OVERVIEW OF NEUTRON MEASUREMENTS IN JET FUSION DEVICE

The design and operation of ITER experimental fusion reactor requires the development of neutron measurement techniques and numerical tools to derive the fusion power and the radiation field in the device and in the surrounding areas. Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case in ITER and power plant studies. The required radiation transport calculations are extremely challenging because of the large physical extent of the reactor plant, the complexity of the geometry, and the combination of deep penetration and streaming paths. This article reports the experimental activities which are carried-out at JET to validate the neutronics measurements methods and numerical tools used in ITER and power plant design. A new deuterium-tritium campaign is proposed in 2019 at JET: the unique 14 MeV neutron yields produced will be exploited as much as possible to validate measurement techniques, codes, procedures and data currently used in ITER design thus reducing the related uncertainties and the associated risks in the machine operation.

Demand driven salt clean-up in a molten salt fast reactor - Defining a priority list

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified.

Evaluation of radioactivity in the bodies of mice induced by neutron exposure from an epi-thermal neutron source of an accelerator-based boron neutron capture therapy system

This study aimed to evaluate the residual radioactivity in mice induced by neutron irradiation with an accelerator-based boron neutron capture therapy (BNCT) system using a solid Li target. The radionuclides and their activities were evaluated using a high-purity germanium (HP-Ge) detector. The saturated radioactivity of the irradiated mouse was estimated to assess the radiation protection needs for using the accelerator-based BNCT system. ²⁴Na, ³⁸Cl, ^80mBr, ⁸²Br, ⁵⁶Mn, and ⁴²K were identified, and their saturated radioactivities were (1.4 ± 0.1) × 10², (2.2 ± 0.1) × 10¹, (3.4 ± 0.4) × 10², 2.8 ± 0.1, 8.0 ± 0.1, and (3.8 ± 0.1) × 10¹ Bq/g/mA, respectively. The ²⁴Na activation rate at a given neutron fluence was found to be consistent with the value reported from nuclear-reactor-based BNCT experiments. The induced activity of each nuclide can be estimated by entering the saturated activity of each nuclide, sample mass, irradiation time, and proton current into the derived activation equation in our accelerator-based BNCT system.

Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis

Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 10(5)-10(6) years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

Nuclear power in the 21st century: Challenges and possibilities

The current situation and possible future developments for nuclear power--including fission and fusion processes--is presented. The fission nuclear power continues to be an essential part of the low-carbon electricity generation in the world for decades to come. There are breakthrough possibilities in the development of new generation nuclear reactors where the life-time of the nuclear waste can be reduced to some hundreds of years instead of the present time-scales of hundred thousand of years. Research on the fourth generation reactors is needed for the realisation of this development. For the fast nuclear reactors, a substantial research and development effort is required in many fields--from material sciences to safety demonstration--to attain the envisaged goals. Fusion provides a long-term vision for an efficient energy production. The fusion option for a nuclear reactor for efficient production of electricity has been set out in a focussed European programme including the international project of ITER after which a fusion electricity DEMO reactor is envisaged.

Neutron flux characterisation of the Pavia TRIGA Mark II research reactor for radiobiological and microdosimetric applications

Nowadays the Pavia TRIGA reactor is available for national and international collaboration in various research fields. The TRIGA Mark II nuclear research reactor of the Pavia University offers different in- and out-core neutron irradiation channels, each characterised by different neutron spectra. In the last two years a campaign of measurements and simulations has been performed in order to guarantee a better characterisation of these different fluxes and to meet the demands of irradiations that require precise information on these spectra in particular for radiobiological and microdosimetric studies. Experimental data on neutron fluxes have been collected analysing and measuring the gamma activity induced in thin target foils of different materials irradiated in different TRIGA experimental channels. The data on the induced gamma activities have been processed with the SAND II deconvolution code and finally compared with the spectra obtained with Monte Carlo simulations. The comparison between simulated and measured spectra showed a good agreement allowing a more precise characterisation of the neutron spectra and a validation of the adopted method.

Radiation protection performance for the dismantling of the WWR-M primary cooling circuit

The WWR-M is a light-water-cooled and moderated heterogonous research reactor with a thermal output of 10 MW. The reactor has been in operation for >50 y and has had an excellent safety record. A non-hermeticity of the inlet line of the primary cooling circuit (PCC) was found, and the only reasonable technical solution was the complete replacement of the PCC inlet and outlet pipe lines. Such a replacement was a challenging technical task due to the necessity to handle large size components with complex geometries under conditions of high-level radiation fields, and therefore, it required detailed planning aiming to reduce staff exposure. This paper describes the dismantling and removal of the PCC components focusing on radiation protection issues.

Study of plasma meniscus and beam halo in negative ion sources using three dimension in real space and three dimension in velocity space particle in cell model

Our previous study by two dimension in real space and three dimension in velocity space-particle in cell model shows that the curvature of the plasma meniscus causes the beam halo in the negative ion sources. The negative ions extracted from the periphery of the meniscus are over-focused in the extractor due to the electrostatic lens effect, and consequently become the beam halo. The purpose of this study is to verify this mechanism with the full 3D model. It is shown that the above mechanism is essentially unchanged even in the 3D model, while the fraction of the beam halo is significantly reduced to 6%. This value reasonably agrees with the experimental result.

