bluuebunny/crossref_metadata_2025 huggingface dataset:
Ringhals Unit 3 and 4—Fluence Determination in a Historic and Future Perspective
Authors: Green Eva-Lena , Roudén Jenny , Efsing Pål
Published: February 2013

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 surveil-lance 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.
Dosimetry Analyses of the Ringhals 3 and 4 Reactor Pressure Vessels
Authors: Kulesza Joel A. , Fero Arnold H. , Roudén Jenny , Green Eva-Lena
Published: February 2013

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.
Modernization of Existing VVER-1000 Surveillance Programs
Authors: Kochkin V. , Erak D. , Makhotin D.
Published: February 2013

According to generally accepted world practice, evaluation of the reactor pressure vessel (RPV) material behavior during operation is carried out using tests of surveillance specimens. The main objective of the surveillance program consists in insurance of safe RPV operation during the design lifetime and lifetime-extension period. At present, the approaches of pressure vessels residual life validation based on the test results of their surveillance specimens have been developed and introduced in Russia and are under consideration in other countries where vodo-vodyanoi energetichesky reactors- (VVER-) 1000 are in operation. In this case, it is necessary to ensure leading irradiation of surveillance specimens (as compared to the pressure vessel wall) and to provide uniformly irradiated specimen groups for mechanical testing. Standard surveillance program of VVER-1000 has several significant shortcomings and does not meet these requirements. Taking into account program of lifetime extension of VVER-1000 operating in Russia, it is necessary to carry out upgrading of the VVER-1000 surveillance program. This paper studies the conditions of a surveillance specimen's irradiation and upgrading of existing sets to provide monitoring and prognosis of RPV material properties for extension of the reactor's lifetime up to 60 years or more.
A New Technique for Dosimetry Reaction Cross-Section Evaluation
Authors: Badikov S. A.
Published: February 2013

An objective of this paper is a unification of the procedure for dosimetry reaction cross-section evaluation. A set of requirements for the unified evaluation procedure is presented. A new code (ORTHO) was developed in order to meet these requirements. A statistical model, an algorithm, and the basic formulae employed in the code are described. The code was used for Ti48(n,p) reaction cross-section evaluation. The results of the evaluation are compared to International Reactor Dosimetry File (IRDF)-2002 data. The evaluated cross-sections and their correlations from this work are in good agreement with the IRDF-2002 evaluated data, whereas the uncertainties of the evaluated cross-sections are inconsistent.
Uncertainties of Responses Calculated with a “Tuned” Library: Geometrical and Algebraic Insights
Authors: Perel R. L.
Published: February 2013

The standard American evaluated nuclear data library ENDF/B-VII was “tuned” based on simple measured critical assemblies. This tuning was not done according to a fully defined mathematical algorithm, such as the adjustment algorithm. In this work, we investigate how tuning affects the uncertainties (covariances) of the cross-section libraries. First, we analyze what happens to cross-section uncertainties as a result of adjustment. The effect of adjustment on the uncertainties is geometrically demonstrated for simple cases. For those parts of the sensitivities of the assembly to be calculated that are parallel to the sensitivities of the assemblies on which adjustment was based, there is significant reduction in the uncertainties. For orthogonal parts, there is no change in the uncertainties. These findings are algebraically proven based on the adjustment algorithm. Then we analyze the differences between tuned libraries and adjusted libraries. We conclude that for tuned libraries, the uncertainties in the direction of sensitivities on which adjustment or tuning were based are improved, similar to the improvement for an adjusted library. However, the displacement of the nominal values of the library parameters to their tuned value, instead of their adjusted value, adds an additional uncertainty. This additional uncertainty is typically small in the direction that was improved by adjustment. The magnitude of the additional uncertainty in perpendicular directions depends on the particular details of the tuning performed.
Updating and Extending the IRDF-2002 Dosimetry Library
Authors: Capote Roberto , Zolotarev Konstantin I. , Pronyaev Vladimir G. , Trkov Andrej
Published: February 2013

