return directory: the TRITON output file (.out); plot files generated by NEWT, KENO, or OPUS (.plt); SAMS sensitivity data files (.sdf), in the case of an S/U calculation; ORIGEN binary concentration files (.f71) and HTML-formatted output (.html), where available. The TRITON output file is a concatenated listing of outputs from TRITON and all modules for which output is kept. Other files of potential interest are not copied, and the user should be aware of these files and their names so that they may be retrieved using a SHELL script after TRITON execution is complete. The following subsections list those files and their purposes. Standard composition restart files ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ At the end of all depletion calculations, standard composition files are automatically produced for each material, listing the nuclides and number densities of the materials at the time the transport calculation (i.e., XSDRN, NEWT, KENO) is performed. Only nuclides for which cross section data are available in the master cross section library are saved in these files. Files are saved using the file naming convention StdCmpMix\ *NNNNN*, where *NNNNN* is the material identifier. The file contains compositions at the final time of the calculation. Additional files are saved with the file naming convention StdCmpMix\ *NNNNN_MMMMM*, where *MMMMM* is an index to a particular time step in the depletion calculation. For example, if a calculation is completed with materials 1 and 40 for two depletion steps, then the following files will be created in the temporary working directory. .. code-block:: none StdCmpMix00001_00000 (t=0) StdCmpMix00001_00001 (midpoint of 1st depletion step) StdCmpMix00001_00002 (midpoint of 2nd depletion step) StdCmpMix00001_00003 (final compositions, end of 2nd depletion step) StdCmpMix00001 (same as StdCmpMix00001_00003) StdCmpMix00040_00000 (t=0) StdCmpMix00040_00001 (midpoint of 1st depletion step) StdCmpMix00040_00002 (midpoint of 2nd depletion step) StdCmpMix00040_00003 (final compositions, end of 2nd depletion step) StdCmpMix00040 (same as StdCmpMix00040_00003) The contents of these files will be a standard composition description of each material by atomic contents-that is, SCALE standard nuclide IDs (e.g., U-235), number density, and temperature (using the temperature of the original material). Using SCALE's external file read capability, these outputs may be automatically included in a follow-on calculation that relies on depleted/decayed number densities. TRITON sample problem 7 (:numref:`3-1-6-7`) provides an example of the use of these restart files. These files are not automatically copied back to the output directory. Users can use a shell block to manually copy back the files, for example: .. code-block:: scale read shell cp ${TMPDIR}/StdCmpMix* ${OUTDIR} end shell .. important:: Standard composition restart files should be used only for follow-on criticality or shielding calculations. Lattice physics parameters ~~~~~~~~~~~~~~~~~~~~~~~~~~ During T-DEPL depletion calculations that use branch states and homogenization, a database of few-group cross sections is saved for each branch state and at each depletion step containing homogenized cross section data and other lattice physics parameters (e.g., discontinuity factors, pin power peaking factors, diffusion coefficients, etc.). The *xfile016* file is intended for post-processing, to be read and written in the desired format for subsequent nodal diffusion core simulator calculations. The *xfile016* file is a binary-formatted file, which is described in detail in Appendix A of TRITON. An auxiliary text-formatted database file (*txtfile16*) is also created that contains the same data as the binary-formatted file. ORIGEN binary library files (.f33) ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ During depletion calculations, ORIGEN binary library files are created to archive cross sections for each depletion material at each depletion subinterval. These files can be used in future depletion calculations in ORIGEN, ORIGAMI, and ARP. For each depletion material, the ORIGEN binary library file is named *${BASENAME}.mixNNNNNN.f33*, where *NNNNN* is the material number for each depleted material. Additionally, the system-average cross section file is saved with the name *${BASENAME}.system.f33*. All f33-files are automatically copied back into the output directory. Note that in SCALE versions up to 6.2, these files were named *ft33f001_mixNNNNN* and *f33f001_cmbined*. They had to be copied manually from the temporary working directory *${TMPDIR}* into the output directory through a shell block following the TRITON input. Since SCALE 6.3, the files are automatically copied back and have more intuitive names. ORIGEN binary concentration file (.f71) ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ During depletion calculations, TRITON creates the ORIGEN binary concentration file (.f71). This file is created in the temporary directory as *ft71f001* and is copied back at the end of the SCALE calculation to the return directory with the name ${OUTBASENAME}.f71. TRITON archives computed concentrations for each depletion material at the beginning and end of each depletion subinterval or decay interval. These files include concentrations and also decay heat term, photon and neutron data, and other quantities or interest computed by ORIGEN. These data may be post-processed by the OPUS module. The .f71 file contains concentrations for each individual material, and it also contains the combined concentrations of the individual material results (i.e., the net response for the entire system). The TRITON output contains an index of the contents of this file (see :numref:`3-1-5-4-5`). Binary mesh response files (.3dmap) ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ If mesh responses are requested, the corresponding binary 3dmap files are generated. If the TRITON-Shift sequences with the definitions and tallies block was used to request mesh responses, the output files follow the following syntax: - flux: *${BASENAME}.flux_time${STEP}_3dmap* - fission_density: *${BASENAME}.fission_density_time${STEP}.3dmap* - fission_source: *${BASENAME}.fission_source_time${STEP}.3dmap* Binary Shift output file (.h5) ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ If the TRITON-Shift sequence was used, detailed results of the Shift calculation can be found in Shift's hdf5 output file. The HDF5 filename follow the following syntax: *${basename}_time${step}.shift-output.h5*. The generation of these files for only selected depletion steps can be controled through the *KEEP_OUTPUT* block. Output