Microdosimetric measurements in the thermal neutron irradiation facility of LENA reactor

A twin TEPC with electric-field guard tubes has been constructed to be used to characterize the BNCT field of the irradiation facility of LENA reactor. One of the two mini TEPC was doped with 50ppm of (10)B in order to simulate the BNC events occurring in BNCT. By properly processing the two microdosimetric spectra, the gamma, neutron and BNC spectral components can be derived with good precision (~6%). However, direct measurements of (10)B in some doped plastic samples, which were used for constructing the cathode walls, point out the scarce accuracy of the nominal (10)B concentration value. The influence of the Boral(®) door, which closes the irradiation channel, has been measured. The gamma dose increases significantly (+51%) when the Boral(®) door is closed. The crypt-cell-regeneration weighting function has been used to measure the quality, namely the RBEµ value, of the radiation field in different conditions. The measured RBEµ values are only partially consistent with the RBE values of other BNCT facilities.

Neutron spectrometry and dosimetry study at two research nuclear reactors using Bonner sphere spectrometer (BSS), rotational spectrometer (ROSPEC) and cylindrical nested neutron spectrometer (NNS)

Neutron spectrometry and subsequent dosimetry measurements were undertaken at the McMaster Nuclear Reactor (MNR) and AECL Chalk River National Research Universal (NRU) Reactor. The instruments used were a Bonner sphere spectrometer (BSS), a cylindrical nested neutron spectrometer (NNS) and a commercially available rotational proton recoil spectrometer. The purposes of these measurements were to: (1) compare the results obtained by three different neutron measuring instruments and (2) quantify neutron fields of interest. The results showed vastly different neutron spectral shapes for the two different reactors. This is not surprising, considering the type of the reactors and the locations where the measurements were performed. MNR is a heavily shielded light water moderated reactor, while NRU is a heavy water moderated reactor. The measurements at MNR were taken at the base of the reactor pool, where a large amount of water and concrete shielding is present, while measurements at NRU were taken at the top of the reactor (TOR) plate, where there is only heavy water and steel between the reactor core and the measuring instrument. As a result, a large component of the thermal neutron fluence was measured at MNR, while a negligible amount of thermal neutrons was measured at NRU. The neutron ambient dose rates at NRU TOR were measured to be between 0.03 and 0.06 mSv h⁻¹, while at MNR, these values were between 0.07 and 2.8 mSv h⁻¹ inside the beam port and <0.2 mSv h⁻¹ between two operating beam ports. The conservative uncertainty of these values is 15 %. The conservative uncertainty of the measured integral neutron fluence is 5 %. It was also found that BSS over-responded slightly due to a non-calibrated response matrix.

Particulate activity accumulated on a moving filter and RCS leak detection

Analytical equations for calculating the particulate activity accumulated on a moving filter of a containment air particulate radiation monitor due to a reactor coolant system leak in the containment, including the noble gas decay products, were presented. The particulate airborne concentration in the containment was treated by assumptions of a constant reactor coolant leak rate at a constant concentration, a given aerosolizing fraction, a constant removal coefficient to account for the loss due to diffusion, settling, diffusiophoresis, and containment air recirculation operation. The ratio of moving-to-fixed filter activity was presented for radionuclide of various half-lives and for different filter moving speeds. The monitor response at one hour after the initiation of a 1-gallon per minute leak at measured reactor coolant concentrations was compared to the standard deviation of the background count rate for an operating pressurized water reactor. The detectability of a 1-gallon per minute leak in an hour can be demonstrated if the aerosolization percentage of the radionuclide in the leaked coolant is at least a few percent.

Applicability of health physics lessons learned from the Three Mile Island Unit 2 accident to the Fukushima Daiichi accident

The TMI-2 and Fukushima Daiichi accidents appear to be dissimilar because they involve different reactor types. However, the health physics related lessons learned from TMI-2 are applicable, and can enhance the Fukushima Daiichi recovery effort.

Experimental simulation of personal dosimetry in production of medical radioisotopes by research reactor

Due to their work conditions, research reactor personnel are exposed to ionising nuclear radiations. Because the absorbed dose values are different for different tissues due to variations in sensitivity, in this work personal dosimetry has been performed under normal working conditions at anatomical locations relevant to more sensitive tissues as well as for the whole body by employing a Rando phantom and thermoluminescent dosemeters (TLDs). Fifty-two TLDs-100H were positioned at high-risk organ locations such as the thyroid, eyes as well as the left breast, which was used to assess the whole-body dose in order to study the absorbed doses originating from selected locations in the vicinity of the reactor. The results have employed the tissue weighting factors based on International Commission on Radiological Protection ICRP 103 and ICRP 60 and the measured results were below the dose limits recommended by ICRP. The mean effective dose rates calculated from ICRP 103 were the following: whole body, 30.64-6.44 µSv h(-1); thyroid, 1.22-0.23 µSv h(-1); prostate, 0.085-0.045 µSv h(-1); gonads, 1.00-0.51 µSv h(-1); breast, 3.68-0.77 µSv h(-1); and eyes, 33.74-7.01 µSv h(-1).

