Neutron activation in interrupted neutron beams.

Modeling and simulating the induced effect on the steel collimator plug used in the PGAA facility of the Moroccan TRIGA Mark-II reactor under different neutron irradiation levels

Boron neutron capture therapy (BNCT) was conducted in a hospital using an accelerator-based neutron source. The neutron beam intensity at the patient position was evaluated offline using a gold-based neutron activation method. During BNCT neutron beam irradiation on patients, the neutron intensity was controlled in real time by measuring the proton beam current irradiated on a lithium neutron target. The neutron intensity at NCCH decreased owing to the degradation of the lithium neutron target during neutron irradiation. The reduction in the neutron beam intensity could not be monitored via proton beam measurement due to the dependence of neutron production on the neutron targetcondition. The duration of BNCT neutron irradiation should be controlled by monitoring the neutron beam intensity with a real-time neutron detector for reliable neutron irradiation on patients. The measurement accuracy of the online neutron beam monitor was experimentally obtained by comparing the gold radioactivity measured at the patient position. Radiation-induced damage was observed from the variation in the pulse height distributions of multichannel analyzer during long-term neutronexposure. Neutron beams were measured during neutron beam irradiation at the BNCT facility of Edogawa hospital in Japan using a neutron beam monitor comprising a 0.07- LiF layer and 40- back-illuminated thin Si pin diode. The proton beam was continuously irradiated until a cumulative total beam charge of approximately 3 kC was achieved. The online neutron beam monitor counting rates on the neutron target unit and gold saturation activities at the patient position were simultaneously measured through the entire duration of proton beamirradiation. The experimental results demonstrated the long-term operation of the online neutron beam monitor positioned on the neutron target unit during the entire duration of the neutron target lifespan without significant performance deterioration. A good synchronization was observed in a correlation distribution measured using the online neutron beam monitor and the gold neutron activation method. A conversion coefficient of 1.199 g with a standard deviation of 2.5% was evaluated. The neutron beam intensity irradiating on patients within an acceptable level of 5% as per the International Commission on Radiation Units and Measurements was evaluated from the online neutron counting rate at the 95% confidence level. The channel numbers of the triton peak and alpha particle edge decreased linearly owing to displacement damage and total ionizing dose effects induced mainly by thermal neutrons andphotons. Neutron doses can be accurately administered by complementing proton beam current measurements with the online neutron beam monitor. The online neutron beam monitoring technique allows monitoring fluctuations in the neutron beam intensity and tracking the degradation of the lithium target through the neutron target lifespan. Using a calibrated online neutron beam monitor, a prescribed dose can be administered in a manner similar to that in x-ray therapy, and the duration of neutron beam irradiation on the patient can be controlled in realtime.

The delayed-gamma neutron activation facility at Brookhaven National Laboratory was originally calibrated using an anthropomorphic hollow phantom filled with solutions containing predetermined amounts of Ca. However, 99% of the total Ca in the human body is not homogeneously distributed but contained within the skeleton. Recently, an artificial skeleton was designed, constructed, and placed in a bottle phantom to better represent the Ca distribution in the human body. Neutron activation measurements of an anthropomorphic and a bottle (with no skeleton) phantom demonstrate that the difference in size and shape between the two phantoms changes the total body calcium results by less than 1%. To test the artificial skeleton, two small polyethylene jerry-can phantoms were made, one with a femur from a cadaver and one with an artificial bone in exactly the same geometry. The femur was ashed following the neutron activation measurements for chemical analysis of Ca. Results indicate that the artificial bone closely simulates the real bone in neutron activation analysis and provides accurate calibration for Ca measurements. Therefore, the calibration of the delayed-gamma neutron activation system is now based on the new bottle phantom containing an artificial skeleton. This change has improved the accuracy of measurement for total body calcium. Also, the simple geometry of this phantom and the artificial skeleton allows us to simulate the neutron activation process using a Monte Carlo code, which enables us to calibrate the system for human subjects larger and smaller than the phantoms used as standards.

