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At least 1,099 records · Page 61

Quasi-differential neutron induced neutron emission reaction measurements at WNR

The quasi-differential neutron induced neutron emission reaction measurement (QD-NINE) method provides an additional benchmark which constrains cross-section evaluations and provides a measure of verification for existing evaluations. Experiments are preformed using an array of EJ-309 liquid scintillators at a variety of angles measuring the neutron scattering response from a sample.. An MCNP simulation is then compared to this experimental data and is run using various sets of evaluated nuclear data. Since the isotopes of interest for this measurement, {sup 235}U and {sup 239}Pu, are difficult to obtain, the WNR facility at LANL was chosen to perform the measurements utilizing the chi-nu detector array. A measurement of carbon has been performed and the high energy cutoff has been extended from 20 to 60 MeV compared with previous measurements using this method. An early stage simulation was compared to the experimental data and the results show good agreement. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

Some Non-Equilibrium Effects of Gamma Dose Deposition in Brick

This work demonstrates the effects of non-equilibrium in gamma dose deposition through brick at multiple energies. These results demonstrate some important effects to consider when modelling dose deposition in materials. The phenomenon is related to the buildup flux when applied to layers which are too thin to generate adequate buildup flux for dose deposition. If retrospective dosimetry techniques are employed to determine historical sources, these are important effects which should be taken into account. This effect is highlighted utilizing the *F8 tally capability in MCNP. (authors)

36 MATERIALS SCIENCE↗

A Comparison of the FW-CADIS and MR-CADIS Variance Reduction Methods

The Consistent Adjoint Driven Importance Sampling (CADIS) method, which uses an adjoint deterministic estimate for making variance reduction parameters that optimize Monte Carlo (MC) shielding calculations, was introduced in 1998. This method was designed to optimize a single tally, typically an integrated response such as a detector count rate or a dose rate. Because both deterministic and stochastic methods are employed, CADIS is referred to as a hybrid method. In 2007, the forward-weighted CADIS (FW-CADIS) method was introduced to optimize several tallies at once, giving more uniform relative uncertainties among the tallies for both high-flux and low-flux areas of the problem. This method can be used for several small detectors, for semi-global problems (mesh tally over a large fraction of the problem), or for global problems (mesh tally over the entire problem). Optimization can be done for an integrated response such as a dose rate D(r) as a function of position, the total flux {phi}(r) as a function of position, or the space/energy flux {phi}(r,E). Both the CADIS and FW-CADIS methods have been in the MAVRIC sequence of SCALE for three versions (6.0 in 2009, 6.1 in 2011, and 6.2 in 2016). The ADVANTG package released in 2015 uses the CADIS and FW-CADIS methods to accelerate MCNP neutron/photon shielding calculations. In 2014, the multi-response CADIS (MR-CADIS) method was presented as a way to obtain even more uniform relative uncertainties for problems with multiple tallies than those obtained with FW-CADIS. The method does not use a forward deterministic calculation but instead uses one adjoint calculation for every tally to be optimized in the Monte Carlo calculation. This work compares the above methods.

42 ENGINEERING↗

Determining Physical Parameters of Shielded Uranium using Gamma Spectroscopy and the Differential Evolution Adaptive Metropolis (DREAM) Method

Inverse radiation transport is a well-established method for determining unknown parameters within a system that exhibits particle leakage. Two significant challenges of solving inverse transport problems are (1) to determine multiple unknowns without false convergence on local minima and (2) to be able to quantify uncertainty. It was recently demonstrated that the Differential Evolution Adaptive Metropolis (DREAM) method effectively meets both of these challenges. In Bledsoe's 2014 article, the DREAM method was only applied to numerical test cases in which measured data were simulated using MCNP. This paper provides results of research conducted at Oak Ridge National Laboratory (ORNL) in which DREAM was applied to two-dimensional inverse transport problems using real-world measurements. The DREAM method uses a Markov chain Monte Carlo (MCMC) method to obtain a posterior distribution of unknown parameters. In this application, the posterior distribution is determined by using a uniform prior distribution and a likelihood function. The goal of the inverse problem is to find the regions in which the likelihood function is at or near its maximum. In the traditional MCMC approach, a single Markov chain is employed. In practice, MCMC approaches have a high acceptance rate to solutions that are near the current point of sampling, but they exhibit slow convergence to solutions far from the sampling point. The DREAM algorithm was designed to alleviate this issue by initiating multiple Markov chains and using the differential evolution algorithm to generate trial points for each chain. Solving for the mass of radioactive material within a voluminous source is a challenging problem for gamma-ray spectroscopy. Inverse gamma-ray transport can supply a range of possible material compositions, but reducing the differences between predicted and actual measurements can lead to false results. DREAM is a rapid technique for determining the most likely solution for several unknown parameters in an inverse transport problem. This work applies DREAM to a series of gamma spectroscopy experiments performed at ORNL.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

