bluuebunny/crossref_metadata_2025 huggingface dataset:
Neutron Fluence Determination for Light Water Reactor Pressure Vessels
Authors: Gold R
Published: March 2009

A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced.
Maine Yankee Dosimetry Capsule and Pressure Vessel Neutron Fluence Calculations
Authors: White JR , Spinney KB , Morrissey KJ , Cacciapouti RJ
Published: March 2009

In-house capability for deterministic neutron and gamma transport analyses has been implemented at Yankee Atomic Electric Company (YAEC). A detailed R-Theta (R-θ) calculational model of Maine 4ankee was developed to help in validation of the methods and to establish appropriate models for support of the ongoing Maine Yankee pressure vessel surveillance program. Several data and modeling sensitivity studies were performed and comparisons to measured dosimetry capsule data were emphasized. The calculated results establish confidence in the YAEC in-house computational methodology for general pressure vessel fluence analyses.
Analysis of PWR Pressure Vessel Surveillance Dosimetry with McBend
Authors: de Wouters RM , Laurent DJ , D'Hondt PJ
Published: March 2009

Fast fluences received by four surveillance capsules in two 900-MWe PWR have been calculated with the Monte Carlo transport program MCBEND and compared with values estimated from activation measurements. First comparisons between calculation (C) and experiment (E) gave C/E values equal to 0.98/0.97/0.98/0.90. The original experimental values were based on saturation activation rates obtained from the measured activities and from the reactor power history, assigning at each operation cycle a contribution proportionnal to its time-integrated power production. A revision was then carried out in order to inlcude the effects of the fuel loadings (low leakage or out-in) in the contribution of each cycle. The revised experimental values led to C/E equal to 0.98/1.00/1.00/0.99, thus considerably reducing the scatter, and confirming the suitability of MCBEND to provide an essentially unbiassed approximation of the fast flux at the level of the surveillance capsules.
Three-Dimensional Discrete Ordinates Radiation Transport Calculations of Neutron Fluxes for Beginnig-of-Cycle at Several Pressure Vessel Surveillance Positions in the High Flux Isotope Reactor
Authors: Pace JV , slater CO , Smith MS
Published: March 2009

The objective of this research was to determine improved thermal, epithermal, and fast fluxes and several responses at mechanical test surveillance location keys 2, 4, 5, and 7 of the pressure vessel of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) for the beginning of the fuel cycle. The purpose of the research was to provide essential flux data in support of radiation embrittlenient studies of the pressure vessel shell and beam tubes at some of the important locations.
Fluence Surveillance by Scraping Samples from the Inner Surface of the Thermal Shield in the Nuclear Power Plant Obrigheim in Germany (KWO).
Authors: Polke E
Published: March 2009

To reduce the end of life fluence of the KWO rpv, the core loading was modified in different steps. The influence of different core loading strategies on the neutron flux distribution at the inner rpv surface was determined by neutron transport calculations with the code DOT in (R-θ)-geometry and by analysis of samples scraped from the inner surface of the thermal shield. Samples were taken during the 7th, 9th, 11th, 18th, and 22nd refueling outages. Up to 38 samples from different azimuthal and axial positions were collected during each sampling campaign. The fluences determined by the Mn54 and Nb93m activities were in good agreement with the corresponding theoretically calculated fluences.
Koeberg Surveillance Capsule Fluence Determination Using PCDC Dosimetry Cross Sections and Activation Measurements
Authors: Stoker CC , Reitsma F , Lamparelli G
Published: March 2009

Neutron fluences were calculated for the first Reactor Vessel Surveillance (RVS) capsules extracted from Koeberg Unit 1 and 2. For the calculation of the Koeberg RVS-capsule dosimetry cross sections a calculational system (called PCDC) was developed on a 32-bit i386/i486 personal computer (PC). This system consists of pre- and post-processors, a 171-group neutron reaction cross section library, a 171-group reactor zone flux spectrum library and an interface with the SAND-II, DOSCROS-84 and IRDF-90 dosimetry libraries. The sensitivity of these cross sections was determined for different reactor model parameters, core conditions and dosimetry libraries. Using the IRDF-90 dosimetry library, a maximum difference of 15% was obtained between the fluences predicted with the non-fissionable dosimeters.
Boron Determinations in Pressure Vessel Steels
Authors: Oliver BM , McElroy WN , Kellogg LS , Farrar H
Published: March 2009

Several studies have suggested that low-energy neutrons contribute to reactor pressure vessel (PV) embrittlement through interactions with boron impurities in the steel. Until now, the available information on boron contents in pressure vessel steels has been based on nominal concentrations or estimates provided by the materials manufacturers. To help resolve the question of boron contribution to PV steel embrittlement, samples of 38 different PV steels were analyzed by high-sensitivity gas mass spectrometry for their helium and boron contents. The boron contents were determined by measuring the increase in helium content in each material as a result of additional thermal neutron exposure. The results of these analyses showed natural boron contents that ranged from 0.23 to 5.11 wt. ppm in the various alloys.
Mono-Energetic Mono-Directional Resonance Neutron Activation of Natural Indium Metal Target
Authors: Ertek C , Oigawa H
Published: March 2009

The mono-energetic neutrons of 1.456eV energy are obtained from 1 MWth TR-I SWIMMING POOL TYPE RESEARCH REACTOR using double collimated beams and BRAGG reflection of pure Beryllium mono-crystal with extremely fine energy resolution.

