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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Deferred Verification: The Role of New Verification Technologies and Approaches

Researchers at UNIDIR have recently proposed a fundamentally new approach to nuclear arms control verification, dubbed "deferred verification." The concept forgoes inspections at sensitive nuclear sites and inspections of nuclear materials or components in classified form. To implement this concept, a state first divides its territory into a closed segment and an open segment. The total fissile material inventory in the closed segment, which includes the weapon complex, is known and declared with very high accuracy. Essentially no inspections take place in the closed segment. Yet, inspectors have access to the open segment, which includes in particular the civilian nuclear sector. The fissile material inventory in the open segment is known with less accuracy, but uncertainties can be reduced over time using methods of nuclear archaeology. Deferred verification relies primarily on established safeguards techniques and avoids many unresolved verification challenges, including for example the need for information barriers for warhead confirmation measurements. At the same time, deferred verification faces some unique challenges. In this work, we explore some of these challenges and offer possible solutions while maintaining the strengths of the concept; to do so, we examine possible noncompliance strategies in which a state would seek to withhold a higher-than-declared inventory.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2011

A new set-up for the simultaneous measurement of neutron-induced capture and fission reactions

The measurement of the capture cross section of fissile elements, of upmost importance for the design of innovative nuclear reactors and the management of nuclear waste, involves particular difficulties related to the {gamma}-ray background produced in the fission reactions. These difficulties are the reason why five out of the six actinide {sigma}(n,{gamma}) measurements in the NEA High Request Priority List are fissile isotopes. At n-TOF we have combined the Total Absorption Calorimeter capture detector with a set of three {sup 235}U loaded MicroMegas fission detectors for measuring simultaneously the two reactions: capture and fission. In a first test measurement we have succeeded in measuring simultaneously with high efficiency the {sup 235}U capture and fission cross sections, disentangling accurately the two types of reactions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Spectral indices measurements using miniature fission chambers at the MINERVE zero-power reactor at CEA using calibration data obtained at the BR1 reactor at SCK.CEN

Spectral indices are parameters of major importance to characterize a multiplying medium and to evaluate performances of neutronics calculation tools as well as quality of associated nuclear data. Experiments performed in 2004 at the CEA MINERVE facility loaded with the R1-UO{sub 2} lattice [1], using homemade miniature fission chambers and calibration data acquired at the SCK.CEN BR1 facility in 2001 [2], resulted in ambivalent conclusions. On one hand, spectral indices involving only fissile isotopes gave relevant results and consistent discrepancies between calculation and experiment. On the other hand, spectral indices involving both fissile and fertile isotopes showed inconsistent results depending on the type of calibration data used. In particular, the {sup 238}U(n,f)/{sup 235}U(n,f) spectral indices resulted in values far from the calculations in all cases. In 2004, those differences between measurement and calculation were mainly attributed to an imprecise knowledge of the isotopic composition of the deposits loaded in the {sup 238}U chambers. At that time, priorities in the development of programs in MINERVE and other considerations about the manufacturing and calibration of new fission chambers did not allow to further investigate these discrepancies. In 2009, the preparation of a new program at the EOLE facility [1] and changes in the regulatory framework for management of fission chambers motivated the manufacturing of new detectors. At the same time, the re-installation of the R1-UO{sub 2} lattice in MINERVE gave the opportunity to carry out a new spectral indices measurement campaign. Nevertheless, although the isotopic composition of active deposits was better known than previously, the comparison between experimental results and calculations still resulted in inconsistent discrepancies. A sensitivity study indicated afterwards that, among all parameters to be considered for the analysis and the assessment of final uncertainty, calibration data have the greatest impact. In April 2010, a new calibration campaign conducted again at the BR1 facility allowed the CEA to reanalyze the spectral indices measurements performed in 2009. With this new set of calibration data, the experimental values of spectral indices finally matched the calculation results within the uncertainty margins. In addition, the confrontation of data from 2001 and 2010 calibration campaigns, both performed with the same experimental conditions, allowed to shed on the light dead time correction issues in the monitoring of calibration measurements. At the end, this also explained the inconsistencies observed during the 2004 campaign. This paper first, presents the two different analysis methods for getting the experimental value of a spectral index and, lists main parameters that influence the determination of the experimental value. It then details the calibration processes that are set-up at the MINERVE and BR1 facilities, describes the various experiments carried out and finally comments the results obtained. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Status of the nuclear measurement stations for the process control of spent fuel reprocessing at AREVA NC/La Hague

Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectroscopy, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality and safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remained after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations no. 1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station no. 1 allows determining the burn-up of the irradiated fuel by gamma-ray spectroscopy with HP Ge (high purity germanium) detectors. The burn-up is correlated to the {sup 137}Cs and {sup 134}Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station no. 3 is dedicated to the control of the correct fuel dissolution, which is performed with a {sup 137}Cs gamma ray measurement with a HP Ge detector. Station no. 7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially in the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional {sup 3}He detectors. The measurement stations have been validated for the reprocessing of Uranium Oxide (UOX) fuels with a burn-up rate up to 60 GWd/t. This paper presents a brief overview of the current status of the nuclear measurement stations. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Fast multiplicity counter featuring stilbene detectors for special nuclear material assay

The Treaty on the Nonproliferation of Nuclear Weapons demands for new instruments to prevent diversion of fissile material and safeguard controls. We present a fast neutron multiplicity counting system based on organic scintillators, i.e. liquid EJ-309 and stilbene. The system is able to detect fission correlated counts and photon and neutron multiplets, needed to quantify fissile mass, without the need of complex electronic circuitry and unfolding procedures, required in moderated systems. Moderated system, typically based on helium-3 detectors, feature a gate time of hundreds of microseconds. Conversely, the use of organic scintillators allows an exceptionally short gate time of hundreds of nanoseconds. This results in a reduction of accidental neutron coincidence. In our system, the superior gamma-neutron discrimination capabilities of stilbene scintillators allow to measure high-order multiplicities (> 3) and the inspection of complex source terms. This counter is thus also suitable for the assay of newer fuels containing plutonium and other fissionable actinides. A prototypal version of the well-counter was assembled and tested at the University of Michigan, using a spontaneous fission and an (a,n) neutron source, i.e. 252Cf and PuBe respectively. Presented measured results show excellent agreement with simulations. The viability of the system to discriminate time-correlated fission neutrons from random, uncorrelated neutrons was proved. An experimental campaign will be carried out at Idaho National Laboratory in August 2015 to characterize weapon grade plutonium samples.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2012

Spectral indices measurements using miniature fission chambers at the MINERVE zero-power reactor at CEA using calibration data obtained at the BR1 reactor at SCK.CEN

Spectral indices measurements performed in 2004 at the CEA MINERVE facility loaded with the R-UO{sub 2} lattice, using calibration data acquired at the SCK center dot CEN BR1 facility in 2001, resulted in ambivalent conclusions. On one hand, spectral indices involving only fissile isotopes gave consistent discrepancies between calculation and experiment. On the other hand, spectral indices involving both fissile and fertile isotopes, in particular the {sup 238}U(n, f)/{sup 235}U(n, f) spectral index, showed inconsistent results depending on the type of calibration data used. For different reasons, no definitive explanation was given at that time. In 2009, the preparation of the AMMON program at the EOLE facility motivated the manufacturing of a new set of detectors. At the same time, the re-installation of the R1-UO{sub 2} lattice in MINERVE provided the opportunity to carry out again a spectral indices measurement campaign. Nevertheless, although the isotopic compositions of active deposits were better known than previously, the comparison between experimental results and calculations still lead to inconsistent discrepancies. In April 2010, a new calibration series conducted again at the BR1 facility allowed the CEA to reanalyze the spectral indices measurements performed in 2009. With these very latest calibration data, experimental values of spectral indices finally matched calculations within the uncertainty margins. This paper also sums up the work that has been achieved to explain the incoherencies observed in 2004. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Effects of Correlated and Uncorrelated Gamma Rays on Neutron Multiplicity Counting

