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FiR 1 is a TRIGA Mark II-type research reactor in Finland. It was in operation between 1962 to 2015 and will be dismantled in 2022 to 2023. Preliminary calculations of the activities in the reactor main structures were performed in an earlier stage of the decommissioning project. Samples of the activated parts of the reactor biological shield concrete were drilled in December 2018 to validate these estimates. This paper describes the calculations and gamma activity measurements performed for the activated concrete samples to determine the boundary between radioactive parts and concrete that can plausibly be free-released from regulatory control. The activities have been estimated with a two-step calculation process using the MCNP and ORIGEN-S calculation codes and measurements using an ISOCS gamma spectrometer with a high-purity germanium detector.

The Jules Horowitz Reactor (JHR) is a European material testing reactor (MTR) under construction at the CEA Cadarache centre. It will be dedicated to material and fuel irradiation tests, as well as to the production of medical isotopes. Gamma and X-Ray benches will be implemented in the reactor pool (RER), the irradiated component storage pool (EPI) and in a shielded hot cell for measuring either the whole underwater test device still containing the experimental sample or just the experimental sample before its extraction in the hot cell. The CEA/Cadarache Nuclear Measurement Laboratory (LMN) has been working in collaboration with VTT (Technical Research Centre in Finland Ltd.) since 2008 under a Finnish in-kind contribution agreement. This agreement focuses on the development of NDE systems implementing gamma-ray spectroscopy and high-energy X-ray imaging of the sample and irradiation device with the highest definition possible (resolution of 100 μm). The CEA-VTT technical specifications led to a European call for tenders launched by VTT. The contract was awarded to the Spanish company IDOM for the design, manufacturing, assembly and commissioning of: - Underwater gamma and X-ray (UGXR) mechanical benches and their associated gamma and X-ray collimation systems for the RER and EPI pools - Hot cell gamma and X-ray (HGXR) bench in the JHR NDE hot cell. The Final Design Reviews (FDR) of the UGXR and HGXR systems were completed in 2016. The design phase has been an iterative process in order to manage interfacing specifications and constraints: - Challenging experimental requirements, mainly to cover the wide diversity of sample shapes, sample activity levels and measurement processes, but also to achieve a level of mechanical accuracy to reach the ambitious geometrical resolution target in X-ray imaging, - Environmental constraints (immersion, radiation, compactness, limited accessibility for maintenance), - Nuclear safety constraints (seism, radiation protection). The whole design process has produced a number of elaborate and innovative mechatronic systems, which is rather unusual in nuclear applications since the resulting solutions have benefited from IDOM’s technological expertise in designing and commissioning large telescopes for the astronomy sector. Once the manufacturing phase and assembly finalised, the site acceptance tests for the UGXR and HGXR mechanical systems will be performed in 2019-2020 in the TOTEM facility at the CEA Cadarache center. The underwater benches will be tested in the CESARINE pool to check their requirements.

The objective of the study has been to estimate the residual activity in the decommissioning waste of the TRIGA Mark II-type research reactor FiR 1 in Finland. Neutron flux distributions were calculated with the Monte Carlo code MCNP. These were used in the ORIGEN-S point-depletion code to calculate the neutron-induced activity of materials at different time points by modeling irradiation history and radioactive decay. Knowledge of the radioactive inventory of irradiated materials is important in the planning of the decommissioning activities and is essential for predicting the radiological impact to personnel and the environment. Decommissioning waste consists mainly of ordinary concrete, aluminum, steel, and graphite parts. Results include uncertainties due to assump-tions on material compositions and lack of some detailed operational history data. Comparison to activity inventory estimates of two other decommissioned research reactors is also presented.

Iodine is one of the most radiotoxic fission product released from fuel during a severe nuclear power plant accident. Within the containment building, iodine compounds can react e.g. on the painted surfaces and form gaseous organic iodides. In this study, it was found out that gaseous methyl iodide (CH3I) is oxidised when exposed to beta radiation in an oxygen containing atmosphere. As a result, nucleation of aerosol particles takes place and the formation of iodine oxide particles is suggested. These particles are highly hygroscopic. They take up water from the air humidity and iodine oxides dissolve within the droplets. In order to mitigate the possible source term, it is of interest to understand the effect of beta radiation on the speciation of iodine.

The most commonly used technology currently used for autoradiography is storage phosphor screens, which has many benefits such as a large field of view but lacks particle-counting detection of the time and energy of each detected radionuclide decay. A number of alternative designs, using either solid state or scintillator detectors, have been developed to address these issues. The aim of this study is to characterize the imaging performance of one such instrument, a double-sided silicon strip detector (DSSD) system for digital autoradiography. A novel aspect of this work is that the instrument, in contrast to previous prototype systems using the same detector type, provides the ability for user accessible imaging with higher throughput. Studies were performed to compare its spatial resolution to that of storage phosphor screens and test the implementation of multiradionuclide ex vivo imaging in a mouse preclinical animal study. (Less)