Foil thickness for 3 foils were 26.,28. and 44.10-4 cm and they were perpendicular to the beam of mono-energetic neutrons and were irradiated in sandwich form. After irradiation, the saturation activities were obtained using Phillips two-pie special “beta-ray” detector in a well controlled and shielded geometry. Counting reproducibility was excellent(better than 0.1%). SPECIAL ATTENTION was paid to the irradiated(side A) and non-irradiated(sideB) sides of the foils. Usual irradiation and decay corrections were applied to obtain the saturation activities.

In this work, the preliminary calculations of reaction rates using Nakazawa M. et al.,JENDL Dosimetry file, JAERI 1325,(1992) were performed. Considerable differences are found between the calculations and experiment and possible reasons are still under investigation. The preliminary calculations of reaction rates using ENDFB/VI are in agreement with JENDL-3 estimates. Absolute reaction rate estimates are not yet ready.

Considerable numbers of research centers are interested in the experiment and very constructive inputs are expressed and obtained from Hiroyuki Oigawa, Shigeaki Okajima and T. Mukaiyama, JAERI, Japan; N.P. Baumann and K.O. Ott, USA; E. Zsolnay and E. Szondy, Hungary; M.C. Lopes and J. Molina,Portugal; F. Bensch, H. Böck Austria; and M. Turgut and A.Ìsyar,Turkey, Investigations using collision theory,multiple scattering and monte-carlo techniques have been undertaken.
Neutron Flux Estimations Based on Niobium Impurities in Reactor Pressure Vessel Steel
Authors: Baers LB , Hasanen EK
Published: March 2009

The use of (ppm level) niobium impurities in reactor pressure vessel (RPV) steel for neutron flux estimations based on the reaction 93Nb (n,n′) 93mNb has been reported previously [1].

The method has now been further investigated and refined.

Small niobium fractions in RPV steel (≈ ppm) and plating (≈1 %) materials have been separated by ion exchange chromatography in one to three steps. The measured Nb fractions in samples from some four pressure vessel (RPV) base materials were 1 to 3 ppm. The purification of tens of milligrams of RPV material provides sufficient amounts of niobium for mass determination with a highly sensitive (10−5 ppm) Inductively Coupled Plasma Mass Spectrometer (ICP-MS).

The 93mNb and small remaining 54Mn activities were measured with a calibrated Liquid Scintillation Counter (LSC) based on dual label technique and almost 100 per cent efficiency to 93mNb. One purification is needed for plating materials (≈1 % Nb) and two purifications of about one gram of steel with Nb impurities in order to resolve the needed activities (≈10 Bq 93mNb/ug Nb).

Other neutron induced reactions in Mo produce interfering 93mNb, 92mNb and 95Nb. With about 0.5 per cent Mo in the RPV steel, several tens of ppm of Nb (Nb/Mo > 0.01) are typically needed in order to obtain a clearly dominant 93mNb activity from the 93Nb (n,n′) 93mNb reaction of interest. RPV plating materials with 0.5...1 % of alloyed Nb have insignificant degradation from Mo on the measured 93mNb activities.

Thus, a sufficient chemical purification of Nb was achieved and the specific 93mNb activities could be measured for all types of available RPV and plating samples. However, other competing and 93mNb producing reactions such as 92Mo(n,γ)93Mo (3500 y), disintegrating to 93mNb, strongly degraded the results in samples with low Nb/Mo ratios. Samples with sufficiently high Nb/Mo ratios may be eventually selected for dosimetric purposes since the Nb and Mo mass fractions vary with position and type of material.

The achieved accuracy of the measured specific 93mNb activities was about ±3 % (lσ) in irradiated RPV plating materials and about ±4 % for Nb ppm impurities.
Neutron Dosimetry and Mechanical Properties Changes Monitoring by Magnetic Response of Fe and Ni Wires
Authors: Shong W-J , Stubbins JF , Williams JG , Rogers JW , Giacobbe M
Published: March 2009

Magnetic properties may be a useful means for monitoring the evolution of irradiation-induced damage in materials which are inherently ferromagnetic. Such materials include pure iron and nickel which are commonly used in a variety of irradiation applications for fluence or spectrum analysis. One such application is for fluence monitoring in Charpy coupon surveillance packets. This study examines the magnetic response of both iron and nickel wires which have been irradiated to doses up to 3.6×1019 cm−2 (E>lMeV) in such surveillance packets at reactor ambient irradiation temperatures of approximately 290°C. The results show that large changes in magnetic response are found for magnetic properties which are sensitive to materials microstructure. The changes are predominantly to decrease the magnetic remanence and coercivity with dose. These changes saturate at the highest doses examined in this study. Though direct microstructural examination is needed to confirm the hypothesis, the results are consistent with the development of a large population of defect structures (vacancy and interstitial clusters) which evolve with irradiation dose.
Materials Prepared at the Institute for Reference Materials and Measurements (IRMM) for Reactor Dosimetry Applications
Authors: Ingelbrecht CD , Peetermans FW , Palmeri S , Robouch PB
Published: March 2009

Thirteen reference materials prepared under the Euratom Working Group on Reactor Dosimetry (EWGRD) monitor materials programme are now available. The determination of trace impurities has recently been completed for rhodium metal (iridium 26.0 ± 0.6 mg·kg−1, platinum < 5 mg·kg−1) and these values will be proposed for certification. Certification analyses of titanium metal, which are continuing, indicate scandium < 0.1 mg·kg−1.