The International Reactor Dosimetry File (IRDF)-2002 released in 2004 by the IAEA (see http://www-nds.iaea.org/irdf2002/) contains cross-section data and corresponding uncertainties for 66 dosimetry reactions. New cross-section evaluations have become available recently that re-define some of these dosimetry reactions including: (1) high-fidelity evaluation work undertaken by one of the authors (KIZ); (2) evaluations from the US ENDF/B-VII.0 and candidate evaluations from the US ENDF/B-VII.1 libraries that cover reactions within the International Evaluation of Neutron Cross-Section Standards; (3) European JEFF3.1 library; and (4) Japanese JENDL-4.0 library. Additional high-threshold reactions not included in IRDF-2002 (e.g.,59Co(n,3n) and 209Bi(n,3n)) have been also evaluated to characterize higher-energy neutron fields. Overall, 37 new evaluations of dosimetry reactions have been assessed and intercomparisons made with integral measurements in reference neutron fields to determine whether they should be adopted to update and improve IRDF-2002. Benchmark calculations performed for newly evaluated reactions using the ENDF/B-VII.0 235U thermal fission and 252Cf spontaneous fission neutron spectra show that calculated integral cross sections exhibit improved agreement with evaluated experimental data when compared with the equivalent data from the IRDF-2002 library. Data inconsistencies or deficiencies of new evaluations have been identified for 63Cu(n,2n), 60Ni(n,p)60m+g Co, 55Mn(n,γ), and 232Th(n,f) reactions. Compared with IRDF-2002, the upper neutron energy boundary was formally increased from the actual maximum energy of typically 20 MeV up to 60 MeV by using the TENDL-2010 cross sections and covariance matrices. This extension would allow the updated IRDF library to be also used in fusion dosimetry applications. Uncertainties in the cross sections for all new evaluations are given in the form of relative covariance matrices. Newly evaluated excitation functions should be considered as suitable candidates in the preparation of an improved version of the IRDF that was planned to be released for testing in December 2011.
Sensitivity of Adjustment to Parameter Correlations and to Response-Parameter Correlations
Authors: Wagschal J. J.
Published: February 2013

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.
Retrospective Dosimetry Analyses of Reactor Vessel Cladding Samples
Authors: Greenwood L. R. , Soderquist C. Z. , Fero A. H.
Published: February 2013

Reactor pressure vessel cladding samples for Ringhals Units 3 and 4 in Sweden were analyzed using retrospective reactor dosimetry techniques. The objective was to provide the best estimates of the neutron fluence for comparison with neutron transport calculations. A total of 51 stainless steel samples consisting of chips weighing approximately 100 to 200 mg were removed from selected locations around the pressure vessel and were sent to Pacific Northwest National Laboratory for analysis. The samples were fully characterized and analyzed for radioactive isotopes, with special interest in the presence of Nb-93 m. The RPV cladding retrospective dosimetry results will be combined with a re-evaluation of the surveillance capsule dosimetry and with ex-vessel neutron dosimetry results to form a comprehensive 3D comparison of measurements to calculations performed with 3D deterministic transport code.
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: February 2013

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.
VENUS-F: A First Fast Lead Critical Core for Benchmarking
Authors: Kochetkov A. , Wagemans J. , Vittiglio G.
Published: February 2013

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.
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: February 2013

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 Measurements and the Associated Unfolding Methodologies to Characterize the Caliban Pulsed Reactor Cavity Neutron Spectrum by the Foil Activation Method
Authors: Jacquet X. , Casoli P. , Authier N. , Rousseau G. , Barsu C.
Published: February 2013

Caliban is a cylindrical metallic core reactor mainly composed of uranium 235. It is operated by the Criticality and Neutron Science Research Laboratory located at the French Atomic Energy Commission research center in Valduc. As with other fast burst reactors, Caliban is used extensively for determining the responses of electronic parts or other objects and materials to neutron-induced displacements. Therefore, Caliban's irradiation characteristics, and especially its central cavity neutron spectrum, have to be very accurately evaluated. The foil activation method has been used in the past by the Criticality and Neutron Science Research Laboratory to evaluate the neutron spectrum of the different facilities it operated, and in particular to characterize the Caliban cavity spectrum. In order to strengthen and to improve our knowledge of the Caliban cavity neutron spectrum and to reduce the uncertainties associated with the available evaluations, new measurements have been performed on the reactor and interpreted by the foil activation method. A sensor set has been selected to sample adequately the studied spectrum. Experimental measured reaction rates have been compared to the results from UMG spectrum unfolding software and to values obtained with the activation code Fispact. Experimental and simulation results are overall in good agreement, although gaps exist for some sensors. UMG software has also been used to rebuild the Caliban cavity neutron spectrum from activation measurements. For this purpose, a default spectrum is needed, and one has been calculated with the Monte-Carlo transport code Tripoli 4 using the benchmarked Caliban description.
Los Alamos National Laboratory Fission Basis
Authors: Keksis A. L. , Chadwick M. B. , Selby H. D. , Mac Innes M. R. , Barr D. W. , Meade R. A. , Burns C. J. , Wallstrom T. C.
Published: February 2013

