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At least 163 records · Page 9

Reply to Comment by Cole-dai Et Al. on "Climatic Impact of the Long-lasting Laki Eruption: Inapplicability of Mass-independent Sulfur Isotope Composition Measurements"

Here we respond to the comments by Cole-Dai et al. [2014] on our article Schmidt et al. [2012]. Specifically, in response to section 2 of their reply, we argued in Schmidt et al. [2012] that based on previously published estimates of the volatile release height during the 1783-1784 C.E. Laki eruption, the lack of a sulfur massindependent fractionation (MIF) anomaly is expected. In other words, no previous study on Laki ever argued that this eruption emitted SO2 into altitudes >13-15 km. In section 2.3, Cole-Dai et al. [2014] argue that the nonzero Δ33S value of their Laki sample 1 may be explained by a short-lived explosive phase at Laki during which volatiles reached the stratosphere. In Schmidt et al. [2012] in section 2, we argued in agreement with Cole-Dai et al. [2014] (section 3.1) that for a MIF anomaly to be preserved, the Laki volatiles would have had to be emitted in >20 km altitude. Our main point is that eruption column heights >20 km are unlikely based on the historical accounts and plume-rise modeling for the Laki eruption [Stothers et al., 1986; Woods, 1993; Thordarson and Self, 2003]. In Schmidt et al. [2012], we argued that to deduce a short-lived climatic impact of the Laki eruption based on the lack of a MIF anomaly and the length of the sulfate deposition in Greenland ice cores may be misleading because the climatic impact will outlast the radiative forcing of the Laki aerosol cloud. Cole-Dai et al. [2014] acknowledge the latter in their reply in section 4.2. We agreewith Cole-Dai et al. [2014] in that themagnitude and length of the climatic impact during the winter of 1783-1784 depends on the altitude of the volatile release during the eruption (sections 2.3 and 4.2). However, even if we assumed that during Laki all sulfur dioxide (SO2) would have been released in the troposphere, then the aerosol cloud would still be present in the upper troposphere during March 1784, as is evident from independent model simulations of this "tropospheric-only" scenario [Stevenson et al., 2003]. We acknowledge that there is uncertainty on the volatile release height for Laki; however, it is worth considering that those climate model simulations that used an injection altitude between 9 km and 13 km for the Laki SO2 [Highwood and Stevenson, 2003; Oman et al., 2006a, 2006b; Schmidt et al., 2012] best match the observed temperature changes during summer of 1783 [Angell and Korshover, 1985; Brázdil et al., 2010; Briffa et al., 1998; D'Arrigo and Jacoby, 1999; Jacoby et al., 1999; Kington, 1988; Manley, 1974; Parker et al., 1992; Thordarson and Self, 2003]. Based on these model simulations, a climatic impact during the winter of 1783-1784, albeit weaker than during the climactic phases of Laki, is expected (and our argument here does not exclude the role of natural variability in contributing to the cold winter of 1783-1784 as discussed in Schmidt et al. [2012]). Therefore, we continue to argue that for high-latitude eruptions such as Laki, the applicability of sulfur isotopic measurements to interpret the climatic relevance has yet to be demonstrated. Itmay transpire that the interpretation of MIF signals for the climate-relevance of an eruption is valid and unambiguous only for short-lived explosive eruptions in the tropics. In terms of the processes producing a MIF anomaly (section 3.3 in Cole-Dai et al. [2014]), the works by Hattori et al. [2013] and Ono et al. [2013] suggest that there are remaining issues not discussed by Cole-Dai et al. [2014], for instance, self-shielding of SO2 due to high column densities typical for eruptions of Pinatubo-scale and greater, and the preservation of the MIF signature in general.