The preparation of alloys for neutron fluence rate measurements are discussed. Satisfactory homogeneity in aluminium alloys can be difficult to achieve on a milligramme scale due to the formation of coarse intermetallic particles, and suitable processing with a high solidification rate, cold working and annealing is essential. The example is given of the alloys Al-15 wt% U and Al-14 wt% Pu, where the inhomogeneity (in terms of relative standard deviation of alloying element) was determined by isotope dilution mass spectrometry as < 0.2% in each case for 20 mg samples.

Some recent examples of dosimeter encapsulation are given including NpO2 and UO2 microspheres in high purity vanadium and Al-Pu and Al-U fission foils in cold-pressed aluminium containers.
Signal Anticipation of Rhodium Self Powered Neutron Detectors for On-Line Applications
Authors: Lacaille R , Jegat A
Published: March 2009

rhodium self powered neutron detectors (SPND) have been used for a long time in the experimental devices at the SILOE reactor.

In a time-dependant neutron flux, the rhodium SPND behaves as a first order linear system. Consequently, the response of the SPND is not similar to the incident flux evolution because time constants are involved. In order to draw-out the true flux variation, a convolution type method is to be used.

As a result of the time discretisation of the linear differential system that couples the SPND current to the neutron flux, one has to use differential equations. The resolution of these equations is equivalent to an approximate inversion of the initial system.

The numerical method described above has been applied in a microcomputer system, in order to treat the SPND signal by a computer-driven millivolmeter.

Nowadays, devices for transient experiments in the SILOE reactor are driven by a P.C. system, based on this numerical approach and using an independent data acquisition; the reliability and the ergonomics aspects have been taken into account. The modular conception of the system allows easy further developments for flux monitoring in various surroundings.
Research for Application of Long Counter with a Position-Sensitive Proportional Counter to Neutron Dosimetry
Authors: Toyokawa H , Uritani A , Mori C , Takeda N , Kudo K
Published: March 2009

Characteristics of a long counter with a position-sensitive 3He proportional counter were examined by means of experiments and Monte Carlo simulations to explore the possibility of applying it to incident neutron energy estimation and dose equivalent evaluation. The long counter was energy sensitive, and can be used for the rough estimation of the energy of beam neutrons. The long counter could be used as an energy- and direction-sensitive dosimeter.
Response Matrices of NE213 Scintillation Detectors for Neutrons
Authors: Guldbakke S , Klein H , Meister A , Pulpan J , Scheler U , Tichy M , Unholzer S
Published: March 2009

Four NE213 detectors of different size have been calibrated at the accelerator facility of the PTB. The response functions were experimentally determined for 3 3 neutron energies between 1 MeV and 16 MeV and compared with Monte Carlo simulations using the NRESP7 code. The light output functions for recoil protons were found to be significantly different for all detectors even if they were of the same size. The neutron fluence determined on the basis of the response functions calculated with the corresponding light output functions agreed to better than ± 2% with reference values if energy independent adjustment factors between 0.98 and 1.03 were applied. The response matrices required for the unfolding of neutron induced pulse height spectra were therefore calculated with the NRESP7 code taking into account the adjustment factors. Similarly, the response matrices for photons were calculated with the EGS4 code, but without any adjustment. Finally, the DIFBAS code was applied for the unfolding of pure neutron- and photon-induced pulse height spectra. The resulting spectral fluences are in reasonable agreement with the results obtained by time-of-flight measurements and by spectrometry with a Ge detector.
Neutron Fluence Determination at Reactor Filters by 3He Proportional Counters: Comparison of Unfolding Algorithms
Authors: Matzke M , Alberts WG , Dietz E
Published: March 2009

Multichannel pulse height measurements with a cylindrical 3He proportional counter obtained at a reactor filter of natural iron are taken to investigate the properties of three algorithms for neutron spectrum unfolding. For a systematic application of uncertainty propagation the covariance matrix of previously determined 3He response functions is evaluated. The calculated filter transmission function together with a covariance matrix estimated from cross-section uncertainties of the filter material is used as fluence pre-information. The results obtained from algorithms with and without pre-information differ in shape and uncertainties for single group fluence values, but there is sufficient agreement when evaluating integrals over neutron energy intervals.
Analysis of the VENUS-1 Ex-Core Neutron Dosimetry Using the Lepricon Code System
Authors: Abderrahim HA , Hort M
Published: March 2009