Purpose: The response of alanine solid state dosimeters to ionizing radiation strongly depends on particle type and energy. Due to nuclear interactions, neutron fields usually also consist of secondary particles such as photons and protons of diverse energies. Various experiments have been carried out in three different neutron beams to explore the alanine dose response behavior and to validate model predictions. Additionally, application in medical neutron fields for boron neutron capture therapy is discussed. Methods: Alanine detectors have been irradiated in the thermal neutron field of the research reactor TRIGA Mainz, Germany, in five experimental conditions, generating different secondary particle spectra. Further irradiations have been made in the epithermal neutron beams at the research reactors FiR 1 in Helsinki, Finland, and Tsing Hua open pool reactor in HsinChu, Taiwan ROC. Readout has been performed with electron spin resonance spectrometry with reference to an absorbed dose standard in a {sup 60}Co gamma ray beam. Absorbed doses and dose components have been calculated using the Monte Carlo codes FLUKA and MCNP. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using the Hansen and Olsen alanine response model. Results: The measured dose response of the alanine detector inmore » the different experiments has been evaluated and compared to model predictions. Therefore, a relative effectiveness has been calculated for each dose component, accounting for its dependence on particle type and energy. Agreement within 5% between model and measurement has been achieved for most irradiated detectors. Significant differences have been observed in response behavior between thermal and epithermal neutron fields, especially regarding dose composition and depth dose curves. The calculated dose components could be verified with the experimental results in the different primary and secondary particle fields. Conclusions: The alanine detector can be used without difficulty in neutron fields. The response has been understood with the model used which includes the relative effectiveness. Results and the corresponding discussion lead to the conclusion that application in neutron fields for medical purpose is limited by its sensitivity but that it is a useful tool as supplement to other detectors and verification of neutron source descriptions.« less

Boron Neutron Capture Therapy (BNCT) is a binary radiotherapy method developed to treat patients with certain malignant tumours. To date, over 300 treatments have been carried out at the Finnish BNCT facility in various on-going and past clinical trials. In this technical review, we discuss our research work in the field of medical physics to form the groundwork for the Finnish BNCT patient treatments, as well as the possibilities to further develop and optimize the method in the future. Accordingly, the following aspects are described: neutron sources, beam dosimetry, treatment planning, boron imaging and determination, and finally the possibilities to detect the efficacy and effects of BNCT on patients.

Purpose: In this work, accuracy of themcnp5 code in the electron transport calculations and its suitability for ionization chamber (IC) response simulations in photonbeams are studied in comparison to egsnrc and penelope codes. Methods: The electron transport is studied by comparing the depth dose distributions in a water phantom subdivided into thin layers using incident energies (0.05, 0.1, 1, and 10 MeV) for the broad parallel electron beams. The IC response simulations are studied in water phantom in three dosimetric gas materials (air, argon, and methane based tissue equivalent gas) for photonbeams (60Co source, 6 MV linear medical accelerator, and mono-energetic 2 MeV photon source). Two optional electron transport models of mcnp5 are evaluated: the ITS-based electron energy indexing (mcnp5ITS) and the new detailed electron energy-loss straggling logic (mcnp5new). The electron substep length (ESTEP parameter) dependency in mcnp5 is investigated as well. Results: For the electron beam studies, large discrepancies (>3%) are observed between themcnp5dose distributions and the reference codes at 1 MeV and lower energies. The discrepancy is especially notable for 0.1 and 0.05 MeV electron beams. The boundary crossing artifacts, which are well known for the mcnp5ITS, are observed for the mcnp5new only at 0.1 and 0.05 MeV beamenergies. If the excessive boundary crossing is eliminated by using single scoring cells, the mcnp5ITS provides dose distributions that agree better with the reference codes than mcnp5new. The mcnp5dose estimates for the gas cavity agree within 1% with the reference codes, if the mcnp5ITS is applied or electron substep length is set adequately for the gas in the cavity using the mcnp5new. The mcnp5new results are found highly dependent on the chosen electron substep length and might lead up to 15% underestimation of the absorbed dose. Conclusions: Since themcnp5 electron transport calculations are not accurate at all energies and in every medium by general clinical standards, caution is needed, if mcnp5 is used with the current electron transport models for dosimetric applications.

Purpose To investigate the efficacy and safety of boron neutron capture therapy (BNCT) in the treatment of inoperable head-and-neck cancers that recur locally after conventional photon radiation therapy. Methods and Materials In this prospective, single-center Phase I/II study, 30 patients with inoperable, locally recurred head-and-neck cancer (29 carcinomas and 1 sarcoma) were treated with BNCT. Prior treatments consisted of surgery and conventionally fractionated photon irradiation to a cumulative dose of 50 to 98 Gy administered with or without concomitant chemotherapy. Tumor responses were assessed by use of the RECIST (Response Evaluation Criteria in Solid Tumors) and adverse effects by use of the National Cancer Institute common terminology criteria version 3.0. Intravenously administered L-boronophenylalanine–fructose (400 mg/kg) was administered as the boron carrier. Each patient was scheduled to be treated twice with BNCT. Results Twenty-six patients received BNCT twice; four were treated once. Of the 29 evaluable patients, 22 (76%) responded to BNCT, 6 (21%) had tumor growth stabilization for 5.1 and 20.3 months, and 1 (3%) progressed. The median progression-free survival time was 7.5 months (95% confidence interval, 5.4–9.6 months). Two-year progression-free survival and overall survival were 20% and 30%, respectively, and 27% of the patients survived for 2 years without locoregional recurrence. The most common acute Grade 3 adverse effects were mucositis (54% of patients), oral pain (54%), and fatigue (32%). Three patients were diagnosed with osteoradionecrosis (each Grade 3) and one patient with soft-tissue necrosis (Grade 4). Late Grade 3 xerostomia was present in 3 of the 15 evaluable patients (20%). Conclusions Most patients who have inoperable, locally advanced head-and-neck carcinoma that has recurred at a previously irradiated site respond to boronophenylalanine-mediated BNCT, but cancer recurrence after BNCT remains frequent. Toxicity was acceptable. Further research on novel modifications of the method is warranted.