Microdosimetric evaluation of the neutron field for BNCT at Kyoto University reactor by using the PHITS code

In this study, microdosimetric energy distributions of secondary charged particles from the (10)B(n,α)(7)Li reaction in boron-neutron capture therapy (BNCT) field were calculated using the Particle and Heavy Ion Transport code System (PHITS). The PHITS simulation was performed to reproduce the geometrical set-up of an experiment that measured the microdosimetric energy distributions at the Kyoto University Reactor where two types of tissue-equivalent proportional counters were used, one with A-150 wall alone and another with a 50-ppm-boron-loaded A-150 wall. It was found that the PHITS code is a useful tool for the simulation of the energy deposited in tissue in BNCT based on the comparisons with experimental results.

Advanced reactors and associated fuel cycle facilities: safety and environmental impacts

The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described.

Neutron fluence in antiproton radiotherapy, measurements and simulations

INTRODUCTION

A significant part of the secondary particle spectrum from antiproton annihilation consists of fast neutrons, which may contribute to a significant dose background found outside the primary beam.

MATERIALS AND METHODS

Using a polystyrene phantom as a moderator, we have performed absolute fluence measurements of the thermalized part of the fast neutron spectrum using Lithium-6 and -7 Fluoride TLD pairs. The results were compared with the Monte Carlo particle transport code FLUKA.

RESULTS

The experimental results are found to be in good agreement with simulations. The thermal neutron kerma resulting from the measured thermal neutron fluence is insignificant compared to the contribution from fast neutrons.

DISCUSSION

The secondary neutron fluences encountered in antiproton therapy are found to be similar to values calculated for pion treatment, however exact modeling under more realistic treatment scenarios is still required to quantitatively compare these treatment modalities.

Dose calculation in biological samples in a mixed neutron-gamma field at the TRIGA reactor of the University of Mainz

To establish Boron Neutron Capture Therapy (BNCT) for non-resectable liver metastases and for in vitro experiments at the TRIGA Mark II reactor at the University of Mainz, Germany, it is necessary to have a reliable dose monitoring system. The in vitro experiments are used to determine the relative biological effectiveness (RBE) of liver and cancer cells in our mixed neutron and gamma field. We work with alanine detectors in combination with Monte Carlo simulations, where we can measure and characterize the dose. To verify our calculations we perform neutron flux measurements using gold foil activation and pin-diodes. Material and methods. When L-α-alanine is irradiated with ionizing radiation, it forms a stable radical which can be detected by electron spin resonance (ESR) spectroscopy. The value of the ESR signal correlates to the amount of absorbed dose. The dose for each pellet is calculated using FLUKA, a multipurpose Monte Carlo transport code. The pin-diode is augmented by a lithium fluoride foil. This foil converts the neutrons into alpha and tritium particles which are products of the (7)Li(n,α)(3)H-reaction. These particles are detected by the diode and their amount correlates to the neutron fluence directly. Results and discussion. Gold foil activation and the pin-diode are reliable fluence measurement systems for the TRIGA reactor, Mainz. Alanine dosimetry of the photon field and charged particle field from secondary reactions can in principle be carried out in combination with MC-calculations for mixed radiation fields and the Hansen & Olsen alanine detector response model. With the acquired data about the background dose and charged particle spectrum, and with the acquired information of the neutron flux, we are capable of calculating the dose to the tissue. Conclusion. Monte Carlo simulation of the mixed neutron and gamma field of the TRIGA Mainz is possible in order to characterize the neutron behavior in the thermal column. Currently we also speculate on sensitizing alanine to thermal neutrons by adding boron compounds.

Estimation of the activity generated by neutron activation in control rods of a BWR

Control rods are activated by neutron reactions into the reactor. The activation is produced mainly in stainless steel and its impurities. The dose produced by this activity is not important inside the reactor, but it has to be taken into account when the rod is withdrawn from the reactor. Activation reactions produced have been modelled by the MCNP5 code based on the Monte Carlo method. The code gives the number of reactions that can be converted into activity.

Development of a cause analysis system for a CPCS trip by using the rule-base deduction method

A Core Protection Calculator System (CPCS) was developed to initiate a Reactor Trip under the circumstance of certain transients by a Combustion Engineering Company. The major function of the Core Protection Calculator System is to generate contact outputs for the Departure from Nucleate Boiling Ratio (DNBR) Trip and a Local Power Density (LPD) Trip. But in a Core Protection Calculator System, a trip cause cannot be identified, thus only trip signals are transferred to the Plant Protection System (PPS) and only the trip status is displayed. It could take a considerable amount of time and effort for a plant operator to analyze the trip causes of a Core Protection Calculator System. So, a Cause Analysis System for a Core Protection Calculator System (CASCPCS) has been developed by using the rule-base deduction method to assist operators in a Nuclear Power Plant. CASCPCS consists of three major parts. Inference engine has a role of controlling the searching knowledge base, executing the rules and tracking the inference process by using the depth-first searching method. Knowledge base consists of four major parts: rules, data base constants, trip buffer variables and causes. And a user interface is implemented by using menu-driven and window display techniques. The advantage of CASCPCS is that it saves time and effort to diagnose the trip causes of a Core Protection Calculator System, it increases a plant's availability and reliability, and it makes it easy to manage CASCPCS because of using only a cursor control.