The VENUS-1 Mock-up ex-core neutron dosimetry is analysed by means of the LEPRICON code system using the ELXSIR 56-group cross section library. Two analyses are performed: the first one using the old iron cross-sections, the second using the new evaluation of the inelastic iron cross-sections present in the ELXSIR library. Results, when compared with those from earlier analysis show the clear improvement of the Calculation-to-Experiment (C/E) ratios. The decreasing radial trend in the C/E observed in the former analysis is drastically reduced due to the finer energy group structure of the ELXSIR library (56 instead of 17 groups). This trend is still improved by the use of the iron cross-section evaluation of FU. This leads to a flat radial C/E (around a value of 1.0) from the edge of the core until the neutron pad. The objectives of this new analysis are triple: i) to quantify the improvement of the calculated detector responses when using the ELXSIR library, ii) to display the consistency of the VENUS-1 ex-core dosimetry results with the intention of their integration in the LEPRICON Associated Benchmark Data Base, iii) to show the drastic reduction of the standard deviation on the adjusted results leading to demonstrate that the LEPRICON methodology is adequate for the neutron surveillance dosimetry analysis.
Evaluation of the Results of the B&W Owners Group Cavity Dosimetry Benchmark Experiment
Authors: Petrusha L , Garat C
Published: March 2009

The B&W Owners Group (B&WOG) Cavity Dosimetry Benchmark Experiment was designed to provide accurate data from diverse neutron dosimetry in the cavity surrounding the reactor vessel and from two standard surveillance capsules located within the vessel. The dosimetry was irradiated for one fuel cycle of operation and analyzed. A detailed analysis using the DOT4. 3 neutron transport code was then performed to obtain calculated dosimeter responses at each dosimeter location. This paper presents the comparison between measured and calculated dosimeter responses for both in-vessel and ex-vessel dosimetry, a description of the procedure to calculate the three-dimensional synthesized fluxes at each cavity dosimetry location, and the methodology to derive a groupwise correction factor for application to the analytical flux to achieve a more accurate prediction of the fast fluence at the pressure vessel.
Pool Critical Assembly Benchmark Solutions Using MCNP and Threedant
Authors: Urban WT , Alcouffe RE , Parsons DK , Marr DR , Clark BA
Published: March 2009

Analyses of pressure vessel damage resulting from neutron irradiation have primarily relied on two-dimensional transport calculations and a spatial-synthesis methodology to accommodate three-dimensional effects in the results of two two-dimensional calculations. In this paper, we report on calculations made on the Pool Critical Assembly (PCA) Benchmark, Configuration 12/13, using the three-dimensional, continuous energy Monte Carlo transport code, MCNP, and the three-dimensional, multigroup, diffusion accelerated discrete ordinates transport code THREEDANT. Neutron fluxes and activation rates as determined from these two calculations are compared to each other and to experimental results in the literature. We also draw some conclusions on the value of 3D calculations on the interpretation of experimental results.
Gamma-Ray Dose Mapping in Operational CANDU Reactor Containment Areas Using MOS Dosimeters
Authors: Mackay G , Thomson I , Hirning CR , Zeya M
Published: March 2009

It is important to determine the radiation dose received by key operational equipment in the containment areas of CANDU reactors. A dosimetry system was developed to evaluate the radiation environment under operational conditions. The key component in the system is a passive MOS (semiconductor) dosimeter which is hermetically sealed and is not affected by the high humidity in the containment area. Doses were measured in a number of Ontario Hydro operating reactor sites over periods of 4 000 to 11 000 hours. As a result of this dose mapping study, several remedial actions have taken place and it is recommended that similar reactors undergo this type of study.
Characterization of the Gamma Field in the B%W Owners Group Cavity Dosimetry Benchmark Experiment
Authors: King SQ , Gilliam DM
Published: March 2009

The B%W Owners Group (BWOG) Cavity Dosimetry Benchmark Experiment included, in addition to various types of neutron dosimeters, lithium-7 fluoride (LiF) optical absorbance gamma dosimeters placed at eight cavity locations. The purpose of these dosimeters was to indirectly measure the gamma field in the vicinity of the neutron dosimeters. These measured data were used in a semi-empirical procedure to characterize the gamma-ray dose rate distribution in the power reactor cavity.

The LiF chip dosimeters were irradiated for one cycle in the high-level mixed neutron-gamma field that existed in the power reactor cavity. The 380.3-day cycle length was sufficient to ensure absorbed doses in the tenths of megagrays. After being removed from the reactor, the LiF chips were shipped to the National Institute of Standards and Technology (NIST) for measurement of the optical absorbance, annealing, and re-calibration in a known gamma ray field. The neutron flux spectrum at each LiF chip location was calculated using a semi-empirical methodology that reconciles measured neutron dosimeter data to calculated neutron spectra. Appropriate energy-dependent normalization factors were developed and applied to the a priori flux calculated by the DOT4.3 discrete ordinates transport code. The neutron flux spectra were then used to determine the total neutron dose absorbed in each LiF chip. Finally, the calculated neutron dose was subtracted from the measured (total) dose, yielding the integrated gamma-ray dose. Calculated-to-measured (C/M) gamma dose ratios were then established for each LiF chip and are reported herein.
Overview of the Activities in Spain on Irradiation Embrittlement of RPV Steel
Authors: Bros J , Ballesteros A
Published: March 2009

Specific problems such as the embrittlement of the WER reactors or the integrity evaluation of the Yankee Rowe reactor pressure vessel, show the necessity of continuing in the understanding of the mechanisms responsible for the embrittlement. Besides, in order to increase the safety and performance of the nuclear power plants, it is important to optimize the surveillance programmes and develop the mitigation techniques of the damage caused by neutron radiation.