This report is an overview of two main publications that provide a comprehensive review of the Los Alamos National Laboratory (LANL) Fission Basis. The first is the experimental paper, “Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on 239Pu,235U, 238U,” [Selby, H. D., et al., Nucl. Data Sheets, Vol. 111 2010, pp. 2891–2922] and the second is the theoretical paper, “Fission Product Yields from Fission Spectrum n+239Pu for ENDF=B-VII.1,” [Chadwick, M. B., et al., Nucl. Data Sheets, Vol. 111, 2010, pp. 2923–2964].
Sensitivity Studies Associated with Dosimetry Experiment Interpretation
Authors: Bourganel S. , Soldevila M. , Ferrer A. , Grégoire G. , Destouches C. , Beretz D.
Published: February 2013

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.
VVER-440 and VVER-1000 Reactor Dosimetry Benchmark—BUGLE-96 Versus ALPAN VII.0
Authors: Duo J. I.
Published: February 2013

Analytical results of the vodo-vodyanoi energetichesky reactor-(VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Institute Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reduced computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10 %) between BUGLE-96 and ALPAN VII.0 libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15 % with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements.
Development and Experimental Validatio of a Calculation Scheme for Nuclear Heating Evaluation in the Core of the OSIRIS Material Testing Reactor
Authors: Malouch F.
Published: February 2013

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.
Analysis of Gamma-Ray Dosimetry Experiments in the Zero Power MINERVE Facility
Authors: Amharrak H. , Di Salvo J. , Lyoussi A. , Roche A. , Masson-Fauchier M. , Bosq J. C. , Carette M.
Published: February 2013

The objective of this study is to develop nuclear heating measurement methods in zero power experimental reactors. These developments contribute to the qualification of photonics calculation schemes for the assessment of gamma heating in the future Jules Horowitz Material Testing Reactor. This paper presents the analysis of thermoluminescent detector (TLD) experiments in the UO2 core of the MINERVE Research Reactor at the French Alternative Energies and Atomic Energy Commission center in Cadarache. The experimental sources of uncertainty in the gamma dose have been reduced by improving the measurement conditions and the repeatability of the calibration step for each individual TLD. The interpretation of these measurements needs to take into account the calculation of cavity correction factors related to calibration and irradiation configurations, as well as neutron correction calculations. These calculations are based on Monte Carlo simulations of neutron-gamma and gamma-electron transport coupled particles. The comparison between calculated and measured integral gamma-ray absorbed doses in the aluminum material surrounding the TLD shows that calculations slightly overestimate the measurement, with a calculated versus experimental ratio equal to 1.04 ± 5.7 % (k = 2).
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: February 2013

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.
Reactor Pulse-Repeatability Studies at the Annular Core Research Reactor
Authors: DePriest K. Russell , Trinh Tri Q. , Luker S. Michael
Published: February 2013

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 the9Be(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.
A Feasibility Study to Determine Cooling Time and Burnup of Advanced Test Reactor Fuel Using a Nondestructive Technique and Three Types of Gamma-ray Detectors
Authors: Navarro J. , Aryaeinejad R. , Nigg D. W.
Published: February 2013

The goal of this work was to perform a feasibility study and establish measurement techniques to determine the burnup of the Advanced Test Reactor (ATR) fuels at the Idaho National Laboratory (INL). Three different detectors of high purity germanium (HPGe), lanthanum bromide (LaBr3), and high pressure xenon (HPXe) in two detection system configurations of below and above the water pool were used in this study. The last two detectors were used for the first time in fuel burnup measurements. The results showed that a better quality spectra can be achieved with the above the water pool configuration. Both short and long cooling time fuels were investigated in order to determine which measurement technique, absolute or fission product ratio, is better suited in each scenario and also to establish what type of detector should be used in each case for the best burnup measurement. The burnup and cooling time calibrations were established using experimental absolute activities or isotopic ratios and ORIGEN burnup calculations. A method was developed to do burnup and cooling time calibrations using fission isotopes activities without the need to know the exact geometry.
The Neutron Standard Fields at the BR1 Reactor at SCK•CEN
Authors: Wagemans J. , Malambu E. , Borms L.
Published: February 2013