Description ------------------ This section contains a brief description and explanation of TRITON output. As with any SCALE module, TRITON output begins with the SCALE header, the job information, the input file, and the program verification information. These outputs are common to all SCALE modules. Likewise, all SCALE calculations report a run-time summary at the end of the output file. Control parameter edit ~~~~~~~~~~~~~~~~~~~~~~ When TRITON control parameters are specified using the parm= command (see :numref:`3-1-3-7`), all specified parameters are echoed following the above output, with an explanation of the meaning of the parameter, as shown below. If no parameters are specified, no edit is provided. .. code-block:: none The following TRITON control parameters were requested: WEIGHT - Weighted collapsed master library option selected for t-newt calculation, based on system-averaged flux. ADDNUX - specifies the set of additional nuclides added in trace quantities for depletion calculations. Set 1 was selected. See TRITON manual for more information. T-XSEC output ~~~~~~~~~~~~~ The T-XSEC sequence performs only cross section processing functions. The XSProc output is written to the output file as the calculation proceeds. T-NEWT and T-XSDRN output ~~~~~~~~~~~~~~~~~~~~~~~~~ By default, the T-NEWT and T-XSDRN outputs include only the NEWT and XSDRN output respectively. The XSProc output can be included by using the *KEEP_OUTPUT* block (see :numref:`3-1-3-6`). Depletion sequence output ~~~~~~~~~~~~~~~~~~~~~~~~~ The output of TRITON depletion sequences contains several depletion edits. The edits are described in the following subsections. These output edits are written to the output file in the order in which they are computed during the calculation. .. _3-1-5-4-1: Burnup history summary (all depletion sequences) ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ TRITON generates the burnup history summary table after processing the *BURNDATA* block. An example of this table is as follows: .. code-block:: none *********************************************************************************************** Based on the supplied burnup history, triton will use the following time history to perform depletion calculations. This breakdown has been calculated so as to permit TRITON depletion steps of no more than 400 days. *********************************************************************************************** 7 Time-dependent libraries will be created Sub-Interval Depletion Sub-interval Specific Burn Length Decay Length Library Burnup No. Interval in interval Power(MW/MTIHM) (d) (d) (MWd/MTIHM) ---------------------------------------------------------------------------------------------------- ---------------------------------------------------------------------------------------------------- 0 ****Initial Bootstrap Calculation**** 0.00000E+00 1 1 1 37.883 42.500 0.000 8.05014e+02 2 1 2 37.883 42.500 15.000 2.41504e+03 3 2 1 32.215 45.000 0.000 3.94489e+03 4 2 2 32.215 45.000 50.000 5.39457e+03 ---------------------------------------------------------------------------------------------------- NOTE: Library Burnup is the cumulative burnup computed at the midpoint of the depletion sub-interval. Specific Power and Library Burnup depend on basis mixture(s) selected in DEPLETION block. ---------------------------------------------------------------------------------------------------- This table shows the results of a burnup history using one depletion interval with 5 depletion subintervals. Column 1 is the cumulative depletion subinterval number. Column 2 is the depletion interval number, and column 3 is the depletion subinterval number within the current depletion interval. Columns 4--6 echo the specific power, depletion interval, and decay interval specified in the *BURNDATA* block. The final column shows the cumulative burnup at the midpoint of each depletion subinterval. Embedded transport model output ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ The output from the initial transport calculation follows the burnup history edit. The output edits for NEWT, XSDRN, KENO-V.a, and KENO-VI are described in their respective manual sections. .. _3-1-5-4-3: System mass balance table ^^^^^^^^^^^^^^^^^^^^^^^^^ After the initial transport calculation output, a summary of system mass information is printed, an example of which is provided as follows. .. code-block:: none ******************************************************************************** ** System total mass is 1.8684e+01 grams heavy metal per unit length. ** ** Masses will be normalized by a factor of 5.3521e+04 cm to obtain a total ** ** system mass of 1.0000e+06 g of heavy metal ** ******************************************************************************** Mix Heavy Metal Normalized HM Fractional HM Heavy Metal Mixture No. Mass (g/cm) Mass (g) Mass (---) Dens. (g/cc) Dens. (g/cc) Depletion Mode 1 1.868420e+01 1.000000e+06 1.000000 9.177679e+00 1.041200e+01 Depleted by power 25 0.000000e+00 0.000000e+00 0.000000 0.000000e+00 6.440000e+00 Not depleted 26 1.951123e-16 1.044264e-11 0.000000 5.964321e-17 6.801399e-01 Depleted by flux System 1.868420e+01 1.000000e+06 1.000000 2.942208e+00 4.746240e+00 This table provides mass and density data for each material used in the transport model. Column 1 provides the material identifier, and columns 5 and 6 provide the material density and material heavy metal density, respectively, in units of grams per cubic centimeter. Heavy metal mass is determined from masses of all nuclides with an atomic number greater than 89. The final column provides the depletion mode for each material (see :numref:`3-1-3-3-4-2`). Column 2 provides the "prenormalized" heavy metal mass of each material. The units for this mass value depend on the transport model. For 2D *xy* NEWT models, the units are grams per centimeter since there is no *z-* dimension in the model. Similarly, the units are grams per centimeter for 1D cylinder XSDRN models, grams per square centimeter for 1D slab XSDRN models, and grams for 1D spherical XSDRN models and 3D KENO models. The total prenormalized heavy metal mass is printed in the final row of the table as well as in the table header. The heavy metal mass is normalized such that a total system mass of 1 MTHM is present. The volume scaling factor used to normalize the system mass is also printed in the table banner. The units of the volume scaling factor depend on the transport model. Column 3 prints the normalized material heavy metal mass in units of grams, which is equal to the prenormalized material heavy metal mass in column 2 multiplied by the