Locally advanced breast cancer treated with neutron beams: Long-term follow-up in 28 patients

Gold wire is commonly used for quality assurance (QA) of the neutron beam in a boron neutron capture therapy (BNCT) system. It is set in water and irradiated with the neutron beam, and then 412-keV gamma photons from the activated gold wire are measured by a semiconductor detector. Since this procedure takes time and labor, a more efficient method is desired. To reduce the time and labor to measure the radioactivity of an activated gold wire, we carried out imaging of 412-keV gamma photons from the activated gold wire using a developed high-energy gamma camera. After the gold wire was set in the depth direction in a water-filled phantom and irradiated with neutron beams using the BNCT system, gamma photon imaging was conducted with the developed high-energy gamma camera. On the measured image, a depth profile was set to obtain the neutron distribution, and this was compared with the profile sequentially measured with a semiconductor detector. An image of the 412-keV gamma photons was obtained with an imaging time of 1.5 hours. The estimated depth profile of the neutron beam from the gamma camera image closely matched that measured with a semiconductor detector. Imaging of the gamma photons emitted from the activated gold wire was possible, and it offers an efficient method to measure the thermal neutron distribution of the BNCT system. This method has the potential to reduce the time and labor for QA of a BNCT system.

Temperature transients induced by beam interruptions of different durations in MOX-fuelled and lead-bismuth cooled experimental ADS are investigated as a WPPT benchmark computational problem. The final report of the second phase of calculations will be published shortly by the NEA. This second phase of calculations aimed to investigate the impact of different fuel power density conditions on the transient results and uncertainties. Temperature variations are investigated assuming fresh fuel conditions, four different fuel power densities and two interruption durations. The main results obtained by the nine participants are presented and compared in this paper. Moreover, their uncertainty component due to the assumptions on models and data recommended in the benchmark specifications is evaluated by means of sensitivity studies carried out by some participants as a further contribution.

Discrepancy between DS86 calculated and measured neutron activation data in Hiroshima N Hunter and M W Charles (345-356) The thermal neutron activation measurements carried out over many years in Hiroshima and Nagasaki have been the subject of ongoing debate in recent years because they indicate that current DS86 neutron doses may have been significantly underestimated in Hiroshima. In this article we show that this discrepancy appears to be reinforced by using the latest thermal neutron activation data. However, some very recent fast neutron activation data suggest that the discrepancy may not be so great as that indicated by the majority of previous thermal neutron data. The extent of the revision needed to the DS86 neutron component remains subject to ongoing neutron activation measurements and re-analysis of existing published measurements.The impact of possible revisions to the DS86 dosimetry on neutron risk and relative biological effectiveness N Hunter and M W Charles (357-370) The current DS86 dosimetry system for the Japanese bomb survivors indicates that neutron doses were too low to directly derive useful neutron risk estimates. We have investigated the impact on neutron risks of increasing neutron doses in Hiroshima in line with the majority of thermal-neutron activation measurements. These increases, if substantiated, would have no significant impact on gamma radiation risk estimates but would allow useful neutron risk estimates to be made for solid cancers and leukaemia. Whether the Japanese bomb survivors can indeed form the basis for useful, directly determined, neutron risks depends on the reliability of existing thermal and fast neutron activation data. This is the subject of current research and debate.New methods for addition of doses from different environmental pathways W C Camplin et al (371-388) A new method for estimating radiation doses to UK critical groups is proposed for discussion. Results of surveillance at nuclear sites by the Food Standards Agency and the Scottish Environment Protection Agency are published in the Radioactivity in Food and the Environment (RIFE) report series. In these reports, doses to critical groups are normally estimated separately for gaseous and liquid discharge pathways. Simple summation of these doses would tend to overestimate doses actually received. Three different methods of combining the effects of both types of discharge in an integrated assessment are considered and ranked according to their ease of application, transparency, scientific rigour and presentational issues. The best method is then further developed using surveillance data for the calendar year 2000.Doses to organs and tissues from radon and its decay products G M Kendall and T J Smith (389-406) It is generally accepted that inhalation of radon decay products leads to a risk of lung cancer. But are there also risks to other organs or tissues? Is it justifiable to ignore the contribution from radon gas? While attention is usually concentrated on inhalation, is there a potential risk from ingestion of water containing radon or from deposition of decay products on the skin? This paper considers these questions, both by comparing the doses calculated for different tissues under different circumstances and also by examining the evidence from epidemiology.Radiological assessment of the level of safety in logging operations in the petroleum industry A A Abison (407-415) A study was initiated in 1993 in the Niger delta area to examine the radiological issues arising from the use of radioactive materials during hydrocarbon exploration and production activities spanning over forty years. The study was aimed at determining the level of radiological safety among various industrial establishments, but it could not make much progress without an assessment protocol. This paper describes such a protocol and the numerical result of its application to establishments involved in logging operations. The individual elements chosen for inclusion in the protocol are common sense and compatible with approaches such as loss prevention philosophy.Ge-doped optical fibre as a thermoluminescent dosimeter Y A Abdulla et al (417-421) This article describes the use of Ge-doped optical fibre as a thermoluminescent dosimeter for measuring the percentage depth dose of 6 and 10 MV x-rays. The results indicate that the Ge-doped optical fibre is in good agreement with the results obtained from a cylindrical ionisation chamber and TLD-100.

