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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which has a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values. After the successful filter installing and a series of measurements, first test neutron radiography attempts with test samples could been carried out. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Advanced instrumentation and analysis methods for in-pile thermal and nuclear measurements: from out-of-pile studies to irradiation campaigns

Research and development on nuclear fuel behavior under irradiations and accelerated ageing of structure materials is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR) currently under construction in the South of France in the CEA Cadarache research centre will offer a real opportunity to perform R and D programs and hence will crucially contribute to the selection, optimization and qualification of innovative materials and fuels. To perform such programs advanced accurate and innovative experiments, irradiation devices that contain material and fuel samples are required to be set up inside or beside the reactor core. These experiments needs beforehand in situ and on line sophisticated measurements to accurately reach specific and determining parameters such as thermal and fast neutron fluxes, nuclear heating and temperature conditions to precisely monitor and control the conducted assays. Consequently, since 2009 CEA and Aix-Marseille University collaborate in order to design and develop a new multi-sensor device which will be dedicated to measuring profiles of such conditions inside the experimental channels of the JHR. These works are performed in the framework of two complementary joint research programs called MAHRI-BETHY and INCORE. These programs couple experimental studies carried out both out-of nuclear fluxes (in laboratory) and under irradiation conditions (in OSIRIS MTR reactor in France and MARIA MTR reactor in Poland) with numerical works realized by thermal simulations (CAST3M code) and Monte Carlo simulations (MCNP code). These programs deal with three main aims. The first one corresponds to the design and/or the test of new in-pile instrumentation. The second one concerns the development of advanced calibration procedures in particular in the case of one specific sensor: a differential calorimeter used to quantify nuclear heating. The last one consists in the development of accurate measurement and analysis methods. The paper will be dedicated to a complete review of the experimental and numerical works performed since 2009 thanks to two parts. The first part will detail a new thermal approach implemented to improve nuclear heating measurements by radiometric calorimeters. New experimental tools (calorimeter prototypes and set-ups such BETHY Bench) developed to perform preliminary out-of-pile studies under suitable conditions will be presented (temperature and velocity of the external cooling fluid, heat source localization and intensity inside the calorimetric cells). Then the response of two kinds of sensors, their calibrations curves and their thermal behaviors will be compared for various parameters. Finally validated numerical thermal and Monte Carlo works will be discussed to propose new improvements. The second parts of the paper will focus on works realized in order to design, develop and test the first prototype of the multi-sensor device called CARMEN [7-9]. The two mock-ups dedicated respectively to neutron measurements and photon measurements will be detailed. The results obtained during two irradiation campaigns inside the periphery of OSIRIS reactor will be shown. The new analysis method will be discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Performance of a Low Activity Beta-Sensitive SR{sup 90} Water Monitor for Fukushima

There are large volumes of contaminated water from the stabilization efforts at the damaged Fukushima Nuclear Power Plants. This water is being processed to remove radioactivity for eventual release to the environment. An on-line continuously operating system to confirm that the clean-up system is working properly, and to provide prompt feedback of the results is required by the system operator. While gamma emitting nuclides allow for the straight forward approach of gamma spectroscopy to identify and quantify radioactivity in water, pure beta emitting nuclides such as Sr{sup 90} pose a challenging problem. The relatively short range of beta radiation in water requires optimization of the measurement geometry in terms of the source-detector distance and source-detector interface while retaining a background sensitivity low enough to meet the Minimum Detectable Concentration (MDC) of 10 Bq/kg in 180 minutes. This issue is complicated by the continuum nature of the beta spectrum which does not allow for simple nuclide identification. The use of the Monte-Carlo code MCNP to estimate system performance before prototyping vastly increases the success of the end product. Various parameters such as detector size and thickness, water chamber size, water chamber construction materials were evaluated to help choose the optimum geometry. The final design was a system consisting of two large-area (16 x 35 cm) and thin (0.15 mm) plastic scintillators placed very close to a sealed a water chamber. The size of the chamber was optimized to obtain the maximum efficiency for the nuclide being measured (Sr/Y{sup 90}) but to minimize the efficiency for possible interferences (Ru/Rh{sup 106}, Cs{sup 137}). A thin carbon fiber window was selected with adequate material and thickness to contain the water under pressure, but also thin enough (0.5 mm) to allow enough beta radiation to pass through to the active detector volume. The entire measurement geometry is then housed in a thick lead shield to reduce contributions from external sources to an acceptable level. Data acquisition is accomplished through customized application-specific software that allows for long counting times to attain a low MDC, but also simultaneously provides alarms on short averaging times to achieve a fast response to sudden changes in activity concentration. Multiple monitors are then linked to supervisory software where real time data and alarms are available for analysis in remote locations. The system also allows for remote operation of the unit; check sources, background checks, systems settings and more may be accessed remotely. Testing of the production devices has shown that we can achieve the 10 Bk/kg MDC requirement for Sr{sup 90} in equilibrium with Y{sup 90} with a count time of approximately 20 minutes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Performance Evaluation of an Intelligent Sensor Platform for Radiation Monitoring Applications