Neutron multiplicity counting relies on time correlation between neutron events to assay the fissile mass, (α,n) to spontaneous fission neutron ratio, and neutron self-multiplication of samples. Gamma-ray sensitive neutron multiplicity counters may misidentify gamma rays as neutrons and therefore miscalculate sample characteristics. Time correlated and uncorrelated gamma-ray-like signals were added into gamma-ray free neutron multiplicity counter data to examine the effects of gamma ray signals being misidentified as neutron signals on assaying sample characteristics. Multiplicity counter measurements with and without gamma-ray-like signals were compared to determine the assay error associated with gamma-ray-like signals at various gamma-ray and neutron rates. Correlated and uncorrelated gamma-ray signals each produced consistent but different measurement errors. Correlated gamma-ray signals most strongly led to fissile mass overestimates, whereas uncorrelated gamma-ray signals most strongly lead to (α,n) neutron overestimates. Gamma-ray sensitive neutron multiplicity counters may be able to account for the effects of gamma-rays on measurements to mitigate measurement uncertainties.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Development of a thin scintillation films fission-fragment detector and a novel neutron source

Here, investigation of prompt fission and neutron-capture Υ rays from fissile actinide samples at the Detector for Advanced Neutron Capture Experiments (DANCE) requires use of a fission-fragment detector to provide a trigger or a veto signal. A fission-fragment detector based on thin scintillating films and silicon photomultipliers has been built to serve as a trigger/veto detector in neutron-induced fission measurements at DANCE. The fissile material is surrounded by scintillating films providing a 4π detection of the fission fragments. The scintillations were registered with silicon photomultipliers. A measurement of the 235 U(n,f) reaction with this detector at DANCE revealed a correct time-of-flight spectrum and provided an estimate for the efficiency of the prototype detector of 11.6(7)%. Design and test measurements with the detector are described. A neutron source with fast timing has been built to help with detector-response measurements. The source is based on the neutron emission from the spontaneous fission of 252 Cf and the same type of scintillating films and silicon photomultipliers. Overall time resolution of the source is 0.3 ns. Design of the source and test measurements with it are described. An example application of the source for determining the neutron/gamma pulse-shape discrimination by a stilbene crystal is given.

42 ENGINEERING↗
DOE OSTI2014

T invariance and T-odd asymmetries for the cold-polarized-neutron-induced fission of nonoriented nuclei

It is shown that the coefficients D{sup exp} for all T-odd asymmetries observed experimentally in the cross sections for the reactions of cold-polarized-neutron-induced fission of nonoriented target nuclei (which involves the emission of prescission and evaporated particles) comply in shape and scale with the coefficients D{sup theor} calculated for the analogous asymmetries on the basis of quantum-mechanical nuclear-fission theory for T-invariant Hamiltonians of fissile systems. It is also shown that the asymmetries in question arise upon taking into account the effect of (i) the interference between the fission amplitudes of s- and p-wave resonances of a polarized fissile compound nucleus formed in the aforementioned reactions; (ii) the collective rotation of the compound nucleus in question (this rotation entails a change in the angular distributions of fission fragments and third particles); and (iii) the wriggling vibrations of this compound nucleus in the vicinity of its scission point, which lead to the appearance of high aligned spins of fission fragments, with the result that the emission of neutrons and photons evaporated from these fragments becomes anisotropic. The possible contribution of T-noninvariant interactions to the formation of the T-odd asymmetries under analysis is estimated by using the results obtained in experimentally testing the detailed-balance principle, (P-A) theorem, and T invariance of cross sections for elastic proton-proton and proton-neutron scattering.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2017

Decisive role of wriggling vibrations in the formation of angular and spin distributions of products originating from binary and ternary fission of oriented nuclei

It is shown that the multiplicities and angular and energy distributions of neutrons and photons evaporated from thermalized fragments originating from the spontaneous and low-energy induced fission of nuclei, the relative yields of ground and isomeric states of final fragments, and the features of delayed neutrons emitted upon the beta decay of the above fragments can successfully be described by employing nonequilibrium distributions of spins and relative orbital angular momenta of fission fragments formed in the vicinity of the scission point for the fissile nucleus being studied. It is also shown that these distributions, which are characterized by large mean values of the spins and orbital angular momenta directed orthogonally to the symmetry axis of the fissioning nucleus are successfully constructed upon simultaneously taking into account zero-mode transverse wriggling and bending vibrations of a fissile compound nucleus in the vicinity of its scission point, the wriggling vibrations being dominant. It is confirmed that the zero-mode wriggling vibrations considered immediately above are directly involved in the formation of the angular distributions of fragments originating from the spontaneous and low-energy fission of nuclei. This makes it possible to describe successfully such distributions for photofission fragments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Photonuclear reaction studies at HIγS: Developing the science of remote detection of nuclear materials