Incineration of spent ion exchange resin was simulated using the ChemSheet chemical calculation program. The simulation of the incineration was modeled for typical spent resin produced by pressurized water reactors (PWRs) and boiling water reactors (BWRs) in Finland. The objective of the study was to find the volume and mass reduction and the chemical compounds formed during incineration. The simulation showed that active elements did not play any role in incineration owing to small amount of Cs, Co, etc. The ash contained metal oxides-mainly hematite, iron oxide Fe2O3. Other products of the incineration were water, carbon dioxide, sulfuric acid, and nitrogen oxides. The volume reductions 1/100 and 1/14 of the spent resin were obtained for PWRs and BWRs, respectively. The annual ash production from incineration was calculated to be 408 kg and 746 kg for the currently operating Finnish PWR and BWR plants in Loviisa and Olkiluoto, respectively.

In this paper, the accuracy of dose planning calculations for boron neutron capture therapy (BNCT) of brain and head and neck cancer was studied at the FiR 1 epithermal neutron beam. A cylindrical water phantom and an anthropomorphic head phantom were applied with two beam aperture-to-surface distances (ASD). The calculations using the simulation environment for radiation application (SERA) treatment planning system were compared to neutron activation measurements with Au and Mn foils, photon dose measurements with an ionization chamber and the reference simulations with the MCNP5 code. Photon dose calculations using SERA differ from the ionization chamber measurements by 2-13% (disagreement increased along the depth in the phantom), but are in agreement with the MCNP5 calculations within 2%. The 55Mn(n,γ) and 197Au(n, γ) reaction rates calculated using SERA agree within 10% and 8%, respectively, with the measurements and within 5% with the MCNP5 calculations at depths >0.5 cm from the phantom surface. The 55Mn(n,γ) reaction rate represents the nitrogen and boron depth dose within 1%. Discrepancy in the SERA fast neutron dose calculation (of up to 37%) is corrected if the biased fast neutron dose calculation option is not applied. Reduced voxel cell size (≤0.5 cm) improves the SERA calculation accuracy on the phantom surface. Despite the slight overestimation of the epithermal neutrons and underestimation of the thermal neutrons in the beam model, neutron calculation accuracy with the SERA system is sufficient for reliable BNCT treatment planning with the two studied treatment distances. The discrepancy between measured and calculated photon dose remains unsatisfactorily high for depths >6 cm from the phantom surface. Increasing discrepancy along the phantom depth is expected to be caused by the inaccurately determined effective point of the ionization chamber.

Purpose Head and neck carcinomas that recur locally after conventional irradiation pose a difficult therapeutic problem. We evaluated safety and efficacy of boron neutron capture therapy (BNCT) in the treatment of such cancers. Methods and Materials Twelve patients with inoperable, recurred, locally advanced (rT3, rT4, or rN2) head and neck cancer were treated with BNCT in a prospective, single-center Phase I-II study. Prior treatments consisted of surgery and conventionally fractionated photon irradiation to a cumulative dose of 56–74 Gy administered with or without concomitant chemotherapy. Tumor responses were assessed using the RECIST (Response Evaluation Criteria in Solid Tumors) criteria and adverse effects using the National Cancer Institute common toxicity grading v3.0. Intravenously administered boronophenylalanine-fructose (BPA-F, 400 mg/kg) was used as the boron carrier. Each patient was scheduled to be treated twice with BNCT. Results Ten patients received BNCT twice; 2 were treated once. Ten (83%) patients responded to BNCT, and 2 (17%) had tumor growth stabilization for 5.5 and 7.6 months. The median duration of response was 12.1 months; six responses were ongoing at the time of analysis or death (range, 4.9–19.2 months). Four (33%) patients were alive without recurrence with a median follow-up of 14.0 months (range, 12.8–19.2 months). The most common acute adverse effects were mucositis, fatigue, and local pain; 2 patients had a severe (Grade 3) late adverse effect (xerostomia, 1; dysphagia, 1). Conclusions Boron neutron capture therapy is effective and safe in the treatment of inoperable, locally advanced head and neck carcinomas that recur at previously irradiated sites.

