• Nuclear Engineering and Technology
• /
• v.2 no.2
• /
• pp.85-95
• /
• 1970

An attempt has been made to do interpretation of the fast neutron dose with two threshold detectors incorporated with the Harwell criticality locket. This method is based on the assumption that the spectral distribution of fission neutrons in criticality accidents may be governed by one spectral parameter. The surface-absorbed dose for a unit fission neutron fluence seems to be insensitive to spectral shifts of the fission neutron spectrum. The average cross-sections for the activation detectors, however, are considerably changed with the neutron spectral shape, which may lead to a large error in calculating the dose from the reaction rate if one uses a fixed value for the average cross sections regardless of the neutron spectral distribution. Besides, the doses calculated from three representative formulae for fission neutron spectra have been compared : these formulae are Watt, Cranberg at al. and Maxwellian forms. The results obtained front the Maxwellian formula show a departure from the Watt and Cranberg's, both being similarly close.

• Journal of radiological science and technology
• /
• v.26 no.3
• /
• pp.51-57
• /
• 2003

Mitigation of fast neutron irradiation damage on reactor vessel and improvement of mechanical integrity are desired for the successful plant life-time extension. In this study, the performance of metallic hydride for this application is reviewed and compared. First, selected prospective metallic hydrides are evaluated by MCNP code and put into the attenuation test using Cf-252 neutron source. Since for the reactor application high moderation and reflection with no absorption are favored, Z factor is introduced for the evaluation. According to the Z value estimation $ZrD_2$ and $TiD_2$ are turned out to be the most favorable fast neutron shielding materials. More thorough evaluation by computer simulation and experimentally, will be followed.

• The Science & Technology
• /
• v.6 no.2 s.45
• /
• pp.28-31
• /
• 1973

원자로제어를 위한 중성자열전대의 응답시간 단축을 목적으로 진공증착된 박모열전대를 이용하여 중성자 열전대를 시작하였다. 이의 실험결과를 선열전대의 것과 비교하였으며, 열중성자동범위 2x(10에 8승)x8x10¹³ neutrons/cm²/sec에서 좋은 선형특성을 가지고 있었다. 시작된 박모중성자열전대를 사용하여 TRIGA MARK-Ⅱ 원자로 로필에서의 열중성자속분포를 측정하였다.

• Nuclear Engineering and Technology
• /
• v.5 no.4
• /
• pp.334-338
• /
• 1973

The horizontal distribution of the fast neutron flux density in the Bulk Shielding Experimental Tank of the TRIGA Mark-II reactor at the steady power of 250 KW has been measured using a solid state track detector which is natural mica placed in contact with $^{232}$ Th fissile foil. The neutron flux density was calculated on the assumption that the fast neutron spectrum is similar to that from the thermal-induced $^{235}$ U fission. The resulting flux density distribution along the horizontal line from the center of the thermalizing column door is presented in tabular and graphical forms.

• Progress in Medical Physics
• /
• v.8 no.2
• /
• pp.45-57
• /
• 1997

It is mandatory to measure accurately the dose distribution and the total absorbed dose of fast neutron for putting it to the clinical use. At present the methods of measurement of fast neutron are proposed largely by American Associations of Physicists in Medicine, European Clinical Neutron Dosimetry Group, and International Commission on Radiation Units and Measurements. The complexity of measurement, however, induces the methodological differences between them. In our study, therefore, we tried to establish a unique technique of measurement by means of measuring the emitted doses and the dose distribution of fast neutron beam from neutron therapy machine, and to invent a standard method of measurement adequate to our situation. For measuring the absorbed doses and the dose distribution of fast neutron beam, we used IC-17 and IC-18 ion chambers manufactured by A-150 plastic(tissue-equivalent material), IC-17M ion chamber manufactured by magnesium, TE gas and Ar gas, and RDM 2A electrometer. The magnitude of gamma-contamination intermingled with fast neutron beam was about 13％ at 5cm depth of standard irradiated field, and increased as the depth was increased. At the central axis the maximum dose depth and 50％ dose depth were 1.32cm and 14.8cm, respectively. The surface dose rate was 41.6-54.1％ throughout the entire irradiated fields and increased as the irradiated fields were increased. Beam profile was that the horn effect of about 7.5％ appeared at 2.5cm depth and the flattest at 10cm depth.