general circulation model↗

Science Instrument Sensitivities to Radioisotope Power System Environment

Radioisotope Power Systems (RPS) have been and will be enabling or significantly enhancing for many missions, including several concepts identified in the 2011 Planetary Science Decadal Survey. Some mission planners and science investigators might have concerns about possible impacts from RPS-induced conditions upon the scientific capabilities of their mission concepts. To alleviate these concerns, this paper looks at existing and potential future RPS designs, and examines their potential radiation, thermal, vibration, electromagnetic interference (EMI), and magnetic fields impacts on representative science instruments and science measurements. Radiation impacts from RPS on science instruments are of potential concern for instruments with optical detectors and instruments with high-voltage electronics. The two main areas of concern are noise effects on the instrument measurements, and long-term effects of instrument damage. While RPS by their nature will contribute to total radiation dose, their addition for most missions should be relatively small. For example, the gamma dose rate from one Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) would be an order of magnitude lower than the environmental dose rate at Mars, and would have a correspondingly lower contribution to instrument noise and to any permanent damage to payload sensors. Increasing the number of General Purpose Heat Source (GPHS) modules used in an RPS would be expected to increase the generated radiation proportionally; however, the effect of more GPHS modules is mitigated from a strictly linear relationship by self-shielding effects. The radiation field of an RPS is anisotropic due to the deviation of the modules from a point-source-geometry. For particularly sensitive instruments the total radiation dose could be mitigated with separation or application of spot shielding. Though a new, higher-power RPS could generate more heat per unit than current designs, thermal impact to the flight system could be mitigated with shading and pointing if required by the mission. Alternatively, excess heat could prove beneficial in providing needed heat to spacecraft components and instruments in some thermal environments. Vibration for a new higher-power Stirling Radioisotope Generator (SRG) would be expected to be similar to the recent Advanced Stirling Radioisotope Generator (ASRG) design. While vibration should be low, it must be considered and addressed during spacecraft and instrument design. EMI and magnetic fields for new RPS concepts are expected to be low as for the current RPS, but must be considered and addressed if the mission includes sensitive instruments such as magnetometers. The assessment conducted for this paper focused on orbiter instrument payloads for two representative mission concepts- a Titan Saturn System Mission (TSSM) and a Uranus Orbiter and Probe (UOP)-since both of these Decadal Survey concepts would include many diverse instruments on board. Quick-look design studies using notional new RPS concepts were carried out for these two mission concepts, and their specific instrument packages were analyzed for their interactions with new RPS designs. The original Decadal Survey TSSM and UOP concepts did not have complete instrument performance requirements so typical measurement requirements were used where needed. Then, the general RPS environments were evaluated for impacts to various types of instruments. This paper describes how the potential impacts of the RPS on science instruments and measurements were assessed, which impacts were addressed, proposed mitigation strategies against those impacts, and provides an overview of future work.

RPS↗

A Physical Model for a Slim Convective Dusty Disk in AGN

An accretion disk in an Active Galactic Nucleus (AGN) harbors and shields dust from external illumination. Our model shows that at the mid-plane of the disk around a MBH = 107M black hole, dust can exist at 0:1pc from the black hole, compared to 0.5pc outside of the disk where such self-shielding is constrained. We construct a physical model of a disk region approximately located between the radius of dust sublimation at the disk mid-plane and the radius at which dust sublimes at the disk surface. Our main conclusion is that for a wide range of model parameters such as local accretion rate and/or opacity, the accretion disk’s own radiation pressure on dust significantly influences its vertical structure. In this region convection plays important role in the vertical transport of energy. When the local accretion rate exceeds 2:5M yr the 10􀀀2pc-scale disk is super-critical with respect to dust opacity. Such a disk puffs up and transforms from geometrically thin to slim. Our model fits into the narrative of a "failed wind" scenario of Czerny & Hryniewicz (2011) and the "compact torus" model of Baskin & Laor (2018), incorporating them as variations of the radiative dusty disk model.