"Radiation-sensitive polymer gels are among the most promising three-dimensional dose verification tools developed to date. We tested the normoxic polymer gel dosimeter known by the acronym MAGIC (methacrylic and ascorbic acid in gelatin initiated by copper) to evaluate its use in boron neutron capture therapy (BNCT) dosimetry. We irradiated a large cylindrical gel phantom (diameter: 10 cm; length: 20 cm) in the epithermal neutron beam of the Finnish BNCT facility at the FiR 1 nuclear reactor. Neutron irradiation was simulated with a Monte Carlo radiation transport code MCNP. To compare dose-response, gel samples from the same production batch were also irradiated with 6 MV photons from a medical linear accelerator. Irradiated gel phantoms then underwent magnetic resonance imaging to determine their R2 relaxation rate maps. The measured and normalized dose distribution in the epithermal neutron beam was compared with the dose distribution calculated by computer simulation. The results support the feasibility of using MAGIC gel in BNCT dosimetry." "Radiation-sensitive polymer gels are among the most promising three-dimensional dose verification tools developed to date. We tested the normoxic polymer gel dosimeter known by the acronym MAGIC (methacrylic and ascorbic acid in gelatin initiated by copper) to evaluate its use in boron neutron capture therapy (BNCT) dosimetry. We irradiated a large cylindrical gel phantom (diameter: 10 cm; length: 20 cm) in the epithermal neutron beam of the Finnish BNCT facility at the FiR 1 nuclear reactor. Neutron irradiation was simulated with a Monte Carlo radiation transport code MCNP. To compare dose-response, gel samples from the same production batch were also irradiated with 6 MV photons from a medical linear accelerator. Irradiated gel phantoms then underwent magnetic resonance imaging to determine their R2 relaxation rate maps. The measured and normalized dose distribution in the epithermal neutron beam was compared with the dose distribution calculated by computer simulation. The results support the feasibility of using MAGIC gel in BNCT dosimetry." "Radiation-sensitive polymer gels are among the most promising three-dimensional dose verification tools developed to date. We tested the normoxic polymer gel dosimeter known by the acronym MAGIC (methacrylic and ascorbic acid in gelatin initiated by copper) to evaluate its use in boron neutron capture therapy (BNCT) dosimetry. We irradiated a large cylindrical gel phantom (diameter: 10 cm; length: 20 cm) in the epithermal neutron beam of the Finnish BNCT facility at the FiR 1 nuclear reactor. Neutron irradiation was simulated with a Monte Carlo radiation transport code MCNP. To compare dose-response, gel samples from the same production batch were also irradiated with 6 MV photons from a medical linear accelerator. Irradiated gel phantoms then underwent magnetic resonance imaging to determine their R2 relaxation rate maps. The measured and normalized dose distribution in the epithermal neutron beam was compared with the dose distribution calculated by computer simulation. The results support the feasibility of using MAGIC gel in BNCT dosimetry."

The first experiments on the radiolytical oxidation of gaseous iodine by beta radiation were conducted with a new BESSEL (Beta irradiation vESSEL) test facility of VTT Technical Research Centre of Finland. The reaction chamber was made of glass and a source of beta radiation (P-32, activity 5 mCi) was placed at the bottom. The residence time of gaseous sample in the radiation field, the composition of gas flow and the temperature of the facility could be adjusted. In the experiments, gaseous methyl iodide (CH3I) was exposed to beta radiation in various mixtures of oxygen and nitrogen. The results of the experiments at 20 °C revealed that as an outcome of radiolytic oxidation the nucleation of particles took place. Their measured diameter was very small, approx. 10 to 50 nm, but a small fraction of particles had a diameter even higher than 100 nm. The formed particles seemed to react with the humidity of air and dissolve very rapidly. Since particles were highly water soluble and volatile, it was very difficult to analyse their properties with SEM-EDX. The only observed gaseous reaction product was carbon dioxide (CO2). It was detected when the residence time of gaseous sample inside the reaction chamber was 1 day or more.

A successful boron neutron capture treatment (BNCT) of a patient with multiple liver metastases has been first given in Italy, by placing the removed organ into the thermal neutron column of the Triga research reactor of the University of Pavia. In Finland, FiR 1 Triga reactor with an epithermal neutron beam well suited for BNCT has been extensively used to irradiate patients with brain tumors such as glioblastoma and recently also head and neck tumors. In this work we have studied by MCNP Monte Carlo simulations, whether it would be beneficial to treat an isolated liver with epithermal neutrons instead of thermal ones. The results show, that the epithermal field penetrates deeper into the liver and creates a build-up distribution of the boron dose. Our results strongly encourage further studying of irradiation arrangement of an isolated liver with epithermal neutron fields.

For treatment of superficially located tumors, such as head and neck cancers that invade the skin, the tumor dose may remain low on the skin when such tumors are treated with epithermal neutrons in boron neutron capture therapy (BNCT). The goal of this study was to examine the effects of bolus material for BNCT of superficial tumors, to verify the calculated 55Mn(n,γ) and the 197Au(n,γ) activation reaction rates and the neutron and the gamma doses in a phantom irradiated with a bolus, to measure the neutron activation of the bolus materials after irradiation, and according to depth dose distribution, to estimate when it is advantageous to use a bolus in BNCT. The present data show that both paraffin and water gel can be used as a bolus material for BNCT. However, we recommend paraffin for clinical use, since it is durable and can be easily shaped. A 5mm paraffin bolus increases the surface dose approximately 50%, and its use may be advantageous for treatment of superficial tumors where the planning target volume (PTV) reaches to 6cm or less in tissue depth.

