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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Analysis of an existing spent fuel pool storage rack using current practices and updated depletion bias and uncertainty values - 14458

In accordance with NRC Information Notice 2011-03: Nonconservative Criticality Safety Analysis for Fuel Storage, a US nuclear power plant reviewed their analysis of record (AOR) for possible nonconservative technical specifications related to the criticality safety of fuel storage. During this review, it was determined that some legacy fuel assemblies did not meet all the assumptions of the AOR, but did meet the plant Technical Specifications. The plant entered into an Operability Evaluation in order to evaluate the moderator temperature and boron concentration assumptions used in the AOR, as well as evaluate the impact of including additional biases and depletion uncertainties for the non-borated case. As this study was part of a response to the NRC Information Notice, a full assessment was not conducted. A re-analysis of the spent fuel rack in accordance with current practices was performed. This analysis used the bounding core conditions, and included all biases and depletion uncertainties. The results challenged the 10CFR50.68(b) requirement that 'the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with un-borated water'. Credits for additional reactivity offsets were then analyzed. Previous assumptions for the measurement uncertainty were evaluated and subsequently reduced. New values for the depletion bias and uncertainty were applied. More detailed, realistic modeling was used to remove conservatisms. The results showed that the storage racks met the 10CFR50.68(b) criteria for the un-borated case. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

ORIGAMI Automator Primer. Automated ORIGEN Source Terms and Spent Fuel Storage Pool Analysis

Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process data to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrations available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Criticality calculation package and its use for a spent fuel storage pool with steam layer - 14324

Criticality studies need several calculations with small variations in a number of parameters. Therefore, a calculation package written at VTT in order to facilitate sensitivity studies in criticality calculations was used. This package is based on MCNP5 [1], and consists of a data transfer and analysis program written in Perl, and a set of pseudo inputs for the transport and storage systems including the fuel assembly. The pseudo inputs contain key words that are replaced by a given value for the specific parameter. In this paper, the package is described in more detail including the new modifications needed for this study. The usability will be demonstrated with a study of the fuel pool in the case of steam gathered under the support plates of the pool. In the new pool model the steam layer was added as a water layer with a very low density. The density of this layer, i.e. the steam density, could be varied as well as the thickness of the layer. This package was then used to study the reactivity effect of the moderator density, the steam density and the thickness of the steam layer. As can be expected the strongest effect was obtained with the moderator density. However, the steam density seemed not to affect the reactivity while the thickness of the layer had a small effect. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Development of an Activation Analysis Methodology to Support the Disposal of the High Flux Isotope Reactor Original Reflector Container

The original beryllium reflector container was installed prior to the first ever High Flux Isotope Reactor (HFIR) cycle and was removed at the end of cycle 382. The original beryllium reflector was part of every HFIR irradiation cycle from September of 1966 to October of 2000. After its service lifetime, the beryllium reflector container has been interim stored at the fuel pool storage, which is located adjacent to the reactor. Due to space limitations in the pool fuel storage, it is necessary to remove the beryllium reflector container and send it to a location that can be stored permanently. Prior to shipping the beryllium reflector container, the radiological activity must be known. The challenge in determining the overall activity as well as the isotopic inventory and volumetric material fractions lays on the fact that some of the original reflector container stainless steel quantities are not known. Therefore, a methodology to determine a conservative activation scenario was developed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Accident analysis of spent fuel storage and transportation cask with Gd based neutron absorber

The shortage of nuclear spent fuel storage pool has been issued. The spent fuel storage and transportation cask is one of the major alternatives for storing spent fuel. The use of advanced neutron absorber is important for effective storage of spent fuel in spent fuel cask. Gadolinium based neutron absorber is chosen since it has high neutron absorption cross section in thermal neutron region compared to boron and it does not produce any gas element when it reacts with a neutron. The major requirement for neutron absorbing materials in the spent fuel cask system is to maintain subcriticality in normal, off-normal and accident conditions. In this study, the criticality calculations under accident conditions are conducted to ensure the safety of spent fuel cask. Major accident conditions are flood and cask tip-over. The full and partial flooded condition is analyzed and 100% flooded condition has the highest reactivity. The water density is selected as 1 g/cm{sup 3} because k{sub eff} is highest at this condition. When the cask tip-over, the flux trap reduces. In this case, a huge increase in reactivity of 800 pcm takes place. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Seismic Performance of Dry Casks Storage for Long- Term Exposure