volume scaling factor in the table header. The total normalized mass is printed in the final row and also in the table header. The fourth column shows the fractional heavy metal mass of all materials, which is equal to the normalized heavy metal mass in column 3, divided by the total normalized system heavy metal mass in the table header. .. _sect-triton-power-balance: Power balance tables ^^^^^^^^^^^^^^^^^^^^ As the TRITON calculation proceeds, the results of the cross section processing and transport calculations are used to calculate fluxes and powers in each mixture. The output segments listed in the next two tables show the results for the first calculation based on the initial mixture compositions. The total power (column 2) represents the mixture-specific power in units of MW/MTHM of initial **system** mass. The fractional power (column 3) is equal to the total power for a mixture divided by the total system power. The mixture power (column 4) represents mixture-specific power in units of MW/MTHM of initial **mixture** mass. The mixture power is equal to the total power of the mixture divided by the fractional heavy metal mass of the mixture, which is provided in the system mass balance table (:numref:`3-1-5-4-3`). Column 5 presents the burnup of the individual mixture. If a mixture does not contain heavy metal, then "N/A" is printed in the mixture power and burnup columns. Columns 6 and 7 show the mixture thermal and total flux values, respectively, in units of neutrons/cm\ :sup:`2`-sec. The thermal flux is determined by integrating multigroup flux values for energy groups below 0.625 eV. If the specific power is normalized to the total system power, the summation of the mixture powers in column 1 should match the input specification in the *BURNDATA* block (in the example given here, 37.883 MW/MTHM): .. code-block:: none --- Mixture powers for depletion pass no. 1 (MW/MITHM) --- Time = 21.25 days ( 0.058 y), Burnup = 0.805 GWd/MTIHM, Transport k= 1.2783 Total Fractional Mixture Mixture Mixture Mixture Mixture Power Power Power Burnup Thermal Flux Total Flux Number (MW/MTIHM) (---) (MW/MTIHM) (GWd/MTIHM) n/(cm^2*sec) n/(cm^2*sec) 1 37.799 0.99779 37.799 0.803 3.2106e+13 3.1724e+14 25 0.041 0.00109 N/A N/A 3.4431e+13 3.1682e+14 26 0.042 0.00112 N/A N/A 3.5114e+13 3.1832e+14 Total 37.883 1.00000 NOTE: Total Power is the Mixture Power per 1 metric ton of HM of the initial system mass. Mixture Power is the Mixture Power per 1 metric ton of HM of the initial mixture mass. Mixture Burnup is the Mixture Burnup per 1 metric ton of HM of the initial mixture mass. Mixture Thermal Flux determined using 0.625 eV cutoff: Groups 214 through 252. --------------------------------------------------------------- If the specific power is normalized to the power to one or more specific mixtures, the mixture powers are slightly different. For the case above, if depletion was performed with input power normalized to mixture 1, then mixture 1 has the input-specified power (37.883 MW/MTHM), and the power in the remainder of the model mixtures is normalized according to this basis mixture: .. code-block:: none --- Mixture powers for depletion pass no. 1 (MW/MITHM) --- Time = 21.25 days ( 0.058 y), Burnup = 0.805 GWd/MTIHM, Transport k= 1.2783 Total Fractional Mixture Mixture Mixture Mixture Mixture Power Power Power Burnup Thermal Flux Total Flux Number (MW/MTIHM) (---) (MW/MTIHM) (GWd/MTIHM) n/(cm^2*sec) n/(cm^2*sec) 1 * 37.883 0.99779 37.883 0.805 3.2176e+13 3.1795e+14 25 0.041 0.00109 N/A N/A 3.4507e+13 3.1753e+14 26 0.042 0.00112 N/A N/A 3.5191e+13 3.1903e+14 Total 37.967 1.00000 * - Power normalized to this mixture. NOTE: Total Power is the Mixture Power per 1 metric ton of HM of the initial system mass. Mixture Power is the Mixture Power per 1 metric ton of HM of the initial mixture mass. Mixture Burnup is the Mixture Burnup per 1 metric ton of HM of the initial mixture mass. Mixture Thermal Flux determined using 0.625 eV cutoff: Groups 214 through 252. --------------------------------------------------------------- .. note:: Note that the above two power balance tables refer to results at the time of the neutron transport calculation, i.e. the middle of a depletion subinterval. Additionally, *after* every depletion subinterval, a summary of the mixture-wise power, flux, fluence, burnup, and initial heavy metal is provided. This is the result from ORIGEN, i.e. results noramlized to 1 ton of initial heavy metal. .. code-block:: none end-of-step summary at time = 42.500 days ( 0.116 y), system-average burnup* = 1.606 GWd/MTIHM mixture power flux fluence burnup* initialhm (-) (MW) (n/cm2-s) (n/cm2) (MWd/MTIHM) (MTIHM) ------- ------------ ------------ ------------ ------------ ------------ 1 3.77957e+01 3.16389e+14 1.16178e+21 1.60632e+03 1.00000e+00 * Burnup is only calculated for mixtures with initial HM mass fraction greater than 1e-6. .. _3-1-5-4-5: ORIGEN binary concentration file listing '''''''''''''''''''''''''''''''''''''''' After all depletion calculations are completed, TRITON creates an ORIGEN binary concentration file (.f71) with isotopic concentrations for each depletion material. The order and content of the .f71 file is provided in the TRITON output. An example of this edit is shown below. For each depletion material, the output gives the location in the file, the ORIGEN time interval number, the depletion interval time in days, the cumulative time in years, and a title. After all materials are added to the library, the system average of all libraries and the average of all fuel mixtures are computed and added to the library. The file *case* numbers correspond to the mixture ids. Special cases are 0, -1, and -2 as indicated in the table below. .. code-block:: none File ft71f001 is the ORIGEN binary concentration file, containing concentrations for - each of the 2 depletion mixtures, - a set for the sum of all depletion mixtures, - a set for the sum of selected mixtures (from optional opus block), - a set for the sum of all fuel mixtures for always 7 time steps. Isotopic data locations are listed according to the following table. Position Time Step Cycle Time (d) Cumulative Time (y) 1 0 0.0000e+00 0.0000e+00 Depletion mixture no. 1 (ft71 case=1) 2 1 4.2500e+01 1.1636e-01 3 2 4.2500e+01 2.3272e-01 4 3 1.5000e+01 2.7379e-01 5 4 4.5000e+01 3.9699e-01 6 5 4.5000e+01 5.2019e-01 7 6 5.0000e+01 6.5708e-01 8 0 0.0000e+00 0.0000e+00 Depletion mixture no. 26 (ft71 case=26) 9 1 4.2500e+01 1.1636e-01 10 2 4.2500e+01 2.3272e-01 11 3 1.5000e+01 2.7379e-01 12 4 4.5000e+01 3.9699e-01 13 5 4.5000e+01 5.2019e-01 14 6 5.0000e+01 6.5708e-01 15 0 0.0000e+00 0.0000e+00 Weighted sum of