County mortality and cancer incidence in relation to living near two former nuclear materials processing facilities in Pennsylvania--an update

A previous county mortality study of populations living near two nuclear materials processing and fabrication facilities in Westmoreland and Armstrong counties in Pennsylvania (1950-1995) was extended through 2004. Noncancer mortality (1996-2004) and cancer incidence (1990-2004) were also evaluated. Among the Westmoreland and Armstrong populations, 10,547 cancer deaths occurred during the period 1996 through 2004 and the relative risk (RR) based on comparisons with six demographically similar counties in western Pennsylvania was 0.97, that is, almost exactly as expected, and no different from our previously published analyses covering the years 1950-1995. The results based on cancer incidence data were very similar to those based on cancer mortality data. Over the years 1990 though 2004, 39,350 incident cancers were reported among residents of Armstrong and Westmoreland counties and the RR based on the six demographically similar counties was 0.99, that is, almost exactly as expected. The number of deaths from nonmalignant conditions was 36,565 and very close to the number expected (RR 1.01). Overall, no increases in cancer or nonmalignant diseases could be attributed to living in counties with nuclear materials processing and fabrication facilities.

Cancer incidence in municipalities near two former nuclear materials processing facilities in Pennsylvania--an update

Previous studies of cancer incidence among persons living in municipalities within one mile of two nuclear materials processing and fabrication plants in Pennsylvania were extended for the years 1998-2004. It had been shown that mailing addresses for residents of rural areas often did not reflect the actual municipality of residence and, if not corrected, would bias study results. The previous studies had corrected for this bias. Accordingly for the extended study, we obtained mailing addresses from the Pennsylvania Department of Health (PDH) for 866 persons with cancer who presumably lived in one of eight minor civil divisions (MCDs) near or encompassing the former nuclear facilities, designated as Area 1 in previous studies conducted by the PDH. Street addresses were geocoded and local postmasters were asked to place rural delivery addresses, post office boxes and street addresses that could not be geocoded into the correct MCD of actual residence. Over 15% of the mailing addresses were found not to be within the boundaries of the Area 1 municipalities. After the mailing addresses of individuals with cancer were placed in their proper MCD of residence, the number of persons diagnosed with cancer (n = 708) and confirmed to have lived in Area 1 was as expected (728.4) based on cancer incidence rates in the general population of Pennsylvania (SIR 0.97; 95% CI 0.90-1.05). To further evaluate the patterns of cancer rates near these nuclear facilities and the influence of improved reporting and geocoding of addresses over time, analyses were conducted of publicly available cancer incidence data from 1990 through 2004. Based on mailing addresses, a steady decrease in the number of cancers reported in the Area 1 proximal MCDs was seen, in contrast to a steady rise in the number of cancers reported in seven adjacent but more distant MCDs from the nuclear facilities, designated as Area 2. These patterns were attributed to improvements over time in the geocoding of residential mailing addresses coupled with the gradual elimination and replacement of rural delivery addresses with street addresses. The incorrect placement of mailing addresses in residential Area 1 municipalities prior to about 2002 overestimated the number of cancers occurring among residents living in close proximity to the nuclear facilities and, correspondingly, underestimated the number among Area 2 residents. Summing Area 1 and Area 2 data showed that there was no change in cancer rates over time. These results are consistent with previous studies indicating that living in municipalities near the former Apollo-Parks nuclear facilities was not associated with an increase in cancer occurrence.

In vitro dissolution of tritium-loaded particles from the JET fusion machine

With the experimental evolution of fusion power the levels of tritium used will increase as will the potential for human exposure. Tritium-loaded carbon particles produced during the experimental operation of the Joint European Torus fusion tokamak have been characterised in terms of size, elemental composition and specific activity of tritium elsewhere. The aim of this study was to characterise the dissolution of tritium from these particles in order to derive dose coefficients for this material and provide guidance on monitoring procedures should it be inhaled accidentally. The dissolution of tritium was measured for 100 d in lung serum simulant from two batches of materials, SG1 and SG2, which were obtained from carbon tiles originating from different positions in the reactor. Retention over this period followed a three-component exponential. About 1-5% dissolved within a minute, and up to a further 20% dissolved over 100 d for the SG1 materials but <1% for the SG2 materials. Dissolution between the SG1 materials varied greatly, whereas the SG2 materials were similar. As a result of this variability, the assessed dose from urinary excretion could be in error by up to two orders of magnitude depending on the material inhaled. It is recommended that (i) the dissolution is measured for a wider range of materials, preferably dusts collected in working areas, and (ii) in vivo studies are performed to characterise fully the urine excretion of tritium from these materials. This information could be used to provide improved guidance on dose assessment after special or routine monitoring, taking account of the likely variation of particle size and biological retention half times.

Production of 89Sr in solution reactor

The new method for medical (89)Sr production in a reactor with solution fuel is proposed which is characterized by simplicity, high production efficiency and low buildup of radioactive waste. The main advantages of the new technology were validated by numerous experiments. The proposed new technology selectively extracts (89)Sr from a fuel of solution reactor and precludes penetration of (90)Sr into the final product. This method is based on the presence of gaseous radionuclide (89)Kr (T(1/2)=190.7s) in the decay chain (89)Se-->(89)Br-->(89)Kr-->(89)Rb-->(89)Sr. The performed experiments on taking the gas probes from internal volume of the solution 20 kW mini-reactor "Argus" have confirmed that the mechanism for (89)Sr delivery to the sorption volume of the reactor experimental loop is based on transport of gaseous (89)Sr predecessor-radionuclide (89)Kr. According to the measurements of radioactive impurities in a final (89)SrCl(2) solution, the filtration of the gas flow with cermet filters followed by cleaning of (89)Sr chloride solution in chromatographic columns with DOWEX-50 x 8 or Sr-Resin ensures reception of (89)Sr fully meeting the requirements for medical application. The experimental estimations have shown that the proposed new technology is multiply more productive than the traditional industrial methods of (89)Sr reception.

