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Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

Description: U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections dete… more
Date: August 12, 2005
Creator: Gauld, I. C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

Description: The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, includi… more
Date: April 25, 2005
Creator: Jardine, L J
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Monte-Carlo Code (MCNP) Modeling of the Advanced Test Reactor Applicable to the Mixed Oxide (MOX) Test Irradiation

Description: Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, and 40 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Dispo… more
Date: July 1, 2005
Creator: Chang, G. S. & Pederson, R. C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Advanced Fuel Cycle Initiative - Projected Linear Heat Generation Rate and Burnup Calculations

Description: This report provides documentation of the physics analysis performed to determine the linear heat generation rate (LHGR) and burnup calculations for the Advanced Fuel Cycle Initiative (AFCI) tests, AFC-1D, AFC-1H, and AFC-1G. The AFC-1D and AFC-1H tests consists of low-fertile metallic fuel compositions and the AFC-1G test consists of non-fertile and low-fertile nitride compositions. These tests will be irradiated in the East Flux Trap (EFT) positions E1, E2, and E3, respectively, during Advanc… more
Date: February 1, 2005
Creator: Ambrosek, Richard G.; Chang, Gray S. & Utterbeck, Debbie J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Verification of a Super Double-Heterogeneous Spherical Lattice Model for Equilibrium Fuel Cycle Analysis

Description: Advanced High Temperature gascooled Reactors (HTR) currently being developed (GFR, VHTR - Very High Temperature gas-cooled Reactor, PBMR, and GT-MHR) are able to achieve a simplification of safety through reliance on innovative features and passive systems. One of the innovative features in these HTRs is reliance on ceramiccoated fuel particles to retain the fission products even under extreme accident conditions. The effect of the random fuel kernel distribution in the HTR is addressed through… more
Date: June 1, 2005
Creator: Cahng, Gray S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Verify Super Double-Heterogeneous Spherical Lattice Model for Equilibrium Fuel Cycle Analysis AND HTR Spherical Super Lattice Model for Equilibrium Fuel Cycle Analysis

Description: The currently being developed advanced High Temperature gas-cooled Reactors (HTR) is able to achieve a simplification of safety through reliance on innovative features and passive systems. One of the innovative features in these HTRs is reliance on ceramic-coated fuel particles to retain the fission products even under extreme accident conditions. Traditionally, the effect of the random fuel kernel distribution in the fuel pebble / block is addressed through the use of the Dancoff correction fa… more
Date: November 1, 2005
Creator: Chang, Gray S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Commercial Spent Nuclear Fuel Waste Package Misload Analysis

Description: The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly… more
Date: July 28, 2005
Creator: Alsaed, A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

The Challenges Associated with High Burnup and High Temperature for UO2 TRISO-Coated Particle Fuel

Description: The fuel service conditions for the DOE Next Generation Nuclear Plant (NGNP) will be challenging. All major fuel related design parameters (burnup, temperature, fast neutron fluence, power density, particle packing fraction) exceed the values that were qualified in the successful German UO2 TRISO-coated particle fuel development program in the 1980s. While TRISO-coated particle fuel has been irradiated at NGNP relevant levels for two or three of the design parameters, no data exist for TRISO-co… more
Date: February 1, 2005
Creator: Petti, David & Maki, John
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis

Description: The UNIX BASH (Bourne Again Shell) script MCWO has been developed at the Idaho National Engineering and Environment Laboratory (INEEL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN2. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. MCWO can handle a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) powers, and irradiation time int… more
Date: April 1, 2005
Creator: Chang, Gray S
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Advanced Fuel Cycle Initiative AFC-1D, AFC-1G and AFC-1H Irradiation Report

Description: The U. S. Advanced Fuel Cycle Initiative (AFCI) seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby dramatically decreasing the volume of material requiring disposition and the long-term radiotoxity and heat load of high-level waste sent to a geologic repository. The AFC-1 irradiation experiments on transmutation fuels are expected to provide irradiation pe… more
Date: September 1, 2005
Creator: Utterbeck, Debra J. & Chang, Gray
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor

Description: A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is… more
Date: August 1, 2005
Creator: Chang, G. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Summary of Generation-IV transmutation impacts.