concentrations of all depleted mixtures (ft71 case=0) 16 1 4.2500e+01 1.1636e-01 17 2 4.2500e+01 2.3272e-01 18 3 1.5000e+01 2.7379e-01 19 4 4.5000e+01 3.9699e-01 20 5 4.5000e+01 5.2019e-01 21 6 5.0000e+01 6.5708e-01 22 0 0.0000e+00 0.0000e+00 Weighted sum of concentrations for selected mixtures (ft71 case=-1) 23 1 4.2500e+01 1.1636e-01 24 2 4.2500e+01 2.3272e-01 25 3 1.5000e+01 2.7379e-01 26 4 4.5000e+01 3.9699e-01 27 5 4.5000e+01 5.2019e-01 28 6 5.0000e+01 6.5708e-01 29 0 0.0000e+00 0.0000e+00 Weighted sum of concentrations for fuel mixtures (ft71 case=-2) 30 1 4.2500e+01 1.1636e-01 31 2 4.2500e+01 2.3272e-01 32 3 1.5000e+01 2.7379e-01 33 4 4.5000e+01 3.9699e-01 34 5 4.5000e+01 5.2019e-01 35 6 5.0000e+01 6.5708e-01 The requested OPUS output edits follow this .f71 file summary edit. .. _SECT-TRITON-SAMPLES: TRITON Sample Cases ------------------- This section provides descriptions of the 13 TRITON sample problems included with SCALE. Note that all of these problems (along with all other SCALE sample problems) are typically executed in the initial SCALE installation to test the performance of various codes and options, for validation of the installation process. Because of the number of problems that are executed, these sample problems are adjusted to run as fast as possible so that all test problems may be completed in relatively short order. To accomplish this, crude modeling approximations (reduced convergence, few histories, simplified cross section processing, low-order quadrature and scattering approximations, coarse computational grids, reduced numbers of libraries per depletion cycle, etc.) may be used. Hence, although these problems provide guidance in setting up and executing TRITON problems, it is generally a good idea to review all control settings to ensure sufficient accuracy in one's own calculations. Additional TRITON input files for several reactor types can be generated with the SCALE/ORIGEN Library Generator (SLIG). The SLIB documentation is available as Appendix B of the ORIGEN chapter. .. _3-1-6-1: TRITON sample problem 1: triton1.inp ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ Sample problem triton1.inp is an example of a T-NEWT transport calculation sequence. Input begins (as with all SCALE sequences) with a title card and cross section library specification; this calculation is performed using the 252-group ENDF/B-7.1 library. After the library specification, three materials are defined in the composition block, followed by a cell specification and the NEWT transport model. This example includes an axial height of 37.1 cm and will therefore do a buckled calculation based on this height. The geometric model consists of a simple pin cell, with cylindrical fuel and clad regions inside a square moderator region, with a 6 :math:`\times` 6 base grid. The NEWT *BOUNDS* block specifies that periodic boundary conditions are used for this model. This simple problem also demonstrates the use of TRITON's automatic cross section collapse capability-\ *parm=weight*. For *T-NEWT* calculations, TRITON uses the NEWT *COLLAPSE* block to define the broad-group energy structure. For this sample problem, the cross sections are collapsed to a 56-group format. The new broad-group library will be identified as filename *newxnlib* in the temporary working directory, which can be used in follow-up SCALE calculations. .. .. code-block:: scale ' THIS SAMPLE PROBLEM TEST THE FOLLOWING: ' ** t-newt sequence ' ** v7-252 group library ' ** centrm cross section processing (default for t-newt calculations) ' ** parm=weight option for the t-newt sequence, which uses the NEWT collapse block to specify a 252 -> 56 group collapse. ' ** latticecell cross section processing option =t-newt parm=weight Buckled pin-cell transport calculation v7-252 read comp u-234 1 0 6.74213e-6 296.15 end u-235 1 0 7.65322e-4 296.15 end u-236 1 0 3.68820e-6 296.15 end u-238 1 0 2.20912e-2 296.15 end o 1 0 4.57338e-2 296.15 end b-10 1 0 3.64042e-9 296.15 end b-11 1 0 1.46531e-8 296.15 end cr 25 0 6.67242e-5 296.15 end fe 25 0 1.25922e-4 296.15 end sn 25 0 4.17642e-4 296.15 end o 25 0 2.63724e-4 296.15 end zr 25 0 3.78392e-2 296.15 end h 26 0 6.68559e-2 296.15 end o 26 0 3.34279e-2 296.15 end end comp read celldata latticecell squarepitch pitch=1.2600 26 fuelr=0.4095 1 cladr=0.4750 25 end end celldata read model 238 group solution read parm dz=37.1 end parm read materials mix=1 com="3.0 enriched fuel, pin location 1" end mix=25 com="cladding" end mix=26 com="water" end end materials read geom global unit 1 cylinder 10 0.4095 cylinder 20 0.4750 cuboid 30 4p0.63 media 1 1 10 media 25 1 20 -10 media 26 1 30 -20 boundary 30 6 6 end geom read collapse 8r1 2r2 3 3r4 5 5r6 6r7 2r8 3r9 4r10 4r11 12 13 10r14 3r15 16 6r17 3r18 18r19 2r20 6r21 22 3r23 24 7r25 26 16r27 2r28 11r29 30 31 14r32 33 2r34 35 3r36 35r37 5r38 7r39 11r40 4r41 2r42 43 44 3r45 2r46 2r47 2r48 2r49 2r50 51 52 2r53 54 3r55 10r56 ' OLD 238G collapse to 49G ' 7r1 2 3 2r4 5 6 7 8 8 8r9 14r10 6r11 10r12 13 7r14 11r15 12r16 30r17 16r18 2r19 ' 6r20 3r21 6r22 14r23 3r24 5r25 4r26 5r27 5r28 5r29 10r30 5r31 32 33 34 2r35 ' 36 37 38 2r39 2r40 3r41 2r42 43 44 45 46 47 3r48 9r49 end collapse read bounds all=periodic end bounds end model end TRITON sample problem 2: triton2.inp ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ Sample problem triton2.inp is an example of a T-XSDRN transport calculation sequence. In this case, the parameter specification *parm=2region* instructs TRITON to perform cross section processing using the CENTRM-based two-region option in place of the default CENTRM-based S\ :sub:`N` option (see :numref:`3-1-2-1`). As in sample problem 1, a simple square-pitched pin cell is modeled but in this case using an XSDRN model block rather than the NEWT model block. The moderator radius was defined in order to preserve the volume of the moderator region. .. code-block:: scale ' THIS SAMPLE PROBLEM TEST THE FOLLOWING: ' ** t-xsdrn sequence ' ** v7-252 group library ' ** 2region cross section processing ' ** latticecell cross section processing option =t-xsdrn parm=2region pin-cell model with MOX v7-252 read comp ' Fuel u-234 1 0 2.5952E-7 900 end pu-238 1 0 4.6610E-5 900 end pu-241 1 0 1.7491E-4 900 end pu-242 1 0 1.3201E-4 900 end o-16 1 0 4.6586E-2 900 end pu-240 1 0 4.8255E-4 900 end pu-239 1 0 1.0156E-3 900 end u-235 1 0 5.4287E-5 900 end u-238 1 0 2.1387E-2 900 end ' zirc zr-90 2 0 3.8657E-2 620 end fe 2 0 1.3345E-4 620 end cr 2 0 6.8254E-5 620 end ' h2o h-1 3 0 4.8414E-2 575 end o-16 3 0 2.4213E-2 575 end b-10 3 0 4.7896E-6 575 end b-11 3 0 1.9424E-5 575 end end comp read cell latticecell squarepitch pitch=1.3127 3 fueld=0.8200 1 