The main activities carried out in Spain are related to the surveillance programmes and to phase III of the IAEA Coordinated Research Programme, although other smaller projects also show the interest of the utilities in the embrittlement issues. The extension of the work performed as well as some of the most relevant results are shown in this paper. Some new projects are being considered. The participation in the recently created European Action Group for RPV Materials Irradiation Effects and Studies (AMES) is of huge value for Spain. This Group try mainly give a European response in the field of the standards. In spite of the nuclear moratorium that has been in vigour in Spain for several years and the present economic crisis, there is a general interest in participation in international projects. That is the case of the Spanish contribution to the certification of Avanced Light-Water Reactors. In particular, regarding the embrittlement problem, in the development of a reactor pressure vessel and support skirt material surveillance programme for the Simplified Boiling Water Reactor (SBWR).
The Irradiation Facility Korpus for Irradiation of the Reactor Structure Materials
Authors: Brodkin EB , Egorov AL , Golovanov VN , Markina NV , Raetsky VM , Sulaberidze VS , Tsikanov VA , Zaritsky SM
Published: March 2009

The results concerning the irradiation embrittlement of WWER vessel materials prove the necessity of organizing experiments in controllable and adjustable conditions ensuring the unequivocal interpretation of the change of material properties under irradiation. The KORPUS facility for RBT-6 reactor was designed and made for the realization of controllable irradiation with the possibility of a simulation of reactor WWER vessel materials working conditions. The KORPUS design and technical characteristics are considered in this paper. Its experimental possibilities from the point of view of material testing and reactor dosimetry execution are discussed. The precomputation results indicative of possible simulation of vessel material working conditions are presented. The proposals on creating dosimetry experimental conditions are formulated.
Intercomparison of Neutron Test Facilities for Equivalent Damage to Electronics
Authors: Heimbach CR
Published: March 2009

Neutron-sensitive diodes are practical monitors for comparing damage delivered to electronic devices among different test facilities. Heat treatment of the diodes before readout is shown to eliminate effects of temperature variations between exposure and readout. This allows the comparisons to be done on a mail-in basis. The results of such a comparison among several neutron facilities is presented.
The Test Reactor Embrittlement Data Base (TR-EDB)
Authors: Stallmann FW , Kam FBK , Wang JA
Published: March 2009

The Test Reactor Embrittlement Data Base (TR-EDB) is part of an ongoing program to collect test data from materials irradiations to aid in the research and evaluation of embrittlement prediction models that are used to ensure the safety of pressure vessels in power reactors. This program is being funded by the US Nuclear Regulatory Commission (NRC) and has resulted in the publication of the Power Reactor Embrittlement Data Base (PR-EDB), the second version of which is currently being released. The TR-EDB is a compatible collection of data from experiments in materials test reactors. These data contain information that is not obtainable from surveillance results, especially, about the effects of annealing after irradiation. Other information that is only available from test reactors is the influence of fluence rates and irradiation temperatures on radiation embrittlement. The first version of the TR-EDB will be released in the fall of 1993 and contains published results from laboratories in many countries. Data collection will continue, and further updates will be published.
A Plant-Specific Upper Shelf Energy Drop Methodology
Authors: Manahan MP , Cuddy LJ , Peterson AJ
Published: March 2009

Appendix G to 10 CFR 50 requires that licensees evaluate all reactor beltline materials to determine whether the Upper Shelf Energy (USE) screening criterion will be exceeded prior to the end-of-license (EOL). Measurement of the unirradiated USE and estimation of the EOL irradiated USE level are key elements of the screening criterion evaluation. The current Regulatory Guide 1.99 (Revision 2) USE drop model does not provide adequate representation of the actual shelf drop behavior for some reactor materials. Therefore, research was conducted to identify the important factors which should be included in the development of a physically-based USE trend curve model and to test the validity of the material-specific modelling approach.

As a result of this investigation, it has been concluded that there are three radiation damage mechanisms which may have significant influence on the ductile fracture process: precipitation and particle coarsening due to radiation enhanced diffusivity; cascade induced matrix damage; and possibly surface active element transport to grain boundaries and particle interfaces. At present, it is believed that the largest micromechanical effect is due to the impact of precipitates and matrix damage on the hardening of the material bridges between voids. However, additional experimental data are needed to confirm the dominant mechanisms which contribute to the USE drop.

In the absence of an experimentally verified microstructural model, work has been performed to test the efficacy of the material-specific approach. The validity of the material-specific approach has been demonstrated by reducing the Light Water Reactor (LWR) database to produce a data set with comparable chemical composition, heat treatment, and neutron flux spectra. Based on the work reported herein, it has been observed that a drop in the unirradiated USE requires an incubation dose. The reduced data set correlates well with the square root of the dose, indicating the importance of the fine scale precipitates and the matrix damage component.
Calculated and Measured Gas Formation in Beryllium Samples Irradiated in the High Flux Materials Testing Reactor BR2
Authors: De Raedt CM , Sannen LF , Vanmechelen PJ , Oliver BM
Published: March 2009

Beryllium samples have been irradiated in BR2, the materials testing reactor of the Nuclear Research Centre SCK/CEN at Mol, Belgium, up to fission fluence values of 5.2 1022 n/cm2 at low temperature.