Development of gamma-ray beam interrogation technologies for remote detection of special nuclear materials and isotope analysis requires comprehensive databases of nuclear structure information and gamma-ray induced nuclear reaction observables. Relevant nuclear structure details include the energy, spin and parity of excited states that have significant probability for electromagnetic transition from the ground state, i.e, the angular momentum transferred in the reaction is Δl ≤ 2. This talk will report recent Nuclear Resonance Fluorescence (NRF) measurements to identify and characterize new low-spin states in actinide nuclei at energies from 1 to 4 MeV, which is the energy range most important for remote analysis methods. Here, these measurements are carried out using the nearly mono-energetic linearly polarized gamma-ray beam at the High Intensity Gamma-ray Source (HIγS) at the Triangle Universities Nuclear Laboratory. Also, studies of the (γ, n) reaction on a variety of nuclei with linearly polarized beams at HIγS indicate that this reaction might be used to discern between fissile and non-fissile materials. This work will be described. In addition, an overview will be given of a concept for a next generation laser Compton-backing scattering gamma-ray source to be implemented as an upgrade to increase the beam intensity at HIγS by more than an order of magnitude.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Non-destructive Assay Measurements Using the RPI Lead Slowing Down Spectrometer

The use of a Lead Slowing-Down Spectrometer (LSDS) is consid- ered as a possible option for non-destructive assay of fissile material of used nuclear fuel. The primary objective is to quantify the 239Pu and 235U fissile content via a direct measurement, distinguishing them through their characteristic fission spectra in the LSDS. In this pa- per, we present several assay measurements performed at the Rensse- laer Polytechnic Institute (RPI) to demonstrate the feasibility of such a method and to provide benchmark experiments for Monte Carlo cal- culations of the assay system. A fresh UOX fuel rod from the RPI Criticality Research Facility, a 239PuBe source and several highly en- riched 235U discs were assayed in the LSDS. The characteristic fission spectra were measured with 238U and 232Th threshold fission cham- bers, which are only sensitive to fission neutron with energy above the threshold. Despite the constant neutron and gamma background from the PuBe source and the intense interrogation neutron flux, the LSDS system was able to measure the characteristic 235U and 239Pu responses. All measurements were compared to Monte Carlo simula- tions. It was shown that the available simulation tools and models are well suited to simulate the assay, and that it is possible to calculate the absolute count rate in all investigated cases.

Lead Slowing Down Spectrometer↗
DOE OSTI2015

Lead Slowing Down Spectrometry Analysis of Data from Measurements on Nuclear Fuel

Improved non-destructive assay of isotopic masses in used nuclear fuel would be valuable for nuclear safeguards operations associated with the transport, storage and reprocessing of used nuclear fuel. Our collaboration is examining the feasibility of using lead slowing down spectrometry techniques to assay the isotopic fissile masses in used nuclear fuel assemblies. We present the application of our analysis algorithms on measurements conducted with a lead spectrometer. The measurements involved a single fresh fuel pin and discrete 239Pu and 235U samples. We are able to describe the isotopic fissile masses with root mean square errors over seven different configurations to 6.35% for 239Pu and 2.7% for 235U over seven different configurations. Funding Source(s):

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Neutron multiplicity counting: Confidence intervals for reconstruction parameters

From nuclear materials accountability to homeland security, the need for improved nuclear material detection, assay, and authentication has grown over the past decades. Starting in the 1940s, neutron multiplicity counting techniques have enabled quantitative evaluation of masses and multiplications of fissile materials. In this paper, we propose a new method to compute uncertainties on these parameters using a model-based sequential Bayesian processor, resulting in credible regions in the fissile material mass and multiplication space. These uncertainties will enable us to evaluate quantitatively proposed improvements to the theoretical fission chain model. Additionally, because the processor can calculate uncertainties in real time, it is a useful tool in applications such as portal monitoring: monitoring can stop as soon as a preset confidence of non-threat is reached.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Caustic Precipitation of Plutonium and Uranium with Gadolinium as a Neutron Poison