cladd=0.9500 2 end end cell read model pin-cell model with MOX read parm sn=16 end parm read materials mix=1 com='fuel' end mix=2 com='clad' end mix=3 com='moderator' end end materials read geom geom=cylinder rightBC=white zoneIDs 1 2 3 end zoneids zoneDimensions 0.41 0.475 0.7406117 end zoneDimensions zoneIntervals 3r10 end zoneIntervals end geom end model end TRITON sample problem 3: triton3.inp ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ Sample problem 3 illustrates the input format for a T-DEPL-1D depletion calculation. In this case, a single square-pitched pin-cell model is depleted, where the fuel composition is comprised of UO\ :sub:`2` fuel homogenized with aluminum and B\ :sub:`4`\ C. Although this is not representative of real fuel, it does allow one to observe the effect of boron depletion during burnup; results will show an increasing multiplication factor as boron is depleted, followed by a decreasing eigenvalue after the fuel depletion becomes the dominant contributor to reactivity change. Three depletion intervals are specified with the same power and no decay intervals. Two depletion subintervals are specified for the first two depletion intervals, with only one depletion subinterval for the final depletion interval. Note that this may be insufficient to capture the effect of boron depletion early in life; fewer depletion subintervals are used here only to reduce the run-time for this sample problem. In this model, power is normalized such that material 1 has a power density of 21.22 MW/MTHM (or MT/MTU for UO\ :sub:`2` fuel), and OPUS output is requested for 35 nuclides. The problem is run using the addnux=3 option set to add trace quantities of 231 nuclides to depletion materials. .. code-block:: scale ' THIS SAMPLE PROBLEM TEST THE FOLLOWING: ' ** t-depl-1d sequence ' ** v7-252 group library ' ** Sn centrm cross section processing (default for t-depl-1d calculations) ' ** latticecell cross section processing option ' ** parm=addnux=3 option to add 231 nuclides to fuel material ' ** deplete-by-constant power ' ** mixture power normalization ' ** opus block =t-depl-1d parm=(addnux=3) Infinite lattice depletion model for a single pincell. v7-252 read comp ' Fuel/AL2O3-B4C uo2 1 den=10.045 1 841 92234 0.022 92235 2.453 92236 0.011 92238 97.514 end b-10 1 0 8.5900E-4 841.0 end b-11 1 0 3.4400E-3 841.0 end c 1 0 1.0700E-3 841.0 end al 1 0 3.9000E-2 841.0 end ' Clad wtptzirc 4 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 620 end ' Moderator h2o 5 den=0.7573 1 557 end end comp read celldata latticecell squarepitch pitch=1.4732 5 fuelr=0.47250 1 cladr=0.5588 4 end end celldata read depletion -1 end depletion read burndata power=21.220 burn=750 down=0 nlib=2 end power=21.220 burn=750 down=0 nlib=2 end power=21.220 burn=375 down=0 nlib=1 end end burndata read opus units=gram symnuc=u-234 u-235 u-236 u-238 pu-238 pu-239 pu-240 pu-241 pu-242 pu-243 np-237 cs-133 cs-134 cs-135 cs-137 nd-143 nd-144 nd-145 nd-146 nd-148 nd-150 pm-147 sm-147 sm-148 sm-149 sm-150 sm-151 sm-152 eu-153 sm-154 eu-154 gd-154 eu-155 gd-155 o-16 end matl=0 1 end end opus read model Infinite-lattice pin model (one-fourth) read parm sn=16 end parm read materials mix=1 com='fuel' end mix=4 com='clad' end mix=5 pn=2 com='water' end end materials read geom geom=cylinder rightBC=white zoneIDs 1 4 5 end zoneids zoneDimensions 0.47250 0.5588 0.83116409 end zoneDimensions zoneIntervals 3r10 end zoneIntervals end geom end model end TRITON sample problem 4: triton4.inp ^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^ Sample problem *triton4.inp* performs a large-scale depletion calculation for a one-fourth PWR assembly, taking advantage of symmetry to reduce the problem size. The same fuel material is used in each fuel rod, which will result in assembly-averaged isotopic compositions for all fuel rods. If one wanted to obtain an isotopic estimate for one or more unique fuel rod locations, then different materials would be specified for different rod positions. Even though all fuel is identical at the beginning of life, unique materials must be specified if one desires to perform tracking of the unique response of each unique fuel position. The problem parameter specification *parm=(weight)* instructs TRITON to perform an automated cross section library collapse. For library collapse automation within depletion calculations (see :numref:`3-1-3-7-3`), TRITON will perform a single 252-group calculation at t = 0 to generate the 56-group cross section library. TRITON will restart the depletion calculation at t = 0 using the broad-group library after it is created. Because *parm=weight* is specified, the *NEWT COLLAPSE* block must comply with the 56-group energy structure and not the 252-group energy structure. The *COLLAPSE* block input is slightly different for the library collapse automation for *T-NEWT* calculations, where the *NEWT COLLAPSE* block must comply with the 252-group energy structure. Problem 4 also uses a timetable to specify boron letdown in the moderator. The initially specified boron concentration in the *COMP* (or *COMPOSITION*) data block is multiplied by a density multiplier at the time of each cross section processing and transport calculation (i.e., the midpoint of depletion subinterval). Linear interpolation is performed between values on the timetable to obtain the multiplier for a given time. Typically a multiplier of 1.0 is used for t = 0, and the beginning-of-life boron concentration is input in the *COMPOSITION* block, but this example demonstrates that this is not necessary. For this calculation, a 500 ppm boron concentration is specified in the standard composition description, and a concentration of (500 ppm)*(1.832), or 916 ppm, would be used in the t = 0 transport calculation. Problem 4 is also an example of a lattice physics calculation for a full fuel assembly. The NEWT model employs coarse-mesh finite-difference acceleration, whole-assembly homogenization, 2-energy-group collapse, and a pin-power print, and computes assembly discontinuity factors. Although this sample problem will create the cross section database file for core calculations, this sample problem does not contain branching calculations, nor do lattice physics calculations typically use boron letdown curves. Additional guidance for TRITON lattice physics calculations can be found in the lattice physics primer. Because only one fuel material is used, only one cell specification is necessary. If multiple fuel materials were used, then a corresponding cell specification would be required for each fuel, with a unique clad and moderator material identifier for each cell. To apply boron letdown