The main goal of this study is to evaluate the long-term seismic performance of freestanding and anchored Dry Storage Casks (DSCs) using experimental tests on a shaking table, as well as comprehensive numerical evaluations that include the cask-pad-soil system. The study focuses on the dynamic performance of vertical DSCs, which can be designed as free-standing structures resting on a reinforced concrete foundation pad, or casks anchored to a foundation pad. The spent nuclear fuel (SNF) at nuclear power plants (NPPs) is initially stored in fuel-storage pools to control the fuel temperature. After several years, the fuel assemblies are transferred to DSCs at sites contiguous to the plant, known as Interim Spent Fuel Storage Installations (ISFSIs). The regulations for these storage systems (10 CFR 72) ensure adequate passive heat removal and radiation shielding during normal operations, off-normal events, and accident scenarios. The integrity of the DSCs is important, even if the overpack does not breach, because eventually the spent fuel-rods need to be shipped either to a reprocessing plant or a repository. DSCs have been considered as a temporary storage solution, and usually are licensed for 20 years, although they can be relicensed for operating periods of up to 60 years. In recent years, DSCs have been reevaluated as a potential mid-term solution, in which the operating period may be extended for up to 300 years. At the same time, recent seismic events have underlined the significant risks DSCs are exposed. The consideration of DCSs for storing spent fuel for hundreds of years has created new challenges. In the case of seismic hazard, longer-term operating periods not only lead to larger horizontal accelerations, but also increase the relative effect of vertical accelerations that usually are disregarded for smaller seismic events. These larger seismic demands could lead to casks sliding and tipping over, impacting the concrete pad or adjacent casks. The casks may also slide and collide with other casks or structural components. Also, the different DSC components may impact each other during these events. This study provides a comprehensive evaluation of DSCs subjected to these extreme demands, including the effect of vertical accelerations, and soilstructure interaction.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Combined Raman imaging and {sup 18}O tracer analysis for the study of high temperature oxidation of pre-oxidized Zircaloy-4 in spent fuel pool accident - 6.4 STP-2016-0037

Oxidation at high temperature of Zr-based cladding materials is expected to be the primary cause of the fuel assemblies' degradation in Spent Fuel Storage Pool loss of cooling accidents. Differently from Loss Of Coolant Accident (LOCA) in reactor vessel, presence of air in the atmosphere in case of a spent fuel pool accident is known to be an aggravating factor because of the 'catalytic' role of nitrogen on the oxidation. 'Low temperature' oxidation scales simulating corrosion oxide formed during normal operation have been observed to have a protective effect regarding subsequent high temperature oxidation, in steam as well as in air and air + steam atmospheres. However, in the presence of nitrogen, this protective effect is lost much earlier than in pure steam or pure oxygen. We present results on oxygen transport through such corrosion oxide layers studied using an {sup 18}O tracer technique. Experiments were performed in {sup 18}O{sub 2} alone as well as in mixed {sup 18}O{sub 2} + N{sub 2} atmospheres, at 850 deg. C, a temperature for which nitrogen is known to have a strong effect on the Zr-based alloys oxidation. Post-test investigations of the specimens were performed by means of micro-Raman profiling or imaging, giving simultaneous information on the {sup 18}O distribution and on structural factors such as composition and stresses in the scales. Raman imaging gave clear evidence for different characteristic distributions of {sup 18}O in the scales. Some of them have been correlated with the development of cracks and porosity in the oxide which allows oxygen to locally penetrate in the scales. Although nitrogen appears to have no or limited influence on the oxygen diffusion, it is observed to reach the metal/oxide interface much faster than oxygen. A mechanism explaining the role of nitrogen on the loss of the protective effect and the kinetic acceleration observed by TGA is proposed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Combined Raman imaging and {sup 18}O tracer analysis for the study of zircaloy-4 high-temperature oxidation in spent fuel pool accident - 2016-0037

