Pressurized water reactor (PWR) burnup credit validation is
demonstrated using the benchmarks for quantifying fuel reactivity
decrements, published as Benchmarks for Quantifying Fuel Reactivity
Depletion Uncertainty, Electric Power Research Institute (EPRI)
report 1022909. This demonstration uses the depletion module
TRITON (Transport Rigor Implemented with Time-Dependent
Operation for Neutronic Depletion) available in the SCALE 6.1
(Standardized Computer Analyses for Licensing Evaluations) code

This paper provides insights into the neutronic similarities between a representative high-capacity rail-transport cask containing typical pressurized water reactor (PWR) spent nuclear fuel assemblies and critical reactor state-points, referred to as commercial reactor critical (CRC) state-points. Forty CRC state-points from five PWRs were analyzed, and the characteristics of CRC state-points that may be applicable for validation of burnup-credit criticality safety calculations for spent fuel transport/storage/disposal systems were identified.

In the 1980s a series of the Haut Taux de Combustion (HTC) critical experiments with fuel pins in a water-moderated lattice was conducted at the Apparatus B experimental facility in Valduc (Commissariat à l'Energie Atomique, France) with the support of the Institut de Radioprotection et de Sûreté Nucléaire and AREVA NC. Four series of experiments were designed to assess profit associated with actinide-only burnup credit in the criticality safety evaluation for fuel handling, pool storage, and spent-fuel cask conditions.

The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure
Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform
postclosure criticality calculations. The validation process applies the criticality analysis methodology
approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report.1
The application systems for this validation consist of waste packages containing transport, aging, and

In the 1980s, a series of critical experiments referred to as the Haut Taux de Combustion (HTC)
experiments was conducted by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) at the
experimental criticality facility in Valduc, France. The plutonium-to- uranium ratio and the isotopic
compositions of both the uranium and plutonium used in the simulated fuel rods were designed to be
similar to what would be found in a typical pressurized-water reactor fuel assembly that initially had an

The purpose of this study is to provide insights into the neutronic similarities that may exist between a
generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-
points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the
type of CRC state-points that may be applicable for validation of burnup credit criticality safety
calculations for spent fuel transport/storage/disposal systems are identified. The study employed cross-

Taking credit for the reduced reactivity of spent nuclear fuel (SNF) in criticality analyses is referred to as burnup credit (BUC). Criticality safety evaluations require validation of the computational methods with critical experiments that are as similar as possible to the safety analysis models, and for which the keff values are known. This poses a challenge for validation of BUC criticality analyses, as critical experiments with actinide and fission product (FP)

The American Nuclear Society (ANS) appreciates the opportunity to comment on the draft Nuclear Waste Administration Act (NWAA). The ANS is a not-for-profit, international, scientific, and educational organization with nearly 12,000 members worldwide. The core purpose of ANS is to promote awareness and understanding of the application of nuclear science and technology. As an organization, it has published a number of position statements regarding the issue of spent fuel and radioactive waste.

The American Nuclear Society (ANS) supports the safe, controlled, licensed, and regulated interim
storage of used nuclear fuel (UNF) (irradiated, spent fuel from a nuclear power reactor) until disposition
can be determined and completed. ANS supports the U.S. Nuclear Regulatory Commission’s (NRC’s)
determination that “spent fuel generated in any reactor can be stored safely and without significant
environmental impacts for at least 30 years beyond the licensed life for operation.

It is increasingly apparent that the United States will require a large expansion of nuclear power
generation capacity to meet its future baseload electricity needs while reducing greenhouse gas
emissions. As a result, Congress and the Administration must act to update U.S. nuclear fuel
cycle policy to address these realities. This will likely require a multifaceted approach involving
some combination of on-site/centralized dry cask interim storage, nuclear fuel recycling, and
emplacement of high-level wastes in long-term geological storage.

More than 45 million shipments of radioactive materials have taken place in the United States
over the last three decades, with a current rate of about three million per year. The majority of
these radioactive shipments consist of radiopharmaceuticals, luminous dials and indicators,
smoke detectors, contaminated clothing and equipment, and research and industrial sources.
Fewer than 3,500, or 0.01%, have been involved in any sort of accident, incident, or anything

The American Nuclear Society (ANS) supports (1) the development and use of geological
repositories for disposal of high-level radioactive wastes and (2) expeditious processing of the
Yucca Mountain license application in an open, technically sound manner. Geological disposal
means placing the wastes hundreds of feet underground and far from the biosphere. The U.S.
Nuclear Regulatory Commission (NRC) is following a legislatively well-defined regulatory
process to evaluate the safety of the proposed Yucca Mountain Site to meet both the scientific

Analytical methods, described in this report, are used to
systematically determine experimental fuel sub-batch
reactivities as a function of burnup. Fuel sub-batch reactivities
are inferred using more than 600 in-core pressurized water
reactor (PWR) flux maps taken during 44 cycles of operation
at the Catawba and McGuire nuclear power plants. The
analytical methods systematically search for fuel sub-batch
reactivities that minimize differences between measured and
computed reaction rates, using Studsvik Scandpower’s