Dose distributions in phantoms irradiated in thermal columns of two different nuclear reactors

In-phantom dosimetry studies have been carried out at the thermal columns of a thermal- and a fast-nuclear reactor for investigating: (a) the spatial distribution of the gamma dose and the thermal neutron fluence and (b) the accuracy at which the boron concentration should be estimated in an explanted organ of a boron neutron capture therapy patient. The phantom was a cylinder (11 cm in diameter and 12 cm in height) of tissue-equivalent gel. Dose images were acquired with gel dosemeters across the axial section of the phantom. The thermal neutron fluence rate was measured with activation foils in a few positions of this phantom. Dose and fluence rate profiles were also calculated with Monte Carlo simulations. The trend of these profiles do not show significant differences for the thermal columns considered in this work.

Preliminary evaluations of the undesirable patient dose from a BNCT treatment at the ENEA-TAPIRO reactor

Boron neutron capture therapy (BNCT) is an experimental technique for the treatment of certain kinds of tumors. Research in BNCT is performed utilizing both thermal and epithermal neutron beams. Epithermal neutrons (0.4 eV-10 keV) penetrate more deeply into tissue and are thus used in non-superficial clinical applications such as the brain glioma. In the last few years, the fast reactor TAPIRO (ENEA-Casaccia Rome) has been employed as a neutron source for research into BNCT applications. Recently, an 'epithermal therapeutic column' has been designed and its construction has been completed. The Monte Carlo code MCNPX was employed to optimize the design of the column and to evaluate the dose profiles and the therapeutic parameters in the cranium of the anthropomorphic phantom ADAM. In the same context, some preliminary evaluations of the undesirable doses to the patient were performed with MCNPX. A hermaphrodite phantom derived from ADAM and EVA was employed to evaluate the energy deposition in some organs during a standard BNCT treatment. The total dose consists of the contributions from the primary neutron beam, the neutron interactions with boron and the neutron induced photons generated in the epithermal column structures and in the patient's tissues. The paper summarizes the computational procedure and provides a general dosimetric framework of the patient radiological protection aspects related to a BNCT treatment scenario at the TAPIRO reactor.

Fast response neutron emission monitor for fusion reactor using stilbene scintillator and Flash-ADC

The stilbene neutron detector which has been used for neutron emission profile monitoring in JT-60U has been improved, to respond to the requirement to observe the high-frequency phenomena in megahertz region such as toroidicity-induced Alfvén Eigen mode in burning plasma as well as the spatial profile and the energy spectrum. This high-frequency phenomenon is of great interest and one of the key issues in plasma physics in recent years. To achieve a fast response in the stilbene detector, a Flash-ADC is applied and the wave form of the anode signal stored directly, and neutron/gamma discrimination was carried out via software with a new scheme for data acquisition mode to extend the count rate limit to MHz region from 1.3 x 10(5) neutron/s in the past, and confirmed the adequacy of the method.

Measurement of the fast neutron flux in the MNSR inner irradiation site

The WIMSD4 code was used to calculate the fast neutron flux spectrum and the fast neutron fission cross-sections for (238)U, using six energy groups ranging from 0.5 to 10 MeV. These results, with the measured radioactivities of the (140)Ba, (131)I, (103)Ru, (95)Zr and (97)Zr fission products emerging from the fission of the (238)U foil covered with a cadmium filter, were used to measure the fast neutron flux in the Syrian Miniature Neutron Source Reactor inner irradiation site.

Plasma–surface interaction in the context of ITER

The decreasing availability of energy and concern about climate change necessitate the development of novel sustainable energy sources. Fusion energy is such a source. Although it will take several decades to develop it into routinely operated power sources, the ultimate potential of fusion energy is very high and badly needed. A major step forward in the development of fusion energy is the decision to construct the experimental test reactor ITER. ITER will stimulate research in many areas of science. This article serves as an introduction to some of those areas. In particular, we discuss research opportunities in the context of plasma-surface interactions. The fusion plasma, with a typical temperature of 10 keV, has to be brought into contact with a physical wall in order to remove the helium produced and drain the excess energy in the fusion plasma. The fusion plasma is far too hot to be brought into direct contact with a physical wall. It would degrade the wall and the debris from the wall would extinguish the plasma. Therefore, schemes are developed to cool down the plasma locally before it impacts on a physical surface. The resulting plasma-surface interaction in ITER is facing several challenges including surface erosion, material redeposition and tritium retention. In this article we introduce how the plasma-surface interaction relevant for ITER can be studied in small scale experiments. The various requirements for such experiments are introduced and examples of present and future experiments will be given. The emphasis in this article will be on the experimental studies of plasma-surface interactions.