Oxidation at high temperature (HT) of cladding materials is expected to be the primary cause of the fuel assemblies degradation in spent fuel storage pool loss of cooling accidents. Unlike a loss-of-coolant accident (LOCA) in a reactor vessel, the presence of air in the atmosphere is expected in a spent fuel pool accident and is known to be an aggravating factor because of the 'catalytic' role of nitrogen on the oxidation. In steam LOCA conditions, a low-temperature (LT) oxidation scale simulating the corrosion oxide formed during in-service use of the assemblies has been observed to have a protective effect regarding subsequent HT oxidation. In the presence of nitrogen, HT oxidation kinetic experiments show that the protective effect of an LT corrosion scale also exists but is lost much earlier than in pure steam. Oxygen transport through LT oxide layers has been studied using the {sup 18}O tracer technique. We performed experiments in {sup 18}O{sub 2} alone as well as in mixed {sup 18}O{sub 2}+N{sub 2} atmospheres, at 850 deg. C. Micro-Raman imaging, both at the specimen surfaces and on cross sections gave clear evidence for different characteristic distributions of {sup 18}O in the scales. Some of these distributions have been correlated with the presence of cracks and porosity in the LT oxide, which allows oxygen to locally penetrate in the scales. In cases in which no radial crack is present, apparent oxygen diffusion coefficients in the oxide were derived from fitting {sup 18}O diffusion profiles. Nitrogen appears to have no or only a limited influence on the oxygen diffusivity, but it is observed to reach the metal-oxide interface faster than oxygen, at which point it reacts with the metal and the sub-stoichiometric oxide to form zirconium nitride.The later conversion of the ZrN into ZrO{sub 2} might lead to the destabilization of the LT scale and to the loss of its protective effect. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Computed Tomography of Dry Cask Storage for Used Fuel

The demand for dry cask storage for used nuclear fuel is on the rise due to the continuous operation of currently existing nuclear plants which are reaching or have reached the capacity of their used fuel pools. Dry cask storage allows used fuel that has already been cooled in the used fuel pool for at least five years to be surrounded by an inert gas inside a specially designed and licensed cask. Currently, 75 of the 78 U.S.A reactor sites either have or are pursuing an independent spent fuel storage installation (ISFSI) for dry cask fuel storage with over 2,000 casks already loaded. Cask loading will only continue to increase until a final repository solution is operational or the U.S. begins to recycle fuel. The challenge of monitoring the condition of the used fuel inside the cask without exposing the used fuel has been apparent since the 1980's. The objective of the University of Florida (UF) project is ultimately to investigate the feasibility of imaging the used fuel assemblies' location and evaluating the inner structural condition using emission source computed tomography (CT). A combination of radiation transport models will be used to simulate gamma-rays and neutrons traversing through the used fuel, structure, and cask body materials. In the simulations, multiple linear detector arrays will record the radiation outside the cask wall and will be placed at various angles in order to model rotating a linear detector array around the entire cask. The combination of data from all the detectors will give the planar radiation projection onto the detectors, more commonly called the Radon transform, of the structure of used fuel assemblies. This project can be processed with filtered back projection or other more advanced reconstructive imaging algorithms to recreate the fuel structures. A majority of the historical work done for used fuel has been on verification, diversion detection, and burnup credit validation in order to meet regulatory guidelines. One of the more commonly used systems for single assembly verification is the Fork Detector developed at Los Alamos National Laboratory in 1988. A group at the Idaho National Laboratory developed a Compton Dry-Cask Imaging Scanner that has the ability to detect empty vs full storage positions but it suffered a setback when transported to the Doel Nuclear Power Plant near Antwerp, Belgium; the system was unable to detect full energy peaks because of the additional scattering caused by a thick ballistic steel shield. As detectors, computational power, and modeling improves, a few groups have leveraged these advancements and started to investigate CT hoping, that despite the difficulties, the technique can give information on fuel condition. So far the groups have mostly limited the tomographic work to a per assembly basis , rather than the full dry cask as the University of Florida plans to investigate. The downsides of CT versus simpler verification and validation are the amount of data required and the time it takes to acquire that data. Given that dry casks are intended to sit undistributed for up to 60 years, a scanning time of several weeks to several months is not unreasonable if such systems do not require human operators beyond the initial setup. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Emergency response and solutions for damaged fuel assembly handling events - Lessons learned from a 2013 fuel positioning issue

