Decay Heat Code Validation Activities at ORNL: Supporting Expansion of NRC Regulatory Guide 3.54

Oak Ridge National Laboratory (ORNL) has a long history of involvement in the development and validation of the ORIGEN series of isotope summation codes and nuclear data libraries, widely recognized and used to predict the decay heat for spent nuclear fuel. In particular, the ORIGEN-S code, the depletion/decay module of the SCALE code system, has been extensively validated using experimental isotopic assay data and decay heat measurements for commercial spent fuel. This work was used in the development of the technical basis for NRC Regulatory Guide 3.54 on spent fuel decay heat. The bulk of the experimental data used to validate spent fuel decay heat predictions are from programs of the 1970s and 1980s and consequently involve older-design fuel assemblies with a relatively low enrichment and burnup. This has led to a situation where the spent fuel now being discharged from operating reactors extends well beyond the regime of the experimental data and area of code applicability based o n the data. The absence of validation data for modern fuel designs has potentially serious consequences for decay heat predictions in terms of added safety factors to account for larger uncertainties and lower volumetric transport and storage capacities

Strategies for Application of Isotopic Uncertainties in Burnup Credit

Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103}Rh have also been included

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

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 determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k{sub eff}) to determine the net importance of cross sections to k{sub eff}. The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: {sup 151}Sm, {sup 103}Rh, {sup 155}Eu, {sup 150}Sm, {sup 152}Sm, {sup 153}Eu, {sup 154}Eu, and {sup 143}Nd

The impact of personalised risk information compared to a positive/negative result on informed choice and intention to undergo colonoscopy following colorectal Cancer screening in Scotland (PERICCS) - a randomised controlled trial:study protocol

Background In Scotland a new, easier to complete bowel screening test, the Faecal Immunochemical Test (FIT), has been introduced. This test gives more accurate information about an individual’s risk of having colorectal cancer (CRC), based on their age and gender, and could lead to fewer missed cancers compared to the current screening test. However, there is no evidence of the effect on colonoscopy uptake of providing individuals with personalised risk information following a positive FIT test. The objectives of the study are: 1) To develop novel methods of presenting personalised risk information in an easy-to-understand format using infographics with involvement of members of the public 2) To assess the impact of different presentations of risk information on informed choice and intention to take up an offer of colonoscopy after FIT 3) To assess participants’ responses to receiving personal risk information (knowledge, attitudes to screening/risk, emotional responses including anxiety). Methods Adults (age range 50–74) registered on the Scottish Bowel Screening database will be invited by letter to take part. Consenting participants will be randomised to one of three groups to receive hypothetical information about their risk of cancer, based on age, gender and faecal haemoglobin concentration: 1) personalised risk information in numeric form (e.g. 1 in 100) with use of infographics, 2) personalised information described as ‘highest’, ‘moderate’ or ‘lowest’ risk with use of infographics, and 3) as a ‘positive’ test result, as is current practice. Groups will be compared on informed choice, intention to have a colonoscopy, and satisfaction with their decision. Follow-up semi-structured qualitative interviews will be conducted, by telephone, with a small number of consenting participants (n = 10 per group) to explore the acceptability/readability and any potential negative impact of the risk information, participants’ understanding of risk factors, attitudes to the different scenarios, and reasons for reported intentions. Discussion Proving personalised risk information and allowing patient choice could lead to improved detection of CRC and increase patient satisfaction by facilitating informed choice over when/whether to undergo further invasive screening. However, we need to determine whether/how informed choice can be achieved and assess the potential impact on the colonoscopy service

Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel

The impact of hypothetical PErsonalised Risk Information on informed choice and intention to undergo Colorectal Cancer screening colonoscopy in Scotland (PERICCS)—a randomised controlled trial

