bluuebunny/crossref_metadata_2025_split huggingface dataset:
Neutron and Gamma Field Investigations in the VVER-1000 Mock-Up Concrete Shielding on the Reactor LR-0
Authors: Zaritsky S. , Egorov A. , Ošmera B. , Mařik M. , Rypar V. , Cvachovec F. , Kolros A. 
Published: May 2012

Abstract
Two sets of neutron and gamma field investigations were carried out in the dismountable model of radiation shielding of the VVER-1000 mock-up on the LR-0 reactor. First, measurements and calculations of the 3He(n,p)T reaction rate and fast neutrons and gamma flux spectra in the operational neutron monitor channel inside a concrete shielding for different shapes and locations of the channel (cylindrical channel in a concrete, channels with collimator in a concrete, cylindrical channel in a graphite). In all cases measurements and calculations of the 3He(n,p)T reaction rate were done with and without an additional moderator-polyethylene insert inside the channel. Second, measurements and calculations of the 3He(n,p)T reaction rate spatial distribution inside a concrete. The 3He(n,p)T reaction rate measurements and calculations were carried out exploring the relative thermal neutron density in the channels and its space distribution in the concrete. Fast neutrons and gamma measurements were carried out with a stilbene (45 × 45 mm) scintillation spectrometer in the energy regions 0.5–10 MeV (neutrons) and 0.2–10 MeV (gammas).
New Experimental Proposal for 235U PFNS Measurement to Answer a Fifty Year Old Question
Authors: Kornilov N. , Massey T. , Grimes S. , Voinov A. 
Published: May 2012

Abstract
The Prompt Fission Neutron Spectrum (PFNS) from 235U(n,f) is very important for various nuclear applications. It has been investigated in different experiments. In spite of ∼50 years of experimental efforts, a continuing conflict exists at thermal neutron energy. Microscopic experimental PFNS cannot describe macroscopic data. In this report we discuss the current status of this problem and suggest a new experiment, which could possibly resolve this problem.
A Three-Dimensional Methodology for the Assessment of Neutron Damage and Nuclear Energy Deposition in Graphite Components of Advanced Gas-Cooled Reactors
Authors: Morgan D. O. , Robinson A. T. , Allen D. A. , Picton D. J. , Thornton D. A , Shaw S. E. 
Published: May 2012

Abstract
This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK’s Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition.
The Current Status of the Shielding Integral Benchmark Archive and Database (SINBAD)
Authors: Kirk B. L. , Grove R. E. , Kodeli I. , Gulliford J. , Sartori E. 
Published: May 2012

Abstract
The shielding integral benchmark archive and database (SINBAD) collection of experiments descriptions was initiated in the early 1990s. SINBAD is an international collaboration between the Organization for Economic Cooperation and Development’s Nuclear Energy Agency Data Bank (OECD/NEADB) and the Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory (ORNL). SINBAD was designed to compile experiments and corresponding computational models with the goal of preserving institutional knowledge and expertise that need to be handed down to future scientists. SINBAD can serve as a learning tool for university students and scientists who need to design experiments or gain expertise in modeling and simulation. The SINBAD database is currently divided into three categories—fission, fusion, and accelerator experiments. Many experiments are described and analyzed using deterministic or stochastic (Monte Carlo) radiation transport software. The nuclear cross sections also play an important role as they are necessary in performing computational analysis.
Development of Neutron Measurement in Intense Gamma Field Using New Type of Nuclear Emulsion
Authors: Kawarabayashi J. , Ishihara K. , Takagi K. , Tomita H. , Iguchi T. , Naka T. , Morishima K. , Maeda S. 
Published: May 2012

