The optimisation of the fast neutron and gamma-ray transmission set-up for moisture measurement of coke.

In the present paper, modelling calculations with the Monte Carlo (MCNP4C) code were performed for the optimisation of the fast neutron and gamma-ray transmission, set-up, used for the humidity measurement of coke. The optimisation focused on maximising the sensitivity of the neutron flux to humidity changes and on lowering neutron-counting error, both leading to higher accuracy of coke moisture determination. Different materials used for the source shielding and neutron collimation, together with different dimensions of the neutron collimators were studied. The results obtained from the Monte Carlo modelling correlate with the real instrument performance.     

Borehole parametric study for neutron induced capture gamma-ray spectrometry using the MCNP code.

The MCNP Monte Carlo code has been used to simulate neutron transport from an Am-Be source into a granite formation surrounding a borehole. The effects of the moisture and the neutron poison on the thermal neutron flux distribution and the capture by the absorbing elements has been calculated. Thermal and nonthermal captures for certain absorbers having resonance structures in the epithermal and fast energy regions such as W and Si were performed. It is shown that for those absorbers having large resonances in the epithermal regions when they are present in dry formation or when accompanied by neutron poisons the resonance captures may be significant compared to the thermal captures.     

Radiochemical separation of carrier-free 204,206 Bi from alpha-irradiated thallium oxide target.

Alpha activation of Tl(2)O(3) target results in the formation of carrier-free 204,206 Bi. Two different radiochemical methods were used for the separation of bismuth radionuclides from the target matrix. A very high separation factor was achieved using liquid-liquid extraction (LLX) method with methyl isobutyl ketone (MIBK)-HCl system. Solid-liquid exchange adsorption was carried out using a novel inorganic ion exchanger, zirconium vanadate from HCl medium. The separation was found to be maximum around pH 2.     

RBE values for colo-rectal injury after caesium 137 gamma-ray and neutron irradiation. I. Single doses

Colo-rectal damage in mice has been assessed after caesium gamma irradiation and 3 MeV neutrons given as single doses. Several assays were used, including body weight changes, faecal deformity and lethality. Dose response curves have been constructed for each assay at different times after irradiation, ranging from 10 days to 16 months. The data have been analysed in terms of the time course of expression of damage and the RBE for neutrons. An initial loss of weight at 10-20 days was presumably related to epithelial denudation, but a dose-dependent weight reduction (compared with controls) persisted over the animal's life span. Mice died progressively after localised pelvic gamma irradiation; there was no sharp demarcation between an early and late phase of lethal injury. Death resulted from intestinal stricture or stenosis. The time course for lethality was qualitatively different after neutrons, with little progression of damage between 5 and 11 months. Faecal deformity was detectable as a higher proportion of small pellets when the rectum became constricted by fibrosis. No significant faecal deformity was observed before 6 months after which time dose response curves could be obtained. The RBE for early damage (assessed at 1-3 months) was 2.2-2.7. This fell to 1.7-1.9 for late damage (determined at 10-15 months) over the range of neutron doses of 7.5-12 Gy. The need for sublethal assays allowing for sequential evaluation of radiation damage within the same animal is stressed, as is the need to compare RBE values from early and late endpoints at equivalent neutron doses.     

Alternative radiochemical heavy ion activation methods for the production and separation of thallium radionuclides.

Heavy ion activation of natural gold foil with a 7Li beam results in the formation of carrier-free 199-201Pb radionuclides along with their corresponding decay products, 199-201Tl, in the matrix. At 49 MeV projectile energy, the production of 199Tl through its precursor 199Pb is maximal, while that of other. thallium radionuclides is minimal. Another way for the direct production of carrier-free thallium radionuclides is to irradiate natural platinum foil with a 7Li beam; on this case, production of 199Tl is maximal at 43 MeV projectile energy. The carrier-free thallium radionuclides have been separated from the bulk target matrix gold and platinum by liquid-liquid extraction using trioctylamine as a liquid anion exchanger.     