Accurate, rugged, and reliable radiation detection systems are important for area and environmental monitoring applications. The desire for spectroscopic capability has increased in monitoring aspects of the nuclear fuel cycle to provide fast characterization of the radiation profile of a situation, such as the planned or unplanned release of material. The reduction or elimination of having to conduct sampling for laboratory analysis can result in significant cost savings for an industry, government agency, or regulatory body. A new system, comprised of a NaI:Tl scintillator and a G-M tube, has been designed and tested, taking into account the following end-user requirements: ease-of-use, capability to network and supervise multiple units, compact form factor, low power consumption, versatility, and stability. The detector sizes were selected to accommodate a dose rate up to 1 Sv/hr. Several algorithms and analysis routines have been developed to incorporate these key needs without sacrificing on accuracy, dynamic range, nuclide identification, and sensitivity. This presentation will introduce the major hardware and software components of the platform, as well as the user interface and data analysis workflow. Key features of the hardware include an environmentally robust housing, low power signal processing electronics, patented LED-based gain stabilization, and an embedded processor for unattended instrument management and data analysis. New and improved algorithms for determining scintillator gamma dose rates, total integrated dose, and nuclide identification will be introduced. The two detector elements were modeled with MCNP and validated experimentally. The results of the radiological testing shall be presented including energy resolution, throughput, dose response, and minimum detectable activities. The dose response has been evaluated in simulations and with measurements to ensure accurate response with respect to energy spectrum of the dose field and direction of the incident radiation. Spectroscopy evaluations, such as nuclide detection probability and false alarm rates, were performed and comparisons to existing algorithms shall also be presented. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

A Prototype {sup 212}Pb Medical Dose Calibrator for Alpha Radioimmunotherapy

AREVA Med, an AREVA group subsidiary, is developing innovative cancer-fighting therapies involving the use of {sup 212}Pb for alpha radioimmunotherapy. Canberra Industries, the nuclear measurement subsidiary of AREVA, has been working with AREVA Med to develop a prototype measurement system to assay syringes containing a {sup 212}Pb solution following production by an elution system. The relative fraction of emitted radiation from the source associated directly with the {sup 212}Pb remains dynamic for approximately 6 hours after the parent is chemically purified. A significant challenge for this measurement application is that the short half-life of the parent nuclide requires assay prior to reaching equilibrium with progeny nuclides. A gross counting detector was developed to minimize system costs and meet the large dynamic range of source activities. Prior to equilibrium, a gross counting system must include the period since the {sup 212}Pb was pure to calculate the count rate attributable to the parent rather than the progeny. The dynamic state is determined by solving the set of differential equations, or Bateman equations, describing the source decay behavior while also applying the component measurement efficiencies for each nuclide. The efficiencies were initially estimated using mathematical modeling (MCNP) but were then benchmarked with source measurements. The goal accuracy of the system was required to be within 5%. Independent measurements of the source using a high resolution spectroscopic detector have shown good agreement with the prototype system results. The prototype design was driven by cost, compactness and simplicity. The detector development costs were minimized by using existing electronics and firmware with a Geiger-Mueller tube derived from Canberra's EcoGamma environmental monitoring product. The acquisition electronics, communications and interface were controlled using Python with the EcoGamma software development kit on a Raspberry Pi Linux computer mounted inside a standard project box. The results of initial calibration measurements are presented. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Methodology comparison for gamma-heating calculations in material-testing reactors