The DT program at the Tokamak Fusion Test Reactor (TFTR) created requirements on 14 MeV neutron measurements to measure from ∼106 n/cm2 (for triton burnup and Ohmic tritium plasmas) to >1012 n/cm2 (characteristic of >10 MW DT plasmas) with an accuracy of 7% (one-sigma).1 To maintain an absolute calibration over this dynamic range with active neutron detectors required one to go from some absolute standard at one fluence level to a measurement at a much higher fluence. Maintaining accuracy requires an extremely linear set of measurements not systematically affected over this dynamic range. Neutron activation can provide such linearity when care is taken with a number of effects such as gamma-ray detection efficiency and sample contamination.2 Absolutely calibrated neutron yield measurements using dosimetric (well-known cross section) reactions with thin (low-mass) elemental foils will be described. This technique makes the detector comparison to an absolute standard of gamma-ray activity correspond to all neutron fluences by reducing the sample mass while keeping the activation detectors operating in a linear counting mode; that is, one always uses low count rates to minimize pileup effects. The International Thermonuclear Experimental Reactor is projected to have 1000 s burn durations at fluxes of few 1013 n/cm2 s, or more neutron fluence per second than entire TFTR discharges. Extrapolating neutron activation to these higher fluences will require yet more care. Some of the issues at such high fluences will be discussed.3 The National Ignition Facility (NIF) is projected to yield 10 MJ of fusion energy, or up to 1012 n/cm2 at the vacuum vessel wall, similar to TFTR DT conditions. However, it is expected that much interesting physics will be performed at yields far less than those from ignition, possibly covering an even greater dynamic range than needed on TFTR. Thin foil techniques do not have the sensitivity required at low fluences. Absolutely calibrated neutron yield measurements using associated particle calibrations of thick (large mass) foils on accelerators4 will be compared. Using both thin and thick foil approaches can cover the dynamic range required for NIF.

The current DS86 dosimetry system for the Japanese bomb survivors indicatesthat neutron doses were so low that they prevent the direct derivation of anyuseful estimates of neutron risk. However, the large body of thermal neutronactivation measurements carried out over many years in Hiroshima and Nagasakiappear to indicate that current DS86 neutron doses may have been significantlyunderestimated in Hiroshima. An earlier companion paper has provided an updateof neutron activation measurements. While a large body of data appears tosupport a significant increase, there is ongoing debate and review regarding itsvalidity. However, as yet, there are no detailed, peer-reviewed, publishedrefutations of the neutron activation data which appear to support an increase inneutron doses. In this paper, we consider the impact of possible futurerevisions in the DS86 dosimetry on radiation risk estimates. We considerthe extreme range of possibilities from maintaining the existing DS86values, to changes in neutron doses in accord with the majority of existingneutron activation data. We have used the latest cancer incidence dataand cancer mortality data for the A-bomb survivors, and neutron doseshave been modified using a neutron revision factor (NRF) in line withthe latest thermal neutron activation measurements in Hiroshima. Incontrast to previous analyses, a nonlinear relationship between log(NRF)and slant range has been used which better represents the data beyondslant ranges of ∼1 km. The impact on the evaluation of neutron relativebiological effectiveness (RBE) and gamma radiation risk estimates has beenassessed. While DS86 neutron doses are too low to allow any useful directevaluation of neutron risk or neutron RBE, it becomes possible to derive moremeaningful values if neutron doses are increased in Hiroshima in linewith the broad range of thermal neutron activation measurements. Theuncertainties are smallest for the cancer incidence data. The best estimates ofneutron RBE give upper 95% confidence limits of about 6 for all solidtumours for the incidence data and about 28 for the mortality data. Theuncertainties in neutron RBE for leukaemia incidence are larger, andestimation at doses below about 0.1 Gy is not possible. There is no significantchange in the excess relative risk for gamma radiation for all solid tumourstaken together, compared with the current DS86 dosimetry. The resultspreclude neutron RBE values significantly greater than the current ICRPradiation weighting factors, which range between 5 and 20, depending onenergy. Whether or not the Japanese bomb survivors can indeed form thebasis for useful, directly determined neutron risks clearly depends on theveracity of existing neutron activation data. This is currently the subjectof careful international scrutiny and the outcome is eagerly awaited.