LANL-SNL Collaboration on NCS Validation

During 2016, nuclear criticality safety (NCS) practitioners from SNL and code developers from LANL collaborated in several areas of interest to the DOE/NNSA Nuclear Criticality Safety Program (NCSP). This collaboration involved: - Testing of the preliminary release of the MCNP6- Whisper methodology, with feedback to the developers, - Sharing of the benchmark catalogs (the collection of MCNP input files and benchmark results), with 1101 cases from LANL and 866 cases from SNL, - Comparison and analysis of 357 benchmarks common to both catalogs, - Investigation of the impact of the different benchmark catalogs on sensitivity-uncertainty based NCS validation results from MCNP6-Whisper, - Investigation of the impact of randomized selections from the benchmark catalog on sensitivity-uncertainty based validation results from MCNP6-Whisper. - Investigation of the use of MCNP6-Whisper in selecting benchmarks for use in NCS validation for unique, nonstandard, legacy fuel applications. This paper summarizes the collaboration work and initial results. It must be noted that the results described herein are preliminary and need further research and detailed analysis. However, the initial results are very interesting, and it is important to share them with the NCSP community.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Testing Nesting DXTRAN Spheres for Neutron Transport in Air

The deterministic transport sphere (DXTRAN sphere) is a technology that can be utilized in MCNP for improving sampling in the vicinity of a particular geometry, when the solid angle, between the source and particular geometry, is small. At a collision event, a DXTRAN particle is created and deterministically transported to the DXTRAN sphere, while the colliding particle (NONDXTRAN particle) continues transporting as if no DXTRAN event had occurred. Weight is balanced by killing the NONDXTRAN particle when traversing the DXTRAN sphere. If set up correctly, the DXTRAN particle placed on the DXTRAN sphere has a high probability of interacting with the geometry within the DXTRAN sphere. The Central Limit Theorem (CLT) states that the estimated mean approaches a normal distribution with known variance as the number of random samples grows sufficiently large. In reality the variance is seldom known and therefore must be estimated. In a Monte Carlo calculation, one must determine when the number of random samples is sufficiently large for the CLT to be satisfied. For the CLT to be satisfied, the first two moments (mean and variance) must exist. The second moment, the variance of a normal distribution is estimated. The slope test, which uses the largest 201 history scores of each tally fluctuation chart (TFC) bin fit to a Pareto function, is used to determine if the largest history scores decrease faster than 1/x{sup 3}. If the slope is less than 3, then enough histories have not yet been sampled such to meet the second moment existence requirement of the CLT. Usually, a slope of less than 3 indicates that the effects of high weight scores have not been satisfactorily sampled; such a consequence can also be manifested in examining trends in the 4. moment, the variance of the variance (VOV). For example, when examining the TFC, if the VOV is monotonically decreasing for a few sets of histories and then drastically increases for the next set of histories, chances are a rare high weight score, significantly greater than the average score, was encountered. As a result of the 1/R{sup 2} term, a point detector can achieve an extremely high weight, much greater than the average, as the distance from collision to point detector approaches zero. To remedy this difficulty, a 'radius of exclusion' is applied, where within the radius of exclusion 'the point detector estimate is assumed to be the average flux uniformly distributed'. A DXTRAN sphere can suffer from the same weight fluctuation as collisions in the vicinity of the DXTRAN sphere produce much larger weights than collisions happening far from the DXTRAN sphere. If the rare large weight put on the DXTRAN sphere then scores in a tally, there will be a significant increase in the VOV and the slope may drop below 3. To compliment that analysis, here we utilize an air composition and density that varies as a function of altitude, detailing the trials and tribulations of attempting to pass the 10 statistical checks for a variety of single and nested DXTRAN sphere approaches. The objective here is to illuminate important considerations for radiation transport problems with small solid angles in a scatter medium of few mean free paths (MFPs)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Using the MCNP6.2 Correlated Fission Multiplicity Models, CGMF and FREYA

