bluuebunny/crossref_metadata_2025_split huggingface dataset:
Reactor Pulse-Repeatability Studies at the Annular Core Research Reactor
Authors: DePriest K. Russell , Trinh Tri Q. , Luker S. Michael 
Published: May 2012

Abstract
The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO2–BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the 9Be(γ, n)8Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times.
Investigation of Load Control Errors for Spectrum Fatigue Testing at High Frequencies
Authors: Rusk David T. , Taylor Robert E. 
Published: May 2012

Abstract
Ultra-high cycle fatigue (gigacycle) tests have shown that a true fatigue endurance limit does not exist for most metallic materials used in high cycle applications. These findings have significant implications for rotorcraft dynamic structural components that have traditionally been designed using endurance limit, stress-life methods. Unfortunately, the gigacycle fatigue test results generated to date cannot be easily applied to rotorcraft component design, because the interaction of different crack nucleating mechanisms under variable amplitude loading is not well understood, and must be studied using spectrum fatigue tests carried out to very long lives. To quantify the effect of spectrum load control errors on the rate of fatigue damage accumulation for a standard servo-hydraulic fatigue test machine, a damage ratio parameter for crack initiation is calculated using the high cycle fatigue portion of the Coffin-Manson strain-life curve with rainflow cycle counting. The fatigue damage parameter allows the level of controller error to be assessed as a function of test frequency, peak load levels, and test spectrum complexity. Also, different command feedback compensation schemes are tested to determine the range of control error that can be expected for a given set of test parameters. A sensitivity study was performed on the command signal response errors as a function of changes in the slope of the mean damage curve beyond 106 cycles. The test results demonstrate that rotorcraft spectrum fatigue tests can be performed at frequencies up to 150 Hz with a level of nominal damage accumulation error less that 1 % under most conditions. When electrical measurement system error is accounted for, the damage accumulation error is less than 2 % under most conditions. The results also show that it is important to consider the shape of the stress-life curves in the gigacycle fatigue range when performing spectrum tests out to long fatigue lives.
The Fast Neutron Fluence and the Activation Detector Activity Calculations Using the Effective Source Method and the Adjoint Function
Authors: Hep J. , Konečná A. , Krýsl V. , Smutný V. 
Published: May 2012

Abstract
This paper describes the application of effective source in forward calculations and the adjoint method to the solution of fast neutron fluence and activation detector activities in the reactor pressure vessel (RPV) and RPV cavity of a VVER-440 reactor. Its objective is the demonstration of both methods on a practical task. The effective source method applies the Boltzmann transport operator to time integrated source data in order to obtain neutron fluence and detector activities. By weighting the source data by time dependent decay of the detector activity, the result of the calculation is the detector activity. Alternatively, if the weighting is uniform with respect to time, the result is the fluence. The approach works because of the inherent linearity of radiation transport in non-multiplying time-invariant media. Integrated in this way, the source data are referred to as the effective source. The effective source in the forward calculations method thereby enables the analyst to replace numerous intensive transport calculations with a single transport calculation in which the time dependence and magnitude of the source are correctly represented. In this work, the effective source method has been expanded slightly in the following way: neutron source data were performed with fewgroup method calculation using the active core calculation code moby-dick. The follow-up neutron transport calculation was performed using the neutron transport code tort to perform multigroup calculations. For comparison, an alternative method of calculation has been used based upon adjoint functions of the Boltzmann transport equation. Calculation of the three-dimensional (3-D) adjoint function for each required computational outcome has been obtained using the deterministic code tort and the cross section library BGL440. Adjoint functions appropriate to the required fast neutron flux density and neutron reaction rates have been calculated for several significant points within the RPV and RPV cavity of the VVER-440 reactor. Both of calculation methods are briefly described and the results are compared.
Dosimetry Assessments for the Reactor Pressure Vessel and Core Barrel in UK Pressurized Water Reactor (PWR) Plant
Authors: Thornton D. A. , Allen D. A. , Huggon A. P. , Picton D. J. , Robinson A. T. , Steadman R. J. , Serén T. , Lipponen M. , Kekki T. 
Published: May 2012

Abstract
Specimens for the Sizewell B reactor pressure vessel (RPV) in-service steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained.
VENUS-F: A First Fast Lead Critical Core for Benchmarking
Authors: Kochetkov A. , Wagemans J. , Vittiglio G. 
Published: May 2012

