MOCUP: MCNP-ORIGEN2 coupled utility program

Moore, Richard L.; Schnitzler, Bruce G.; Wemple, C.A.

doi:10.2172/130667

Cited by 69 publications

(28 citation statements)

References 2 publications

(2 reference statements)

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“…The calculation of temperatures on the outer surface of the kernel, and all other radii beyond the kernel, rely on the quasi steady-state assumption that implies: Based on the notation shown in Figure 3-3, Equation (3)(4)(5)(6)(7)(8)(9)(10)(11)(12)(13)(14)(15)(16)(17)(18)(19) and Fourier's law (for spherical coordinates) can be used to determine the temperature at the SiC / OPyC interface, given the input for the temperature of the exterior surface of the TRISO-coated fuel particle. Specifically, (3)(4)(5)(6)(7)(8)(9)(10)(11)(12)(13)(14)(15)(16)(17)(18)(19)(20) ( [3][4][5][6][7][8][9][10][11][12][13][14]…”

Section: Micro Temperaturesmentioning

confidence: 99%

“…Specifically, (3)(4)(5)(6)(7)(8)(9)(10)(11)(12)(13)(14)(15)(16)(17)(18)(19)(20) ( [3][4][5][6][7][8][9][10][11][12][13][14][15][16][17][18][19][20][21] Note that in instances where computational nodes are placed at the interface of each fuel particle layer (i.e., Figure 3-3), with no interior layer nodes, Equation (3)(4)(5)(6)(7)(8)(9)(10)(11)(12)(13)(14)(15)…”

Section: Micro Temperaturesmentioning

confidence: 99%

“…This method of solution, though, can be applied as well to a system with any number of coating layers. The result is a set of simultaneous differential equations of the form of Equations (4)(5)(6)(7)(8)(9)(10)(11)(12)(13)(14)(15)(16)(17)(18)(19) and (4)(5)(6)(7)(8)(9)(10)(11)(12)(13)(14)(15)(16)(17)(18)(19)(20), where there is an equation for each interface surface. These can be solved using matrix analysis, which yields a set of eigenvalues and eigenvectors for the system.…”

Section: Radial Stresses At Layer Interfacesmentioning

confidence: 99%

“…18 This CO production model considers burnup, temperature, uranium enrichment, and fuel composition in the calculation. Input to the HSC code consisted of elemental fission product inventories generated by the MOCUP 19 code which couples the MCNP 20 and ORIGEN2 21 codes.…”

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confidence: 99%

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PARFUME Theory and Model basis Report

Miller¹,

Petti²,

Maki³

et al. 2009

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The success of gas reactors depends upon the safety and quality of the coated particle fuel. The fuel performance modeling code PARFUME simulates the mechanical, thermal and physico-chemical behavior of fuel particles during irradiation. This report documents the theory and material properties behind various capabilities of the code, which include: 1) various options for calculating CO production and fission product gas release, 2) an analytical solution for stresses in the coating layers that accounts for irradiation-induced creep and swelling of the pyrocarbon layers, 3) a thermal model that calculates a time-dependent temperature profile through a pebble bed sphere or a prismatic block core, as well as through the layers of each analyzed particle, 4) simulation of multi-dimensional particle behavior associated with cracking in the IPyC layer, partial debonding of the IPyC from the SiC, particle asphericity, and kernel migration (or amoeba effect), 5) two independent methods for determining particle failure probabilities, 6) a model for calculating release-to-birth ratios of gaseous fission products that accounts for particle failures and uranium contamination in the fuel matrix, and 7)

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Section: Micro Temperaturesmentioning

confidence: 99%

Section: Micro Temperaturesmentioning

confidence: 99%

Section: Radial Stresses At Layer Interfacesmentioning

confidence: 99%

mentioning

confidence: 99%

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PARFUME Theory and Model basis Report

Miller¹,

Petti²,

Maki³

et al. 2009

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“…That is, k eff and the neutron flux are calculated and input into a coupled depletion/fission-product generation code. Although not inclusive, some of these codes include SAS2H (Hermann and Parks, 2000) from the SCALE Modular Code System (Parks, 1987), VENTURE (Shapiro and Huria, 1990), (Vondy, et al, 1981), COSMO (Uchikawa and Takeda, 1974), LASER (Poncelet, 1966), EPRI-CELL (ARMP-02, 1988a), PDQ7 (ARMP-02, 1988b), CORSYS (Caner, et al, 1994), MOCUP (Moore, et al, 1995), and MONTEBURNS (Poston and Trellue, 1999). These and other codes have been used extensively and perform adequately for their intended purpose.…”

Section: Introductionmentioning

confidence: 99%

BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies

Parma¹

2002

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BURNCAL is a Fortran computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. The code uses output parameters generated by the Monte Carlo neutronics code MCNP to determine the isotopic inventory as a function of time and power density. The code allows for multiple fueled regions to be analyzed. The companion code, RELOAD, can be used to shuffle fueled regions or reload regions with fresh fuel. BURNCAL can be used to study the reactivity effects and isotopic inventory as a function of time for a nuclear reactor system. Neutron transmutation, fission, and radioactive decay are included in the modeling of the production and removal terms for each isotope of interest. For a fueled region, neutron transmutation, fuel depletion, fission-product poisoning, actinide generation, and burnable poison loading and depletion effects are included in the calculation. Fueled and un-fueled regions, such as cladding and moderator, can be analyzed simultaneously. The nuclides analyzed are limited only by the neutron cross section availability in the MCNP cross-section library. BURNCAL is unique in comparison to other burnup codes in that it does not use the calculated neutron flux as input to other computer codes to generate the nuclide mixture for the next time step. Instead, BURNCAL directly uses the neutron absorption tally/reaction information generated by MCNP for each nuclide of interest to determine the nuclide inventory for that region. This allows for the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. EXECUTIVE SUMMARYThere are many coupled neutronic/isotope-composition codes available to the nuclear community that are capable of performing burnup and depletion calculations on nuclear reactor fuels or fuel element assemblies. In general, these coupled codes all work in a similar fashion -a neutronics code performs a calculation to determine the multiplication factor, k eff , and the energy dependent neutron flux for a given geometry and fuel loading; another code then uses the neutron flux, power density, and material composition to determine the new material composition after a given burnup time. The material composition can include transmuted isotopes, fission products, actinides, and burnable poisons, as well as the fissile component of the fuel. Many of these codes have been used extensively by the nuclear industry, perform adequately, and give perfectly acceptable answers for their intended purpose. However, using a neutronics code such as MCNP to directly determine the neutron absorption rate for each individual nuclide as part of the neutronics calculation, thereby eliminating the need to include neutron flux information in the burnup calculation, is a more straightforward and accurate solution method. To the author's knowledge, applying this type of a methodology on a modern neutronics analysis code has never been accomplished. The goal of this work was to develop a ...

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