Abstract
To precisely measure the neutron emissions from a spent fuel assembly of a fast breeder reactor, we formed nuclear emulsions based on a non-sensitized Oscillation Project with Emulsion tRacking Apparatus (OPERA) film with AgBr grain sizes of 60, 90, and 160 nm. The efficiency for 252Cf neutron detection of the new emulsion was calculated to be 0.7 × 10−4, which corresponded to an energy range from 0.3 to 2 MeV and was consistent with a preliminary estimate based on experimental results. The sensitivity of the new emulsion was also experimentally estimated by irradiating with 565 keV and 14 MeV neutrons. The emulsion with an AgBr grain size of 60 nm had the lowest sensitivity among the above three emulsions but was still sensitive enough to detect protons. Furthermore, the experimental data suggested that there was a threshold linear energy transfer of 15 keV/μm for the new emulsion, below which no silver clusters developed. Further development of nuclear emulsion with an AgBr grain size of a few tens of nanometers will be the next stage of the present study.
Sensitivity studies associated with dosimetry experiment interpretation
Authors: Bourganel S. , Soldevila M. , Ferrer A. , Grégoire G. , Destouches C. , Beretz D. 
Published: May 2012

Abstract
Interpretation of reactor dosimetry experiments with C/E comparison requires precise knowledge of parameters involved in modeling. Some parameters have more weight than others on the calculated values. So, sensitivity studies should be conducted to verify the importance of these parameters. The conclusions of these studies are used to refine the experiment modeling, or to correct uncertainty calculations. The results of these sensitivity studies allow a post-irradiation analysis, which can justify the discarding of some atypical C/M values. Derived uncertainties may be improved by the sensitivity analyses. Beyond classical parameters as geometry or composition, this paper describes some specific sensitivity studies conducted for dosimetry irradiation in reactor, and presents conclusions. These studies are based on dosimeters irradiated in the EOLE reactor facility at Cadarache CEA center. Conclusions drawn from these studies are generic and can be applied to any dosimetry study. Calculations performed for these studies were realized using TRIPOLI-4 Monte Carlo code.
Ringhals Unit 3 and 4—Fluence Determination in a Historic and Future Perspective
Authors: Green Eva-Lena , Roudén Jenny , Efsing Pål 
Published: May 2012

Abstract
The Ringhals site is situated on the Swedish southwest coastline. At the site, there are four operating nuclear power plants. Historically, the Swedish policy has been that the nuclear power plants were to be closed in 2010. The present position is to operate the units until their technical and economic lifetime has run out. The units shall be maintained and invested in to ensure a lifetime of at least 50 years, but the actions taken shall not limit the time to this date. When the initial surveillance capsules were evaluated, it was noted that the material properties of the weld material of unit 3 and 4 showed some deviations from the expected behaviour. Currently there is an extensive project running for re-evaluating the embrittlement situation from a long-term operating perspective. One part of the project is aimed at more accurately determining the fluence levels of the reactor pressure vessels (RPVs). The basis for the early evaluations of the dosimeters in the surveillance capsules and the corresponding fluence evaluation had an operating lifetime of 25 years as a target value. Therefore, the accuracy and refinement of the measurement and calculation were taken to be good enough to suit this life span. Looking back at the results from the dosimetry measurements there are a few discrepancies. Some of the dosimeters were disintegrated and some measurements had comparatively large uncertainties. When starting this project there were some re-evaluations done with the old fluence prediction model. For every new run and refinement there appeared new difficulties, and the decision was to start the evaluation from scratch.
Then there are two questions remaining regarding the fluence: What is the current fluence level? What will the resulting fluence be after 60 years of operation, when we have up-rated output power of both reactors? This paper aims to describe the view of the fluence evaluation in the surveillance program of the RPV, both in a historical and prospective view.
Measurements of Actinide-Fission Product Yields in Caliban and Prospero Metallic Core Reactor Fission-Neutron Fields
Authors: Casoli P. , Authier N. , Laurec J. , Bauge E. , Granier T. 
Published: May 2012

Abstract
In the 1970s and early 1980s, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on 233U, 235U, and 239Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments.
Application of Ex-vessel Neutron Dosimetry Combined with In-core Measurements for Correction of Neutron Source Used for Reactor Pressure Vessel Fluence Calculations
Authors: Borodkin P. G. , Borodkin G. I. , Khrennikov N. N. , Konheiser J. 
Published: May 2012