The Jules Horowitz Reactor (JHR) is a Material-Testing Reactor (MTR) under construction in the south of France at CEA Cadarache (French Alternative Energies and Atomic Energy Commission). It will typically host about 20 simultaneous irradiation experiments in the core and in the beryllium reflector. These experiments will help us better understand the complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuels. Gamma heating, i.e. photon energy deposition, is mainly responsible for temperature rise in non-fuelled zones of nuclear reactors, including JHR internal structures and irradiation devices. As temperature is a key parameter for physical models describing the behavior of material, accurate control of temperature, and hence gamma heating, is required in irradiation devices and samples in order to perform an advanced suitable analysis of future experimental results. From a broader point of view, JHR global attractiveness as a MTR depends on its ability to monitor experimental parameters with high accuracy, including gamma heating. Strict control of temperature levels is also necessary in terms of safety. As JHR structures are warmed up by gamma heating, they must be appropriately cooled down to prevent creep deformation or melting. Cooling-power sizing is based on calculated levels of gamma heating in the JHR. Due to these safety concerns, accurate calculation of gamma heating with well-controlled bias and associated uncertainty as low as possible is all the more important. There are two main kinds of calculation bias: bias coming from nuclear data on the one hand and bias coming from physical approximations assumed by computer codes and by general calculation route on the other hand. The former must be determined by comparison between calculation and experimental data; the latter by calculation comparisons between codes and between methodologies. In this presentation, we focus on this latter kind of bias. Nuclear heating is represented by the physical quantity called absorbed dose (energy deposition induced by particle-matter interactions, divided by mass). Its calculation with Monte Carlo codes is possible but computationally expensive as it requires transport simulation of charged particles, along with neutrons and photons. For that reason, the calculation of another physical quantity, called KERMA, is often preferred, as KERMA calculation with Monte Carlo codes only requires transport of neutral particles. However, KERMA is only an estimator of the absorbed dose and many conditions must be fulfilled for KERMA to be equal to absorbed dose, including so-called condition of electronic equilibrium. Also, Monte Carlo computations of absorbed dose still present some physical approximations, even though there is only a limited number of them. Some of these approximations are linked to the way how Monte Carlo codes apprehend the transport simulation of charged particles and the productive and destructive interactions between photons, electrons and positrons. There exists a huge variety of electromagnetic shower models which tackle this topic. Differences in the implementation of these models can lead to discrepancies in calculated values of absorbed dose between different Monte Carlo codes. The magnitude of order of such potential discrepancies should be quantified for JHR gamma-heating calculations. We consequently present a two-pronged plan. In a first phase, we intend to perform compared absorbed dose / KERMA Monte Carlo calculations in the JHR. This way, we will study the presence or absence of electronic equilibrium in the different JHR structures and experimental devices and we will give recommendations for the choice of KERMA or absorbed dose when calculating gamma heating in the JHR. In a second phase, we intend to perform compared TRIPOLI4 / MCNP absorbed dose calculations in a simplified JHR-representative geometry. For this comparison, we will use the same nuclear data library for both codes (the European library JEFF3.1.1 and photon library EPDL97) so as to isolate the effects from electromagnetic shower models on absorbed dose calculation. This way, we hope to get insightful feedback on these models and their implementation in Monte Carlo codes. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

First Industrial Tests of a Drum Monitor Matrix Correction for the Fissile Mass Measurement in Large Volume Historic Metallic Residues with the Differential Die-away Technique