In 2013, during the refueling of a French 1300 MWe reactor, a fuel assembly (FX3599) was damaged by a fuel assembly positioning tool in the reactor vessel. The upper part of a corner rod of the fuel assembly was twisted and the top end grid was torn off. Because of the protruding fuel rod, the assembly could not be easily transferred for repair to the fuel storage pool via the transfer tube. This unexpected and rarely encountered situation was assessed jointly by EDF and AREVA NP. The chosen option was to directly perform the field service inside the reactor building in cells generally used for intermediate storage during refueling. This operation, performed by EDF and AREVA, involved removing the twisted rod and positioning it in a special-purpose canister by means of a gripper-type tool for removal (KIPCI gripper) and a gripper for holding the rod in place. The special-purpose canister was designed, ordered and supplied by a sub-supplier within one week. The operation went through 3 phases. First a preparation step in the fuel building: The stripper was disassembled and technical checks carried out to ensure correct operation. The gripper tightening force was checked in real-life conditions with the gripper suspended from auxiliary crane. The fuel rod removal was simulated several times by trained operators. Second, the equipment was installed in the reactor building. And third, the operation took place in the reactor cavity: the special-purpose canister was opened, the radioactive waste canister placed in the special-purpose canister, the twisted fuel rod was then grasped by the KIPCI gripper and removed from the fuel assembly. The twisted fuel rod was positioned above the canister and dropped into it. Once the fuel rod had been removed, the canister and fuel assembly FX3599 were transferred to the fuel building and refueling resumed. The canister containing the fuel rod is currently in interim storage in the reactor pool until appropriate action is taken for it to be shipped and eventually reprocessed. EDF has performed improvement in its handling procedures and in its handling tools to avoid a similar new event.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Probabilistic Risk Assessment of Spent Fuel Pool Cooling System Failure Considering Safety and Security Initiating Events

In order to quantify the probable risk associated with a safety system, nuclear safety risk evaluation process is applied which is generally performed in three steps. The first step is to calculate the damage frequency of a critical system. This step is followed by the radioactive source term estimation due to the potential system damage. The final step is the assessment of off-site consequences. The safety risk formula is shown by Eq. (1). P{sub F} is the probability of safety system failure (the probability used describes a random, independent event), and C is the consequence value. R{sub SAF} = P{sub F} * C (1) The security risk associated with an attack on critical infrastructure and/or special nuclear materials theft is usually evaluated using three- term risk equation (probability of attack, probability of adversary success, and consequence). A security threat and the related initiating events are not considered in a pure safety risk analysis. The vulnerability of the physical protection system is performed independently even though the consequence resulting from a security initiating event can be similar to a safety related initiating event and thus gets double counted. The security risk can be calculated using Eq. (2) R{sub SEC} = P{sub A} * P{sub S} * C (2) PA is the probability of attack, P{sub S} is the probability of adversary success (P{sub S} = 1 - P{sub I} P{sub N}). P{sub I} being the probability of interruption which accounts for the probabilities of detection of all timely detection elements along an adversary path. PN being the probability of neutralization: the result of engagement of the response force with the adversaries after the interruption has occurred. C is the consequence magnitude from a successful attack. The two security probabilities do not describe random, independent events. As a result, because of the differences in the randomness and independence of the respective safety and security probabilities, the safety and the security risks are currently calculated separately. The consequence term in nuclear safety and security risk evaluation have the same definition. An effort is made here to study the effect on the overall risk by considering security initiating events along with the subsequent failure of the intermediate safety element. The target selected is a spent fuel storage pool for analyzing the public risk arising from the interplay between nuclear security and safety initiating events. EASI (Estimate of Adversary Sequence Interruption) model is used to calculate the probability of interruption related to a security system that leads to the calculation of the security initiating event frequency. Further progression of this initiating security event to a top event is analyzed using a Probabilistic Safety Assessment (PSA) tool, SAPHIRE (version 8) to obtain a joint pool damage frequency. For the progression of events leading to a top event, the failure of the intermediate safety elements is assumed to be external random events such as heavy load drop and fire events. The objectives of this study is to (1) determine the double accounting of frequency of failure, risk and allocated resources in the scenario of a non-random security initiating event leading to a safety random event chain (2) compare the damage frequency from pure security with the frequency of combined security-safety failure event and pure safety failure events. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Radiological Surveys of 1600 km of In-Use Roads in Ichinoseki City, Japan Following the Fukushima Dai-ichi Reactor Accident - 17467

