MC/sup 2/-2: a code to calculate fast neutron spectra and multigroup cross sections. [LMFBR]

Henryson, H.; Toppel, B.J.; Stenberg, C.G.

doi:10.2172/7143331

Cited by 65 publications

(50 citation statements)

References 0 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Inelastic and (n,2n) scattering sources, fission sources at energies above the upper limit of the RABANL energy range are treated as external sources, being supplied from the SRATES file [3] which is obtained from the ultra-fine group calculation. Options are also available for inhomogeneous group-dependent sources, group-dependent buckling, and buckling search calculations.…”

Section: -Methodologiesmentioning

confidence: 99%

See 1 more Smart Citation

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

Yang¹,

Lee²

2008

View full text Add to dashboard Cite

Section: -Methodologiesmentioning

confidence: 99%

“…For future extension to thermal reactor applications, the methodologies of the CENTRM/PMC system [3,4] were also selected to be investigated. It is planed to use the ultra-fine group methodologies of MC 2 -2 [5] for the above resolved resonance energy range and the CENTRM methodologies for the thermal energy range.…”

Section: Introductionmentioning

confidence: 99%

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

Yang¹,

Lee²

2008

View full text Add to dashboard Cite

“…The Argonne National Laboratory fast reactor codes MC 2 -2, DIF3D and REBUS were used for the reactor physics and fuel cycle calculations [4,5,6]. The MC 2 -2 code was used to generate a 33 group cross section set for each driver fuel enrichment zone, the targets, reflectors and shields.…”

Section: Reactor Physics and Fuel Cycle Simulationmentioning

confidence: 99%

“…The MC 2 -2 code was used to generate a 33 group cross section set for each driver fuel enrichment zone, the targets, reflectors and shields. Starting with an ultra-fine group ENDF-V/B cross section library, MC 2 -2 creates a collapsed cross section set by performing a zero dimensional infinite dilution critical buckling search using the extended P1 method [4]. Using this collapsed cross section set, the DIF3D diffusion code was used to solve the multi-group steady state neutron diffusion equation using a hexagonal-z nodal coordinate system [5].…”

Section: Reactor Physics and Fuel Cycle Simulationmentioning

confidence: 99%

Evaluation of Heterogeneous Options: Effects of MgO versus UO2 Matrix Selection for Minor Actinide Targets in a Sodium Fast Reactor

Pope¹,

Bays²,

Ferrer³

2008

View full text Add to dashboard Cite

Date Kathryn McCarthy, Systems Analysis Campaign Director Date iii Executive SummaryThe primary focus of this work is a comparison between MgO and UO 2 as matrix material options for burning minor actinides (MA) in transmutation targets within a sodium cooled fast reactor. The heterogeneous sodium fast reactor considered in this analysis was adapted from the homogeneous Advanced Burner Reactor.[2] In a previous work by Idaho National Laboratory, MgO and UO 2 target matrix options were compared based on identical target geometry.[3] Thus, the TRU conversion ratio (CR) was allowed to change due to excess TRU breeding in the UO 2 matrix. In this work, the fuel volume fraction was changed in order to match the CR between the UO 2 and MgO cases at approximately 0.75. This was done in order to provide a comparison of the two matrix materials based on similar net TRU destruction rates. The number of target assemblies was kept the same in each heterogeneous case (48) and approximately the same cycle length was used since it was limited by the fluence in the outer driver zone. The cycle length was set by limiting each assembly in the core to 200 dpa during its life. Also, the mass ratio of MAs (Am+Cm+Cf+Bk) to Np+Pu in the external supply of TRU was held constant and equal to that found in LWR SNF. For the purpose of this study, neptunium is excluded from the definition of "minor actinide".The UO 2 matrix-based targets required a 30% higher MA loading in order to achieve the same net MA destruction rate as the targets with an inert MgO matrix. The larger MA loading for the UO 2 case is due to the fact that the UO 2 case has a smaller transmutation efficiency than the MgO case. The MA destruction efficiency (EMA) of the targets is defined as the absolute value of the MA mass in a target assembly at discharge minus the amount charged in a fresh assembly, divided by the amount in the fresh assembly. The transmutation efficiency is 33% and 43%, for the UO 2 and MgO case respectively. The reduced transmutation efficiency of UO 2 is caused by the breeding of TRU from U-238 during the target's 10 cycle irradiation. During the irradiation time, the successive neutron captures from this bred TRU, eventually produce and accumulate MAs within the target. Due to the introduction of this internal source of MAs in the fuel cycle of the UO 2 matrix, the MA charge rate must be increased in order to maintain the same net MA destruction rate as in the MgO case. This is achieved by increasing the MA concentration in the targets (target enrichment).Assuming that the two matrix materials release the same fraction of fission gasses produced, the higher MA actinide inventory present in the targets having UO 2 matrix leads to higher fission gas pressures in the target pins, since they will have more fissions occurring during their core residence time. In this analysis, the calculated gas pressures in the UO 2 matrix targets were higher still due to the lower fuel volume fraction in these targets. However, even if compared on a basis of equal sp...

show abstract

“…The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems [2][3][4][5][6][7][8]. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code [9].…”

Section: Introductionmentioning

confidence: 99%

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

Heidet¹,

Kim²,

Grandy³

2012

View full text Add to dashboard Cite

show abstract

MC/sup 2/-2: a code to calculate fast neutron spectra and multigroup cross sections. [LMFBR]

Cited by 65 publications

References 0 publications

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

Evaluation of Heterogeneous Options: Effects of MgO versus UO2 Matrix Selection for Minor Actinide Targets in a Sodium Fast Reactor

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

Contact Info

Product

Resources

About