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.
Photon Spectrum behind Biological Shielding of the LVR-15 Research Reactor
Authors: Klupák V. , Viererbl L. , Lahodová Z. , Marek M. , Vinš M.
Published: February 2013

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.
Development and Testing of the VITAMIN-B7/BUGLE-B7 Coupled Neutron-Gamma Multigroup Cross-Section Libraries
Authors: Risner J. M. , Wiarda D. , Miller T. M. , Peplow D. E. , Patton B. W. , Dunn M. E. , Parks B. T.
Published: February 2013

The U.S. Nuclear Regulatory Commission's Regulatory Guide 1.190 states that calculational methods used to estimate reactor pressure vessel (RPV) fluence should use the latest version of the evaluated nuclear data file (ENDF). The VITAMIN-B6 fine-group library and BUGLE-96 broad-group library, which are widely used for RPV fluence calculations, were generated using ENDF/B-VI.3 data, which was the most current data when Regulatory Guide 1.190 was issued. We have developed new fine-group (VITAMIN-B7) and broad-group (BUGLE-B7) libraries based on ENDF/ B-VII.0. These new libraries, which were processed using the AMPX code system, maintain the same group structures as the VITAMIN-B6 and BUGLE-96 libraries. Verification and validation of the new libraries were accomplished using diagnostic checks in AMPX, “unit tests” for each element in VITAMIN-B7, and a diverse set of benchmark experiments including critical evaluations for fast and thermal systems, a set of experimental benchmarks that are used for SCALE regression tests, and three RPV fluence benchmarks. The benchmark evaluation results demonstrate that VITAMIN-B7 and BUGLE-B7 are appropriate for use in RPV fluence calculations and meet the calculational uncertainty criterion in Regulatory Guide 1.190.
Neutron Flux Measurements in the Side-Core Region of Hunterston B Advanced Gas-Cooled Reactor
Authors: Allen D. A. , Shaw S. E. , Huggon A. P. , Steadman R. J. , Thornton D. A. , Whiley G. S.
Published: February 2013

The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., “A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,” Proceedings of the 13th International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debar-beris, and P. D'hondt, Eds., pp. 679–687] using a detailed 3D model and a Monte Carlo radiation transport program, MCBEND. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the MCBEND model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement “stringers” within the side-core region of one of the Hunterston B reactors for the purpose of validating the MCBEND model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed MCBEND flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which the vulnerable steel restraint components are exposed.
Analysis of Dosimetry from the H. B. Robinson Unit 2 Pressure Vessel Benchmark Using RAPTOR-M3G and ALPAN
Authors: Fischer Greg A.
Published: February 2013

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 surveil-lance 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.
A Database-Informed Approach to New Plant Shielding Design
Authors: Lloyd Timothy M.
Published: February 2013

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.
EPR/PTFE Dosimetry for Test Reactor Environments
Authors: Vehar D. W. , Griffin P. J. , Quirk T. J.
Published: February 2013

In 1992 the Radiation Metrology Laboratory (RML) at Sandia National Laboratories implemented EPR/Alanine capabilities for use in routine and calibration activities at its Co-60 and pulsed-power facilities. At that time it also investigated the usefulness of the system for measurement of absorbed dose in the mixed neutron/photon environments of reactors such as the Sandia Pulsed Reactor and the Annular Core Research Reactor used for hardness testing of electronics. The RML concluded that the neutron response of alanine was a sufficiently high fraction of the overall dosimeter response that the resulting uncertainties in the photon dose would be unacceptably large for silicon-device testing. However, it also suggested that nonhydrogenous materials such as polytetrafluoroethylene (PTFE) would exhibit smaller neutron response and might be useful in mixed environments. Preliminary research with PTFE in photon environments indicated considerable promise, but further development was not pursued at that time. Because of renewed interest in absorbed dose measurements that could better define the individual contributions of photon and neutron components to the overall dose delivered to a test object, the RML has re-initiated the development of an EPR/PTFE dosimetry system. This paper presents a summary of the research, a description of the EPR/PTFE dosimetry system, and recommendations for preparation and fielding of the dosimetry in photon and mixed neutron/photon environments.
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: February 2013

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.
Design and Testing of a Boron Carbide Capsule for Spectral Tailoring in Mixed-Spectrum Reactors
Authors: Greenwood L. R. , Wittman R. , Pierson B. P. , Metz L. A. , Payne R. , Finn E. C. , Friese J. I.
Published: February 2013