中子和γ射线照射对小鼠骨髓造血细胞死亡方式的探讨

目的探讨中子和γ射线照射后小鼠骨髓造血细胞的死亡方式及其意义。方法采用不同剂量中子和^60Coγ射线全身照射350只BALB／c小鼠，通过光镜、电镜、流式细胞术、原位末端标记和DNA凝胶电泳等观察造血细胞凋亡与坏死情况。结果中子照射后小鼠骨髓中坏死和凋亡的造血细胞均明显增多，5．5Gy组以坏死为主；而γ射线照射后则以凋亡为主，且呈剂量相关性。凋亡的细胞表现为核染色质浓缩、边移，呈半月型、环状或不规则状，凋亡小体形成，电泳下见DNA梯状图谱。坏死的细胞表现为核肿胀、溶解，线粒体膜、核膜等膜性结构破坏。结论2．5～5．5Gy中子及5．5～12Gy γ射线照射可使骨髓造血细胞死亡方式不同：中子照射见凋亡与坏死并存，5．5Gy组以坏死为主；而γ射线照射则以凋亡为主。     

Monte Carlo assessment of soil moisture effect on high-energy thermal neutron capture gamma-ray by 14N.

Among many conventional techniques, nuclear techniques have shown to be faster, more reliable, and more effective in detecting explosives. In the present work, neutrons from a 5 Ci Am-Be neutron source being in water tank are captured by elements of soil and landmine (TNT), namely (14)N, H, C, and O. The prompt capture gamma-ray spectrum taken by a NaI (Tl) scintillation detector indicates the characteristic photo peaks of the elements in soil and landmine. In the high-energy region of the gamma-ray spectrum, besides 10.829 MeV of (15)N, single escape (SE) and double escape (DE) peaks are unmistakable photo peaks, which make the detection of concealed explosive possible. The soil has the property of moderating neutrons as well as diffusing the thermal neutron flux. Among many elements in soil, silicon is more abundant and (29)Si emits 10.607 MeV prompt capture gamma-ray, which makes 10.829 MeV detection difficult. The Monte Carlo simulation was used to adjust source-target-detector distances and soil moisture content to yield the best result. Therefore, we applied MCNP4C for configuration very close to reality of a hidden landmine in soil.     

Comparison of a digital and an analog signal processing system for neutron inelastic gamma-ray spectrometry

We compared the value of using a digital signal processing unit for gamma-ray spectroscopy with that of an analog one for in situ measurements of gamma-rays generated by inelastic neutron scattering reactions with soil elements. A large cylindrical NaI(Tl) scintillation detector, 15.24 cm high by 15.24 cm diameter was used to measure carbon (C) and oxygen (O). The performance of the systems was assessed as a function of input count rate (ICR) by monitoring the peak areas of the C, 4.43 MeV, and O, 6.13 MeV, gamma-rays. In separate experiments, the digital and the analog systems were also compared using an intense 10.3 mCi ¹³⁷Cs source to vary the ICR, and the 1.17 MeV peak area of ⁶⁰Co was used as the reference.     

Analysis of cadmium in food by multiple prompt gamma-ray spectroscopy.

The Cd concentration in food is a public concern related to the human health. In order to remove Cd-polluted food, the development and validation of a rapid and sensitive method of Cd analysis is required. By applying the multiple gamma-ray detection method to prompt gamma-ray analysis (PGA), the influence from nuclei which emit only one prompt gamma-ray at a time at every neutron capture reaction can be reduced, therefore the quantification limit of Cd is improved significantly. The limit of Cd contained in rice in the case of MPGA was evaluated, and under our proposed experimental conditions, it may be possible to quantify Cd content in rice to within 0.2 ppm in 10 min.     

Determination of natural radioactivity in building materials used in the Rawalpindi/Islamabad area by gamma-ray spectrometry and instrumental neutron activation analysis.

Samples of sand and manufactured building materials collected from the Rawalpindi/Islamabad area have been analyzed for the primordial natural radionuclides 238U, 232Th and 40K using gamma-ray spectrometry. The uranium and thorium contents have also been determined by instrumental neutron activation analysis (INAA). The specific activities of these radionuclides in the samples are compared with those of the world averages for soil. The comparison shows that, of the nine samples analyzed, only the three brick samples have higher activities than the world averages for soil. The specific activities in these materials, having a radium equivalent activity of less than 370 Bq kg-1, when evaluated for radiological effects show that all materials meet the external gamma-ray dose limitation of 1.5 mSv y-1. The gamma-spectrometric and INAA techniques complemented each other well in this study.     