Background There is currently no existing evidence on the effects of personalised risk information on uptake of colonoscopy following first line screening for colorectal cancer. This study aimed to measure the impact of providing risk information based on faecal haemoglobin concentration to allow a fully informed choice around whether or not to undergo colonoscopy. Methods Two thousand seven hundred sixty-seven participants from the Scottish Bowel Screening Programme (SBoSP) database, who had not recently been invited for screening, were randomised to receive one of three types of hypothetical risk information materials: (1) numerical risk information (risk categories of one in 40, one in 1600 and one in 3500), (2) categorical risk information (highest, moderate and lowest risk), or (3) positive screening result letter (control group). The primary outcome was the impact of the risk materials on intention to undergo colonoscopy, to allow comparison with the current colonoscopy uptake of 77% for those with a positive screening result in the SBoSP. Secondary outcomes were knowledge, attitudes and emotional responses to the materials. Results Four hundred thirty-four (15.7%) agreed to participate with 100 from the numerical risk group (69.0%), 104 from the categorical risk group (72.2%) and 104 from the control group (71.7%) returning completed materials. Intention to undergo colonoscopy was highest in the highest risk groups for the numerical and categorical study arms (96.8% and 95.3%, respectively), but even in the lowest risk groups was > 50% (58.1% and 60.7%, respectively). Adequate knowledge of colorectal screening and the risks and benefits of colonoscopy was found in ≥ 98% of participants in all three arms. All participants reported that they found the information easy-to-understand. 19.1%, 24.0% and 29.6% of those in the numerical, categorical and control group, respectively, reported that they found the information distressing (p > 0.05). Conclusions Applying the risk categories to existing SBoSP data shows that if all participants were offered an informed choice to have colonoscopy, over two thirds of participants would intend to have the test. Equating to an increase in the number of screening colonoscopies from approx. 14,000 to 400,000 per annum, this would place an unmanageable demand on colonoscopy services, with a very small proportion of cancers and pre-cancers detected. However, the response to the materials were very positive, suggesting that providing risk information to those in lowest and moderate risk groups along with advice that colonoscopy is not currently recommended may be an option. Future research would be required to examine actual uptake

OrigenArp Primer: How to Perform Isotopic Depletion and Decay Calculations with SCALE/ORIGEN

The SCALE (Standardized Computer Analyses for Licensing Evaluation) computer software system developed at Oak Ridge National Laboratory is widely used and accepted around the world for nuclear analyses. ORIGEN-ARP is a SCALE isotopic depletion and decay analysis sequence used to perform point-depletion calculations with the well-known ORIGEN-S code using problem-dependent cross sections. Problem-dependent cross-section libraries are generated using the ARP (Automatic Rapid Processing) module using an interpolation algorithm that operates on pre-generated libraries created for a range of fuel properties and operating conditions. Methods are provided in SCALE to generate these libraries using one-, two-, and three-dimensional transport codes. The interpolation of cross sections for uranium-based fuels may be performed for the variables burnup, enrichment, and water density. An option is also available to interpolate cross sections for mixed-oxide (MOX) fuels using the variables burnup, plutonium content, plutonium isotopic vector, and water moderator density. This primer is designed to help a new user understand and use ORIGEN-ARP with the OrigenArp Windows graphical user interface in SCALE. It assumes that the user has a college education in a technical field. There is no assumption of familiarity with nuclear depletion codes in general or with SCALE/ORIGEN-ARP in particular. The primer is based on SCALE 6 but should be applicable to earlier or later versions of SCALE. Information is included to help new users, along with several sample problems that walk the user through the different input forms and menus and illustrate the basic features. References to related documentation are provided. The primer provides a starting point for the nuclear analyst who uses SCALE/ORIGEN-ARP. Complete descriptions are provided in the SCALE documentation. Although the primer is self-contained, it is intended as a companion volume to the SCALE documentation. The SCALE Manual is provided on the SCALE installation DVD

Prompt atmospheric neutrino fluxes: perturbative QCD models and nuclear effects

We evaluate the prompt atmospheric neutrino flux at high energies using three different frameworks for calculating the heavy quark production cross section in QCD: NLO perturbative QCD, $k_T$ factorization including low-$x$ resummation, and the dipole model including parton saturation. We use QCD parameters, the value for the charm quark mass and the range for the factorization and renormalization scales that provide the best description of the total charm cross section measured at fixed target experiments, at RHIC and at LHC. Using these parameters we calculate differential cross sections for charm and bottom production and compare with the latest data on forward charm meson production from LHCb at $7$ TeV and at $13$ TeV, finding good agreement with the data. In addition, we investigate the role of nuclear shadowing by including nuclear parton distribution functions (PDF) for the target air nucleus using two different nuclear PDF schemes. Depending on the scheme used, we find the reduction of the flux due to nuclear effects varies from $10\%$ to $50 \%$ at the highest energies. Finally, we compare our results with the IceCube limit on the prompt neutrino flux, which is already providing valuable information about some of the QCD models.Comment: 61 pages, 25 figures, 11 table