properly, the moderator present in each cell specification would need to have the same letdown curve specified. Hence, a letdown timetable would need to be specified for each moderator (even if the moderators are not all used in the NEWT *model* block). If multiple fuel materials are used, requiring corresponding multiple clad, moderation, cell, and timetable specifications, the use of an *alias* specification can simplify input. Aliases are described in :numref:`3-1-3-5`; sample problems triton6.inp (:numref:`3-1-6-6`), triton8.inp (:numref:`3-1-6-8`), and triton12.inp (:numref:`3-1-6-12`) demonstrate the use of aliases. This case also illustrates the use of stacked OPUS cases within a single TRITON input file. Here, an OPUS calculation is requested to obtain the mass in grams of 26 actinides and fission products for material 1 and for the entire system; since material 1 is the entire set of depletion materials, the system output will be identical to the material 1 output. A second OPUS calculation is also specified, which requests a ranked output of the top 20 nuclides in terms of decay heat (in watts).), TRITON will perform a single 252-group calculation at t = 0 to generate the 56-group cross section library. TRITON will restart the depletion calculation at t = 0 using the broad-group library after it is created. Because *parm=weight* is specified, the *NEWT COLLAPSE* block must comply with the 56-group energy structure and not the 252-group energy structure. The *COLLAPSE* block input is slightly different for the library collapse automation for *T-NEWT* calculations, where the *NEWT COLLAPSE* block must comply with the 252-group energy structure. Problem 4 also uses a timetable to specify boron letdown in the moderator. The initially specified boron concentration in the *COMP* (or *COMPOSITION*) data block is multiplied by a density multiplier at the time of each cross section processing and transport calculation (i.e., the midpoint of depletion subinterval). Linear interpolation is performed between values on the timetable to obtain the multiplier for a given time. Typically a multiplier of 1.0 is used for t = 0, and the beginning-of-life boron concentration is input in the *COMPOSITION* block, but this example demonstrates that this is not necessary. For this calculation, a 500 ppm boron concentration is specified in the standard composition description, and a concentration of (500 ppm)*(1.832), or 916 ppm, would be used in the t = 0 transport calculation. Problem 4 is also an example of a lattice physics calculation for a full fuel assembly. The NEWT model employs coarse-mesh finite-difference acceleration, whole-assembly homogenization, 2-energy-group collapse, and a pin-power print, and computes assembly discontinuity factors. Although this sample problem will create the cross section database file for core calculations, this sample problem does not contain branching calculations, nor do lattice physics calculations typically use boron letdown curves. Additional guidance for TRITON lattice physics calculations can be found in the lattice physics primer. Because only one fuel material is used, only one cell specification is necessary. If multiple fuel materials were used, then a corresponding cell specification would be required for each fuel, with a unique clad and moderator material identifier for each cell. To apply boron letdown properly, the moderator present in each cell specification would need to have the same letdown curve specified. Hence, a letdown timetable would need to be specified for each moderator (even if the moderators are not all used in the NEWT *model* block). If multiple fuel materials are used, requiring corresponding multiple clad, moderation, cell, and timetable specifications, the use of an *alias* specification can simplify input. Aliases are described in :numref:`3-1-3-5`; sample problems triton6.inp (:numref:`3-1-6-6`), triton8.inp (:numref:`3-1-6-8`), and triton12.inp (:numref:`3-1-6-12`) demonstrate the use of aliases. This case also illustrates the use of stacked OPUS cases within a single TRITON input file. Here, an OPUS calculation is requested to obtain the mass in grams of 26 actinides and fission products for material 1 and for the entire system; since material 1 is the entire set of depletion materials, the system output will be identical to the material 1 output. A second OPUS calculation is also specified, which requests a ranked output of the top 20 nuclides in terms of decay heat (in watts). .. code-block:: scale ' THIS SAMPLE PROBLEM TEST THE FOLLOWING: ' ** t-depl sequence ' ** v7-252 group library ' ** 2region cross section processing ' ** parm=weight option for the t-depl sequence, which uses builtin 49-group collapse ' ** latticecell cross section processing option ' ** deplete-by-constant power ' ** system power normalization ' ** timetable block using density multiplier ' ** opus block defining multiple plots =t-depl parm=(2region,weight) Large scale 2-D depletion model with a boron letdown curve v7-252 read comp uo2 1 den=10.412 1 900 92234 0.04 92235 4.11 92238 95.85 end wtptzirc 25 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 600 end h2o 26 den=0.6798 1 593 end wtptbor 26 0.6798 1 5000 100 500e-6 593 end end comp read celldata latticecell squarepitch pitch=1.2600 26 fuelr=0.4025 1 cladr=0.4750 25 end end celldata read depletion 1 end depletion read timetable densmult 26 2 5010 5011 0.0 1.832 106 1.419 205 1.033 306 0.641 385 0.611 473 1.797 592 1.371 704 0.973 817 0.568 875 0.362 end end timetable read burndata power=37.883 burn=385 down=88 nlib=1 end power=32.215 burn=402 down=158 nlib=1 end end burndata read opus units=gram symnuc=u-234 u-235 u-236 u-238 pu-238 pu-239 pu-240 pu-241 pu-242 np-237 am-241 am-243 cm-242 cm-243 cs-134 cs-137 nd-143 nd-144 nd-145 nd-146 cm-244 cm-245 cm-246 cm-247 ru-106 am-242m end matl=0 1 end newcase units=watts sort=yes nrank=20 time=years end opus read model One-fourth fuel assembly read parm drawit=yes cmfd=yes xycmfd=0 echo=yes collapse=yes sn=4 inners=3 outers=200 epsilon=1e-3 end parm read materials mix=1 com='4.11 wt % enriched fuel' end mix=25 com='cladding' end mix=26 com='water' end end materials read collapse 40r1 16r2 end collapse read homog 500 whole_assm 1 25 26 end end homog read adf 1 500 n=10.71 e=10.71 end adf read geom ' unit 25 is a right-half water hole unit 25 cylinder 10 .4500 chord +x=0.0 cylinder 20 .4950 chord +x=0.0 cuboid 30 0.63 0.0 0.63 -0.63 media 26 1 10 media 25 1 20 -10 media 26 1 30 -20 boundary 30 2 4 ' unit 45 is top-half water hole unit 45 cylinder 10 .4500 chord +y=0.0 cylinder 20 .4950 chord +y=0.0 cuboid 30 0.63 -0.63 0.63 0.0 media 26 1 10 media 25 1 20 -10 media 26 1 30 -20 boundary 30 4 2 ' unit 46 is a 1/4 water hole unit 46 cylinder 10 .4500 chord +x=0 chord +y=0 cylinder 20 .495 chord +x=0 chord +y=0 cuboid 30 0.63 0. 