A Dorodnitsyn↗

Coordinated analyses of hydrated Kuiper Belt dust

We report on coordinated mineralogical, chemical, and isotopic analyses of hydrated carbon-rich interplanetary dust particles (IDPs). We focus on hydrated IDPs that contain high solar flare track densities consistent with an origin in outer Solar System bodies such as Edgeworth-Kuiper belt objects (EKBOs) [1]. We are searching for evidence of the heavy water reservoir predicted by self-shielding models [2]. The IDPs were embedded in sulfur and microtome thin sections were obtained for Fourier-transform infrared (FTIR), transmission electron microscopy (TEM), and nanoSIMS analyses. The IDPs were only partly microtomed and the remainder was extracted from the sulfur and pressed into In or Au for high precision O isotopic analyses. The hydrated IDPs in this study are compact low-porosity objects with a mineralogy dominated by fine-grained saponite and lesser serpentine, Fe-Ni sulfides, Mg-Fe carbonates, rare crystaline silicates, and abundant organic matter. The C in the IDPs is abundant (typically >4X CI) and occurs as isolated nanoglobules and finely disseminated within the phyllosilicate matrix of the IDPs. TEM analyses of the nanoglobules show O-rich, S-depleted rims with higher carbonyl functionality than their cores. FTIR spectra show strong aliphatic C-H stretches consistent with the presence of short-chain aliphatic hydrocarbons in these IDPs. The oxygen isotopic compositions measured in 6 IDPs all plot within error of the CCAM line [3]. For several of the IDPs we were able to obtain multiple analyses. The ∆^(17)O values range from -23‰ up to +3.4‰ with most analyses clustering above the terrestrial fractionation line. The O isotopic compositions and the high C contents of these IDPs are distinct from known hydrated carbonaceous chondrites. The track-rich hydrated IDPs record parent body aqueous alteration effects through interactions with 16O-poor H2O on a EKBO parent body. The aqueous alteration activity likely resulted from the heat generated by collisional processes among EKBOs given that these objects are believed to have accreted late after the decay of 26Al [4].

Lindsay P. Keller↗

Bulk Oxygen Isotopic Compositions of Anhydrous Interplanetary Dust Particles: Indication of an 16O-Poor Reservoir In The Outer Solar System

The oxygen isotopic ratios of ancient solar system materials reflect the isotopically distinct reservoirs in which they formed or were altered. These compositions are generally explained by mixing of material derived from an 16O-rich reservoir and material from an 16O-poor reservoir thought to be isotopically heavy H2O. Evidence of this 16O-poor reservoir is scarce and includes cosmic symplectite (COS), probable silicates in an interplanetary dust particle (IDP), and sulfates in a cometary xenolith. These materials have 17O and 18O enrichments between ~150–200 ‰, which agree with predictions for self-shielding and photodissociation of CO. The ‘cometary’ 16O-poor materials indicate alteration or formation in the outer solar system. The bulk O isotopic compositions of IDPs thought to have cometary origins may also harbor evidence of this primordial 16O-poor reservoir. Here we report the bulk O compositions of anhydrous IDPs and discuss the nebular reservoirs from which they derived

A N Nguyen↗

Measurement of the 235U Induced Fission Gamma-ray Spectrum as an Active Non-destructive Assay of Fresh Nucleear Fuel

Non-destructive assay is a powerful tool the International Atomic Energy Agency (IAEA) employs to verify adherence to safeguards agreements. Current IAEA veri- cation techniques for fresh nuclear fuel include passive gamma-ray spectroscopy to determine fuel enrichment. This technique suers from self-shielding and lakes the percision to detect diversion of central fuel rods. The aim of this research is to develop a new, more capable non-destructive analysis technique using active neutron interroga- tion of fuel assemblies and determining the yields of short-lived ssion products from high-resolution gamma-ray spectroscopy using high-purity germanium (HPGe). This paper reports results from irradiation of a one meter tall mock fresh fuel assembly with low enriched uranium (LEU) or depleted uranium (DU) rods using a down-scattered deuterium-tritium (D-T) neutron source. Both prompt and delayed gamma-ray spec- tra were collected as time-stamped list-mode data in a coax detector and without list mode data in a planar strip detector. No dierentiating signatures were observed in the prompt spectra in either detector; however, both detectors observed several short-lived ssion product signatures in LEU and not DU fuel, indicating that this technique has potential for determination of enrichment of fresh fuel assemblies. There were eight unique ssion products observed in the LEU spectra with the coax detector spectra, and three ssion products were observed in the LEU spectra with the strip detector.