Abstract
The zero-power thermal neutron water-moderated facility VENUS at SCK•CEN has been extensively used for benchmarking in the past. In accordance with GEN-IV design tasks (fast reactor systems and accelerator driven systems), the VENUS facility was modified in 2007–2010 into the fast neutron facility VENUS-F with solid core components. This paper introduces the projects GUINEVERE and FREYA, which are being conducted at the VENUS-F facility, and it presents the measurement results obtained at the first critical core. Throughout the projects other fast lead benchmarks also will be investigated. The measurement results of the different configurations can all be used as fast neutron benchmarks.
Comparison of Regulatory Guide 1.99 Fluence Attenuation Methods
Authors: Jones Eric N. 
Published: May 2012

Abstract
U.S. Regulatory Guide 1.99 Revision 2 (U.S. Nuclear Regulatory Commission, 1988, “Radiation Embrittlement of Reactor Vessel Materials,” Regulatory Guide 1.99, Revision 2, Washington, D.C.) provides for the use of two substantially different methods for determining through-wall fluence in nuclear reactor pressure vessels. One method is a generic attenuation curve based on a simplistic exponential decay equation. Partly due to the simplicity of its application, the generic attenuation method is predominantly used for licensing calculations. However, it has a limitation in that at increasing distances away from the core beltline, it becomes increasingly less accurate because it cannot account for neutron streaming effects in the cavity region surrounding the pressure vessel. The other attenuation method is based on a displacement per atom (dpa) calculation specific to the reactor vessel structure. The dpa method provides a more accurate representation of fluence attenuation through the reactor pressure vessel (RPV) wall at all elevations of the pressure vessel because it does account for neutron streaming in the cavity region. A requirement for using the dpa method, however, is an accurate flux solution through the RPV wall. This requirement has limited the use of traditional transport methods, such as discrete ordinates, that are limited by their treatment of cavity regions (i.e., air) outside the pressure vessel wall. TransWare Enterprises, under the sponsorship of EPRI and BWRVIP, has developed an advanced three-dimensional transport methodology capable of producing fully converged flux solutions throughout the entire reactor system, including in the cavity region and primary shield structures. This methodology provides an accurate and reliable determination of through-wall fluence in boiling water reactor (BWR) and pressurized water reactor (PWR) pressure vessels, thus allowing the dpa method to be implemented with high reliability. Using this advanced 3-D methodology, this paper presents comparisons of the generic and dpa attenuation methods at critical locations in both BWR and PWR pressure vessel walls.
Dosimetry Analyses of the Ringhals 3 and 4 Reactor Pressure Vessels
Authors: Kulesza Joel A. , Fero Arnold H. , Roudén Jenny , Green Eva-Lena 
Published: May 2012

Abstract
A comprehensive series of neutron dosimetry measurements consisting of surveillance capsules, reactor pressure vessel cladding samples, and ex-vessel neutron dosimetry has been analyzed and compared to the results of three-dimensional, cycle-specific neutron transport calculations for the Ringhals Unit 3 and Unit 4 reactors in Sweden. The comparisons show excellent agreement between calculations and measurements. The measurements also demonstrate that it is possible to perform retrospective dosimetry measurements using the 93Nb (n,n′) 93mNb reaction on samples of 18-8 austenitic stainless steel with only trace amounts of elemental niobium.
CALMOS: Innovative Device for the Measurement of Nuclear Heating in Material Testing Reactors
Authors: Carcreff H. 
Published: May 2012

Abstract
An R&D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA’s Nuclear Energy Division at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2 W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed.
Sensitivity of Adjustment to Parameter Correlations and to Response-Parameter Correlations
Authors: Wagschal J. J. 
Published: May 2012

Abstract
The adjusted parameters and response, and their respective posterior uncertainties and correlations, are presented explicitly as functions of all relevant prior correlations for the two parameters, one response case. The dependence of these adjusted entities on the various prior correlations is analyzed and portrayed graphically for various valid correlation combinations on a simple criticality problem.
Exposure Conditions of Reactor Internals of Rovno VVER-440 Nuclear Power Plant Units 1 and 2
Authors: Grytsenko O. V. , Pugach S. M. , Diemokhin V. L. , Bukanov V. N. , Marek M. , Vandlik S. 
Published: May 2012