Improvements have been made at the FiR 1 BNCT facility to ease the positioning of the patient with a tumor in the head and neck region into a lateral neutron beam. Shoulder recesses were constructed horizontally on both sides of the beam aperture. When shoulder recesses are not needed, they are filled with neutron attenuating filling blocks. MCNP simulations using an anthropomorphic human model BOMAB phantom showed that the main contribution to the increase in the effective dose to the patient's body due to the shoulder recesses was from the neutron dose of the arm. In a position when one arm is inside the shoulder recess, the maximal effective dose of the patient was estimated to be 0.7Sv/h. Dose measurements using the twin ionization chamber technique showed that the neutron dose increased on the sides as predicted by the MCNP model but there was no noticeable change in the gamma doses. When making the recesses into the lithium containing neutron shield material tritium contamination was confined using an underpressurized glove box and machine tools with local exhaust. The shoulder recesses give space for more flexible patient positioning and can be considered as a significant improvement of the Finnish BNCT facility.

Two clinical trials are currently running at the Finnish dedicated boron neutron capture therapy (BNCT) facility. Between May 1999 and December 2001, 18 patients with supratentorial glioblastoma were treated with boronophenylalanine (BPA)-based BNCT within a context of a prospective clinical trial (protocol P-01). All patients underwent prior surgery, but none had received conventional radiotherapy or cancer chemotherapy before BNCT. BPA-fructose was given as 2-h infusion at BPA-dosages ranging from 290 to 400 mg/kg prior to neutron beam irradiation, which was given as a single fraction from two fields. The average planning target volume dose ranged from 30 to 61 Gy (W), and the average normal brain dose from 3 to 6 Gy (W). The treatment was generally well tolerated, and none of the patients have died during the first months following BNCT. The estimated 1-year overall survival is 61%. In another trial (protocol P-03), three patients with recurring or progressing glioblastoma following surgery and conventional cranial radiotherapy to 50-60 Gy, were treated with BPA-based BNCT using the BPA dosage of 290 mg/kg. The average planning target dose in these patients was 25-29 Gy (W), and the average whole brain dose 2-3 Gy (W). All three patients tolerated brain reirradiation with BNCT, and none died during the first three months following BNCT. We conclude that BPA-based BNCT has been relatively well tolerated both in previously irradiated and unirradiated glioblastoma patients. Efficacy comparisons with conventional photon radiation are difficult due to patient selection and confounding factors such as other treatments given, but the results support continuation of clinical research on BPA-based BNCT.

Dose planning in boronneutron capture therapy (BNCT) is a complex problem and requires sophisticated numerical methods. In the framework of the Finnish BNCT project, new deterministic three-dimensional radiationtransport code MultiTrans SP 3 has been developed at VTT Chemical Technology, based on a novel application of the tree multigrid technique. To test the applicability of this new code in a realistic BNCT dose planning problem, cylindrical PMMA (polymethyl-methacrylate) phantom was chosen as a benchmark case. It is a convenient benchmark, as it has been modeled by several different codes, including well-known DORT and MCNP. Extensive measured data also exist. In this paper, a comparison of the new MultiTrans SP 3 code with other methods is presented for the PMMA phantom case. Results show that the total neutrondose rate to ICRU adult brain calculated by the MultiTrans SP 3 code differs less than 4% in 2 cm depth in phantom (in thermal maximum) from the DORT calculation. Results also show that the calculated 197 Au (n,γ) and 55 Mn (n,γ) reaction rates in 2 cm depth in phantom differ less than 4% and 1% from the measured values, respectively. However, the photondose calculated by the MultiTrans SP 3 code seems to be incorrect in this PMMA phantom case, which requires further studying. As expected, the deterministic MultiTrans SP 3 code is over an order of magnitude faster than stochastic Monte Carlo codes (with similar resolution), thus providing a very efficient tool for BNCT dose planning.

Calculation of neutron flux in three-dimensions is a complex problem. A novel approach for solving complicated neutron transport problems is presented, based on the tree multigrid technique and the Simplified P3 (SP3) approximation. Discretization of the second-order elliptic SP3 equations is performed for a regular three-dimensional octree grid by using an integrated scheme. The octree grid is generated directly from STL files, which can be exported from practically all computer-aided design-systems. Marshak-like boundary conditions are utilized. Iterative algorithms are constructed for SP3 approximation with simple coarse-to-fine prolongation and fine-to-coarse restriction operations of the tree multigrid technique. Numerical results are presented for a simple cylindrical homogeneous one-group test case and for a simplistic two-group pressurized water reactor pressure vessel fluence calculation benchmark. In the former homogeneous test case, a very good agreement with 1.6% maximal deviation compared with DORT results was obtained. In the latter test case, however, notable discrepancies were observed. These comparisons show that the tree multigrid technique is capable of solving three-dimensional neutron transport problems with a very low computational cost, but that the SP3 approximation itself is not satisfactory for all problems. However, the tree multigrid technique is a very promising new method for neutron transport.

The new 100 MW Jules Horowitz material testing reactor will be built in Cadarache, France. It will support, for instance, research on new types of innovative nuclear fuel. As a Finnish in-kind contribution, 3D emission and transmission tomography equipment will be delivered for both the reactor and the active component storage pool. The image reconstruction of activities inside the used nuclear fuel will be based on gamma spectrometry measurements. A new type of underwater digital X-ray linear detector array is under development for transmission imaging, based on GaAs and direct conversion of X-rays into an electrical signal. A shared collimator will be used for both emission and transmission measurements. Some preliminary design has been performed. For the current design, the expected gamma spectrometric response of a typical high-purity germanium detector has been simulated with MCNP for minimum and maximum source activities (specified by CEA) to be measured in future.