Fission Gamma-ray Spectrum↗

ADVANCED TEST REACTOR (ATR) CONVERSION TO LOW ENRICHED URANIUM (LEU) FUEL

The United States Department of Energy (DOE), National Nuclear Security Administration (NNSA), Office of Material Management and Minimization (MMM) is working to convert research reactors globally from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. MMM and its predecessor programs have converted or verified the shutdown of 88 HEU research reactors and isotope production facilities, using LEU fuel developed in the 1980s. However, there is a small set of high performance research reactors which require a new high density LEU fuel for conversion, including six U.S. high performance research reactors (USHPRRs), which include the Advanced Test Reactor (ATR) and the ATR Critical (ATRC) Facility at the Idaho National Laboratory (INL) near Idaho Falls, ID; University of Missouri Research Reactor (MURR) at Columbia, MO; Massachusetts Institute of Technology Reactor (MITR) at MIT in Cambridge, MA; High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) near Oak Ridge, TN; and National Bureau of Standards Reactor (NBSR) at the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR) in Gaithersburg, MA. These reactors will share the same uranium-molybdenum (U-Mo) fuel plate design. The U-Mo fuel meat foil will have a zirconium (Zr) diffusion barrier with aluminum cladding. Three different application methods for the Zr diffusion barrier layer are being pursued – co-rolled, electroplating and plasma spraying. Fabrication of the higher density LEU U-Mo fuel has presented several challenges. Because the U-Mo fuel fabrication process uses the existing stockpile of 93% highly enriched uranium (HEU) to be blended with molybdenum and depleted uranium (DU) to form 19.75% LEU, cost benefits of enriching to 19.75% will not be realized. Homogeneity of the U-235 and molybdenum in the metal foil is a challenge due to U-Mo casting process results in variations between adjacent cast U-Mo plates. LEU will not require less U-235 as previously believed due to self-shielding properties of U-238 requiring additional U-235 to be loaded in each fuel element. Although the overall dimensions of the ATR fuel element remain unchanged, the weight of the LEU ATR fuel element is 72% greater than a HEU ATR fuel element. During waste recovery and recycling efforts, separation of aluminum cladding from the HEU fuel meat is a relatively simple chemical dissolution, however, separation of the Zr diffusion barrier from the U-Mo fuel meat is significantly more challenging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Summary: Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-group Libraries for Advanced Reactor Analysis

The SCALE/AMPX multigroup (MG) cross section processing procedure has been updated to minimize reactivity differences for various advanced thermal and fast reactor designs, as observed for the current MG libraries, resulting in excellent agreement between the calculations with the new MG libraries and the continuous-energy reference calculations. The SCALE MG calculations are widely applied to thermal spectrum light-water reactor systems, as well as fast spectrum metallic systems. Due to growing interest from industry and regulators in applying SCALE for the design of fast spectrum reactors—both sodium and molten salt—it was desirable to review and improve the SCALE/AMPX procedure for unresolved resonance self-shielded data and high-energy neutron spectra. The data were improved by generating MG unresolved resonance data based on the analytic probability table method with the narrow resonance approximation and by using very fine and intermediate group structures that are typical for fast system analysis. This study focused on verifying the improved SCALE/AMPX MG cross section processing procedure and the new AMPX 1597-group library with the ENDF/B-VII.1 and VIII.0 evaluated nuclear data. The verification was made by performing reaction rate analysis and benchmark calculations for various thermal and fast reactor systems. Results indicate that the improved SCALE/AMPX MG cross section processing and libraries provide excellent results for advanced reactor analysis.