0.63 0. media 26 1 10 media 25 1 20 -10 media 26 1 30 -20 boundary 30 2 2 ' unit 1 is a full material #1 rod unit 1 cylinder 10 .4025 cylinder 20 .4950 cuboid 30 0.63 -0.63 0.63 -0.63 media 1 1 10 media 25 1 20 -10 media 26 1 30 -20 boundary 30 4 4 ' unit 2 is a top-half material #1 rod unit 2 cylinder 10 .4025 chord +y=0 cylinder 20 .4950 chord +y=0 cuboid 30 0.63 -0.63 0.63 0.0 media 1 1 10 media 25 1 20 -10 media 26 1 30 -20 boundary 30 4 2 ' unit 3 is a right-half material #1 rod unit 3 cylinder 10 .4025 chord +x=0 cylinder 20 .4950 chord +x=0 cuboid 30 0.63 0.0 0.63 -0.63 media 1 1 10 media 25 1 20 -10 media 26 1 30 -20 boundary 30 2 4 global unit 100 cuboid 1 10.71 0.0 10.71 0.0 array 10 1 media 26 1 1 boundary 1 end geom read array ara=10 nux=9 nuy=9 pinpow=yes typ=cuboidal fill 46 2 2 45 2 2 45 2 2 3 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 25 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 25 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 end fill end array read bounds all=refl end bounds end model end TRITON sample problem 5: triton5.inp ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ Sample problem *triton5.inp* is similar to *triton4.inp*, except that it is a T5-DEPL case; thus, a KENO V.a transport model is used in place of the NEWT model of the earlier case. The KENO V.a model, although 3D, is axially uniform with reflecting boundary conditions, so it is effectively the same model as the 2D model of *triton4.inp*. Moreover, the KENO V.a model represents the full assembly rather than a one-fourth model. Hence, both cases will generate similar results. In the KENO model, only 300,000 neutron histories are retained, which is somewhat low to obtain good statistics on fluxes. The 238 ENDF/B-VII library is used for this sample problem compared to the 252 ENDF/B-VII.1 library utilized in *triton4.inp*. .. code-block:: scale ' THIS SAMPLE PROBLEM TEST THE FOLLOWING: ' ** t-depl sequence ' ** v7-252 group library ' ** 2region cross section processing ' ** latticecell cross section processing option ' ** deplete-by-constant power ' ** system power normalization ' ** timetable block using density multiplier =t5-depl parm=2region Large scale 2-D depletion model with boron density change. V7-238 read comp uo2 1 den=10.412 1 900 92234 0.04 92235 4.11 92238 95.85 end wtptzirc 25 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 600 end h2o 26 den=0.6798 1 593 end wtptbor 26 0.6798 1 5000 100 500e-6 593 end end comp read celldata latticecell squarepitch pitch=1.2600 26 fuelr=0.4025 1 cladr=0.4750 25 end end celldata read depletion 1 end depletion read timetable densmult 26 2 5010 5011 0.0 1.832 106 1.419 205 1.033 306 0.641 385 0.611 473 1.797 592 1.371 704 0.973 817 0.568 875 0.362 end end timetable read burndata power=37.883 burn=385 down=88 nlib=1 end power=32.215 burn=402 down=158 nlib=1 end end burndata read model read parm cfx=yes gen=620 nsk=20 npg=500 plt=no htm=no end parm read geom ' unit 2 is a water hole unit 2 cylinder 26 1 .4500 10.0 0.0 cylinder 25 1 .4950 10.0 0.0 cuboid 26 1 0.63 -0.63 0.63 -0.63 10.0 0.0 ' unit 1 is a material #1 rod unit 1 cylinder 1 1 .4025 10.0 0.0 cylinder 25 1 .4950 10.0 0.0 cuboid 26 1 0.63 -0.63 0.63 -0.63 10.0 0.0 global unit 100 array 10 0.0 0.0 0.0 end geom read array ara=10 nux=17 nuy=17 nuz=1 typ=cuboidal fill 17r1 17r1 8r1 2 8r1 17r1 17r1 8r1 2 8r1 17r1 17r1 2r1 2 2r1 2 2r1 2 2r1 2 2r1 2 2r1 17r1 17r1 8r1 2 8r1 17r1 17r1 8r1 2 8r1 17r1 17r1 end fill end array read bounds all=refl end bounds end data end model end .. _3-1-6-6: TRITON sample problem 6: triton6.inp ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ Sample problem *triton6.inp* performs T-DEPL depletion in a pin cell model; however, the pin is discretized into five equal-volume rings of fuel. Thus, CENTRM-based S\ :sub:`N` cross section processing is necessary to capture the radial burnup of the pin cell. A *multiregion* cell specification is given to allow specification of the varying radii for the fuel regions. Because the multiregion cell is cylindrical, the moderator volume is represented in terms of a radius that corresponds to the volume associated with the pin pitch. The right boundary condition for the cell is set to *white*; this is important, as the default right boundary condition for a multiregion cylinder is vacuum. In this case, addnux=1 is also requested in the parameter specification, simply for a faster (but less accurate) calculation. Material aliases are used to simplify input. The calculation is performed with the 238 ENDF/B-VII library. The TRITON *TIMETABLE* block is used to demonstrate time-dependent temperature changes to the moderator material. .. code-block:: scale ' THIS SAMPLE PROBLEM TEST THE FOLLOWING: ' ** t-depl sequence ' ** v7-252 group library ' ** centrm cross section processing ' ** multiregion cross section processing option ' ** deplete-by-constant power ' ** parm=addnux=1 option to add 15 nuclides to fuel material ' ** system power normalization ' ** timetable block using temperature change ' ** alias block definition ' ** opus block =t-depl parm=(centrm,addnux=1) Pin-cell depleted in rings v7-252 read alias $fuel 1-5 end end alias read comp uo2 $fuel den=9.459 1 829.0 92234 0.027 92235 3.038 92236 0.014 92238 96.921 end wtptzirc 10 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 620 end h2o 11 den=0.7575 1 557 end wtptbor 11 0.7575 1 5000 100 654e-6 557 end end comp read celldata multiregion cylindrical right=white end 1 0.16425 2 0.28449 3 0.36727 4 0.43456 5 0.49275 10 0.55880 11 .83120 end zone end celldata read depletion $fuel end depletion read timetable temperature 11 ' cycle 1 0.0 557.0 306.0 557.0 ' cycle 2 377.0 540.0 838.1 557.0 end end timetable read burndata power=27.24 burn=306.0 down=71 nlib=1 end power=34.57 burn=461.1 down=1870 nlib=1 end end burndata read opus units=gram symnuc=u-235 u-238 pu-239 pu-241 nd-148 end matl=0 1 2 3 4 5 end end opus read model Infinite lattice PWR pin cell read parm drawit=yes prtbroad=yes epsilon=1e-3 soln=b1 converg=matl end parm read materials mix=$fuel com='3.038 wt % enriched fuel' end mix=10 pn=0 com='cladding' end mix=11 com='water' end end materials read geom global unit 1 cylinder 1 .16425 chord +x=0 chord +y=0 cylinder 2 .28449 chord +x=0 chord +y=0 cylinder 3 .36727 chord +x=0 chord +y=0 cylinder 4 .43456 chord +x=0 chord +y=0 cylinder 5 .49275 chord +x=0 chord +y=0 cylinder 20 .5588 chord +x=0 chord +y=0 cuboid 30 0.7366 