Abstract
This paper deals with calculated and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Neutron activation measurements analyzed in the paper were carried out in an ex-vessel air cavity at different nuclear power plant units with VVER-1000 during different fuel cycles. The time-integrated neutron source distributions used for DORT calculations were prepared via two different approaches based on (a) calculated fuel burnup (standard routine procedure) and (b) in-core measurements by means of self-powered detectors (SPDs) and thermocouples (TCs) (new approach). Considering that fuel burnup distributions in operating VVER may be evaluated now by the use of analytical methods (calculations) only, it is necessary to develop new approaches for the testing and correction of calculated evaluations of a neutron source. The results presented in this paper allow one to consider the reverse task of the alternative estimation of fuel burnup distributions. The proposed approach is based on the adjustment (fitting) of time-integrated neutron source distributions, and thus fuel burnup patterns, in some part of the reactor core, taking into account neutron leakage measurements, neutron-physical calculations, and in-core SPD and TC measurement data.
The Neutron Standard Fields at the BR1 Reactor at SCK•CEN
Authors: Wagemans J. , Malambu E. , Borms L. 
Published: May 2012

Abstract
The BR1 research reactor at SCK•CEN is characterized by a wide variety of irradiation possibilities, a large reactor core, and strong flexibility in its operation. A full MCNP model of BR1 has been recently developed in order to complement the results that can be obtained from activation dosimetry. After a general presentation of the reactor, this paper pays particular attention to its standard 235U(n,f) fast neutron field MARK III. This irradiation field is a useful tool for integral measurements and for detector calibrations. With the support of MCNP calculations, irradiations in MARK III can be directly referred to the pure 235U(n,f) fast neutron spectrum.
Development and Experimental Validation of a Calculation Scheme for Nuclear Heating Evaluation in the Core of the OSIRIS Material Testing Reactor
Authors: Malouch F. 
Published: May 2012

Abstract
The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron–photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation.
Ex-Vessel Neutron Dosimetry Analysis for Westinghouse 4-Loop XL Pressurized Water Reactor Plant Using 3D Parallel Discrete Ordinates Code RAPTOR-M3G
Authors: Chen J. , Alpan F. A. , Fischer G. A. , Fero A. H. 
Published: May 2012

Abstract
Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (&gt;1.0 MeV) fluence exposure to reactor pressure vessel in the beltline region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locations and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure.
An Alternative Calibration Method for Counting P-32 Reactor Monitors
Authors: Quirk T. J. , Vehar D. W. 
Published: May 2012

Abstract
Radioactivation of sulfur is a common technique used to measure fast neutron fluences in test and research reactors. Elemental sulfur can be pressed into pellets and used as monitors. The 32S(n, p)32P reaction has a practical threshold of about 3 MeV and its cross section and associated uncertainties are well characterized [1]. The product 32P emits a beta particle with a maximum energy of 1710 keV [2]. This energetic beta particle allows pellets to be counted intact. ASTM Standard Test Method for Measuring Reaction Rates and Fast-Neutron Fluences by Radioactivation of Sulfur-32 (E265) [3] details a method of calibration for counting systems and subsequent analysis of results. This method requires irradiation of sulfur monitors in a fast-neutron field whose spectrum and intensity are well known. The resultant decay-corrected count rate is then correlated to the known fast neutron fluence. The Radiation Metrology Laboratory (RML) at Sandia has traditionally performed calibration irradiations of sulfur pellets using the 252Cf spontaneous fission neutron source at the National Institute of Standards and Technology (NIST) [4] as a transfer standard. However, decay has reduced the intensity of NIST’s source; thus lowering the practical upper limits of available fluence. As of May 2010, neutron emission rates have decayed to approximately 3e8 n/s. In practice, this degradation of capabilities precludes calibrations at the highest fluence levels produced at test reactors and limits the useful range of count rates that can be measured. Furthermore, the reduced availability of replacement 252Cf threatens the long-term viability of the NIST 252Cf facility for sulfur pellet calibrations. In lieu of correlating count rate to neutron fluence in a reference field the total quantity of 32P produced in a pellet can be determined by absolute counting methods. This offers an attractive alternative to extended 252Cf exposures because it can be performed regardless of the characterization of the exposure environment. Count rates produced by sulfur pellets are correlated to the measured quantity of separated 32P. A posteriori spectral and cross section determination can be used to correlate the quantity of phosphorus back to a neutron fluence in a reference field. This paper outlines a method for the setup, calibration, and use of the detector systems, 32P sample preparation, and analysis of the beta spectrum. An uncertainty analysis and comparison to ASTM E265 is also included.
Comparison of Attenuation Coefficients for VVER-440 and VVER-1000 Pressure Vessels
Authors: Marek Milan , Rataj Jan , Vandlik Stanislav 
Published: May 2012