The culmination of a multi-year, collaborative effort by researchers at Los Alamos National Laboratory (LANL), Lawrence Berkeley National Laboratory (LBNL), Lawrence Livermore National Laboratory (LLNL) and the University of Michigan, led to the delivery of two new correlated fission multiplicity models available within the latest release of MCNP{sup R} version 6.2. The needs within the nuclear nonproliferation community, particularly in simulating and predicting the neutron and gamma-ray signatures of special nuclear materials (SNM), provided the motivation for this major collaborative effort. From LANL, the CGMF code has been implemented as an upgrade to the original Cascading Gamma-ray Multiplicity (CGM) code. Additionally, the LLNL Fission Library has been upgraded from the original implementation released in prior versions of MCNP6/X to include the LBNL/LLNL Fission Reaction Event Yield Algorithm (FREYA) code. Both new codes are capable of simulating fission events with correlated neutron and gamma-ray emissions. Due to the release of these new models within MCNP6.2, the purpose of this paper is to describe the differences between these new models and existing capabilities, demonstrate their use in the context of simulation inputs and outputs, and to discuss some of the ongoing work directed toward the next generation of these features.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Improvements in Electron-Photon-Relaxation Data for MCNP6

The Los Alamos Monte Carlo transport code MCNP6 has recently benefitted from significant work to improve the photon-electron transport and atomic-relaxation capabilities of the code. The improvements include extension of existing photon data to lower energies than previously supported; addition of new data such as subshell-specific photoelectric cross sections; completion of the incoherent and coherent form factor data to cover all scattering angles at all supported energies; new data for electron interactions including Bremsstrahlung, elastic scattering, subshell-specific electroionization, and atomic excitation; a much more detailed representation for atomic relaxation; and an entirely new (to MCNP) approach to electron transport: the single-event method. These improvements were made available with the initial production release of MCNP6 along with an extended electron, photon, and relaxation data file eprdata12 to support the new capabilities. More recently new data have become available and have now been incorporated into MCNP6, correcting a few issues with the previous work, and adding new sets of cross sections that were not included in eprdata12. The purpose of this paper is to introduce these newer changes and enhancements, which will be included with the next public release of the MCNP6 code system. The database that supports these developments relies on an evolving set of libraries originally released by D. E. Cullen and collaborators as separate data for photo-atomic interactions (EPDL), atomic relaxation (EADL), and electro-atomic interactions (EEDL). An updated version of this release was officially adopted in the ENDF/B VI.8 release of the Evaluated Nuclear Data File and it was this source that was used to produce the ACE (A Compact ENDF) format library eprdata12 for use in the initial production release of MCNP6.1 (and the subsequent interim release MCNP6.1.1). The latest improved version of this database was released by Cullen in 2014. This is the EPICS2014 collection, which includes both an ENDF-formatted file and separate EPDL, EADL, and EEDL files. In the area of photon-electron transport and atomic-relaxation, the newest enhancements to the current version of MCNP6 (scheduled for release as MCNP6.2) are primarily based on improvements and additions drawn from the EPICS2014 data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Verification of High-Fidelity Neutronics Code PROTEUS for C5G7 Benchmark Problems

Under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the 3D whole core transport code PROTEUS has been developed at Argonne National Laboratory to perform the detailed 3D transport calculation for various reactor types including irregular or complex geometric reactors. PROTEUS includes two distinct transport solvers: SN and MOC. Among them, the MOC solver based on the extruded geometry, thus named MOCEX, can produce the faithful 3-D transport solutions by employing the rigorous and robust 2D/3D MOC formulation. This formulation overcomes the drawbacks of a full 3D MOC method requiring enormous memory and computational resources and resolves the instability issues of the 2D/1D method as well. The solution of MOCEX converges to that of a Monte Carlo code as the discretization of angle, space and energy is refined, unlike the 2D/1D method. As a verification and validation (V and V) effort of PROTEUS-MOCEX, this paper presents verification test results using the well-known 3D C5G7 benchmark problems by solving and comparing eigenvalues and pin powers with MCNP solutions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Study on Neutronics Calculation Method of SCWR with Annular Fuel Assembly