Kim, Kang Seog↗

Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-Group Libraries for Advanced Reactor Analysis

The SCALE/AMPX multigroup (MG) cross section processing procedure has been updated to minimize reactivity differences for various advanced thermal and fast reactor designs, as observed for the current MG libraries, resulting in excellent agreement between the calculations with the new MG libraries and the continuous-energy reference calculations. The SCALE MG calculations are widely applied to thermal spectrum light-water reactor systems, as well as fast spectrum metallic systems. Due to growing interest from industry and regulators in applying SCALE for the design of fast spectrum reactors—both sodium and molten salt—it was desirable to review and improve the SCALE/AMPX procedure for unresolved resonance self-shielded data and high-energy neutron spectra. The data were improved by generating MG unresolved resonance data based on the analytic probability table method with the narrow resonance approximation and by using very fine and intermediate group structures that are typical for fast system analysis. This study focused on verifying the improved SCALE/AMPX MG cross section processing procedure and the new AMPX 1597-group library with the ENDF/B-VII.1 and VIII.0 evaluated nuclear data. The verification was made by performing reaction rate analysis and benchmark calculations for various thermal and fast reactor systems. Results indicate that the improved SCALE/AMPX MG cross section processing and libraries provide excellent results for advanced reactor analysis.

Kim, Kang Seog↗

PROGRESS ON 140,142CE NEUTRON CROSS SECTION RESOLVED RESONANCE REGION EVALUATIONS

Oak Ridge National Laboratory is working on completing the resonance parameter evaluation of 140,142Ce in the neutron energy range up to 200 keV. Performed with the support of the US Nuclear Criticality Safety Program, this evaluation aims to generate high-fidelity cerium cross-section and covariance data. A point-wise representation of the cross sections derived from the resonance parameters will provide improved calculations of self-shielding factors for nuclear criticality safety applications and additional evaluation support for continuous-energy radiation transport methodologies. The evaluation procedure uses the Reich Moore approximation of the R-matrix theory implemented in the SAMMY code system to fit resonance parameters to high-resolution transmission and neutron capture measurements of natCe performed in 2016 by the JRC-GEEL instrument scientists at the Geel Linear Accelerator facility as well as other experimental data sets on both natural and highly-enriched cerium samples available in the experimental library EXFOR. In the analyzed energy range this work aims to improve and extend the resolved resonance region present in the latest US nuclear data library ENDF/B-VIII.0 for 140,142Ce isotopes. This paper will present the preliminary results of the R-matrix analysis based on recently measured natCe transmission data.

Chapman, Chris↗

VERIFICATION OF TRISO FUEL BURNUP USING MACHINE LEARNING ALGORITHMS

Pebble Bed Reactors are fueled with fuel pebbles that are circulated multiple times through the reactor vessel before discharge. During the normal operation of a PBR, ejected pebbles are returned to the reactor or discharged depending on the fuel burnup and physical condition of the pebbles. The burnup measurement is usually based on detected radiation signatures of fission products accumulated in the pebble fuel over burnup. Previous research has shown that height of photopeaks of fission products, such as 134Cs, 137Cs, 154Eu, etc., can be used independently or in combination to infer or predict the level of burnup in the fuel. However, it remains challenging to measure such complex sources due to self-shielding effects, strong radiation background and intervening materials. Another operational challenge is the required high throughput of burnup measurement, which necessitates limited measurement time and thus impacts quality of measured gamma-ray spectra. Hence, advanced spectral analysis methods are needed to analyze the noisy gamma spectra and predict the burnup values. We propose to use machine learning (ML) method to interpret gamma-ray spectra and predict the burnup values of the pebbles. ML has achieved widespread success and adoption across a few domains that require pattern recognition and analysis in varied data types. In this work, we apply three proven ML approaches - multilayer perceptrons, convolutional neural networks, and transformers - to the task of predicting fuel burnup from measured gamma spectra, and compile a dataset of simulated spectra for training and validation of the ML models. In this paper, we will discuss the network architecture of these three ML approaches and compare the performance of the simplest of these (MLP) to a standard linear regression.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Initial study on cross section generation requirements for a PBR equilibrium core