The fissile mass in radioactive waste drums filled with compacted metallic residues (spent fuel hulls and nozzles) produced at AREVA La Hague reprocessing plant is measured by neutron interrogation with the Differential Die-away measurement Technique (DDT. In the next years, old hulls and nozzles mixed with Ion-Exchange Resins will be measured. The ion-exchange resins increase neutron moderation in the matrix, compared to the waste measured in the current process. In this context, the Nuclear Measurement Laboratory (NML) of CEA Cadarache has studied a matrix effect correction method, based on a drum monitor ({sup 3}He proportional counter inside the measurement cavity). A previous study performed with the NML R and D measurement cell PROMETHEE 6 has shown the feasibility of method, and the capability of MCNP simulations to correctly reproduce experimental data and to assess the performances of the proposed correction. A next step of the study has focused on the performance assessment of the method on the industrial station using numerical simulation. A correlation between the prompt calibration coefficient of the {sup 239}Pu signal and the drum monitor signal was established using the MCNPX computer code and a fractional factorial experimental design composed of matrix parameters representative of the variation range of historical waste. Calculations have showed that the method allows the assay of the fissile mass with an uncertainty within a factor of 2, while the matrix effect without correction ranges on 2 decades. In this paper, we present and discuss the first experimental tests on the industrial ACC measurement system. A calculation vs. experiment benchmark has been achieved by performing dedicated calibration measurement with a representative drum and {sup 235}U samples. The preliminary comparison between calculation and experiment shows a satisfactory agreement for the drum monitor. The final objective of this work is to confirm the reliability of the modeling approach and the industrial feasibility of the method, which will be implemented on the industrial station for the measurement of historical wastes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Detection Of Special Nuclear Materials Tagged Neutrons

In order to detect Special Nuclear Materials (SNM) in unattended luggage or cargo containers in the field of homeland security, fissions are induced by 14 MeV neutrons produced by an associated particle DT neutron generator, and prompt fission particles correlated with tagged neutron are detected by plastic scintillators. SMN produce high multiplicity events due to induced fissions, whereas nonnuclear materials produce low multiplicity events due to cross-talk, (n,2n) or (n,n'γ) reactions. The data acquisition electronics is made of compact FPGA boards. The coincidence window is triggered by the alpha particle detection, allowing to tag the emission date and direction of the 14 MeV interrogating neutron. The first part of the paper presents experiment vs. calculation comparisons to validate MCNP-PoliMi simulations and the post-processing tools developed with the data analysis framework ROOT. Measurements have been performed using different targets (iron, lead, graphite), first with small plastic scintillators (10 x 10 x 10 cm{sup 3}) and then with large detectors (10 x 10 x 100 cm{sup 3}) to demonstrate that nuclear materials can be differentiated from nonnuclear dense materials (iron, lead) in iron and wood matrixes. A special attention is paid on SNM detection in abandoned luggage. In the second part of the paper, the performances of a cargo container inspection system are studied by numerical simulation, following previous work reported in. Detectors dimensions and shielding against the neutron generator background are optimized for container inspection. Events not correlated to an alpha particle (uncorrelated background), counting statistics, time and energy resolutions of the data acquisition system are all taken into account in a realistic numerical model. The impact of the container matrix (iron, ceramic, wood) has been investigated by studying the system capability to detect a few kilograms of SNM in different positions in the cargo container, within 10 min acquisitions. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2015

Impact Study on the Methodology used for Photon- Heating Determination in Material-Testing Reactors

Determination of photon heating by calculation is an important issue for the Jules Horowitz Reactor (JHR), the next international Material-Testing Reactor (MTR) under construction at CEA Cadarache research center in the south of France. Accurate knowledge of photon heating in structure materials and irradiation devices is necessary for JHR design and safety studies. In this paper, we quantify the impact of different photon-heating calculation routes by comparing absorbed dose and KERMA calculations (Kinetic Energy Released per Mass) from two different Monte Carlo codes, TRIPOLI-4.9{sup R} and MCNP (Monte Carlo N-Particle transport code). These calculations are carried out in JHR-representative geometries with the nuclear-data library JEFF3.1.1 and the photon-data library EPDL97. Discrepancies amounting to up to 18% between absorbed dose and KERMA are found in JHR irradiation devices and are linked to charged-particle transport effects taking place in heterogeneous materials of small dimensions. In a JHR assembly cell, discrepancies of about 1% on photon KERMA and of about 3% on absorbed dose are highlighted between the two Monte Carlo codes. These latter discrepancies are small compared to typical sources of uncertainty for Monte Carlo calculation (for instance, nuclear data uncertainty) and are supposed to be due to differences in the processing of gamma-production data by neutron interactions and to differences in electromagnetic-shower models and implementation between the two codes. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of a New Clearance Monitor for Exemption of NPP Radioactive Wastes and its Performance Test - 17200