Shielding design for the PANDA spectrometer at the Munich high-flux reactor FRM-II

The start-up of the Munich high-flux reactor FRM-II is in progress on. At the beam tube SR-2 the spectrometer PANDA has been installed. It is at three-axis neutron spectrometer looking onto a slightly under-moderated cold neutron source. For polarisation analysis, PANDA is equipped with a vertical cryomagnet producing fields up to 14.5 T for the sample. To get an appropriate shielding of the high-intensity instrument, one has to take into account the large cross section of the primary beam, several restrictions using magnetic materials, limitations in loading the site and finally, has to keep the lateral extent of the shielding small to allow for high-scattering angles. The shielding has been designed on the basis of the results, which were achieved by the combined use of both the Monte Carlo code MCNP-4B2 and an analytical method based on one-dimensional dose transmission functions.

Investigation of radiation fields outside the Sub-critical Assembly in Dubna

The radiation fields outside the planned experimental Sub-critical Assembly in Dubna (SAD) have been studied in order to provide a basis for the design of the concrete shielding that cover the reactor core. The effective doses around the reactor, induced by leakage of neutrons and photons through the shielding, have been determined for a shielding thickness varying from 100 to 200 cm. It was shown that the neutron flux and the effective dose is higher above the shielding than at the side of it, owing to the higher fraction of high-energy spallation neutrons emitted in the direction of the incident beam protons. At the top, the effective dose was found to be -150 microSv s(-1) for a concrete thickness of 100 cm, while -2.5 microSv s(-1) for a concrete thickness of 200 cm. It was also shown that the high-energy neutrons (> 10 MeV), which are created in the proton-induced spallation interactions in the target, contribute for the major part of the effective doses outside the reactor.

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

Monte Carlo optimisation of a BNCT facility for treating brain gliomas at the TAPIRO reactor

An epithermal boron neutron capture therapy facility for treating brain gliomas is currently under construction at the 5 kW fast-flux reactor TAPIRO located at ENEA, Casaccia, near Rome. In this work, the sensitivity of the results to the boron concentrations in healthy tissue and tumour is investigated and the change in beam quality on modifying the moderator thickness (within design limits) is studied. The Monte Carlo codes MCNP and MCNPX were used together with the DSA in-house variance reduction patch. Both usual free beam parameters and the in-phantom treatment planning figures-of-merit have been calculated in a realistic anthropomorphic phantom ('ADAM').

Monte Carlo simulation of neutron backscattering from concrete walls in the dense plasma focus laboratory of Bologna University

Between 2001 and 2003 a 3.2 kJ dense plasma focus (DPF) device has been built at the Montecuccolino Laboratory of the Department of Energy, Nuclear and Environmental Control Engineering (DIENCA) of the University of Bologna. A DPF is a pulsed device in which deuterium nuclear fusion reactions can be obtained through the pinching effects of electromagnetic fields upon a dense plasma. The empirical scale law that governs the total D-D neutron yield from a single pulse of a DPF predicts for this machine a figure of approximately 10(7) fast neutrons per shot. The aim of the present work is to evaluate the role of backscattering of neutrons from the concrete walls surrounding the Montecuccolino DPF in total neutron yield measurements. The evaluation is performed by MCNP-5 simulations that are aimed at estimating the neutron spectra at a few points of interest in the laboratory, where neutron detectors will be placed during the experimental campaigns. Spectral information from the simulations is essential because the response of detectors is influenced by neutron energy. Comparisons are made with the simple r(-2) law, which holds for a DPF in infinite vacuum. The results from the simulations will ultimately be used both in the design and optimisation of the neutron detectors and in their final calibration and placement inside the laboratory.

Evaluation of induced radioactivity in structural material of Toshiba Training Reactor 'TTR1'

A decommissioning programme for the Toshiba Training Reactor (TTR1), a swimming pool type reactor used for reactor physics experiments and material irradiation, was started in August 2001. As a part of the programme, induced radioactivity in structural material was evaluated using neutron flux data obtained with the three-dimensional Sn code TORT. Induced activity was calculated with the isotope generation code ORIGEN-79 using activation cross section data created from multi-group library based on JENDL-3. The obtained results for radioactivities such as 60Co, 65Zn, 54Mn and 152Eu were compared with measured ones, and the present calculational method was confirmed to have enough accuracy.

ISFSI site boundary radiation dose rate analyses

Across the globe nuclear utilities are in the process of designing and analysing Independent Spent Fuel Storage Installations (ISFSI) for the purpose of above ground spent-fuel storage primarily to mitigate the filling of spent-fuel pools. Using a conjoining of discrete ordinates transport theory (DORT) and Monte Carlo (MCNP) techniques, an ISFSI was analysed to determine neutron and photon dose rates for a generic overpack, and ISFSI pad configuration and design at distances ranging from 1 to -1700 m from the ISFSI array. The calculated dose rates are used to address the requirements of 10CFR72.104, which provides limits to be enforced for the protection of the public by the NRC in regard to ISFSI facilities. For this overpack, dose rates decrease by three orders of magnitude through the first 200 m moving away from the ISFSI. In addition, the contributions from different source terms changes over distance. It can be observed that although side photons provide the majority of dose rate in this calculation, scattered photons and side neutrons take on more importance as the distance from the ISFSI is increased.