A boron carbide capsule has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. Irradiations were conducted in pulsed mode and in continuous operation for up to 4 h. A cadmium cover was used to reduce thermal heating. The neutron spectrum calculated with the Monte Carlo N-particle transport code was found to be in good agreement with reactor dosimetry measurements using the STAY'SL computer code. The neutron spectrum resembles that of a fast reactor. The design of a capsule using boron carbide fully enriched in 10B shows that it is possible to produce a neutron spectrum similar to that of 235U fission.
Development of an Active Detector for the Characterization of the Late-Time Radiation Environment from a Reactor Pulse
Authors: Luker S. M. , Griffin P. J. , Kolb N. R. , Naranjo G. N. , Suo-Anttila A. J.
Published: February 2013

This paper discusses the use of a commercially available 235U fission chamber, with a matching compensating ion chamber, originally sold as a single-ended detector with the signal conducted over the shield of a coaxial cable. The authors designed an aluminum housing that isolates the two detectors and converts the signals to full differential mode as a noise-reduction technique. The signals are processed using the switched resistor technique to extend the signal range to longer times from the peak of the pulse [Luker, S. M., Griffin, P. J., King, D. B., and Suo-Anttila, A. J., “Improved Diagnostics for Analysis of a Reactor Pulse Radiation Environment,” 13th International Symposium on Reactor Dosimetry, Akersloot, Netherlands, May 25, 2008, pp. 4–6.]. The newly configured fission chamber assembly has been used at the annular core research reactor at Sandia National Laboratories to provide a high-fidelity characterization of the neutron time profile from a pulsed operation.
Crystal structure of tris(1,10-phenanthroline-N,N')manganese(II)dichloro octahydrate, C36H40Cl2MnN6O8
Authors: Wang Xinfang
Published: February 2013

C36H40Cl2MnN6O8, tetragonal, I41/a (no. 88), a = 35.922(1) Å, c = 11.8977(8) Å, V = 15352.4(12) Å3, Z = 16, Rgt(F) = 0.0801, wRref(F2) = 0.2670, T = 293 K.
Crystal structure of bis(2,2’bipyridine)(tetrafluorophthalato-κO,κO) zink(II)tetrafluorophthalate, [Zn(C8HF4O4)(C10H8N2)2]C8HF4O4, C36H18F8N4O8Zn
Authors: Liu Yuan-Ying , Guo Jin-Fu , Li Hui-Yun , Zhao Guo-Yan
Published: February 2013

C36H18F8N4O8Zn, triclinic, P1̄ (no. 2), a = 10.618(2) Å, b = 12.026(2) Å, c = 15.399(2) Å, α = 81.574(2)°, β = 85.344(2)°, γ = 87.642(2)°, V = 1937.7 Å3, Z = 2, Rgt(F) = 0.0566, wRref(F2) = 0.1366, T = 293 K.
Remittances from Migrants
Authors: La Hai Anh , Leung Suiwah
Published: February 2013

By analyzing data from the three Vietnam Household Living Standards Surveys between 2002 and 2006, the authors found that a very large proportion of Vietnamese households received remittances during those years and that these remittances constituted a significant share of household incomes. The majority of households received remittances from within Vietnam, but for those able to access remittances from abroad, the amounts were very much larger. In general, remittances do seem to play an “insurance role” for the poor, cushioning any sudden falls in household incomes. There is also some evidence that public transfers can crowd out private remittances, but the effect is by no means complete.
The Socio-Spatial Constellation of a Central Vietnamese Village and its Emigrants
Authors: Carruthers Ashley , Dang Trung Dinh
Published: February 2013

This paper seeks to understand the patterns and experiences of migration from a subsistence rice-farming village in Quảng Nam Province. Emigration from Bình Yên ranges from circular and seasonal migration to the cities and central highlands for construction, forestry and plantation work; semi-permanent migration to Đà Nẵng and the south for work in the manufacturing and service sectors; and more permanent migration to Hà Lam, Tam Kỳ, Đà Nẵng, Hồ Chí Minh City and elsewhere for education, trading, business, and professional employment. In addition, people from Bình Yên have engaged in assisted settlement migration to frontier zones in Đắk Lắk Province since the time of the Ngô Đình Diệm regime. By examining migration from the perspective of the village, the paper aims to illuminate the central Vietnamese peasant experience of the policies and socioeconomic conditions that have driven some of the nation's major migratory movements. The authors also map the social geography that ties members of the Bình Yên diaspora back to the village and to each other. Emigrants participate in the cash economy and endure alienated social relations in the cities and factories. When they return to Bình Yên they return to a world of tradition, warm social ties and a moral economy. Given this relationship, the paper asks, how do Bình Yên's locals and emigrants conceptualize their places in the socio-spatial “constellation” that connects the village to the locations in which its members have settled or found temporary work?
Long-term remission and recurrence rates in Cushing's disease: predictive factors in a single-centre study
Authors: Alexandraki Krystallenia I , Kaltsas Gregory A , Isidori Andrea M , Storr Helen L , Afshar Farhad , Sabin Ian , Akker Scott A , Chew Shern L , Drake William M , Monson John P , Besser G Michael , Grossman Ashley B
Published: February 2013