Abstract
The paper summarizes the attenuation coefficient of the neutron fluence with E &gt; 0.5 MeV through a reactor pressure vessel for vodo-vodyanoi energetichesky reactor (VVER) reactor types measured and/or calculated for mock-up experiments, as well as for operated nuclear power plant (NPP) units. The attenuation coefficient is possible to evaluate directly only by using the retro-dosimetry, based on a combination of the measured activities from the weld sample and concurrent ex-vessel measurement. The available neutron fluence attenuation coefficients (E &gt; 0.5 MeV), calculated and measured at a mock-up experiment simulating the VVER-440-unit conditions, vary from 3.5 to 6.15. A similar situation is used for the calculations and mock-up experiment measurements for the VVER-1000 RPV, where the attenuation coefficient of the neutron fluence varies from 5.99 to 8.85. Because of the difference in calculations for the real units and the mock-up experiments, the necessity to design and perform calculation benchmarks both for VVER-440 and VVER-1000 would be meaningful if the calculation model is designed adequately to a given unit.
A Database-Informed Approach to New Plant Shielding Design
Authors: Lloyd Timothy M. 
Published: May 2012

Abstract
To facilitate the definition and description of radiation dose rates in the numerous rooms and areas of a new nuclear power plant, a database approach was developed. This approach offers a number of benefits over more manual methods. A key benefit is that the selection of an appropriate shielding method to use in each area of the plant is greatly facilitated by virtue of the team’s improved ability to grasp the significance of each of the individual sources that are candidates for making a significant contribution to the dose rate in each area. By understanding the level of relevant contribution—if any—of each of these candidate sources, an analyst is able to select a method that will define the contribution without becoming enmired in a model representing inappropriately high degrees of accuracy and modeling time. This database method, by allowing for an evolving understanding of dose rates and sources in the neighboring rooms for each portion of the plant, leads to substantial reductions in the effort of characterizing a plant’s radiation environment. As an additional benefit, the database serves as a tool for documenting the shielding calculations themselves, automatically generating formatted sections including drawing and source term references, shielding calculation types, key dimensions, and results; these sections can form the starting point of a full calculation package. The approach offers a final project management benefit: estimating, tracking, and predicting the effort associated with the many calculations involved in such a project are greatly systematized, leading to more reliable manpower estimates.
Outline of Testing and Evaluation Program Used in Selection of Adhesives for Transparent Adhesive Joints in All-Glass Load-Bearing Structures
Authors: Weller Bernhard , Nicklisch Felix , Prautzsch Volker , Vogt Iris 
Published: May 2012

Abstract
Architects are increasingly demanding all-glass load-bearing structures with fully transparent adhesive joints. Usually, such structures are classed as nonregulated forms of construction. The Institute of Building Construction at Dresden’s Technische Universität has now obtained the first individual approvals in Germany for all-glass structures with transparent adhesive joints for two buildings in Dresden and Grimma. In these cases, the loads are carried via load-bearing glued frames that rely on a material bond between the individual parts without any metal fixings. This solution is based on the results of many years of development. Currently, various frame corners covering a wide range of parameters are being studied in a follow-up project. The aim is to optimize the adhesive joints and the bonding technology. The requirements placed on the joint are being identified and corresponding adhesive systems researched. Preliminary studies of numerous material specimens form the starting point for determining the material parameters. Small-scale specimens are being tested under various boundary conditions and aging scenarios to establish the strengths of a number of suitable adhesives. These results will enable prototypes to be designed for numerical simulation. Numerical calculations and experimental investigations are being carried out in parallel to optimize the geometry, load-carrying capacity, long-term reliability, and durability of the glued all-glass frame corners. Specimen components are loaded in a testing machine to study the structural effect of these glued glass frame corners. The findings will be incorporated in the design of the adhesive joint and in the development of a numerical simulation for the glued connection. Data for executing an adhesive joint, specific to each adhesive and crucial for the quality of the transparency, conclude the project.