Interim storage of spent and disused sealed sources: optimisation of external dose distribution in waste grids using the MCNPX code

Radioactive sealed sources are in use worldwide in different fields of application. When no further use is foreseen for these sources, they become spent or disused sealed sources and are subject to a specific waste management scheme. Portugal does have a Radioactive Waste Interim Storage Facility where spent or disused sealed sources are conditioned in a cement matrix inside concrete drums and following the geometrical disposition of a grid. The gamma dose values around each grid depend on the drum's enclosed activity and radionuclides considered, as well as on the drums distribution in the various layers of the grid. This work proposes a method based on the Monte Carlo simulation using the MCNPX code to estimate the best drum arrangement through the optimisation of dose distribution in a grid. Measured dose rate values at 1 m distance from the surface of the chosen optimised grid were used to validate the corresponding computational grid model.

Radiation protection and shielding design--strengthening the link

The improvement in quality and flexibility of shielding methods and data has been progressive and beneficial in opening up new opportunities for optimising radiation protection in design. The paper describes how these opportunities can best be seized by taking a holistic view of radiation protection, with shielding design being an important component part. This view is best achieved by enhancing the role of 'shielding assessors' so that they truly become 'radiation protection designers'. The increase in speed and efficiency of shielding calculations has been enormous over the past decades. This has raised the issue of how the assessor's time now can be best utilised; pursuing ever greater precision and accuracy in shielding/dose assessments, or improving the contribution that shielding assessment makes to radiological protection and cost-effective design. It is argued in this paper that the latter option is of great importance and will give considerable benefits. Shielding design needs to form part of a larger radiation protection perspective based on a deep understanding/appreciation of the opportunities and constraints of operators and designers, enabling minimal design iterations, cost optimisation of alternative designs (with a 'lifetime' perspective) and improved realisation of design intent in operations. The future of shielding design development is argued to be not in improving the 'toolkit', but in enhanced understanding of the 'product' and the 'process' for achieving it. The holistic processes being developed in BNFL to realise these benefits are described in the paper and will be illustrated by case studies.

Characterisation of radioactive waste products associated with plant decommissioning

The inventory of radioactivity that must be considered in the decommissioning of a typical 1000 MWe Spanish pressurised water reactor (PWR) was investigated as part of a generic plant decommissioning study. Analyses based on DORT models (in both R-Z and R-theta geometries) were used with representative plant operating history and core power distribution data in defining the expected neutron environment in regions near the reactor core. The activation analyses were performed by multiplying the DORT scalar fluxes by energy-dependent reaction cross sections (based on ENDF/B-VI data) to generate reaction rates on a per atom basis. The results from the ORIGEN2 computer code were also used for determining the activities associated with certain nuclides where multi-group cross section data were not available. In addition to the bulk material activation of equipment and structures near the reactor, the activated corrosion-product (or 'crud') deposits on system and equipment surfaces were considered. The projected activities associated with these sources were primarily based on plant data and experience from operating PWR plants.

Radiological characterization of the pressure vessel internals of the BNL High Flux Beam Reactor

In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, measurements and calculations of the decay gamma-ray dose-rate were performed in the reactor pressure vessel and on vessel internal structures such as the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. Measurements of gamma-ray dose rates were made using Red Perspex polymethyl methacrylate high-dose film, a Radcal "peanut" ion chamber, and Eberline's RO-7 high-range ion chamber. As a comparison, the Monte Carlo MCNP code and MicroShield code were used to model the gamma-ray transport and dose buildup. The gamma-ray dose rate at 8 cm above the center of the Transition Plate was measured to be 160 Gy h (using an RO-7) and 88 Gy h at 8 cm above and about 5 cm lateral to the Transition Plate (using Red Perspex film). This compares with a calculated dose rate of 172 Gy h using Micro-Shield. The gamma-ray dose rate was 16.2 Gy h measured at 76 cm from the reactor core (using the "peanut" ion chamber) and 16.3 Gy h at 87 cm from the core (using Red Perspex film). The similarity of dose rates measured with different instruments indicates that using different methods and instruments is acceptable if the measurement (and calculation) parameters are well defined. Different measurement techniques may be necessary due to constraints such as size restrictions.

Neutron flux characterization of a peripheral target position in the High Flux Isotope Reactor

The High Flux Isotope Reactor at the Oak Ridge National Laboratory provides the highest steady-state thermal neutron flux in the western world for a wide range of experiments and for isotope production. The highest available fluxes are located in a flux trap region created inside the nested fuel elements. The experimentally determined thermal and the empirically obtained epithermal flux values along the vertical axis of the peripheral target position were fit to cosine curves, with the thermal flux ranging from 1.1 x 10(15)ns(-1)cm(-2) at outer positions to 1.5 x 10(15)ns(-1)cm(-2) at the center. The corresponding epithermal flux ranged from 3.5 x 10(13) to 7.5 x 10(13)ns(-1)cm(-2), respectively. The fast neutron flux (En > or = 0.32 MeV in two positions and En > or = 1.5 MeV in two other positions) was approximately 6 x 10(14)ns(-1)cm(-2), corresponding to a fast to thermal ratio of approximately 0.4.