The Jules Horowitz Reactor (JHR) is a European experimental reactor under construction in CEA Cadarache. It will be dedicated to material and fuel irradiation tests, and to medical isotopes production. Non-Destructive nuclear Examinations systems (NDE) will be implemented in pools to analyse the irradiated fuel or tested material in their supporting experimental irradiation devices extracted from the core or its immediate periphery. The Nuclear Measurement Laboratory (NML) of CEA Cadarache is working in collaboration with VTT (Technical Research Centre in Finland) in designing and developing NDE systems implementing gamma-ray spectroscopy and high energy X-ray imaging of the sample and irradiation device. CEA is also designing a neutron radiography system for which NML is working on the detection system. Design studies are performed with Monte Carlo transport codes and specific simulation tools developed by the NML for X-ray and neutron imaging.

The mixed neutron–photon beam of FiR 1 reactor is used for boron–neutron capture therapy (BNCT) in Finland. A beam model has been defined for patient treatment planning and dosimetric calculations. The neutron beam model has been validated with an activation foil measurements. The photon beam model has not been thoroughly validated against measurements, due to the fact that the beam photon dose rate is low, at most only 2% of the total weighted patient dose at FiR 1. However, improvement of the photon dose detection accuracy is worthwhile, since the beam photon dose is of concern in the beam dosimetry. In this study, we have performed ionization chamber measurements with multiple build-up caps of different thickness to adjust the calculated photon spectrum of a FiR 1 beam model.

In this work, accuracy of the mcnp5 code in the electron transport calculations and its suitability for ionization chamber (IC) response simulations in photon beams are studied in comparison to egsnrc and penelope codes.The electron transport is studied by comparing the depth dose distributions in a water phantom subdivided into thin layers using incident energies (0.05, 0.1, 1, and 10 MeV) for the broad parallel electron beams. The IC response simulations are studied in water phantom in three dosimetric gas materials (air, argon, and methane based tissue equivalent gas) for photon beams ((60)Co source, 6 MV linear medical accelerator, and mono-energetic 2 MeV photon source). Two optional electron transport models of mcnp5 are evaluated: the ITS-based electron energy indexing (mcnp5(ITS)) and the new detailed electron energy-loss straggling logic (mcnp5(new)). The electron substep length (ESTEP parameter) dependency in mcnp5 is investigated as well.For the electron beam studies, large discrepancies (3%) are observed between the MCNP5 dose distributions and the reference codes at 1 MeV and lower energies. The discrepancy is especially notable for 0.1 and 0.05 MeV electron beams. The boundary crossing artifacts, which are well known for the mcnp5(ITS), are observed for the mcnp5(new) only at 0.1 and 0.05 MeV beam energies. If the excessive boundary crossing is eliminated by using single scoring cells, the mcnp5(ITS) provides dose distributions that agree better with the reference codes than mcnp5(new). The mcnp5 dose estimates for the gas cavity agree within 1% with the reference codes, if the mcnp5(ITS) is applied or electron substep length is set adequately for the gas in the cavity using the mcnp5(new). The mcnp5(new) results are found highly dependent on the chosen electron substep length and might lead up to 15% underestimation of the absorbed dose.Since the mcnp5 electron transport calculations are not accurate at all energies and in every medium by general clinical standards, caution is needed, if mcnp5 is used with the current electron transport models for dosimetric applications.

The purpose of the TOPAS project has been to maintain and develop stationary reactor physics code system, covering a wide range of calculation needs. Main results have been the development of a new Monte Carlo code Serpent, advanced methods for burnup calculations, and the development of sensitivity and uncertainty analysis methods. There has also been a strong interest in educating new competent people to work in field of reactor physics and nuclear safety. Several university degrees have been taken during the project, and the educational objectives have been well fulfilled. International co-operation has also been active. The results of the project will serve the needs of both the safety authorities and the power utilities.

Three treatment planning systems developed for clinical boron neutron capture therapy (BNCT) use are SERA developed by INL/Montana State University, NCTPlan developed by the Harvard-MIT and the CNEA group and JAEA computational dosimetry system (JCDS) developed by Japan Atomic Energy Agency (JAEA) in Japan. Previously, performance of the SERA and NCTPlan has been compared in various studies. In this preliminary study, the dose calculations performed with SERA and JCDS systems were compared in single brain cancer patient case with the FiR 1 epithermal neutron beam. A two-field brain cancer treatment plan was performed with the both codes. The dose components to normal brain, tumor and planning target volume (PTV) were calculated and compared in case of one radiation field and combined two fields. The depth dose distributions and the maximum doses in regions of interest were compared. Calculations with the treatment planning systems for the thermal neutron induced ((10)B and nitrogen) dose components and photon dose were in good agreement. Higher discrepancy in the fast neutron dose calculations was found. In case of combined two-field treatment plan, overall discrepancy of the maximum weighted dose was approximately 3% for normal brain and PTV and approximately 4% for tumor dose.