In nuclear power plants, it is common practice to clear radioactive wastes for free release by using clearance monitor with plastic scintillation detectors so as to reduce disposal costs. These monitors are possible in part thanks to the excellent detection levels achievable with high energy and high yield photons emitted from contaminants such as Cs-137 and Co-60. However, it is difficult for the monitors to separate Naturally Occurring Radioactive Materials (NORM) from the radioactive wastes because the plastic scintillation detectors have poor resolution compared to NaI or HPGe detectors. KHNP has developed a new clearance monitor for determining whether or not the radioactivity level of metal scraps satisfies regulatory criteria. The monitor is composed of 12 large area plastic scintillation detectors and 3 NaI scintillation detectors. The measuring chamber is surrounded by a 5 cm-thick lead shield with chamber dimensions of 1.5 m x 1.0 m x 1.0 m. Its algorithm has a function of measuring total gamma activity while analyzing gamma nuclides and identifying 27 hotspots in radioactive wastes. The detection limit of this system, according to the MCNP simulation, was determined to be about 45 Bq for Co-60 and 78 Bq for Cs-137 with the condition of an empty chamber and measuring time of 1 minute. We evaluated the self - shielding effect of waste for background correction factor and waste height (volume in the chamber) correction factor. We also tested its MDA, location determination of hot-spots, temperature compensation function and convenience of operation. In the future, we will plan to install it at a nuclear site for more tests including evaluating its durability. (authors)

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

Evaluation of Neutron Flux on the Surface of a PWR Dry Storage Cask - 17574

Since the construction of a dry storage facility for spent nuclear fuel is expected in Korea, several methods have been studied for assuring the safety of pressurized-water reactor (PWR) spent fuels in dry storage casks. Existing evaluation methods are mostly based on detection technology of gamma-rays. However, the gamma-ray detection methods have high uncertainties due to background radiation. Therefore, we developed an evaluation method for the spent nuclear fuel monitoring based on neutron detection. In this study, we evaluated neutron flux on the surface of a PWR dry storage cask using computational simulation. Neutron release rate and energy spectra from a Westinghouse (WH) 17x17 type spent fuel with 10 years cooling time were calculated using ORIGEN-ARP code. MCNP code was used to simulate geometry and material properties of the spent fuel assemblies and casks. Neutron flux after 10 years of cooling time of the spent fuels was calculated on the surface of top, side, and bottom of the casks. The metal cask and the concrete cask, called as KORAD-21 casks which newly developed by Korea Radioactive Waste Agency (KORAD), were selected for this study. As a result, for the metal cask, average neutron flux were 4.79x10{sup 2} neutrons/cm{sup 2} sec, 5.29x10{sup 2} neutrons/cm{sup 2} sec, and 1.47x10{sup 3} neutrons/cm{sup 2} sec on the surface of top, side, and bottom of the cask, respectively. For the concrete cask, average neutron flux on the surface of top, side and bottom of the casks were 5.07x10{sup 1} neutrons/cm{sup 2} sec, 1.18x10{sup 1} neutrons/cm{sup 2} sec, and 1.46x10{sup 3} neutrons/cm{sup 2} sec, respectively. The study results can be applied to evaluate characteristics of PWR spent fuels and the neutron flux of dry storage cask. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Validation testing of the Canberra mobile Feed Roll Assay system - 15404

Japanese farmers grow various grasses as feed for cattle. These are rolled up and sealed for future use. Canberra has created a special truck to monitor these Feed Rolls, called 'Feed Roll Assay System [FRAS]'. On the bed of the truck is the shield containing a large NaI detector. The detector is 3x5x16'' in size, the same size as used in the Canberra FastScan Whole Body Counter. The detector is also LED-stabilized, to keep the gain constant under the wide range of temperature variations expected for this outdoor measurement instrument. The Canberra Osprey digital MCA is connected to the detector. The detector and the MCA are housed in a water proof box, which is surrounded on the bottom and all sides by a 15 cm thick steel shield. A portion of the shield at the top is open where the detector can view the Feed Rolls placed there. A single ethernet cable goes to a laptop PC which provides power to the MCA and detector, and also transmits the signal to the PC. The system was calibrated using MCNP for a variety of different densities, and Feed Roll sizes. The Japanese version of the Canberra Genie software is used to analyze the spectra and compute the Bq/kg concentration of the Feed Roll. The shielding is sufficient so that at a 5 uSv/hr background dose rate, no Cs-137 or Cs-134 peaks were detected in a 1 hour count. Measurements were performed in 0.2 uSv/hr dose rate, using uncontaminated Feed Rolls at various counting times. The MDA for Cs-137 was 11.9 Bq/kg at 30 seconds, 8.2 Bq/kg for 60 seconds, and 1.1 Bq/kg at 3600 seconds count time. The relative measurement uncertainty of the FRAS measurements was 9.4 to 13%, as compared to assays of samples extracted [30-45%] and in-situ Ge measurements [6.0 - 7.1%]. The FRAS measurement results have considerably better precision than measurements of samples extracted from the Feed Rolls, and are much faster than the in-situ Ge measurements and have just about the same uncertainty. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Using Green's Functions to Compute 3 Dimensional Neutron Flux Maps - 16016