A Serpent model of the HTR-PM equilibrium core was developed for use in cross section prepa-ration studies in order to guide methods development for the Griffin reactor multiphysics applica-tion. The model includes detailed isotopics for 10 distinct pebble burnup groups in 126 core zoneswith unique fuel and moderator temperatures obtained from a coupled neutronics-thermal-fluidsequilibrium core calculation using Griffin-Pronghorn. A sensitivity study of the fuel and mod-erator temperatures for various core regions was performed with the MOOSE stochastic tools.The results show that the uncertainties are, not unexpectedly, dominated by the value of the fluidtemperature and that the power level, heat transfer coefficient and effective conduction to neigh-boring pebbles and fluid constitute, at best, second order effects. The temperature uncertaintyrange varies from 28 K to 57 K at the core entry and exit planes, respectively, but these val-ues are probably higher. We still have to quantify the significance of these uncertainties in thepreparation of cross sections, which will be postponed for future work. In addition, we verifythat the single effective pebble approach works well for the preparation of region averaged crosssections in the infinite domain approximation. Nevertheless, there are significant discrepanciesin the cross sections when compared to the multi-pebble model. This could affect the predictionof peak values and in the depletion calculation. We conclude that is highly desirable for futurestudies with Griffin to be able to handle both the ?effective? pebble approximation and the multi-pebble approach for the various pebble burnup groups. This enables Griffin users the flexibilityto perform higher-fidelity studies. Finally, we initiate the preparation of cross sections for variouscore regions from the full core Serpent reference model. We quantify the differences in 26 groupcross sections from infinite domain models. These reference cross sections will serve to validatethe double heterogeneity, self-shielding, and spectrum-correction methods in Griffin.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

SENSITIVITY STUDIES OF HEATER ROD DESIGN FOR TRANSIENT CRITICAL HEAT FLUX EXPERIMENTS IN TREAT

An important area of reactor safety is the cladding-to-coolant heat transfer under fast transient conditions. These conditions can occur during events such as a reactivity-initiated accident (RIA). In this paper we present the work performed under a laboratory directed research and development (LDRD) project, headed by Idaho National Laboratory (INL). The goal was to develop a test matrix to inform on the design of a borated steel heater rod that will induce critical heat flux (CHF) within an experimental capsule filled with water, using the TREAT facility, when submitted to a power pulse. The energy deposition in the rod and occurrence of CHF were identified as the most important key Figures of Merit (FoMs) used in the design process. Energy deposition was varied by using different pulse characteristics and various boron concentrations in the rod. A power coupling factor (PCF) was determined for each boron concentration. A sensitivity study determined the lowest limiting PCF needed to induce CHF in water with different degrees of subcooling while keeping pressure constant. Boron concentrations between 0.1-1.54 wt.% were considered. Additionally, the value of CHF is also known to increase during a rapid transient. A CHF multiplier sensitivity study focused on an example case concluded that CHF is no longer exceeded when the value is between 2.6- and 2.9-times higher. Lastly, a self-shielding study was performed to determine whether a borated tube could be used in place of a solid borated rod. The study determined that the inner 1mm of the rod can essentially be excluded or instrumented without heat generation penalties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

New Approach in Fissile Assay Using LSDS - 17076

A quantitative assay of isotopic fissile materials (U{sup 235}, Pu{sup 239}, Pu{sup 241}) was conducted at the Korea Atomic Energy Research Institute (KAERI), using a lead slowing down spectrometer (LSDS). Using a uranium rod and plutonium, the actual measurement of induced fission neutrons was performed in the RPI (Rensselaer Polytechnic Institute) LSDS facility. In the simulation, a content of U{sup 235}, Pu{sup 239}, and Pu{sup 241} was changed. The detected signal has a direct relationship with the mass of the existing fissile isotopes. The mathematical model for fissile assay was applied to the measured and simulated detection data. The assay results show that the isotopic content was consistent with 2 to 3% uncertainty for Pu{sup 239}. The simulation indicated that the assay result was affected by the self-shielding correction. Usually, in highly enriched plutonium, the correction plays an important role and good understanding of it is required in order to increase assay accuracy. The combination of LSDS technique measurement and simulation is a very powerful and direct way to analyze the isotopic fissile content. The information provided by such analysis is applicable to nuclear fuel cycle development and spent fuel management for safety and economics. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Reactor physics benchmark experiments at the JSI TRIGA MARK II reactor - Current status and future outlook