The gas formation (helium, tritium), as measured by SCK/CEN, as well as the induced swelling of the beryllium samples and the enhancement of the swelling due to annealing have been presented at the 17th SOFT Conference (Rome, 14–18 Sept., 1992).

Since this conference, helium measurements on the same samples have been carried out at RI and calculations of the gas production have been performed, taking into account the various formation schemes.

The experimental results from SCK/CEN and from RI are compared with the calculated gas formations.
The Spallation Neutron Source SINQ and Related Dosimetry Problems
Authors: Janett A , Atchison F , Bauer GS
Published: March 2009

The spallation neutron source SINQ, presently under construction at Switzerland's Paul Scherrer Institut, will handle the highest proton current of any comparable facility in the world: a continuous beam of 1.5 mA, 590 MeV protons from an isochronous ring cyclotron. For the users, SINQ as a neutron source should resemble closely a medium flux research reactor; the presence of high and medium energy particles creates new technical problems in design and operation. The engineering design for the major components is based on the results of neutronic calculations, using a code package built around the HETC program.

At present, opportunities to verify the theoretical calculations experimentally are very limited. Safety factors have to be built in which conflict with the optimation of SINQ as a neutron source. To benchmark the calculational methods, a wide ranging diagnostic system will be required.
Dosimetry Aspects of the New Canadian MAPLE-X10 Reactor
Authors: Lidstone RF , Wilkin GB
Published: March 2009

Atomic Energy of Canada Limited is building the 10-MWt MAPLE-X10 reactor facility as a dedicated producer of medical and industrial radioisotopes. Dosimetry aspects of the MAPLE-X10 nuclear design include the calculated thermal and fast neutron flux distributions throughout the reactor assembly and the rate of heat generation in reactor materials and components. Examples of the resolution of design issues are also presented, such as the use of fission counters and ion chambers to provide diverse methods of detecting neutron flux levels and the use of the difference between photon and neutron signals to guard against the effects of downgrading of the heavy-water reflector. Computer codes employed in the calculations include MCNP, ONEDANT, WIMS-AECL, and 3DDT.
Neutron Metrology in the Fuel Assemblies of a Material Test Reactor.
Authors: Voorbraak WP , Paardekooper A , Polle BN , Freudenreich WE
Published: March 2009

Results are presented of detailed thermal and fast neutron measurements performed in all fuel and control assemblies of the HFR in Petten. The results give information about deviations of a general shape of vertical thermal and fast fluence rate distributions due to material transitions in the reactor core and different control assembly settings. Further it is demonstrated that the ratio of fast and thermal fluence rate at the various monitor positions in the assemblies give useful information for the (relative) local burn-up of the fuel.
Two-Dimensional Neutron and Gamma Calculations for Devices Irradiated in BR2 Reflector Positions
Authors: De Raedt CM , Bodart SF , Ponsard BJ , Wéber MJ , Maldague T
Published: March 2009

For two typical LWR fuel rod irradiation devices introduced into the BR2 reflector two-dimensional neutron and gamma calculations are being carried out.

In the first case a single fuel rod surrounded by several pressure tubes and by a time-variable neutron screen (He−3 gas at various pressures) is studied. The calculations are carried out in R,θ geometry, centered on the irradiated fuel rod.

In the second case a bundle of nine fuel rods in a 3×3 square lattice is irradiated. Here the calculations are performed in X,Y geometry.

The necessity is discussed of performing two-dimensional calculations for these irradiations in the BR2 reflector, to improve the modelling of the fuel behaviour.
Determination of the Neutron Flux for the Yankee Rowe Experiment in the Ford Nuclear Reactor
Authors: Cacciapouti RJ , Petrusha L
Published: March 2009

Yankee Atomic Electric Company undertook a Test Irradiation Program at the Ford Nuclear Reactor of the University of Michigan. The program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials. The program was also intended to remove uncertainties in the existing reactor vessel fluence and damage predictions on the Yankee Rowe reactor vessel steel. Since this is the first in-core experiment of this type for the Ford Nuclear Reactor, the measurement of the reaction rate and the estimate of the fluence are presented.
Testing of the IRDF-90 Cross-Section Library in Benchmark Neutron Spectra
Authors: Nolthenius HJ , Zsolnay éM , Szondi EJ
Published: March 2009

The new version of the International Reactor Dosimetry File IRDF-90 (called “Version April 1993”) has been tested by calculatio of average cross-sections and their uncertainties in a coarse three energy group structure and by neutron spectrum adjustments in reference neutron spectra. This paper presents the results and compares them with the corresponding ones of the old Reactor Dosimetry File IRDF-85 and with the data of the Nuclear Data Guide for Reactor Neutron Metrology. Finally, the applicability of the new library in the field of neutron metrology is discussed.
Production and Testing of the Vitamin-B6 Fine-Group and the BUGLE-93 Broad-Group Neutron/Photon Cross-Section Libraries Derived from ENDF/B-VI Nuclear Data
Authors: Ingersoll DT , White JE , Wright RQ , Hunter HT , Slater CO , MacFarlane RE , Greene NM
Published: March 2009