The knowledge of the radionuclide content of radioactive waste is of utmost importance for safety and waste management reasons. Numerical simulations are used by EDF-DP2D to anticipate dismantling and radioactive waste management. The main numerical activation schemes include the computing of a 3 dimensional neutron flux map, obtained on the basis of a neutron propagation calculation. The codes used solve the transport equation called the Boltzmann equation without simplification. The Monte Carlo method is used to solve the Boltzmann equation. Random series of numbers are used to simulate the lives of millions of neutrons. The codes follow each neutron individually, from its birth to its disappearance by leakage, absorption, or fission, and then the neutron flux map is calculated at the nominal power rating conditions. The main drawback of the Monte Carlo codes is the large computing time needed to converge within acceptable statistical criteria. This extended time period is incompatible with the multiple fuel managements used by EDF to operate its NPPs because it is necessary to make a Monte Carlo calculation for each fuel loading pattern. The solution developed by EDF-DP2D is to use the importance factors of Green's Functions to simplify the computing of the 3 dimensional neutron flux maps. Physicists talk about Green's Functions where mathematicians talk about elementary solutions of linear equations with constant coefficients. With an activation scheme based on the Green's Functions approach the 3D neutron flux is obtained on the basis of the convolution of the normalized source contributions (i.e. normalized response matrices) with the neutrons emitted by the fuel assemblies. The response matrices to unitary sources are computed with the Monte Carlo code MCNP for different tallies of interest and the neutron sources emitted by the fuel are computed with a core code for different fuel loading patterns. The principal interest of this approach is the calculation of the response matrices to unitary sources that can be used for all power distributions, thus limiting the Monte Carlo computing time. (authors)

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

Characterization of Sr-90 Drums at the Hanford 618-10 Burial Ground - 16132

The 618-10 Burial Ground on the Hanford Reservation was an early trench waste site used for the disposal of waste from the 300 Area at Hanford. Much of this waste consisted of residues from evaluations of failed fuel as well as various process and separations studies associated with the Department of Energy (DOE) mission which were performed in the 300 Area between March 1954 and September 1963. Historical records indicated that there were also several drums which contained significant quantities of separated Sr-90. Cleanup of the 618-10 burial ground is covered under the Washington Closure Hanford (WCH) contract. Due to the proximity of the 618-10 site to the Columbia River, waste is being removed from the 618-10 site, stabilized, and disposed of into either the Environmental Restoration Disposal Facility (ERDF) if the containers are determined to be LLW or transferred to the Hanford Central Waste Complex (CWC) if the containers are suspected to be TRU waste. In order to determine the correct disposal path, Non Destructive Assay (NDA) must be performed on all of the containers excavated from the trenches. The NDA of these containers is performed using germanium gamma spectroscopy with an In Situ Object Counting System (ISOCS) and with a neutron slab counter. Occasional ISOCS spectra exhibited a characteristic Sr-90/Y-90 Bremsstrahlung continuum. In some cases there were measurable gammas present, however in other cases only the Bremsstrahlung continuum was present. Typically Sr-90 is correlated to Cs-137 as they are normally produced in approximately a 1:1 ratio during the fission process. For some of these drums there was no Cs-137 detectable and therefore no way to quantify the Sr-90. Due to concerns about providing a reasonable and conservative nuclide inventory it was decided to evaluate a technique to utilize data from the Bremsstrahlung continuum to quantify the Sr-90 activity in these drums. The development of the technique combined MCNP modeling and comparison of measured gammas to the continuum for a well-defined waste stream. The final result showed reasonable correlation to the expected scaling factors for many drums, while also providing a reasonable measured value for drums with significantly enhanced Sr-90 concentrations. Ultimately all drums were demonstrated to be within the ERDF waste acceptance criteria. (authors)