Supercritical Water-Cooled Reactor (SCWR) is one of Gen-IV reactor systems for its advantages over LWRs, such as high thermal efficiency, simplified reactor system and economics. Annular fuel was proposed by MIT to supplant conventional solid fuel and improve power density in LWRs. Annular fuel is cooled from both inside and outside of the fuel. Thus a lower fuel temperature and cladding surface temperature can be achieved with higher power density. SCWRs will benefit from this advantage because high cladding temperature (usually over 650 degree C) is one challenge of SCWRs. Using annular fuel in SCWRs is very promising. This is also because the inner hole of the annual fuel can provide extra neutron moderation. Supercritical water density is too low to provide enough moderation. Different analysis codes have been developed all over the world to carry out SCWR pre-conceptual core design. Yamaji et al. developed SRAC/ASMBURN/COREBN to study Japan version of SCWR. Monti et al. used ERANOS to analyze HPLWR (High Performance Light Water Reactor). WIMS-D4/CITATION was also used to calculate HPLWR by Tashakor et al.. These codes employ a two-step method used in PWRs calculation. Lattice/Assembly calculation is carried out as the first step. In the second step, few group constant of different fuel assembly types is then used in core calculation. MCNP had also been used to study SCWR by Watta et al.. Due to large computation time cost, it's used mostly to analyze fuel assembly or core performance rather than pre-conceptual design, in which much optimization is needed. In this research, a neutronics calculation code for SCWR is developed using two-step method. This code is verified by comparing with Monte-Carlo code. Fuel assembly homogenization method is studied to improve the code accuracy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Protection factors for French dwellings - 25573

IRSN's current method of assessment of public exposure from atmospheric releases of radioactivity from nuclear installations assumes an exposed person outside of any building; even though the majority of time is spent inside. A more realistic assessment would take into account the protection from atmospheric contamination and deposits provided by dwellings. Therefore, IRSN carried out a study to estimate protection factors for gamma ray emitting radionuclides. This was done with the MCNP calculation code and assumed several French dwelling types. Air contamination (outside and inside) and deposits on soil (outside and inside) and on a roof are considered. Method and results are set out for a few radionuclides. The results show the influence of dwelling type, and also of room sealing (considering inner contamination or not) in case of radiological emergency. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

Generating shield response functions using a source independent method - 25587

Generalized questions about shield materials are common and yet difficult to answer analytically. Most forms of quick analysis use simple attenuation. The drawback to such methods is the use of mono-energetic averaged cross sections. Averaging a cross section requires knowledge of the flux a priori. If the flux was known, then some more in-depth calculation was performed negating the need for quick scoping calculations. Source independent methods for assessing the transport medium provide a means of creating response functions that yield general statements about the effectiveness of a shielding material. The following paper uses MCNP 6.2 to generate source independent, energy dependent, response functions for the transport medium that allow for quick scoping of effectiveness of various neutron shields. The source independent method here leveraged FT SCX tally treatment, to assess the performance of various shields such that energy response is preserved. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

Testing of ENDF/B-VIII.0 in the GNDS format with LLNL transport codes - 25606

ENDF/B-VIII.0 was released in the Evaluated Nuclear Data Format and the Generalized Nuclear Data Structure format (GNDS), a new international standard for storing nuclear data developed under the auspices of the Nuclear Energy Agency Working Party on Evaluation Cooperation Subgroup 38. GNDS relies on modern coding practices, and offers a structured, flexible and physics-based standard for storing and exchanging nuclear data to address limitations in legacy nuclear data formats such as ENDF-6. GNDS supports the storage of evaluated nuclear data as well as processed data suitable for both deterministic (i.e., multi-group) and Monte Carlo transport codes. Lawrence Livermore National Laboratory has implemented a complete suite of codes to test GNDS data. Simulations for the ENDF/B-VIII.0 cross-section library in GNDS format are presented. Deterministic and Monte Carlo radiation transport simulations with the Ardra and Mercury codes respectively, are validated against reactivities in fast criticality safety benchmark assemblies. Finally, Mercury simulations of {sup 238}U(n,f), {sup 237}Np(n,f) and {sup 239}Pu(n,f) reaction rates in Godiva, Jezebel and Flattop are compared to MCNP results and measured data. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