Neutron contamination of 10 MV X-rays: its relevance to treatment room door and maze design

10 MV X-rays produced by linear accelerators are commonly used in radiotherapy. This paper demonstrates that neutron leakage cannot be neglected at 10 MV when direct access doors are used or when short mazes, typically less than 7 m in length, are employed. Measurements and calculations were made in terms of the operational quantities and personal and ambient dose equivalents, and the use of neutron fluence or contentious issues, such as effective dose and radiation weighting factors, were avoided. Neutron leakage from the head of an Elekta Precise linear accelerator was measured and compared with published data. The neutron protection afforded by a direct access door was considered and the use of borated polyethylene to remove thermal neutrons was found to be unnecessary. Extensive measurements were made in two treatment room mazes using survey meters of the Andersson-Braun and Leake type. Acceptable agreement was found between the individual instrument readings and calculations. However, one meter unexpectedly appeared to both respond to photons and over-respond at high neutron dose rates.     

Air kerma transmission factors of Scattered X-rays in the maze of a Linac room for lead shield

Spectra of scattered X-rays in the maze of a Linac (X-ray energies of 4, 6, and 10 MV) room were estimated by means of the Monte Carlo simulation, and air kerma transmission factors of the X-rays scattered through a lead shield were evaluated based on those spectra. Spectra of scattered X-rays showed a maximum in the energy area below 200 keV. The higher the accelerated electron energy, also, the smaller the scattering angle that tended to spread to the higher energy area of the distribution of spectra. The air kerma transmission factor of 120 degrees scattered X-rays of 4 MV X-rays obtained in this study was larger than the transmission factors of 124 degrees scattered photons of (60)Co gamma rays through a lead shield given in ICRP. The air kerma transmission factors of 120 degrees scattered X-rays of 6 MV X-rays were smaller than the transmission factors of 90 degrees scattered photons of (60)Co gamma rays. The air kerma transmission factors of 120 degrees scattered X-rays of 10 MV X-rays was slightly larger than transmission factors of 90 degrees scattered photons of (60)Co gamma rays. Therefore, in the case of a 4 MV X-ray Linac room, the calculation method given in the "Manual of Practical Shield Calculation of Radiation Facilities (2000)" causes underestimation of leakage doses.     

Analysis of epithermal neutron production by near-threshold (p,n) reactions

A recent advance in portable accelerator neutron source development was research on production of epithermal neutrons by near-threshold charged-particle reactions. When the projectile energy is accurately controlled at an energy close to the reaction threshold, the neutrons produced will have energies less than or around 100 keV and can be used with little or no moderation or filtration in neutron capture therapy. Although the total neutron yield is lower than at higher proton energies, the epithermal neutron flux may be sufficiently intense because of the softer energy spectrum and the requirement for less neutron moderation. This paper presents an analysis of the main characteristics of epithermal neutron production by this method using the Li (p,n) reaction as an example. The energy, yield and angular characteristics of neutron emission are discussed. The achievable epithermal fluxes are computed from experimental data. The results are used to assess the feasibility of near-threshold production of epithermal neutrons for neutron capture therapy with compact accelerators such as a RFQ proton acceelerator. The results indicated that, using a Li₃N target, 1 mA of 2 MeV protons will produce 10⁹ n/cm²/s with an average energy of 83 keV while 5.6 mA of 1.91 MeV protons can produce 10⁹ n/cm²/s with an average energy of 45 keV.     

Sympathetic denervation and reinnervation after the maze procedure.