A new multigroup cross-section library based on ENDF/B-VI data has been produced and tested for light water reactor shielding and reactor pressure vessel dosimetry applications. The broad-group library is designated BUGLE-93. The processing methodology is consistent with ANSI/ANS 6.1.2, since the ENDF data were first processed into a fine-group, “pseudo problem-independent” format and then collapsed into the final broad-group format. The fine-group library is designated VITAMIN-B6. An extensive integral data testing effort was also performed. In general, results using the new data show significant improvements relative to earlier ENDF data.
Status of Neutron Dosimetry Cross Sections
Authors: Griffin PJ , Kelly JJG
Published: March 2009

Several new cross section libraries, such as ENDF/B-VI (revision 2), IRDF-90, JEF-2.2, and JENDL Dosimetry, have recently been made available to the dosimetry community. The Sandia National Laboratories (SNL) Radiation Metrology Laboratory (RML) has worked with these libraries since pre-release versions were available. This paper summarizes the results of the intercomparison and testing of dosimetry cross sections. As a result of this analysis, a compendium of the best dosimetry cross sections was assembled from the available libraries for use within the SNL RML. This library, referred to as the SNLRML, contains 66 general dosimetry sensors and 3 special dosimeters unique to the RML sensor inventory. The SNLRML cross sections have been put into a format compatible with commonly used neutron spectrum determination codes.
Evaluation of the 93Nb(n,n') 93mNb and 93Nb(n,2n) 92mNb Reactions Cross-Sections from Threshold to 20 MEV.
Authors: Badikov SA , Zolotarev KI
Published: March 2009

The evaluation of the 93Nb(n,n') 93mNb and 93mNb(n,2n) 92mNb reactions cross-sections important for reactor dosimetry was carried out by means of Pade-approximation. The experimental data bases comprise the results of cross-sections measurements performed up to January 1993. The preliminary analysis of experimental cross-sections has been made. The evaluated cross-sections are presented in an analytical form convenient for the use in applications. The files of evaluated cross-sections (File 3) and covariances (Files 30 and 33) in the ENDF-VI format were prepared. The comparison of evaluations from this work and IRDF-90 was performed.
Measurements of Thermal Neutron Cross Section and Resonance Integral for the 237Np(n,γ)238Np Reaction
Authors: Kobayashi K , Yamanaka A , Kimura I
Published: March 2009

Making use of a standard neutron spectrum field with a pure Maxwellian distribution at the heavy water thermal neutron facility of the Kyoto University Reactor (KUR), the thermal neutron cross section for the 237Np(n,γ)238Np reaction was measured by the activation method, using a high purity Ge detector. The result is 158±3 b, which is obtained relative to the reference value of 98.65±0.09 b for the 197Au(n,γ)198Au reaction. The present value is lower by about 13% than that of the ENDF/B-VI data. The data given by Mughabghab and obtained from JENDL-3 are larger by 11 to 14% than the present measurement.

The resonance integral for the 237Np(n,γ)238Np reaction was measured with a 1/E standard neutron spectrum in the cavity of the central graphite reflector region between the two-divided cores of the Kinki University Reactor (UTR-KINKI), relative to the reference value of 1550±28 b for the 197Au(n,γ)198Au reaction. By defining the Cd cut-off energy as 0.5 eV, the present resonance integral is 652±24 b, which is in good agreement with the JENDL-3, ENDF/B-VI and Mughabghab data. However, most of the old experimental data are, in general, larger by 24 to 38% than the present measurement.
Measurement and Analysis of the Nitrogen Total and Differential Scattering Cross Sections in the Resonance Region
Authors: Beasley CO , Larson NM , Harvey JA , Larson DC , Hale GM
Published: March 2009

High-resolution neutron transmission and differential elastic scattering measurements have been made on samples of nitrogen from 0.5 eV to 10 MeV at the Oak Ridge Electron Linear Accelerator (ORELA). For the transmission measurement several different detectors were used at various flight path lengths, with high-purity nitrogen gas samples. The data have been corrected for the dead time of the time digitizer and for several small backgrounds. For the scattering measurements six detectors were placed at various angles inside an evacuated scattering chamber at 200 meters from the neutron producing target. The sample was silicon nitride held in a thin graphite container. Scattering from carbon and silicon samples was also measured to correct for these materials and calibrate the relative efficiencies of the scattering detectors. The data have been corrected for dead time and for a constant background. A preliminary analysis has been completed to obtain parameters of the resonances up to 4.8 MeV.
Effect of Gadolinium Covers on Fast Dosimeter Response
Authors: Rutherford MA , King SQ
Published: March 2009

Gadolinium (Gd) has been used extensively at BSW Nuclear Technologies (BWNT) as a dosimetry shielding material for several years. Gadolinium has a significant cross section throughout the energy range of interest (i.e. above 1 MeV) in vessel fluence calculations. While Gd has a number of advantages as a dosimetry cover, it also has the potential for introducing a bias of unknown magnitude into the measured activities of covered fast response dosimeters such as U238 and iron. The Gd cover causes a reduction in the measured dosimeter activity, which means that the calculated fluxes normalized to those activities would be under-predicted. This is, of course, in the non-conservative direction. Since dosimeter covers are generally not included in the typical transport analysis model, corrections for this effect are not generally made. The calculated activities would not include the effects of the Gd while, of course, the measured activities would. The resulting calculation to measurement ratios, (C/M), which are used to remove biases in the calculated fluxes, would therefore be in error, in the non-conservative direction, by some undetermined amount.