The most accurate determination of neutron yields from fusion reactors may be obtained from neutron activation measurements of elemental foils. On the Tokamak Fusion Test Reactor (TFTR), a re-entrant irradiation end has been installed to provide a low-scattering environment close to the plasma for neutron activation measurements. The ratio of energy-dependent fluence to total fusion yield is calculated using a fully three-dimensional Monte Carlo calculation with the Monte Carlo code for neutron and photon transport (MCNP). Corrections to the `virgin` fluence from attenuation and scattering are only 10 to 20% for deuterium-tritium (D-T) reactions and 30 to 40% for deuterium-deuterium reactions. A total 1-sigma accuracy of {+-}8% is achieved for D-T neutron yields over a wide dynamic range. This paper documents the response coefficients (hits per source neutron, where hits are activated nuclei per target nuclei) for use by the neutron activation system on TFTR; describes the possible systematic corrections needed (such as major radial variations or the impact of ion temperature on reactions with high-energy thresholds); and estimates uncertainties in the response coefficients. Results from in situ use of a D-T neutron generator are also analyzed using the MCNP modeling as an approximate benchmarking experiment; only 20% accuracy inmore » the comparison is possible because of poor counting statistics in the calibration experiment. 21 refs., 9 figs., 3 tabs.« less

The primary advantages that accelerator driven systems have over critical reactors are: (1) Greater flexibility regarding the composition and placement of fissile, fertile, or fission product waste within the blanket surrounding the target, and (2) Potentially enhanced safety brought about by operating at a sufficiently low value of the multiplication factor to preclude reactivity induced events. The control of the power production can be achieved by vary the accelerator beam current. Furthermore, once the beam is shut off the system shuts down. The primary difference between the operation of an accelerator driven system and a critical system is the issue of beam interruptions of the accelerator. These beam interruptions impose thermo-mechanical loads on the fuel and mechanical components not found in critical systems. Studies have been performed to estimate an acceptable number of trips, and the value is significantly less stringent than had been previously estimated. The number of acceptable beam interruptions is a function of the length of the interruption and the mission of the system. Thus, for demonstration type systems and interruption durations of 1sec 5mins 2500/yr and 50/yr are deemed acceptable. However, for industrial scale power generation without energy storagemore » type systems and interruption durations of t 5mins, the acceptable number of interruptions are 25000, 2500, 250, and 3 respectively. However, it has also been concluded that further development is required to reduce the number of trips. It is with this in mind that the following study was undertaken. The primary focus of this study will be the merit of a multi-beam target system, which allows for multiple spallation sources within the target/blanket assembly. In this manner it is possible to ameliorate the effects of sudden accelerator beam interruption on the surrounding reactor, since the remaining beams will still be supplying source neutrons. The proton beam will be assumed to have an energy of 1 GeV, and the target material will be natural lead, which will also be the coolant for the reactor assembly. Three proton beam arrangements will be considered, first a single beam (the traditional arrangement) with an entry at the assembly center, two more options will consist of three and six entry locations. The reactor fuel assembly parameters will be based on those of the S-PRISM fast reactor proposed by GE, and the fuel composition and type will be based on that proposed by Aker Solutions for use in their accelerator driven thorium reactor. The following table summarizes the parameters to be used in this study. The isotopic composition of the fertile material is 100% Th-232, and the plutonium isotopic distribution corresponds to that characteristic of the discharge from a typical LWR, following five years of decay. Thus, the isotopic distribution for the plutonium is; Pu-238 2.5%, Pu-239 53.3%, Pu-240 25.1%, Pu-241 11.8%, and Pu-242 7.3%.« less

Measurement of the applicability of various experimental materials in a medically relevant reactor neutron source part two: Study of H3BO3 and B-DTPA under neutron irradiation

Cholangiocarcinoma (CCA) is an aggressive cancer that is prevalent in the northeastern part of Thailand. Surgical treatment is the gold standard for CCA treatment, but some CCA patients are inoperable. Chemotherapy and radiotherapy are alternative treatments to improve the quality of life of patients. However, the effect of radiotherapy on CCA treatment is still unclear. In this study, we aimed to investigate the effect of X-rays and neutron beams on the human CCA cell line (KKU-055). First, KKU-055 cells were irradiated using 6 MV X-rays with a dose range of 0–5 Gy at King Chulalongkorn Memorial Hospital (KCMH) to obtain reference data. Next, cells were exposed to thermal neutron beams with doses ranging from 0 to 5 Gy using the Thai Research Reactor-1/Modification 1 (TRR-1/M1) at the Thailand Institute of Nuclear Technology (TINT). After neutron irradiation, survival curves were studied, and the relative biological effectiveness (RBE) was investigated. The findings revealed that the survival rate of the KKU-055 cells under X-ray irradiation is lower than that of neutron beams. To increase neutron interaction with the DNA of CCA cell lines, we plan to introduce boron compounds to CCA cell lines prior to neutron irradiation. This technique is referred to as boron neutron capture therapy (BNCT).