In March 2011, the Japanese Fukushima Dai-ichi nuclear reactor was damaged by a 9.0 magnitude earth quake and subsequent 15-meter tsunami. As a result of the nuclear reactor accident, radioactive material from the nuclear reactor and fuel storage pool (primarily radioactive iodine and cesium) was released to the atmosphere and carried by the wind as a plume that spread radioactive contamination across roughly 13,000 square kilometers of the countryside. Amec Foster Wheeler's Radiological Services and Engineering Group mobilized a team of Engineers and Operators to Japan to implement a radiological survey (scan) of roads located in Ichinoseki City, Iwate Prefecture, north of Fukushima, using three of Amec Foster Wheeler's Orion ScanPlotSM advanced radiological scanning systems. The survey was performed in less than two months while the roads were kept open for public use, and was conducted with motorized scooters and battery-assisted bicycles. The survey was performed with operational support from Obayashi, Inc., a large Japanese construction company headquartered in Tokyo. Over a seven-week period, the ScanPlotSM systems covered approximately 1,600 kilometers of roads within Ichinoseki City and surrounding areas of Iwate Prefecture. ScanPlotSM systems recorded GPS locations and radiation dose rate levels from Cs-137 on the road's surface at one-second increments during the survey and displayed the dose rate reading in micro-Sieverts per hour (μSv/h) on the back of the bicycle or motor scooter in real-time on a results monitor so that members of the public would be able to see real-time data. Radiological survey data was provided to Obayashi daily for development of survey result maps for the city. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Acoustic sensors for fission gas characterization: R and D skills devoted to innovative instrumentation in MTR, non-destructive devices in hot lab facilities and specific transducers for measurements of LWR rods in nuclear plants

First of all, we will present the main principle of the method. A piezoelectric transducer, driven by a pulse generator, generates the acoustic waves in a cavity that may be the fuel rod or a chamber connected to an instrumented rod. The composition determination consists in measuring the time of flight of the acoustic signal emitted. The pressure can be estimated by a calibration process, above the measurement of the amplitude of the signal. Two projects will then be detailed. The first project consists in the development of advanced instrumentation for in-pile experiments in Material Testing Reactor. It constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). As a first step of the development program, we performed in-pile tests on the most sensitive component, i.e., the piezoelectric transducer. For this purpose, the active part of this sensor has been qualified on gamma and neutron radiations and at high temperature. Various industrial piezo-ceramics were exposed to a high activity Cobalt source for few days. The cumulated dose was ranged from 50 kGy up to 2 MGy. Next, these devices were placed inside a Material Test Reactor to investigate their reliability towards neutron fluence. The final fluence after 150 days of irradiation was up to 1.6.10{sup 21}n/cm{sup 2} (for thermal neutron). Irreversible variations have been measured. Next, a specific sensor has been implemented on an instrumented fuel rod and tested in the frame of a REMORA 3 Irradiation test. It was the first experiment under high mixed, temperature neutron and gamma flux. A first irradiation phase took place in March 2010 in the OSIRIS reactor and in November 2010 for the second step of the irradiation. The instrumented fuel rod incorporating the ultrasonic gas composition sensor was finally irradiated during 2 weeks in nominal conditions. Neutronics calculation will be performed in order to calculate the thermal and fast neutron fluence and the gamma dose absorbed by acoustic sensor. A first evaluation gives a thermal fluence about 4,5.10{sup 19} n/cm{sup 2}, a fast neutrons fluence about 4,5.1018 n/cm{sup 2} and a total gamma dose up to 0,25 MGy The maximal temperature during the irradiation test was about 150 C. Although the ultrasonic sensor appears to be damaged, the optimization of the electrical attack parameters and the development of a new signal processing maintain the measurement feasibility up the end of the irradiation campaign. It was the first time that the composition of fission gas has been monitored all along an irradiation experiment in a MTR, giving access to the gas release kinetics. New researches involve thick film transducers produced by screen-printing process in order to propose piezoelectric structures for harsh temperature and irradiation measurements. The second project consists in the development of a non-destructive device that can be directly applied on a LWR fuel rod. The problem to be solved relates to the measurement of the fission gas pressure and composition in a fuel rod using a non-destructive method. Fuel rod internal pressure is one of the safety criteria applied in nuclear power analyses. This criterion must be verified in order to avoid any fuel-cladding gap reopening risk and therefore any local clad ballooning. Apart from the safety implications, this parameter is also a fuel behaviour indicator and reflects the overall fuel performance in operation, but also during shipping and long-term storage. Rod internal pressure is one criterion amongst others, like cladding corrosion, against which the acceptable fuel burn-up limit is set. A sensor has been achieved in 2007. A full-scale hot cell test of the internal gas pressure and composition measurement by an acoustic sensor was conducted successfully between 2008 and 2010 on 5 high burn-up MOX fuel rods and 2 very high burn-up UO{sub 2} fuel rods in LECA Facility at Cadarache Centre. An improvement of this sensor has been proposed, allowing us to divide by two the uncertainty on the pressure measurement. In the case of hot-cell measurements, viscous liquid can be used to couple the sensor with the rod. For gas content with a pressure exceeding 15 bars and a 10% Xe/Kr ratio, such coupling may reduce relative acoustic method accuracy by ±7% for pressure measurement result and ±0.25 % for the assessment of gas composition. These results make it possible to demonstrate the feasibility of the technique on LWR fuel rods. The transducer and the associated methodology are now operational for non-destructive measurements in hot lab facilities and allow characterising the fission gas without puncturing the fuel rods. Up to now, any other non-destructive method can be proposed. A next step will be the development of an industrial application in a fuel storage pool in order to perform a large number of measurements on a fuel assembly in nuclear plants.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Overview of zion comparative analysis project for assessment of Boral{sup R} neutron absorber material performance and monitoring in spent fuel pools - 14618