Full text of publication follows. With the development of new high-fidelity computational methods, improvement of nuclear data, and multiphysics modelling, there is an increased need for benchmark experiments to experimentally validate the models, methods and input data. Many of the nuclear facilities designed to perform reactor physics benchmark experiments have been shut down. Therefore, research reactors offer a great opportunity for benchmark experiments, if they are well designed and performed with great care and accuracy. In this presentation we provide an overview of the past and ongoing activities related to benchmark experiments at the Jozef Stefan Institute TRIGA Mark II research reactor. The following experiments have been performed: criticality with fresh fuel, {sup 197}Au(n,γ) and {sup 27}Al(n,α) reaction rates in irradiation channels, absolute and relative {sup 197}Au(n,γ), {sup 235}U(n,f) and {sup 238}U(n,f) reaction rates in the core, burnup, kinetic parameters, control rod worth, isothermal reactivity coefficient, self-shielding, slow and fast (pulse) transients, nuclear heating, delayed and prompt gamma ray production, temperature profiles for multi-physics. Since the existing fleet of research reactors is ageing very rapidly and new experiments are needed, new research reactors should be designed and built to meet the needs of future advanced reactors, education and training, and other technologies in the coming years. We will review planned activities at the JSI TRIGA reactors and plans for the new research reactor in Slovenia. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

An influence of manufacturing tolerances on Pin-Cell k-infinity of MOX fuel using data from the FUBILA experiment program

An influence of manufacturing tolerances (MTs) was evaluated on the pin-cell k-infinity of MOX fuel through the random sampling of CASMO5 calculations. The data from the FUBILA experiment program was used as the manufacturing parameters. For the uncertainties of element/isotope mass fractions, their covariance matrices were calculated by the generalized least square method. The total k-infinity uncertainty was 120-250 pcm (percent mille). From the breakdown of k-infinity uncertainty per materials, the MTs of fuel pellet had a dominant influence. The individual influences were also evaluated for element/isotope mass fractions, an inner/outer diameter, and a density. Those of element/isotope mass fractions were less than a few dozen pcm. Since the perturbations of fuel pellet diameter and density caused large variations to the total amount of heavy metals and effect of spatial self-shielding, they had a negative correlation with the k-infinity perturbation. On the AG3 (Al-Mg alloy) over-cladding, the inner/outer diameter perturbations had both negative and positive correlations. The negative one was due to decrease in amount of light water and the positive one was due to increase in amount of AG3. Since the effect of neutron slowing-down by light water had a dominant influence, the influence of the former negative correlation was larger than that of the latter positive correlation. It is, therefore, concluded that the consideration of the MTs that had a large influence on the effect of neutron slowing-down is important for precise quantification of the pin-cell k-infinity uncertainty of MOX fuel. (author)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

3D full core BEAVRS using OpenMOC with transport equivalence

Using an optimized implementation of the three-dimension method of characteristics for neutron transport, as well as a novel equivalence method for transport calculations de- signed to recover self-shielding errors from neglecting the angular dependence of reso- nant group absorption, a 3D full core light water reactor hybrid stochastic-deterministic eigenvalue calculation was achieved. This paper presents the optimizations developed and compares the transport solutions obtained. It realizes a runtime near 6,000 CPU hours for the statepoint, an improvement of an order of magnitude on previous works, with less than 1% error on pin fission/U-238 capture ratios and a few dozen pcms on the eigenvalue.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