The primary motivation of this investigative study is trying to find an alternative treatment that can be used to slow down or treat glioblastoma due to the witnessed toxic side effects of the current drugs coupled with limited effectiveness in overall treatment. Consequently, a Chinese plant extract emodin proves to play a critical role in this investigative study since results from the Western blot and the other accompanying assays for anti-cancer effects indicate that it cannot work a lot to suppress cell migration and possible invasion, but rather emodin can be combined with radiation to give desired outcomes. Our result shows that the kind of radiation which acts well with emodin is neutron radiation rather than gamma radiation. Emodin significantly enhanced the radiosensitivity of LN18 and LN428 cells to γ-rays through MTT assay and cell counting. Accordingly, exposure to neutron radiation in the presence of emodin induced apoptotic cell death and autophagic cell death to a significantly higher extent, and suppressed cell migration and invasiveness more robustly. These effects are presumably due to the ability of emodin to amplify the effective dose from neutron radiation more efficiently. Thus, the study below is one such trial towards new interventional discovery and development in relation to glioblastoma treatment.

Silicon photo-multipliers, often called SiPM, are semiconductor photon detectors built from a square matrix of avalanche photo-diodes on common silicon substrate. SiPM have been proposed for several different applications in High Energy Physics, in particular where a large detection granularity is needed. In this presentation the results of a radiation hardness test performed at the Frascati Neutron Generator are presented. Several SiPM of different manufacturers have been irradiated integrating up to 7 1010 1-MeV-equivalent neutrons per cm2. For the first time, their performance have been recorded during the neutron irradiation and a gradual deterioration of their properties was found to happen already after an integrated dose of the order of 108 1-MeV-equivalent neutrons per cm2. The Frascati Neutron Generator (FNG) FNG uses a deuteron beam accelerated up to 300 keV impinging on a deuteron target to produce a nearly isotropic 2.5 MeV neutron output via the D(d,n)3He fusion reaction. The beam current at the target can be regulated up to 1 mA resulting in a maximum neutron production rate of 5 108 neutrons on the whole solid angle per second. Through the monitoring of the rate of associated emitted particles, protons or alpha, the neutron emission rate can be monitored on-line. This gives the unique possibility of measuring the effect of neutrons as long as the irradiation takes place. On-Line Measurements Six devices produced by the IRST and four produced by the Hamamatsu have been tested with neutrons. Depending on the distance from the production point, in four days of test, the SiPM integrated between 0.18 and 7.32 1-MeV-equivalent neutron per cm2. The current drawn by each device and its dark counting rate were continuously monitored and recorded while being irradiated. Fig. 1 shows that the current drawn by the SiPM starts to increase soon after the beginning of the irradiation. No differences between the current behavior of tested devices were found. The effects of the different neutron fluences are not visible at the level we operated. The neutron flux was kept off for a whole night while the currents were recorded. No significant recovery effects appeared. The absolute value of the current and the increasing rate, once the flux was back on, didn't change. The neutron beam has been paused several times in order to perform low voltage scans during the irradiation runs and to measure the effects on the dark currents and on the dark counting rates for different bias values. In the low voltage scans the current behavior changed rapidly with the integrated dose as it is shown in Fig.2. Off-Line Measurements The SiPM have been tested with cosmic rays before and after the neutron irradiation and the charge spectra obtained are shown in Fig 3. After the neutron irradiation, the gain was found to be about the half of the initial one (Fig.3 Bottom) and the noise pedestals (Fig. 3 Top) are much broader. The main effect is an important reduction of the detection efficiency from more than 95% to about 70%. Fig2: Measured currents as a function of the low voltage supply after different integrated doses Fig3: SiPM charge spectra with cosmic rays before (top) and after (bottom) the neutron irradiation. Fig1: Increasing factor of the current drawn by the SiPM as a function of the integrated neutron dose.