This study has determined the magnitude of the bias in the fast response dosimeters, U238 and iron caused by the absorption of fast (E>1.0 MeV) neutrons in 2 mm (0.002 m) Gd covers. The results indicate that the bias is in the 2.9–3.4% range and that it is only a weak function of dosimeter material.
The DS86 Neutron Dosimetry Enigma: Some Missing Pieces to the Puzzle
Authors: Gold R
Published: March 2009

International programs have been conducted over the last four decades to quantify the exposure of atom bomb survivors from Hiroshima and Nagasaki. Unfortunately, the quest for accurate gamma-ray and neutron exposure doses of atom bomb survivors has proven illusive. Efforts in the most recent of these programs, designated as Dosimetry System 1986 (DS86), have revealed a serious and persistent discrepancy between neutron transport calculations and thermal neutron activation measurements at the Hiroshima site, which will be called the DS86 neutron dosimetry enigma. It is established that this enigma is a complex puzzle that precludes simple solutions. This conclusion is deduced through the identification of a number of missing pieces to the puzzle. Implications and conclusions that can be inferred from these missing puzzle pieces are advanced.
Spallation Radiation Damage and Dosimetry for Accelerator Transmutation of Waste Applications
Authors: Wechsler MS , Ferguson PD , Lin C , Sommer WF
Published: March 2009

Proposals are currently being made for systems to treat radioactive waste based on the use of accelerator-driven neutron sources. A linear proton accelerator with energies as high as 1600 MeV and currents up to 250 ma are anticipated for the driver. The neutron fluxes may reach up to 1020 neutrons/m2s as generated by the spallation reactions that occur when the protons strike target materials. Calculations are described to determine radiation fluxes and flux spectra inherent in such systems and to estimate likely radiation effects on system components. The calculations use LAHET, a Monte Carlo high-energy transport code, and MCNP, a generalized-geometry, coupled neutron-photon Monte Carlo transport code. Cross sections for displacement and helium production are presented for spallation neutrons of energies from 21 MeV to 1600 MeV for Inconel 718 (Ni plus 18.5, 18.5, 5.1, and 3 wt% of Cr, Fe, Nb, and Mo, respectively), an alloy that is used for the proton beam entry window in several accelerators. In addition, calculations for this alloy are presented for the primary knocked-on atom (PKA) spectrum and the transmutation yield for 1600 MeV incident neutrons.
Dosimetry of Medium Energy Neutrons and Protons in Material Irradiation Experiments
Authors: Hegedüs F , Victoria M
Published: March 2009

A program is in progress at PSI to study radiation damage in first wall materials of future fusion reactors. The program uses mainly the Proton IRradiation Experiment (PIREX) at the PSI 590 MeV ring accelerator. However, other radiation sources have been considered such as spallation neutron sources. An important aspect of this program is that of dosimetry and its relation to spallation impurity production. The neutron or proton fluence and the atomic displacement (dpa) was determined from the 22Na reaction rate measured in an aluminium detector. Neutron spectra, calculated by the HETC code, were used in the evaluation procedure. The concentration of transmutational elements was measured by means of the total reflection X-ray fluorescence (TXRF) method using synchrotron radiation as the primary X-ray source.
Neutron Cross Sections for the Interpretation of a Spallation Experiment
Authors: Bortignon PF , Mariani F , Perini A , Sangiust V
Published: March 2009

An irradiation was carried out in a spallation neutron spectrum at the beam stop of LAMPF (Los Alamos); many activation detectors were used in order to obtain a fluence mapping inside the capsule volume. The cross sections were derived from ENDF B/V Dosimetry File up to 20 MeV and were based on calculations with the code ALICE up to 200 MeV. From 200 to 800 MeV an empirical extrapolation was employed since no data, calculated or measured, were available at the moment. An attempt has been made in this paper to revisit the interpretation of the experiment reconsidering the cross sections in the energy range from 200 to 800 MeV as given by the semiempirical models of Rudstam, Silberger& Tsao and Hufner.
Development of Zirconium-Barrier Fuel Cladding
Authors: Armijo JS , Coffin LF , Rosenbaum HS
Published: March 2009

This paper was prepared for the 1991 Kroll Award. A review is presented of the development of barrier fuel. It includes the recognition of the pellet-cladding interaction (PCI) fuel failure mode and of a coordinated program to develop understanding, mitigating strategies, and a fuel that is resistant to this failure mode. The efforts to understand PCI led to the conclusion that the dominant mechanism is stress-corrosion cracking of the Zircaloy. The invention and development of zirconium-barrier fuel was intended to provide a materials solution to this fuel failure mode. This review includes the work to understand the failure mechanism as well as the program to develop PCI-resistant fuel designs. Ultimately, the zirconium-barrier fuel was tested in power ramps to ascertain and to quantify the resistance to PCI under expected service conditions in commercial boiling water reactors (BWRs). The program that led to a large-scale demonstration in a commercial power plant (Quad Cities-2) is described briefly. Subsequent to that, program work continued with in-reactor load following and experiments in a test reactor on power cycling of barrier fuel. Finally, the performance of failed fuel is discussed briefly.