Three treatment planning systems developed for clinical BNCT use are SERA developed by INL/Montana State University, NCTPlan developed by the HarvardMIT and the CNEA group and JCDS developed by JAEA in Japan. Previously, performance of the SERA and NCTPlan has been compared in various studies. In this preliminary study, the dose calculations performed with SERA and JCDS systems were compared in single brain cancer patient case with the FiR 1 epithermal neutron beam. A twofield brain cancer treatment plan was performed with the both codes. The dose components to normal brain, tumor and planning target volume (PTV) were calculated and compared in case of one radiation field and combined two fields. The depth dose distributions and the maximum doses in regions of interest were compared. Calculations with the treatment planning systems for the thermal neutron induced (10B and nitrogen) dose components and photon dose were in good agreement. Higher discrepancy in the fast neutron dose calculations was found. In case of combined twofield treatment plan, overall discrepancy of the total maximum weighted dose was ~3% for normal brain and PTV and ~4% for tumor dose.

The MultiTrans software has been developed at VTT Technical Research Centre of Finland for 3D radiation transport problems. Adaptive tree multigrid technique is used as a deterministic solution method. This enables local refinement of the calculation grid combined with the use of effective multigrid acceleration on tree-structured nested grids: starting from a fast solution on coarse grid, successive solutions are obtained on finer and finer grids. In the MultiTrans code, simplified spherical harmonics (SP3) radiation transport approximation is used. In order to test the applicability of the new MultiTrans code to reactor dosimetry problems, light water reactor pressure vessel steel (LWR-PVS) benchmark experiment VENUS-3 (with partial length shielded assemblies) was chosen. The results show good agreement to the experimental reaction rates of the VENUS-3 benchmark, demonstrating the applicability of the new MultiTrans code in reactor dosimetry.

An international collaboration was organized to undertake a dosimetry exchange to enable the future combination of clinical data from different centers conducting neutron capture therapy trials. As a first step (Part I) the dosimetry group from the Americas, represented by MIT, visited the clinical centers at Studsvik (Sweden), VTT Espoo (Finland), and the Nuclear Research Institute (NRI) at Rez (Czech Republic). A combined VTT/NRI group reciprocated with a visit to MIT. Each participant performed a series of dosimetry measurements under equivalent irradiation conditions using methods appropriate to their clinical protocols. This entailed in-air measurements and dose versus depth measurements in a large water phantom. Thermal neutron flux as well as fast neutron and photon absorbed dose rates were measured. Satisfactory agreement in determining absorbed dose within the experimental uncertainties was obtained between the different groups although the measurement uncertainties are large, ranging between 3% and 30% depending upon the dose component and the depth of measurement. To improve the precision in the specification of absorbed dose amongst the participants, the individually measured dose components were normalized to the results from a single method. Assuming a boron concentration of 15 mu g g(-1) that is typical of concentrations realized clinically with the boron delivery compound boronophenylalanine-fructose, systematic discrepancies in the specification of the total biologically weighted dose of up to 10% were apparent between the different groups. The results from these measurements will be used in future to normalize treatment plan calculations between the different clinical dosimetry protocols as Part II of this study. (c) 2005 American Association of Physicists in Medicine. (Less)

VTT, the Technical Research Centre of Finland, has participated in to the on-going international blind benchmark on 3-D VENUS-2 MOX-fuelled reactor dosimetry calculations with the deterministic discrete-ordinate code TORT and the stochastic Monte Carlo code MCNP. Calculations have also been performed by in-house code MultiTrans, which is a deterministic 3-D radiation transport code under development. For both deterministic transport codes, the BUGLE-96 cross-section library was used. MCNP cross-sections were taken from the endf60 library based on ENDF/B-VI. For calculation of 58Ni(n,p), 115In(n,n’), 103Rh(n,n’), 64Zn(n,p), 237Np(n,f), and 27Al(n,α) responses, the IRDF-90 version 2 dosimetry cross-section library was used. With MCNP, this data was directly utilised in the SAND-II scheme, but for deterministic codes the data was condensed into the 47 BUGLE neutron groups. Corresponding fission flux values have been calculated for all the VENUS-2 detector positions. The comparison of the calculated fission fluxes between TORT and MCNP shows rather good agreement: 75 % of the values agree within 10 %. The maximum difference between TORT and MCNP results is 26 % for fission flux values of the 27Al(n,α) reaction inside the water gap. The MultiTrans results, on the other hand, show very large discrepancies inside the neutron pad when compared to the other codes, with 32 % maximum difference between MultiTrans and TORT, and 40 % between MultiTrans and MCNP. The disagreement inside the neutron pad was to some extent anticipated due to more approximative radiation transport method used in MultiTrans. The measured values are at the moment not yet open to the participants, and a comparative analysis between the calculated and measured values will remain as a future work.

The gamma dose determination using thermoluminescent (TL) dosimeters in mixed neutron-gamma fields, such as in boron neutron capture therapy (BNCT), is difficult due to the thermal neutron sensitivity of the detectors; especially when equipment capable of glow curve analysis is not available. The two TL analysis methods used previously in Finnish BNCT to correct the measured TL signal to obtain absorbed gamma dose in vivo were studied and compared, and an enhanced method was introduced. The three TL methods were found surprisingly consistent despite, e.g. the rough estimate made in the first method.