Abstract
Results of determination of irradiation conditions for vessel internals of VVER-440 reactor No. 1 and 2 at Rovno Nuclear Power Plant, obtained by specialists at Institute for Nuclear Research Kyiv, Ukraine, and Nuclear Research Institute Rez, Czech Republic, are presented. To calculate neutron transport, detailed calculation models of these reactors were prepared. Distribution of neutron flux functionals on the surface of reactor VVER-440 baffle and core barrel for different core loads was studied. Agreement between results obtained by specialists at Institute for Nuclear Research and at Nuclear Research Institute is shown.
Analysis of Dosimetry from the H. B. Robinson Unit 2 Pressure Vessel Benchmark Using raptor-m3g and ALPAN
Authors: Fischer Greg A. 
Published: May 2012

Abstract
The dosimetry from the H. B. Robinson Unit 2 Pressure Vessel Benchmark is analyzed with a suite of Westinghouse-developed codes and data libraries. The radiation transport from the reactor core to the surveillance capsule and ex-vessel locations is performed by raptor-m3g, a parallel deterministic radiation transport code that calculates high-resolution neutron flux information in three dimensions. The cross-section library used in this analysis is the ALPAN library, an Evaluated Nuclear Data File (ENDF)/B-VII.0-based library designed for reactor dosimetry and fluence analysis applications. Dosimetry is evaluated with the industry-standard snlrml reactor dosimetry cross-section data library.
Development of INSPCT-S for Inspection of Spent Fuel Pool
Authors: Walters William , Haghighat Alireza , Sitaraman Shivakumar , Ham Young 
Published: May 2012

Abstract
In this paper, we discuss an accurate and fast software tool (INSPCT-S, Inspection of Nuclear Spent fuel-Pool Calculation Tool, version Spreadsheet) developed for calculation of the response of fission chambers placed in a spent fuel pool, such as Atucha-I. INSPCT-S is developed for identification of suspicious regions of the pool that may have missing or substitute assemblies. INSPCT-S uses a hybrid algorithm based on the adjoint function methodology. The neutron source is comprised of spontaneous fission, (α, n) interactions, and subcritical multiplication. The former is evaluated using the ORIGEN-ARP code, and the latter is obtained with the fission matrix (FM) formulation. The FM coefficients are determined using the MCNP Monte Carlo code, and the importance function is determined using the PENTRAN 3-D parallel Sn code. Three databases for the neutron source, FM elements, and adjoint flux are prepared as functions of different parameters including burnup, cooling time, enrichment, and pool lattice size. INSPCT-S uses the aforementioned databases and systems of equations to calculate detector responses, which are subsequently compared with normalized experimental data.
Fitting Method for Spectrum Deduction in High-Energy Neutron Field Induced by GeV-Protons Using Experimental Reaction-Rate Data
Authors: Kasugai Y. , Matsuda N. , Sakamoto Y. , Nakashima H. , Yashima H. , Matsumura H. , Iwase H. , Hirayama H. , Mokhov N. , Leveling A. , Boehnlein D. , Vaziri K. , Lauten G. , Oishi K. , Nakamura T. 
Published: May 2012

Abstract
The Japanese and American Study of Muon Interaction and Neutron detection (JASMIN) collaboration, has been conducting shielding experiments using the Fermilab anti-proton target station (Pbar) shielding assembly. A multi-foil technique was used to measure the high energy neutron spectra, in the range of 1 to 100 MeV, for the target station shielding configuration. The neutron spectra were de-convoluted using a new fitting method. This method is based on the assumption that a neutron spectrum can be expressed as a simple sum of two exponentials. The validity of the fitting method was confirmed by comparison with the results obtained using SAND-II computer code and theoretical calculations. Finally, it was found that there are simple correlations between reaction rates and the adjustable parameters in the fitting function.
Photon Spectrum behind Biological Shielding of the LVR-15 Research Reactor
Authors: Klupák V. , Viererbl L. , Lahodová Z. , Marek M. , Vinš M. 
Published: May 2012

Abstract
The LVR-15 reactor is a light water research reactor situated at the Research Centre Řež, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of 235U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified.
Apparatus for Use in Evaluating Protection from Low Pressure Hot Water Jets
Authors: Jalbani S. H. , Ackerman M. Y. , Crown E. M. , van Keulen M. , Song G. 
Published: May 2012