Description: An assessment of the potential role of Generation IV nuclear systems in an advanced fuel cycle has been performed. The Generation IV systems considered are the thermal-spectrum VHTR and SCWR, and the fast-spectrum GFR, LFR, and SFR. This report addresses the impact of each system on advanced fuel cycle goals, particularly related to waste management and resource utilization. The transmutation impact of each system was also assessed, along with variant designs for transuranics (TRU) burning. The… more
Date: August 3, 2005
Creator: Taiwo, T. A. & Hill, R. N.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Generalization of the FRAM's Bias

Description: The Fixed-Energy Response-Function Analysis with Multiple Efficiency (FRAM) code was developed at Los Alamos National Laboratory to measure the gamma-ray spectrometry of the isotopic composition of plutonium, uranium, and other actinides. Its reported uncertainties of the results come from the propagation of the statistics in the peak areas only. No systematic error components are included in the reported uncertainties. We have done several studies and found that the FRAM's statistical precisio… more
Date: October 1, 2005
Creator: Vo, Duc T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

A Feasibility Study of Reactor-Based Deep-Burn Concepts.

Description: A systematic assessment of the General Atomics (GA) proposed Deep-Burn concept based on the Modular Helium-Cooled Reactor design (DB-MHR) has been performed. Preliminary benchmarking of deterministic physics codes was done by comparing code results to those from MONTEBURNS (MCNP-ORIGEN) calculations. Detailed fuel cycle analyses were performed in order to provide an independent evaluation of the physics and transmutation performance of the one-pass and two-pass concepts. Key performance paramet… more
Date: September 16, 2005
Creator: Kim, T. K.; Taiwo, T. A.; Hill, R. N. & Yang, W. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Preliminary Neutronic Studies for the Liquid-Salt-Cooled Very High Temperature Reactor (Ls-Vhtr).

Description: Preliminary neutronic studies have been performed in order to provide guidelines to the design of a liquid-salt cooled Very High Temperature Reactor (LS-VHTR) using Li{sub 2}BeF{sub 4} (FLiBe) as coolant and a solid cylindrical core. The studies were done using the lattice codes (WIMS8 and DRAGON) and the linear reactivity model to estimate the core reactivity balance, fuel composition, discharge burnup, and reactivity coefficients. An evaluation of the lattice codes revealed that they give ver… more
Date: October 5, 2005
Creator: Kim, T. K.; Taiwo, T. A. & Yang, W. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Sensitivity and Uncertainty Analysis to Burn-up Estimates on ADS Using ACAB Code

Description: Within the scope of the Accelerator Driven System (ADS) concept for nuclear waste management applications, the burnup uncertainty estimates due to uncertainty in the activation cross sections (XSs) are important regarding both the safety and the efficiency of the waste burning process. We have applied both sensitivity analysis and Monte Carlo methodology to actinides burnup calculations in a lead-bismuth cooled subcritical ADS. The sensitivity analysis is used to identify the reaction XSs and t… more
Date: February 11, 2005
Creator: Cabellos, O; Sanz, J; Rodriguez, A; Gonzalez, E; Embid, M; Alvarez, F et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Nuclear Forensics Attributing the Source of Spent Fuel Used in an RDD Event

Description: An RDD attack against the U.S. is something America needs to prepare against. If such an event occurs the ability to quickly identify the source of the radiological material used in an RDD would aid investigators in identifying the perpetrators. Spent fuel is one of the most dangerous possible radiological sources for an RDD. In this work, a forensics methodology was developed and implemented to attribute spent fuel to a source reactor. The specific attributes determined are the spent fuel burn… more
Date: June 1, 2005
Creator: Scott, M.R.
Item Type: Refine your search to only Thesis or Dissertation
Partner: UNT Libraries Government Documents Department
open access

HTR Spherical Super Lattice Model for Equilibrium Fuel Cycle Analysis

Description: Advanced High Temperature gas-cooled Reactors (HTR) currently being developed (GFR, VHTR - Very High Temperature gas-cooled Reactor, PBMR, and GT-MHR) are able to achieve a simplification of safety through reliance on innovative features and passive systems. One of the innovative features in these HTRs is reliance on ceramic-coated fuel particles to retain the fission products even under extreme accident conditions. The effect of the random fuel kernel distribution in the fuel pebble / block is… more
Date: September 1, 2005
Creator: Cahng, Gray S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Preliminary Advanced Test Reactor LEU Fuel Conversion Feasibility Study

Description: The Advanced Test Reactor (ATR) is a high power density, high neutron flux research reactor operating in the United States. The ATR has large irradiation test volumes located in high flux areas. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. As a result, the ATR is a representative candidate for assessing the necessary modifications and evaluating the subsequent oper… more
Date: November 1, 2005
Creator: Chang, G. S. & Ambrosek, R. G.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)

Description: A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to coolant voiding. Three organizations participated in the study: Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), an… more
Date: December 15, 2005
Creator: Ingersoll, DT
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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