Distinguishing Research Reactor Fuels Using Neutron Multiplicity - 25615

Neutron measurements are one of the reliable means of interrogation of fissile material. Neutron measurements are used for total counting (pure metal fissile samples) or counting coincidence/multiplicity (for impure fissile samples). Various flavors of neutron measurement are used routinely by the IAEA to determine the fissile mass of an unknown sample. The objective of this work is to design an experiment to distinguish the two different fuel types using neutron multiplicity method. The design of experiment is performed using Monte Carlo N- Particle (MCNP) transport code. General Electric (GE) designed 1-MWt pool-type research reactor at UMass Lowell (UMLRR) uses Materials Testing Reactor (MTR) plate-type fuel consisting of 19 full and 2 partial fuel assemblies, controlled by four boron control blades and one regulating control rod. Two different fuel types are available at the UMLRR facility. In 2008, UMass Lowell obtained slightly burnt fuel assemblies from the decommissioned WPI research reactor. The simulation results discussed in this paper show that within the limits of statistical error, the single and double moments for the WPI and UMLRR fuels can be measured. The two fuels have very similar composition and enrichments resulting in very similar neutron multiplicity moments. This implies that neutron multiplicity measurements alone would not be sufficient to distinguish between the two fuel types at low burnups. The change in the observed single-double ratio between each fuel type is less than 6 % which falls below the uncertainty in the efficiency of the JCC71 slab detectors. The neutron signature from low burnup and low enriched uranium fuel are not strong, however, these can be measured under low background and long measurement times. This would be an important consideration in designing the experiments and parallel gamma measurements will also be planned. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Estimating consistent biased source probabilities over arbitrary bins - 25621

This paper describes computational algorithms for estimating consistent biased source probabilities for the CADIS and FW-CADIS methods that are compatible with most of the general fixed-source capabilities of the MCNP continuous-energy transport code. The basic algorithm is a novel Monte Carlo sampling technique that enables the straightforward estimation of biased source probabilities over arbitrarily complex distribution trees. In recent work, a method has been developed for generating correct biased spatial source probabilities in the presence of rejection from an arbitrary number of geometry cells. Two numerical test problems that demonstrate the correctness of this new algorithm are presented, including a simplified model of a cooling water system with an N-17 neutron source. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

Application of the GBFP Method to Electron Partial-Wave Expansion Elastic Scattering Differential Cross Sections Within the Geant4 Toolkit - Paper 69

Analog Monte Carlo simulation of electrons in complex material geometries is generally impractical because the long range Coulomb interactions result in highly peaked differential cross sections (DCS) for scattering and energy-loss and extremely small collision mean free paths (mfp). The condensed history (CH) Monte Carlo method attempts to alleviate this limitation by advancing electrons in fixed steps and sampling angular deflections and energy losses from precomputed distributions obtained from approximate solutions to the underlying transport equation that are valid only for thin layers. Though CH is efficient and is widely employed in production codes (ITS, MCNP, EGS, Geant), step size restrictions and inconsistent handling of material and free surface boundaries limit the ultimate accuracy possible with this approach, especially in applications to highly heterogeneous media. Recently proposed alternatives to the CH method, such as the Generalized Boltzmann Fokker-Planck (GBFP) method, restore the correct transport mechanics by simulating single events based on a Boltzmann transport equation but with efficiency and accuracy achieved through the use of reduced-physics interaction models. Motivated by Lewis theory, various moment-preserving strategies are employed in the GBFP approach to construct DCSs that are less singular, and with mfps shorter than analog. Specifically, the approximate cross sections are constructed to rigorously preserve a finite number of angular and energy-loss moments of the analog DCSs while all higher moments are approximated in terms of the exact lower moments. By systematically increasing the number of moments retained, any desired accuracy can be obtained for integrated quantities, such as dose, reflection and transmission yields, as well as distributed quantities such as energy and angular distributions. For relativistic Rutherford cross section models, the GBFP method has been demonstrated in Monte Carlo simulation to yield accuracy comparable to analog but at a fraction of the cost. One advantage of the GBFP approach is that angular deflection and energy-loss interactions depend only on Legendre moments and energy-loss moments and not the detailed form of the analog DCSs. This feature allows moment-preservation to be enforced through DCSs expressed as a superposition of discrete (i.e., using delta-functions) angular deflections and discrete energy losses. This promotes extremely efficient computation when compared to analog and completely continuous or hybrid discrete-continuous moment-based DCS reconstructions, while yielding highly accurate results for integrated and distributed quantities. In this paper, we demonstrate that the moment-preserving algorithms central to our fully discrete GBFP method can be readily extended to apply to tabulated DCS data, specifically to the partial-wave expansion (PWE) elastic scattering DCSs. Since data libraries for the PWE DCSs for energy and material combinations of interest can be prohibitively large, the GBFP method provides a means to reduce data storage requirements. We implemented our algorithm in the Geant4 Monte Carlo code and all numerical results are obtained using this implementation. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