Here, the caustic precipitation of plutonium (Pu) and uranium (U) from Pu and U containing waste solutions has been investigated to determine whether gadolinium (Gd) could be used as a neutron poison for precipitation with greater than a fissile mass containing both Pu and enriched U. Precipitation experiments were performed using both actual samples and simulant solutions with a range of 2.6-5.16 g/L U and 0-4.3 to 1 U to Pu. Analyses were performed on solutions at intermediate pH to determine the partitioning of elements for accident scenarios. When both Pu and U were present in the solution, precipitation began at pH 4.5 and by pH 7, 99% of Pu and U had precipitated. When complete neutralization was achieved at pH greater than 14 with 1.2 M excess OH - , greater than 99% of Pu, U, and Gd had precipitated. At pH greater than 14, the particles sizes were larger and the distribution was a single mode. The ratio of hydrogen to fissile atoms in the precipitate was determined after both settling and centrifuging and indicates that sufficient water was associated with the precipitates to provide the needed neutron moderation for Gd to prevent a criticality in solutions containing up to 4.3 to 1 U to Pu and up to 5.16 g/L U.

08 HYDROGEN↗
DOE OSTI2016

Liquid fuel molten salt reactors for thorium utilization

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233 U/thorium fuel cycle, as the nuclear properties of 233 U combined with the online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing thorium-based MSRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Proposed Plutonium-Nickel Critical Experiment

The Los Alamos Critical Experiments Facility (LACEF) and subsequently its descendant, the National Criticality Experiments Research Center (NCERC) located at the Nevada National Security Site, have been performing critical experiments for over 75 years. The work done continues to build upon lessons learned from the past. It also builds upon knowledge gained from past experiments. The LACEF and NCERC experiments “provide nuclear criticality benchmark data for a wide variety of fissile and non-fissile materials in all neutron energy spectra systems”. In 2014, a subcritical benchmark with plutonium surrounded by nickel was issued in the International Criticality Safety Benchmark Evaluation Project (ICSBEP). Two years later, a new benchmark experiment was performed, commonly known as Subcritical Copper Reflected $\mathcal{a}$-Phase Plutonium SCR$\mathcal{a}$P. The uncertainties in the SCR$\mathcal{a}$P experiment were significantly reduced from the Pu-Ni experiment and the methods were refined. Additionally, a Pu-lead critical experiment was performed in 2017 which followed different methods and with different objectives. Now, an experiment is proposed that builds on both the subcritical nickel experiment and the critical lead experiment. The proposed experiment has interleaved nickel and plutonium, in a thermal arrangement. This particular experiment would support criticality safety at the Hanford Tank Farms, whose tanks contain a significant amount of nickel credited as a neutron absorber in the criticality safety evaluation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

Lead Slowing-Down Spectrometry Time Spectral Analysis for Spent Fuel Assay: FY11 Status Report

Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) from next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT supports a multi-institutional collaboration, of which PNNL is a part, to study the feasibility of Lead Slowing Down Spectroscopy (LSDS). This technique is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertainty considerably lower than the approximately 10% typical of today's confirmatory assay methods. This document is a progress report for FY2011 PNNL analysis and algorithm development. Progress made by PNNL in FY2011 continues to indicate the promise of LSDS analysis and algorithms applied to used fuel. PNNL developed an empirical model based on calibration of the LSDS to responses generated from well-characterized used fuel. The empirical model, which accounts for self-shielding effects using empirical basis vectors calculated from the singular value decomposition (SVD) of a matrix containing the true self-shielding functions of the used fuel assembly models. The potential for the direct and independent assay of the sum of the masses of 239Pu and 241Pu to within approximately 3% over a wide used fuel parameter space was demonstrated. Also, in FY2011, PNNL continued to develop an analytical model. Such efforts included the addition of six more non-fissile absorbers in the analytical shielding function and the non-uniformity of the neutron flux across the LSDS assay chamber. A hybrid analytical-empirical approach was developed to determine the mass of total Pu (sum of the masses of 239Pu, 240Pu, and 241Pu), which is an important quantity in safeguards. Results using this hybrid method were of approximately the same accuracy as the pure empirical approach. In addition, total Pu with much better accuracy with the hybrid approach than the pure analytical approach. In FY2012, PNNL will continue efforts to optimize its empirical model and minimize its reliance on calibration data. In addition, PNNL will continue to develop an analytical model, considering effects such as neutron-scattering in the fuel and cladding, as well as neutrons streaming through gaps between fuel pins in the fuel assembly.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