We evaluated serial changes in cardiac sympathetic nerve distribution using 123I-metaiodobenzylguanidine (123I-MIBG) after the Maze procedure. The Maze procedure, in which multiple incisions are made in the atrium, has been concomitantly performed with mitral valve (MV) surgery in an attempt to eliminate atrial fibrillation (AF). Although attenuation of the sinoatrial node response to exercise and a reduction of left ventricular function (left ventricular ejection fraction [LVEF]) in early stages after the Maze procedure have been suggested, factors leading to these changes have not been clarified. METHODS: Thirteen patients with MV disease were enrolled in this study. Six of them had undergone MV surgery and the Maze procedure (Maze+), and 7 had undergone MV surgery without the Maze procedure (Maze-). All patients underwent cardiac 123I-MIBG imaging preoperatively and 10 d and 1 y after surgery to assess 123I-MIBG uptake (heart-to-mediastinum count ratio of early planar images [H/M]) and the washout rate (WR). Radionuclide ventriculography was also performed to calculate LVEF 3 d after each 123I-MIBG imaging. RESULTS: The LVEF of the Maze+ group significantly decreased 10 d after surgery (44.2 +/- 4.8; mean +/- SD) compared with that before surgery (60.3 +/- 6.9; P < 0.05) and significantly increased at 1 y (65.2 +/- 2.9) compared with that at 10 d (P < 0.05). In the Maze- group, there was no significant change 10 d (53.0 +/- 12.3) and 1 y (58.6 +/- 4.8) after surgery compared with that before surgery (60.4 +/- 4.6) (P = not significant, each). In the Maze+ group, the H/M (1.51 +/- 0.18) was significantly lower at 10 d after than that at the preoperative stage (1.90 +/- 0.25; P < 0.05) but significantly recovered at 1 y (2.23 +/- 0.18; P < 0.05) with a similar transient increase in the WR (36.7% +/- 6.1% at preoperative stage; 46.9% +/- 3.4% at 10 d; 39.9% +/- 6.5% at 1 y; P < 0.05, each). On the other hand, the Maze- group did not show a significant change in the H/M (1.94 +/- 0.32, 2.06 +/- 0.18, and 2.13 +/- 0.17, respectively; P = not significant, each) but did exhibit a significant decrease in the WR (40.4% +/- 5.1%, 37.0% +/- 5.1%, and 32.9% +/- 2.5%, respectively; P < 0.05, each). Changes in the H/M of both groups significantly correlated with the change in LVEF (r = 0.82; P < 0.05), and the WR showed a significant inverse correlation with changes in the LVEF (r = -0.81; P < 0.05). CONCLUSION: Cardiac sympathetic nerves were denervated at early stage and reinnervated at late stage after the Maze procedure. Such adrenergic nerve changes may be correlated, at least in part, with changes in left ventricular function after this procedure.     

Optimization of the target of an accelerator-driven neutron source through Monte Carlo numerical simulation of neutron and gamma transport by the PRIZMA code

At Budker Institute of Nuclear Physics, epithermal neutron source for neutron-capture therapy was built and neutron generation was realized. Source is based on tandem accelerator and uses near-threshold neutron generation from the reaction ⁷Li(p,n)⁷Be. The paper describes target optimization through the numerical simulation of proton, neutron and gamma transport by Monte Carlo method (PRIZMA code). It is shown that the near-threshold mode attractive due low activation provides high efficiency of the dose and acceptable therapeutic ratio and advantage depth.     

The additivity of the fast neutron and gamma-ray induced radiophotoluminescence and re-usability of Toshiba FD-7 and SEI High-Z RPL glass dosimeters

The radiophotoluminescent response of Toshiba FD-7 and SEI Schulman High-Z glass dosimeters to fast neutrons and to γ-rays is found to be additive. It is also demonstrated that they may be conveniently annealed and reused; the ratio of the fast neutron to γ-ray sensitivity remaining fixed.     

Study of material changes of SINQ target rods after long-term exposure by neutron radiography methods.

This paper describes the results of non-destructive investigations by indirect neutron radiography methods obtained at the facility NEUTRA [Nondestruct. Testing Eval. 16 (2000b) 203], spallation neutron source SINQ [Operating experience and development projects at SINQ, PSI Report 98-04, ISSN 1019-0643]. Target rods from the second SINQ metal target were removed after 6 Ah of proton beam exposure and studied under well-shielded conditions. No real damage was found at one of the 11 observed rods and one tube. However, hydrogen accumulation could be identified inside the zircaloy rods and the steel rods as well. Whereas the hydrogen has a homogenous distribution in Zr (with the peak value near the centre of the applied beam), the steel samples show clusters of hydrogen near the edge of the Zr cladding. Lead (in steel cladding) was found modified by accumulations of spallation products, mainly mercury. In the radiography images, a depression of the neutron field was observed due to the absorption by mercury. The applied method with Dy and In as neutron converters and imaging plates [Nucl. Instrum. Methods 377 (1996) 119] as secondary detectors seems to be optimal for such kind of investigations, especially when quantitative considerations have to be made.     

On the invariability of the total-to-peak ratio in gamma-ray spectrometry.

It is demonstrated that the assumption on the independence of the total-to-peak ratio of the source position in gamma-ray spectrometry holds for coaxial detectors with sufficient accuracy to make it applicable to the determination of the full energy peak efficiencies in environmental measurements via the so-called efficiency transfer method. The calculations show that for planar detectors the assumption breaks down at higher gamma-ray energies and that for such detectors the use of the efficiency transfer method is only warranted for gamma-ray energies up to 200 keV.