Objective

To investigate the early and late outcomes of patients with Cushing's disease (CD) submitted to a neurosurgical procedure as first-line treatment.

Design

In this single-centre retrospective case notes study, 131 patients with CD with a minimum follow-up period of 6 years (124 operated by transsphenoidal surgery (TSS) and seven by the transcranial approach) were studied. Apparent immediate cure: post-operative 0900 h serum cortisol level <50 nmol/l; remission: cortisol insufficiency or restoration of ‘normal’ cortisol levels with resolution of clinical features; and recurrence: dexamethasone resistance and relapse of hypercortisolaemic features.

Results

In patients operated by TSS, remission of hypercortisolaemia was found in 72.8% of 103 microadenomas and 42.9% of 21 macroadenomas, with recurrence rates 22.7 and 33.3% respectively with a 15-year mean follow-up (range, 6–29 years). Of 27 patients with microadenomas operated after 1991, with positive imaging and pathology, 93% obtained remission with 12% recurrence. In multivariate analysis, the time needed to achieve recovery of hypothalamo-pituitary–adrenal axis was the only significant predictor of recurrence; all patients who recurred showed recovery within 3 years from surgery: 31.3% of patients had total hypophysectomy with no recurrence; 42% of patients with selective adenomectomy and 26.5% with hemi–hypophysectomy showed recurrence rates of 31 and 13% respectively (χ2=6.275, P=0.03). Strict remission criteria were not superior in terms of the probability of recurrence compared with post-operative normocortisolaemia.

Conclusions

Lifelong follow-up for patients with CD appears essential, particularly for patients who have shown rapid recovery of their axis. The strict criteria previously used for ‘apparent cure’ do not appear to necessarily predict a lower recurrence rate.
Topics in Neonatal Informatics: Infants and Data in the Electronic Health Record Era
Authors: Pallotto Eugenia K. , Hunt Patricia G. , Dykes Francine D. , Durand David J. , Murthy Karna
Published: February 2013

Adoption of electronic health records continues to proceed rapidly. Neonatology offers unique barriers that must be addressed within the continued development of the electronic health record. Reducing these barriers can lead to improvements in the quality and safety of health-care delivery for the neonatal patient.
Neonatal Cholestasis
Authors: Feldman Amy G. , Sokol Ronald J.
Published: February 2013

Cholestatic jaundice is a common presenting feature of neonatal hepatobiliary and metabolic dysfunction. Any infant who remains jaundiced beyond age 2 to 3 weeks should have the serum bilirubin level fractionated into a conjugated (direct) and unconjugated (indirect) portion. Conjugated hyperbilirubinemia is never physiologic or normal. The differential diagnosis of cholestasis is extensive, and a step-wise approach based on the initial history and physical examination is useful to rapidly identify the underlying etiology. Early recognition of neonatal cholestasis is essential to ensure timely treatment and optimal prognosis. Even when specific treatment is not available, infants who have cholestasis benefit from early medical management and optimization of nutrition. Future studies are necessary to determine the most reliable and cost-effective method of universal screening for neonatal cholestasis.
Neonatal Thrombocytopenia
Authors: Fernández Karen S. , de Alarcón Pedro
Published: February 2013

Thrombocytopenia is one of the most common hematologic problems in the neonate. It affects up to 30% of all patients admitted to the neonatal intensive care unit (NICU). The causes of thrombocytopenia in neonates are diverse and include immune, inherited, and acquired disorders. The evaluation of the neonate with thrombocytopenia may be challenging. Developing a diagnostic strategy to evaluate the neonate with thrombocytopenia is key for the practicing clinician. Here, we provide a practical approach to the evaluation of the neonate with thrombocytopenia and an overview of its most common etiologies.