Abstract
Use of hot water has become extensive, especially in bitumen extraction from oil sands and in the production of heavy oil. The hot water is often under pressure and is 80–90°C, which is well above temperatures that result in immediate, potentially severe burn injuries. The ASTM F2701-08 apparatus consists of a funnel through which hot liquid is hand-poured to produce a 10 s exposure. Two 40 mm diameter copper calorimeters, mounted in an insulating sheet are positioned beneath the funnel outlet and are intended to measure the energy transfer through the fabric from the hot liquid. For this research, changes were made to the apparatus and procedures to more closely simulate low pressure hot water streams found in the oil industry and to improve reproducibility. The funnel producing the liquid splash was replaced with a small pipe directly fed by a circulating hot water bath via a small pump, through a hose and valve system, allowing for consistent application of a given quantity of water at a consistent temperature and flow rate. Water temperature, flow rate, and pressure can be altered as desired. A series of fabrics varying systematically on several parameters were tested with the modified equipment. Resulting heat transfer data suggest the system differentiates well among both semi-permeable and impermeable fabrics. Specifications for hot water protection are proposed.
Path Forward for Dosimetry Cross Sections
Authors: Griffin P. J. , Peters C. D. 
Published: May 2012

Abstract
In the 1980’s the dosimetry community embraced the need for a high fidelity quantification of uncertainty in nuclear data used for dosimetry applications. This led to the adoption of energy-dependent covariance matrices as the accepted manner of quantifying the uncertainty data. The trend for the dosimetry community to require high fidelity treatment of uncertainty estimates has continued to the current time where requirements on nuclear data are codified in standards such as ASTM E 1018. This paper surveys the current state of the dosimetry cross sections and investigates the quality of the current dosimetry cross section evaluations by examining calculated-to-experimental ratios in neutron benchmark fields. In recent years more nuclear-related technical areas are placing an emphasis on uncertainty quantification. With the availability of model-based cross sections and covariance matrices produced by nuclear data codes, some nuclear-related communities are considering the role these covariance matrices should play. While funding within the dosimetry community for cross section evaluations has been very meager, other areas, such as the solar-related astrophysics community and the US Nuclear Criticality Safety Program, have been supporting research in the area of neutron cross sections. The Cross Section Evaluation Working Group (CSEWG) is responsible for the creation and maintenance of the ENDF/B library which has been the mainstay for the reactor dosimetry community. Given the new trends in cross section evaluations, this paper explores the path forward for the US nuclear reactor dosimetry community and its use of the ENDF/B cross-sections. The major concern is maintenance of the sufficiency and accuracy of the uncertainty estimate when used for dosimetry applications. The two major areas of deficiency in the proposed ENDF/B approach are: 1) the use of unrelated covariance matrices in ENDF/B evaluations and 2) the lack of “due consideration” of experimental data in the evaluation.
A New Formulation of the Unified Monte Carlo Approach (UMC-B) and Cross-Section Evaluation for the Dosimetry Reaction 55Mn(n,γ)56Mn
Authors: Capote Roberto , Smith Donald L. , Trkov Andrej , Meghzifene Mehdi 
Published: May 2012

Abstract
Two relatively new approaches to neutron cross section data evaluation are described. They are known collectively as Unified Monte Carlo (versions UMC-G and UMC-B). Comparisons are made between these two methods, as well as with the well-known generalized least-squares (GLSQ) technique, through the use of simple, hypothetical (toy) examples. These new Monte Carlo methods are based on stochastic sampling of probability functions that are constructed with the use of theoretical and experimental data by applying the principle of maximum entropy. No further assumptions are involved in either UMC-G or UMC-B. However, the GLSQ procedure requires the linearization of non-linear terms, such as those that occur when cross section ratio data are included in an evaluation. It is shown that these two stochastic techniques yield results that agree well with each other, and with the GLSQ method, when linear data are involved, or when the perturbations due to data discrepancies and nonlinearity effects are small. Otherwise, there can be noticeable differences. The present investigation also demonstrates, as observed in earlier work, that the least-squares approach breaks down when these conditions are not satisfied. This paper also presents an actual evaluation of the 55Mn(n,γ)56Mn neutron dosimetry reaction cross section in the energy range from 100 keV to 20 MeV, which was performed using both GLSQ and UMC-G approaches.