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.