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

Release of MCNP6.2 and WHISPER-1.1 - Guidance for NCS users

In 2017, a new version of the MCNP6 Monte Carlo code package was released: MCNP6.2. This is the first version of MCNP to include the WHISPER package to support sensitivity-uncertainty based nuclear criticality safety validation. The coding changes to MCNP6.2 physics are relatively insignificant. Corrections to the S(α,β) thermal scattering numerics are generally negligible relative to problem statistics (or, in rare cases, prevent aborts). Similarly, the changes to adjoint-weighting for computing kinetics parameters may result in small differences, generally negligible compared to problem statistics. The change to the MCNP6.2 geometry treatment to correctly handle coincident surfaces in problems with universe/fill features will produce different round-off in the geometry tracking. This will produce differences in results relative to previous versions, but those differences should be small relative to problem statistics, and are not a concern. An immediate benefit of using MCNP6.2 (rather than MCNP6.1) is that the new version is typically 1.5-2 times faster.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Validation and bias quantification of criticality safety codes for SRNS operations

The validation and bias quantification methods employed by the Criticality Safety Engineering group at Savannah River Nuclear Solutions (SRNS) are discussed. Both SCALE and MCNP are validated against benchmark experiments, where these benchmark models have been divided into subgroups representative of the types of fissile material operations at SRNS. A quantification of the bias is made for each subgroup for each code and the three methods employed for quantification are discussed. The bias information is employed directly in the establishment of the Upper Safety Limit (USL) value of the neutron multiplication factor when using standardized documented SRNS methods. A simple example of validation and bias quantification is given and the procedure for setting the USL using the derived bias information is also presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Efficiency Calculation for an In-situ Beta Nuclide Monitoring System Based on a Plastic Scintillator - 18193

A plastic scintillator performance for an in-situ measurement system was analyzed using both simulation and experimental methods. The measurement results of four major pure beta-emitting radionuclides - namely H-3. C-14, P-32 and Sr-90/Y-90 - were compared with those from a Monte Carlo N-particle (MCNP) code simulation. Good agreement between the simulation and experimental results was observed for P-32 and Sr-90/Y-90. However, due to the short range of their beta particles in water, the efficiency for both H-3 and C-14 was rather low, and these radionuclides could not be detected. The optimal thickness for the plastic scintillator was identified, as was its suggested diameter. The effect of water depth on system performance was also considered. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

On-board Radiation Sensor and Low-Altitude Contour Mapping of Radiation Fields using UAS Swarm - 18239

This paper addresses the design of light weight radiation sensors for the small-scale unmanned aerial system (UAS), and its implementation for low-altitude radiation source localization and contour mapping. The compact, high-resolution gamma-ray CZT sensors were integrated to UAS platforms as plug-and-play components using Robot Operation System (ROS). In this work, the UAS swarm consists of three platforms positioned in circular formation. The onboard data analysis provides time and position stamped intensities of gamma-ray peaks for each UAS-based sensor that are used as the input data for the swarm flight control algorithm. Each UAS measures radiation simultaneously for estimating gradient of radiation to compute a bulk heading angle of the swarm for contour mapping, which can provide relatively precise boundary of the safe area for potential human exploration. The UAS radiation sensors allow measurements from just few meters up to hundreds of meters thus filling the gap between land- and helicopter- based platforms. The swarm of UAS has advantages over a single agent based approach in detecting radiative sources and effectively mapping the area. The proposed method can locate sources of radiation as well as mapping the contaminated area for enhancing situation awareness capabilities for first responders. Computer simulation is presented for UAS swarm based source localization and contour mapping in the simulated radiation fields created by the MCNP code. Selected key gradient estimation and contour mapping algorithms are validated experimentally with small scale multicopter UAS in the indoor flight test bed. (authors)

42 ENGINEERING↗