In order to assure the stability of the beam, the reliability of the beam monitoring system and the quality of the patient dose delivered, several procedures are followed at the FiR 1 epithermal beam in Finland. Routine procedures include in-phantom activation measurements before each patient treatment and a long-term follow-up of the results. The sensitivity of the beam monitors to external objects in the beam and to variations in the control rod positions in the reactor has been checked and found insignificant. The linearity of the beam monitor channels has been checked with activation measurements. It was found that due to saturation effects a correction of 11% has to be applied when extrapolating results from experiments at low power to full power using the reference monitor channel. The correction is even larger for other channels with higher count rates.

The minimum size of a water phantom used for calibration of an epithermal neutron beam of the boron neutron capture therapy (BNCT) facility at the VTT FiR 1 research reactor is studied by Monte Carlo simulations. The criteria for the size of the phantom were established relative to the neutron and photon radiation fields present at the thermal neutron fluence maximum in the central beam axis (considered as the reference point). At the reference point, for the most commonly used beam aperture size at FiR 1 (14 cm diameter), less than 1% disturbance of the neutron and gamma radiation fields in a phantom were achieved with a minimum a 30 cm × 30 cm cross section of the phantom. For the largest 20 cm diameter beam aperture size, a minimum 40 cm × 40 cm cross-section of the phantom and depth of 20 cm was required to achieve undisturbed radiation field. This size can be considered as the minimum requirement for a reference phantom for dosimetry at FiR 1. The secondary objective was to determine the phantom dimensions for full characterization of the FiR 1 beam in a rectangular water phantom. In the water scanning phantom, isodoses down to the 5% level are measured for the verifications of the beam model in the dosimetric and treatment planning calculations. The dose distribution results without effects caused by the limited phantom size were achieved for the maximum aperture diameter (20 cm) with a 56 cm × 56 cm × 28 cm rectangular phantom. A similar approach to study the required minimum dimensions of the reference and water scanning phantoms can be used for epithermal neutron beams at the other BNCT facilities.

Polymer gels have been reported as a new, potential tool for dosimetry in mixed neutron-gamma radiation fields. In this work, BANG-3 (MGS Research Inc.) gel vials from three production batches were irradiated with 6 MV photons of a Varian Clinac 2100 C linear accelerator and with the epithermal neutron beam of the Finnish boron neutron capture therapy (BNCT) facility at the FiR 1 nuclear reactor. The gel is tissue equivalent in main elemental composition and density and its T2 relaxation time is dependent on the absorbed dose. The T2 relaxation time map of the irradiated gel vials was measured with a 1.5 T magnetic resonance (MR) scanner using spin echo sequence. The absorbed doses of neutron irradiation were calculated using DORT computer code, and the accuracy of the calculational model was verified by measuring gamma ray dose rate with thermoluminescent dosimeters and 55Mn(n,γ) activation reaction rate with activation detectors. The response of the BANG-3 gel dosimeter for total absorbed dose in the neutron irradiation was linear, and the magnitude of the response relative to the response in the photon irradiation was observed to vary between different gel batches. The results support the potential of polymer gels in BNCT dosimetry, especially for the verification of two- or three-dimensional dose distributions.

In the novel MultiTrans SP3 radiation transport code the advanced tree multigrid technique is applied to the simplified P3 (SP3) transport approximation. The tree multigrid is generated directly from stereolitography (STL) files exported by computer-aided design (CAD) systems, thus allowing an easy interface for construction and upgrading of the geometry. The deterministic MultiTrans code allows fast solution of complicated three-dimensional transport problems in detail, offering a new tool for calculation of quantities of dosimetric interest. In order to determine the feasibility of a new code, computational benchmarks need to be carried out. In this paper, an application of the MultiTrans code to criticality problem is for the first time reported.

The aim of the Finnish BNCT-project is to start BNC-treatments of malignant brain tumors. The first clinical trial is planned to start in early 1999 at the treatment facility of the 250kW FiR 1 TRIGA research reactor. Excellent patient treatment facilities have been built at the reactor which is located only 5 km from the Helsinki University Central Hospital making the treatment facility very easy to reach.

In any radiotherapy the physical dose must have a metrologically traceable link to the national and international dosimetry standards. In addition, the uncertainty of the dose delivered to the patient has to be low enough for estimating the effects of the treatment beforehand and analyze then afterwards. Accurately characterized spectrum, fluence and dose distributions in a phantom form the basis for a reliable dose delivery to a patient and is essential also for biological studies at the BNCT beam. There is an obvious need for standardization in medical application of BNCT.1, 2, 3, 4, 5

The Finnish BNCT facility with a new epithermal beam is close to complete.1 The neutron spectrum of the beam were determined by calculations and activation measurements free in air.2 Consequently, as the FiR 1 BNCT beam was renewed, also the new beam model FiR(K63) to be used by the actual dose planning system was created and verified.3 The verification was done by comparison of the calculated and the measured dose distributions in phantoms using different dosimetric methods which together can provide appropriate reliability.

Using FiR 1, a 250kW TRIGA reactor as a neutron source for BNCT was screened as a viable option in 1990, as the future of the reactor was questioned. By the initiative of the medical radio isotope group at VTT the Finnish BNCT project started incorporating early on also the medical and medical physics sides.1,2 At first a thermal neutron source only was considered conceivable, but quite soon it was realized that using aluminum-aluminum fluoride moderator a high quality epithermal neutron source was quite feasible.3 After gaining support both from the medical community as well as from private and state financing sources a decision was made in 1994 to accomplish the necessary changes in the reactor and the reactor hall.