Neutron absorber materials are used in spent fuel pool (SFP) storage racks to control reactivity. In the United States, about half of the operating plants currently use BORAL{sup R} neutron absorber material in SFP racks. Although operational experience spanning more than 25 years has not shown any loss of the BORAL{sup R} core material, corrosion effects, including blisters and pits, have been observed and reported. To improve understanding for the behavior of BORAL{sup R} in SFPs and evaluate the performance of monitoring approaches, EPRI initiated the Zion comparative analysis project in 2014. The Zion project involves three phases. During the first phase of the project, in situ measurements were performed on selected SFP rack panels in December 2014. Panels for in situ measurements were selected after an extensive analysis of each panel in the pool to ensure that the selected panels represented the pool in terms of variation in neutron and gamma dose values. As part of the second phase of the project, the remaining 10 coupons from the Zion SFP were shipped to a Penn State University (PSU) laboratory for analysis in January 2015. During the final phase of the project, a subset of the panels for which in situ measurements were performed will be harvested from the Zion SFP, cut into samples, and analyzed at PSU. The overall scope of this project will allow comparison of in situ measurements to real panel data as well as data obtained from coupon analysis. In this paper, an overview of the Zion comparative analysis project, the criteria and procedure for panel selection for in situ measurements and harvesting, preliminary coupon analysis results and in situ measurement results are presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Evaluation of an unexpected reflector temperature effect for water-isolated array configurations - 14500

It has been conventional wisdom in the criticality safety discipline that for spent fuel systems like water pit storage and transportation, modeling water at its peak density will maximize reactivity. Thus, modeling water at 3.96 deg. C will produce a conservatively bounding k{sub eff} for that configuration. This assumption has been found to be incorrect for certain high-leakage water-isolated array configurations. In fact, temperatures up to the boiling point of water can in fact be more reactive, and the most reactive temperature is highly dependent upon the specific fuel configuration being analyzed. In light of the accident at Fukushima-Daiichi - where the spent fuel pool temperature increased - knowledge of this unexpected behavior is essential to ensuring that criticality safety evaluations are conservatively bounding. To illustrate the reflector temperature effect, MC21 calculations are provided for a sensitivity study based on LEU-COMP-THERM-044, which models the IPEN/MB-01 reactor at various temperatures between 3.9 deg. C and 93.3 deg. C. (authors)

42 ENGINEERING↗
DOE OSTI2014

Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application of the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Potential applications of microbial genomics in nuclear non-proliferation

