The "Summary Report of Commercial Reactor Critical Analyses Perfonned for the Disposal Criticality Analysis Methodology" contains a summary of the commercial reactor critical (CRC) analyses used to support the validation of the criticality models for spent commercial light water reactor (LWR) fuel.

The objective of this analysis is to characterize a codisposal canister containing MIT or ORR fuel in the Five-Pack Defense High-Level Waste (5-DHLW) Waste Package (WP) to demonstrate concept viability related to use in the Mined Geologic Disposal System (MGDS) environment for the postclosure time frame. The purpose of this analysis is to investigate the disposal criticality and shielding issues for the DHLW WP and establish DHLW WP and codisposal canister compatibility with the MGDS, and to provide criticality and shielding evaluations for the preliminary DHLW WP design.

The objective of this calculation is to perform additional degraded mode criticality evaluations of the Department of Energy's (DOE) Fast Flux Test Facility (FFTF) Spent Nuclear Fuel (SNF) codisposed in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP). The scope of this calculation is limited to the most reactive degraded configurations of the codisposal WP with an almost intact Ident-69 container (breached and flooded but otherwise non-degraded) containing intact FFTF SNF pins.

The purpose of this calculation is to perform waste-form specific nuclear criticality safety calculations to aid in establishing criticality safety design criteria, and to identify design and process parameters that are potentially important to the criticality safety of Department of Energy (DOE) standardized Spent Nuclear Fuel (SNF) canisters.

The purpose of this document is to summarize the degraded waste package disposal criticality evaluations which were performed in fiscal years I995 and I996. These evaluations were described in detail in 4 previous documents (Refs. I through 4). The initial version of this summary has been described in the I996 Disposal Criticality Analysis Methodology Technical Report (Ref. 5). A topical report planned for 1998 will present the methodology in its final form for approval by the US Nuclear Regulatory Commission.

Isotopic characterization of spent fuel via depletion and decay calculations is necessary for
determination of source terms for subsequent system analyses involving heat transfer, radiation
shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality
safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and
decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in

The purpose of this document is to present the methodology to be used for development of the Subcritical Limit (SL) for post closure conditions for the Yucca Mountain repository. The SL is a value based on a set of benchmark criticality multiplier, keff> results that are outputs of the MCNP calculation method. This SL accounts for calculational biases and associated uncertainties resulting from the use of MCNP as the method of assessing kerr·

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

The purpose of this calculation is to evaluate the transient behavior and consequences of a worst- case criticality event involving intact pressurized water reactor (PWR) mixed-oxide (MOX) spent nuclear fuel (SNF) in a degraded basket configuration inside a 21 PWR waste package (WP). This calculation will provide information necessary for demonstrating that the consequences of a worst-case criticality event involving intact PWR MOX SNF are insignificant in their effect on the overall radioisotopic inventory and on the integrity of the repository.

This report summarizes the results of EPA's review of the AEC
draft environmental statement, "Management of Commercial High-Level
and Transuranium-Contaminated Radioactive Waste" (WASH-1539). The
means by which high-level and long-lived radioactive wastes are
managed constitutes one of the most important questions upon which
the public acceptability of nuclear power, with its social and economic
benefits, will be determined. While the generation of power by
nuclear means offers certain benefits from the environmental viewpoint,

The purpose of this design analysis is to determine the accuracy of the SAS2H module of SCALE 4.3 in predicting isotopic concentrations of spent fuel assemblies. The objective is to develop a methodology for modeling assemblies similar to those evaluated within this analysis and to establish the consistency of SAS2H predictions. The results of this analysis may then be applied to future depletion calculations using SAS2H in which no measurements are available.

The 7th Forum on Stakeholder Confidence (FSC) National Workshop and Community Visit was held on 7-9 April 2009 in Bar-le-Duc, France.

The Total System Model (TSM) is a planning tool that estimates the logistic and cost impacts of
various operational assumptions in accepting radioactive wastes. Waste forms currently tracked
are Commercial Spent Nuclear Fuel (CSNF), U.S. Department of Energy (DOE) spent nuclear
fuel (DOE SNF), and defense high-level (radioactive) waste (HLW). The TSM uses a TSM
Preprocessor (TSMPP) to generate the cask loads and target dates for shipments from waste
sites. The TSM then tracks these wastes from pickup at the waste sites until repository

With the Department of Energy’s (DOE) recent submittal of a license application to the U.S. Nuclear Regulatory Commission, the development of Yucca Mountain, Nevada, as a national nuclear waste repository moves one step closer to reality. An operating site at Yucca Mountain could receive between 38-106 rail shipments and 53-89 truck shipments annually over a period of 50 years.

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

The purpose of this calculation is to perform criticality evaluations for mixed oxide spent nuclear fuel in 12 and 21 pressurized water reactor waste packages for both intact and degraded configurations. The MOX assembly design considered in previous studies on Pu disposition in commercial reactors is based on the Westinghouse 17x17 Vantage 5 assembly (Ref. 7.2). Depletion analyses of four Pu enrichment and burnup(expressed as gigawatt days/metric ton heavy metal; GWd/MTHM)) combinations were performed in Ref. 7.4.

Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since
1990 under the contract with Science and Technology Agency of Japan entitled ‘Technical Development on
Criticality Safety Management for Spent LWR Fuels.’ Main purposes of this work are to obtain the
experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup
and criticality calculation codes. In this work three major experiments of exponential experiments for

This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors.

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to determine the viability of the UCF waste package concept with respect to criticality regulatory requirements in compliance with the goals of the Waste Package Implementation Plan5·1 for conceptual design. These design calculations are performed in sufficient detail to provide a comprehensive comparison base with other design alternatives.

This Memorandum analyzes issues related to the Standard Contract between the U.S. Department of Energy (“DOE”) and the “utilities.” Beginning with a discussion of specific provisions of the Standard Contract, this Memorandum then analyzes the status of lawsuits involving the Standard Contract, reviews issues related to on-site storage of spent fuel and HLW, and assesses the prospects for modifying the current waste-disposal regime through Federal legislation or amendments to the Standard Contract.

The purpose of the material degradation and release (MDR) model is to predict the fate of the waste package materials, specifically the retention or mobilization of the radionuclides and the neutron-absorbing material as a function of time after the breach of a waste package during the 10,000 years after repository closure. The output of this model is used directly to assess the potential for a criticality event inside the waste package due to the retention of the radionuclides combined with a loss of the neutron-absorbing material.

Confidence in the long-term safety of deep geological disposal, and the ways in which this
confidence can be obtained and communicated, are topics of great importance to the radioactive waste
management community.1
The technical aspects of confidence have been the subject of considerable debate, especially
the concept of model validation. It has, for example, been pointed out that it is impossible to describe
fully the evolution of an open system, such as a repository and its environment, that cannot be