Fission reactor neutron sources for neutron capture therapy--a critical review

The status of fission reactor-based neutron beams for neutron capture therapy (NCT) is reviewed critically. Epithermal neutron beams, which are favored for treatment of deep-seated tumors, have been constructed or are under construction at a number of reactors worldwide. Some of the most recently constructed epithermal neutron beams approach the theoretical optimum for beam purity. Of these higher quality beams, at least one is suitable for use in high through-put routine therapy. It is concluded that reactor-based epithermal neutron beams with near optimum characteristics are currently available and more can be constructed at existing reactors. Suitable reactors include relatively low power reactors using the core directly as a source of neutrons or a fission converter if core neutrons are difficult to access. Thermal neutron beams for NCT studies with small animals or for shallow tumor treatments, with near optimum properties have been available at reactors for many years. Additional high quality thermal beams can also be constructed at existing reactors or at new, small reactors. Furthermore, it should be possible to design and construct new low power reactors specifically for NCT, which meet all requirements for routine therapy and which are based on proven and highly safe reactor technology.

The reactor facility that was built at Columbia University but never used

A large and heavy experimental plasma vessel is located on the second floor of the Engineering Building at Columbia University. It sits atop the concrete shell of the old nuclear reactor facility. The reactor facility was built many years ago but no nuclear fuel was ever loaded into it. It was designed to contain a 250 kW reactor core. However, due to certain circumstances, it was never fueled or operated. This paper describes the events leading to the decision to not put the reactor into operation.

Consistency and efficiency of standard swipe procedures taken on slightly radioactive contaminated metal surfaces

In radiation work areas, a standard "swipe" procedure is widely used to evaluate the extent of contamination on surfaces. This report documents the variability in results of swipes carried out on various metal surfaces and the variability between different experienced health physics technicians. Also, there is an issue of the efficiency of the first swipe in terms of what fraction of the total absorbed surface contamination is detected by a swipe. The samples used for this study were metal surfaces uniformly exposed in the spent fuel pool of a nuclear power plant The primary surfaces studied were those usually found on spent fuel transportation casks (mainly 304 stainless steel in the U.S.), which are submerged in the spent fuel pools for loading or unloading of the highly radioactive fuel assemblies from nuclear power plants. These surfaces become contaminated with suspended and dissolved radionuclides, primarily 137Cs, 134Cs, and 60Co, in the spent fuel pool. A detailed evaluation was conducted of variations in the swipe measurements made on these metal samples using repeated swipes of the same area by the same technician and comparing swipes of one technician to those of another on similar surfaces. Rough surface finishes showed considerable inconsistency (approximately 30% variation) from one technician to another, but smooth surface finishes show substantially better consistency (<10% variation) between technicians. The "efficiencies" of a single swipe, particularly the initial swipe, expressed as a fraction of total "removable" contamination, ranged from approximately 10% to 20% for the stainless steel and titanium surfaces. Aluminum surfaces, on the other hand, showed much higher efficiencies on the initial swipe. However, in terms of the total contamination imbedded in the surfaces, the first swipe picked up only between 0.5% and 3% of the total adsorbed contamination. The overall results show the wide variations that routinely occur in swipe results on portions of surfaces that would be expected to give consistent results. These difference are an order of magnitude or more greater than the counting statistical errors.

Steam generator hand hole shielding

Seabrook Station is an 1198 MWE Pressurized Water Reactor (PWR) that began commercial operation in 1990. Expensive and dose intensive Steam Generator Replacement Projects among PWR operators have led to an increase in steam generator preventative maintenance. Most of this preventative maintenance is performed through access ports in the shell of the steam generator just above the tube sheet known as secondary side hand holes. Secondary side work activities performed through the hand holes are typically performed without the shielding benefit of water in the secondary side of the steam generator. An increase in cleaning and inspection work scope has led to an increase in dose attributed to steam generator secondary side maintenance. This increased work scope and the station goal of maintaining personnel radiation dose ALARA led to the development of the shielding concept described in this article. This shield design saved an estimated 2.5 person-rem (25 person-Smv) the first time it was deployed and is expected to save an additional 50 person-rem (500 person-mSv) over the remaining life of the plant.

Comparison of the MCNP calculated and measured radiation field quantities near the RB reactor

The RB experimental reactor has operated at Vinca Institute of Nuclear Sciences since the end of April 1958. In this paper, neutron and gamma-ray spectra and corresponding dose quantities near the reactor, calculated by using the MCNP code, are compared to the measured values during the Third International Intercomparison Experiment on Nuclear Accident Dosimetry carried out at the RB reactor in 1973. Discrepancies in the correlation declared power of the reactor-dose rates are found. Good agreements are obtained between measured and calculated neutron and gamma-ray spectra, and corresponding absorbed doses in air, but only after the reactor declared power is multiplied by a correction factor, determined in this study.

Design of a high-flux epithermal neutron beam using 235U fission plates at the Brookhaven Medical Research Reactor

Beams of epithermal neutrons are being used in the development of boron neutron capture therapy for cancer. This report describes a design study in which 235U fission plates and moderators are used to produce an epithermal neutron beam with higher intensity and better quality than the beam currently in use at the Brookhaven Medical Research Reactor (BMRR). Monte Carlo calculations are used to predict the neutron and gamma fluxes and absorbed doses produced by the proposed design. Neutron flux measurements at the present epithermal treatment facility (ETF) were made to verify and compare with the computed results where feasible. The calculations indicate that an epithermal neutron beam produced by a fission-plate converter could have an epithermal neutron intensity of 1.2 x 10(10) n/cm2.s and a fast neutron dose per epithermal neutron of 2.8 x 10(-11) cGy.cm2/nepi plus being forward directed. This beam would be built into the beam shutter of the ETF at the BMRR. The feasibility of remodeling the facility is discussed.