• Journal of the Korean Institute of Telematics and Electronics
• /
• v.9 no.5
• /
• pp.1-5
• /
• 1972

In order to improve the response time of nelltron theromopile for reactor control a neutron thermopile made use of a vacuunl evaporated thin film thor mocouple was fablicated and tested. The test results were compared with a wire-type neutron thermopile. Good linearities between the response of the neutron thermopile and the thermal flux has been shown in the ranges from n/$\textrm{cm}^2$/sec. Thermal neutron flux distributions in the core of TRIGA Mark-II reactor were measured using the fabricated neutron thermopile, and the results were conpared with data obtained by the acrivatin foil measurement.

• 대한방사선치료학회:학술대회논문집
• /
• 1997.03a
• /
• pp.13-14
• /
• 1997

• Journal of Radiation Protection and Research
• /
• v.32 no.4
• /
• pp.168-177
• /
• 2007

A neural network model to evaluate the neutron exposure on the reactor pressure vessel inner diameter was developed. By using the three dimensional synthesis method described in Regulatory Guide 1.190, a simple linear equation to calculate the neutron spectrum on the reactor pressure vessel was constructed. This model can be used in a quick estimation of fast neutron flux which is the most important parameter in the assessment of embrittlement of reactor pressure vessel. This model also used in the selection of an optimum core loading pattern without the neutron transport calculation. The maximum relative error of this model was less than 3.4% compared to the transport calculation for the calculations from cycle 1 to cycle 23 of Kori unit 1.

• Analytical Science and Technology
• /
• v.5 no.4
• /
• pp.367-372
• /
• 1992

In order to measure the energy spectrum of a radioactive neutron source, the pulse shape discrimination (PSD) system with organic scintillator, NE-213, was characterized by using some of the gamma ray sources and neutron source, Am-Be. The figure of merit of the rise time spectrum of AmBe source measured by this system was about 1.13. This value agrees well with the value of 1.3 which is measured for monoenergetic source, $^{12}C(d,\;n)^{13}N$. The results of present experiment for performance test of NE213-PSD system will provide the useful technique to measure the spectrum of neutron-gamma mixed field and to establish the neutron energy spectrum and flux density standards.

• Journal of IKEEE
• /
• v.20 no.1
• /
• pp.61-67
• /
• 2016

In this paper, a hardware design method is proposed for high speed high precision neutron radiation measurements. Our system is fabricated to use a high performance A/D Converter for digital data conversion of high precision and high speed analog signals. Using a neutron sensor, incident neutron radiation particles are detected; a precision microcurrent measurement module is also included: this module allows for more precise and rapid neutron radiation measurement design. The high speed high precision neutron measurement hardware system is composed of the neutron sensor, variable high voltage generator, microcurrent precision measurement component, embedded system, and display screen. The neutron sensor detects neutron radiation using high density polyethylene. The variable high voltage generator functions as a 0 ~ 2KV variable high voltage generator that is robust against heat and noise; this generator allows the neutron sensor to perform normally. The microcurrent precision measurement component employs a high performance A/D Converter to precisely and swiftly measure the high precision high speed microcurrent signal from the neutron sensor and to convert this analog signal into a digital one. The embedded system component performs multiple functions including neutron radiation measurement for high speed high precision neutron measurements, variable high voltage generator control, wired and wireless communications control, and data recording. Experiments using the proposed high speed high precision neutron measurement hardware shows that the hardware exhibits superior performance compared to that of conventional equipment with regard to measurement uncertainty, neutron measurement rate, accuracy, and neutron measurement range.