Methodological Aspects of Shielding and Activation Analyses for ITER Diagnostics Systems - Paper 125

It is expected that the diagnostics systems of the ITER tokamak will provide valuable measurement data about the first in its kind D-T fusion plasma with emission of 14 MeV neutrons. These systems must withstand the neutron and secondary photon radiation along the ITER operation. Under the neutron irradiation the ITER materials are getting radioactive and therefore an adequate shielding must be provided along all the time of ITER operation and shut-down for the reliable machine operation and for the safe personnel access to its components, such as the diagnostics ports. These are the prerequisites for the shielding and activation analysis of the ITER diagnostics systems discussed in this paper. This paper provides insight into the current tasks of neutronics support for the development of ITER diagnostics systems. The tasks themselves range from relatively straightforward ones such as Monte Carlo radiation transport calculations for shielding analyses of the diagnostics systems, to more advanced tasks of Shut-Down Dose Rate (SDDR) analyses for which the dedicated Rigorous 2-Step mesh-based (R2Smesh) and Direct 1-Step (D1S) methodologies were applied. The accuracy and precision of the ITER neutronics analysis is dependent on the methodological approaches and the implying modeling assumptions. In addition to the SDDR challenges, another distinguishing feature of fusion neutronics is a wide use of CAD-based neutronics modeling to design components of fusion devices such as ITER and DEMO. Basically, there are two possibilities of CAD geometry settings in Monte Carlo (MC) radiation transport models, either by the geometry format conversion of CAD models into the format of MC models by means of the interface programs such as McCad and MCAM or by the direct use of CAD geometry engine in MC calculations such as it was realized in the DAGMC code approach. Several conversion interfaces and the DAGMC software library have been tested on the 3D ITER model developed for the CAD-MCNP benchmark. It was concluded that the considered approaches have reached the maturity level to allow their application to real ITER design analyses.

61 RADIATION PROTECTION AND DOSIMETRY↗

Implementation and Validation of Reference Sensitivity Profile Calculations in TRIPOLI 4 - 14247

In current neutron transport calculations, particularly on continuous-energy Monte-Carlo codes, the major source of uncertainty is coming from nuclear data. In order to improve nuclear data evaluations, integral experiments are used. Feedback is then precisely obtained by re-estimating cross-section levels and correlations according to their sensitivity. k{sub eff}-sensitivity coefficients are traditionally obtained using deterministic codes since hands-on methods were not yet implemented in continuous energy Monte Carlo codes. In this paper, an Iterated-Fission-Probability-based method newly developed in the continuous energy Monte Carlo code TRIPOLI-4 will be used to calculate sensitivity coefficient breakdown on isotopes and reactions and results will be compared to those of ERANOS deterministic code. Objective is twofold: to show the reliability of the new TRIPOLI-4 method and to identify where and why profiles are slightly different. Sensitivity profiles, forward and adjoint fluxes are thus meticulously compared using a fine 1968-group energy structure. An OECD international benchmark (Uncertainty Analysis for Criticality Safety Assessment - UACSA - Phase III) is taken as working example to enforce validation and to compare to independent results (e.g. MCNP-6). This benchmark is constituted of two bare spheres filled with homogeneous mixture of uranium and polyethylene (respectively 2% and 50% U-235 enrichments) and gives an overview of issues encountered in thermal reactor cores. Several energy-group structures (1968G, 172G, 33G) are also tested in ERANOS to identify multi-group and implicit effects. Lessons learnt are that continuous Monte Carlo methods seem to be mature and compare well with deterministic codes. Only damper the U-238 elastic - and on certain conditions the U-238 capture - sensitivity coefficient for which difference can reach 30% between TRIPOLI-4 and ERANOS. Indeed concerning U-238, implicit effects are emphasized and multi-group theory remains limited by hypothesis; even a 1968-group structure is not refined enough to properly calculate U-238 elastic scattering k{sub eff}-sensitivity. This reference 3D continuous-energy sensitivity capability in TRIPOLI4-IFP will be extensively used in the uncertainty automated tool RIB of the French Criticality-Safety package CRISTAL. (authors)

42 ENGINEERING↗