As nuclear technology evolves in response to increased demand for diversification and decarbonization of the energy sector, new and innovative approaches are needed to effectively identify and deter the proliferation of nuclear arms, while ensuring safe development of global nuclear energy resources. Preventing the use of nuclear material and technology for unsanctioned development of nuclear weapons has been a long-standing challenge for the International Atomic Energy Agency and signatories of the Treaty on the Non-Proliferation of Nuclear Weapons. Environmental swipe sampling has proven to be an effective technique for characterizing clandestine proliferation activities within and around known locations of nuclear facilities and sites. However, limited tools and techniques exist for detecting nuclear proliferation in unknown locations beyond the boundaries of declared nuclear fuel cycle facilities, representing a critical gap in non-proliferation safeguards. Microbiomes, defined as “characteristic communities of microorganisms” found in specific habitats with distinct physical and chemical properties, can provide valuable information about the conditions and activities occurring in the surrounding environment. Microorganisms are known to inhabit radionuclide-contaminated sites, spent nuclear fuel storage pools, and cooling systems of water-cooled nuclear reactors, where they can cause radionuclide migration and corrosion of critical structures. Microbial transformation of radionuclides is a well-established process that has been documented in numerous field and laboratory studies. These studies helped to identify key bacterial taxa and microbially-mediated processes that directly and indirectly control the transformation, mobility, and fate of radionuclides in the environment. Expanding on this work, other studies have used microbial genomics integrated with machine learning models to successfully monitor and predict the occurrence of heavy metals, radionuclides, and other process wastes in the environment, indicating the potential role of nuclear activities in shaping microbial community structure and function. Results of this previous body of work suggest fundamental geochemical-microbial interactions occurring at nuclear fuel cycle facilities could give rise to microbiomes that are characteristic of nuclear activities. These microbiomes could provide valuable information for monitoring nuclear fuel cycle facilities, planning environmental sampling campaigns, and developing biosensor technology for the detection of undisclosed fuel cycle activities and proliferation concerns.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2023

Preliminary Design of Critical Experiments Involving Commercially Available B 4 C Neutron Absorber Plates with Low-Enriched UO 2 Fuel

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is an initiative to provide high-quality benchmark data in a standardized format for criticality safety analysts to use to validate calculational tools and nuclear data. In the last released 2022 version, the ICSBEP handbook contains over 5,000 critical experiment descriptions, results, and associated models. This paper describes the critical configurations proposed and calculations performed as part of the conceptual design phase of Integral Experiment Request (IER) 554 (IER-554). The goal of IER-554 is to add to the benchmark experiments available to the criticality safety community by designing critical experiments that can analyze how adding commercially available Boralcan plates to an assembly of low-enriched UO2 fuel rods affects the effective neutron multiplication factor (k eff ). Boralcan, a neutron absorber product developed by Rio Tinto, is made of B4C in an Al 1100 matrix and is commonly used for criticality suppression in fuel storage pools. In the current version of the ICSBEP handbook, only a few dated critical experiments involve B4C materials, and none involve Boralcan in a thin plate shape. The experiments designed as part of IER-554 are intended to be performed at the Sandia Pulsed Reactor Facility/Critical Experiments (SPRF/CX) apparatus at Sandia National Laboratories. SPRF/CX is a well-characterized assembly considered trustworthy by the benchmarking community because of the numerous high-quality evaluations with very low experimental uncertainties included in the ICSBEP handbook. Figure 1 illustrates the assembly for a particular configuration from LCT-078, one of the published benchmarks in the ICSBEP handbook. The experiments will be moderated and reflected by light water at ambient atmospheric pressure. Before the plates are inserted in the critical assembly, they will be characterized with x-ray computed tomography (XCT) to identify the sizes and distribution of the B 4 C powder particles inside the plates. The preliminary design calculations were performed with SCALE 6.3.0 using the KENO V.a sequence for the criticality calculations, the CE-TSUNAMI-3D sequence for the sensitivity and uncertainty studies, and the ENDF/B-VII.1 Continuous Energy cross section library. All the calculation results presented have 10 pcm statistical uncertainties. A set of critical experiments with Boralcan plates could increase the confidence of the criticality safety community in its modeling methods when using this type of neutron absorber material. Experiments would also help validate k eff calculations, and the industry could use these validations to change the B loading credit limits from the US Nuclear Regulatory Commission standard review plan for dry cask storage of spent nuclear fuel. This paper summarizes only part of the analysis documented in the preliminary design report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