User's guide for the REBUS-3 fuel cycle analysis capability

Toppel, B.J.

doi:10.2172/6068571

Cited by 120 publications

(92 citation statements)

References 5 publications

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“…However, the feasibility of the Th-LEU fuel option was also assessed as an alternative option. The density of 232 Th is taken to be 11.7 g/cm 3 and that of Th-TRU is taken to be 13.4 g/cm 3 , which corresponds to approximately 20wt% of TRU.…”

Section: Design Objectives and Requirementsmentioning

confidence: 99%

“…Due to the lack of data regarding Th-TRU metal fuel characteristics, its density is assumed to be 13.4 g/cm 3 which corresponds approximately to 20wt% TRU. Subsequently, the specific power density is increasing as LEU fuel is replaced with Th-TRU fuel.…”

Section: Figure 35 Th-tru Fueled Afr-100 Core Layout For a 4-batchesmentioning

confidence: 99%

“…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%

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

Heidet¹,

Kim²,

Grandy³

2012

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Section: Design Objectives and Requirementsmentioning

confidence: 99%

Section: Figure 35 Th-tru Fueled Afr-100 Core Layout For a 4-batchesmentioning

confidence: 99%

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

Heidet¹,

Kim²,

Grandy³

2012

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“…al. REBUS-3 was used to perform the core neutronics simulations, depletion, in-core fuel management and out-of-core fuel reprocessing operations pertinent to the ARR's closed fuel cycle [6,7,8]. These calculations were performed for each incremental cooling time from Table 2 for both Case 1 and Case 2 scenarios.…”

Section: Reactor Calculationsmentioning

confidence: 99%

Deep Burn Fuel Cycle Integration: Evaluation of Two-Tier Scenarios

Bays¹,

Zhang²,

Pope³

2009

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The use of a deep burn strategy using VHTRs (or DB-MHR), as a means of burning transuranics produced by LWRs, was compared to performing this task with LWR MOX. The spent DB-MHR fuel was recycled for ultimate final recycle in fast reactors (ARRs). This report summarizes the preliminary findings of the support ratio (in terms of MWth installed) between LWRs, DB-MHRs and ARRs in an equilibrium "two-tier" fuel cycle scenario. Values from literature were used to represent the LWR and DB-MHR isotopic compositions. A reactor physics simulation of the ARR was analyzed to determine the effect that the DB-MHR spent fuel cooling time on the ARR transuranic consumption rate.These results suggest that the cooling time has some but not a significant impact on the ARRs conversion ratio and transuranic consumption rate. This is attributed to fissile worth being derived from non-fissile or "threshold-fissioning" isotopes in the ARR's fast spectrum.The fraction of installed thermal capacity of each reactor in the DB-MHR 2-tier fuel cycle was compared with that of an equivalent MOX 2-tier fuel cycle, assuming fuel supply and demand are in equilibrium. The use of DB-MHRs in the 1 st -tier allows for a 10% increase in the fraction of fleet installed capacity of UO2-fueled LWRs compared to using a MOX 1 st -tier. Also, it was found that because the DB-MHR derives more power per unit mass of transuranics charged to the fresh fuel, the "front-end" reprocessing demand is less than MOX. Therefore, more fleet installed capacity of DB-MHR would be required to support a given fleet of UO2 LWRs than would be required of MOX plants. However, the transuranic deep burn achieved by DB-MHRs reduces the number of fast reactors in the 2 nd -tier to support the DB-MHRs "back-end" transuranic output than if MOX plants were used.

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“…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. 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 [7]. In the nodal discretization, each hexagonal node in the lateral direction represents an assembly.…”

Section: Fast Reactor Calculationsmentioning

confidence: 99%

Summary Report on New Transmutation Analysis for the Evaluation of Homogeneous and Heterogeneous Options in Fast Reactors

Ferrer¹,

Bays²,

Pope³

et al. 2008

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User's guide for the REBUS-3 fuel cycle analysis capability

Abstract: LIST OF TABLES 1. Example of Priority System for Selection of Fuel Charge Batches. 11 2. BCD Input Datasets for REBUS-3 J.4 3. REBUS-3 BCD Input Error Checking. .. , 4. CPU Execution Times for Sample Problem on the IBM 3033 48 5. Modules Invoked by Standard Path Driver STP027 50 6.

Cited by 120 publications

References 5 publications

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

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

Deep Burn Fuel Cycle Integration: Evaluation of Two-Tier Scenarios

Summary Report on New Transmutation Analysis for the Evaluation of Homogeneous and Heterogeneous Options in Fast Reactors

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