0.0 0.7366 0.0 media 1 1 1 media 2 1 2 -1 media 3 1 3 -2 media 4 1 4 -3 media 5 1 5 -4 media 10 1 20 -5 media 11 1 30 -20 boundary 30 4 4 end geom read bounds all=refl end bounds end model end .. _3-1-6-7: TRITON sample problem 7: triton7.inp ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ Sample problem triton7.inp is an example of a T-DEPL depletion calculation for a full PWR fuel assembly model. Depletion is performed on the basis of material 7, which is located in a single fuel pin for which destructive assay measurements were performed. All other fuel is modeled as a single (average) material, material 1. The parameter specification *parm=(2region,addnux=1,weight)* was chosen to reduce the run-time of the sample problem. This sample problem also demonstrates the use of TRITON's standard composition restart files and SCALE external file reading capabilities to represent the time-dependent behavior of an assembly in which burnable poisons are removed after the first cycle of operation. This problem consists of two TRITON 2D depletion cases. In the first case, the full assembly model contains borosilicate glass burnable poison rods (BPRs), material 4, which are included in the list of materials to be depleted. The calculation is run for the entirety of the first operational cycle, which included a 40-day mid-cycle decay interval. The model also includes a 64-day decay interval after the end of the operational cycle. When this calculation is completed, TRITON creates in the temporary working directory a standard composition file for each material containing the isotopic inventories for each depletion material at the end of the 64-day decay interval. The second TRITON calculation reads the standard composition specifications for materials 1 and 7 as part of the input to provide the fuel state for the second calculation. In the second model, the BPRs are removed and replaced with the moderator in the embedded NEWT model. The initial depletion calculation uses the 252 ENDF/B-VII.1 library. With the *parm=(…,weight)* option, a 56 group library is created in the temporary directory called *newxnlib*. This library is used for the second *T-NEWT* calculation. .. code-block:: scale ' THIS SAMPLE PROBLEM TEST THE FOLLOWING: ' ** t-depl sequence ' ** v7-252 group library ' ** 2region cross section processing ' ** latticecell cross section processing option ' ** deplete-by-constant flux ' ** parm=addnux=1 option to add 15 nuclides to fuel material ' ** mixture power normalization ' ** timetable block using density multiplier ' ** composition restart files. ' ** weight used to collapse library for reuse in restart calculation =t-depl parm=(2region,addnux=1,weight) ASSEMBLY model with BPRs with depletion v7-252 read comp uo2 1 den=9.550 1 743 92234 0.023 92235 2.561 92236 0.013 92238 97.403 end wtptzirc 2 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 620 end h2o 3 den=0.7544 1 559 end wtptbor 3 0.7544 1 5000 100 652.5e-6 559 end wtptbpr 4 2.081 6 8016 53.58 11000 2.82 13027 1.758 14000 37.63 19000 0.33 5000 3.882 1 559 end wtptair 5 0.00129 2 7000 78.0 8016 22.0 1 559.0 end ss304 6 1 559.0 end uo2 7 den=9.550 1 743 92234 0.023 92235 2.561 92236 0.013 92238 97.403 end wtptzirc 8 6.44 4 40000 97.91 26000 0.5 50118 0.64 50120 0.95 1 595 end h2o 9 den=0.7544 1 559 end wtptbor 9 0.7544 1 5000 100 652.5e-6 559 end end comp read celldata latticecell squarepitch pitch=1.43 3 fueld=0.9484 1 cladd=1.0719 2 end latticecell squarepitch pitch=1.43 9 fueld=0.9484 7 cladd=1.0719 8 end end celldata read depletion 1 -7 flux 4 end depletion read timetable density 3 2 5010 5011 0.00 1.000 243.5 1.000 283.5 0.379 527.0 0.379 end density 9 2 5010 5011 0.00 1.000 243.5 1.000 283.5 0.379 527.0 0.379 end end timetable read burndata power=20.86 burn=243.5 down=40.0 nlib=1 end power=20.15 burn=243.5 down=64.0 nlib=1 end end burndata read model ASSEMBLY model with BPRs with depletion read parm drawit=yes inners=2 epsilon=-5e-2 cmfd=1 xycmfd=0 echo=yes solntype=b1 timed=yes end parm read materials mix=6 pn=1 com="SS-304 - BPR clad" end mix=5 pn=1 com="air in BPRs" end mix=4 pn=1 com="borosilicate glass" end mix=3 pn=2 com="water" end mix=2 pn=1 com="cladding" end mix=1 pn=1 com="2.561 wt % enriched fuel " end mix=7 pn=1 com="rod N-9" end end materials read geom global unit 10 cuboid 13 10.725 0.0 10.725 0.0 array 101 13 place 1 1 -0.715 -0.715 media 3 1 13 boundary 13 15 15 unit 1 cuboid 13 1.43 0.0 1.43 0.0 cylinder 12 0.53595 origin x=0.715 y=0.715 cylinder 11 0.4742 origin x=0.715 y=0.715 media 3 1 13 -12 media 2 1 12 -11 media 1 1 11 boundary 13 2 2 unit 2 cuboid 13 1.43 0.0 1.43 0.0 cylinder 14 0.28385 origin x=0.715 y=0.715 cylinder 15 0.30035 origin x=0.715 y=0.715 cylinder 16 0.50865 origin x=0.715 y=0.715 cylinder 17 0.55755 origin x=0.715 y=0.715 media 3 1 13 -17 media 6 1 17 -16 media 4 1 16 -15 media 6 1 15 -14 media 5 1 14 boundary 13 2 2 unit 3 cuboid 13 1.43 0.0 1.43 0.0 cylinder 12 0.6934 origin x=0.715 y=0.715 cylinder 11 0.6502 origin x=0.715 y=0.715 media 3 1 13 -12 media 2 1 12 -11 media 3 1 11 boundary 13 2 2 unit 4 cuboid 13 1.43 0.715 1.43 0.0 cylinder 12 0.53595 origin x=0.715 y=0.715 chord +x=0.715 cylinder 11 0.4742 origin x=0.715 y=0.715 chord +x=0.715 media 3 1 13 -12 media 2 1 12 -11 media 1 1 11 boundary 13 1 2 unit 5 cuboid 13 1.43 0.0 1.43 0.715 cylinder 12 0.53595 origin x=0.715 y=0.715 chord +y=0.715 cylinder 11 0.4742 origin x=0.715 y=0.715 chord +y=0.715 media 3 1 13 -12 media 2 1 12 -11 media 1 1 11 boundary 13 2 1 unit 6 cuboid 13 1.43 0.715 1.43 0.0 cylinder 12 0.6934 origin x=0.715 y=0.715 chord +x=0.715 cylinder 11 0.6502 origin x=0.715 y=0.715 chord +x=0.715 media 3 1 13 -12 media 2 1 12 -11 media 3 1 11 boundary 13 1 2 unit 7 cuboid 13 1.43 0.0 1.43 0.715 cylinder 12 0.6934 origin x=0.715 y=0.715 chord +y=0.715 cylinder 11 0.6502 origin x=0.715 y=0.715 chord +y=0.715 media 3 1 13 -12 media 2 1 12 -11 media 3 1 11 boundary 13 2 1 unit 8 cuboid 13 1.43 0.715 1.43 0.715 cylinder 12 0.6934 origin x=0.715 y=0.715 chord +x=0.715 chord +y=0.715 cylinder 11 0.6502 origin x=0.715 y=0.715 chord +x=0.715 chord +y=0.715 media 3 1 13 -12 media 2 1 12 -11 media 3 1 11 boundary 13 1 1 unit 9 cuboid 13 1.43 0.0 1.43 0.0 cylinder 12 0.53595 origin x=0.715 y=0.715 cylinder 11 0.4742 origin x=0.715 y=0.715 media 3 1 13 -12 media 2 1 12 -11 media 7 1 11 boundary 13 2 2 end geom read array ara=101 nux=8 nuy=8 typ=cuboidal fill 8 5 5 5 7 5 5 5 4 1 1 1 1 1 1 1 4 1 1 1 1 2 1 1 4 1 1 3 1 1 1 1 6 1 1 1 1 1 1 1 4 9 2 1 1 2 1 1 4 1 1 1 1 1 1 1 4 1 1 1 1 1 1 1 end fill end array end model end .. code-block:: scale =t-newt parm=(2region) ASSEMBLY model without BPRs newxnlib read comp