Self-powered detectors sensitivity determination

International Nuclear Information System (INIS)

Surkov, V.; Soares, A.J.

1994-01-01

The determination of the initial sensitivity of Self Powered Detectors (SPDs) was performed. Measurements of thermal, epithermal and to gamma flux sensitivities were made with Vanadium, Cobalt, Rhodium, Silver and Platinum SPDs and, when possible, the values are compared with the ones from the existing literature. The determination of neutron sensitivity was realized using the IEA-R1 Reactor from IPEN. The thermal and epithermal neutron flux were determined with bare and Cadmium covered activation foils. (GOLD). (author)

Radiological safety aspects associated with the handling, storage and disposal of self power neutron detectors in TAPS - 3 and 4

International Nuclear Information System (INIS)

Parida, B.K.; Mudgal, B.; Ghadigoankar, V.R.; Niraj; Ashok; Pati, C.K.; Patil, P.M.; Pawar, S.K.; Varadhan, R.S.

2006-01-01

At Tarapur Atomic Power Station 3 and 4, 540 MWe Pressurised Heavy Water Reactors, core being large in size requires a continuous in core monitoring for local flux disturbances. Nearly 200 Self Powered Neutron Detectors (SPNDs) of the Straight Individually Replaceable (SIR) type are distributed in the reactor core. For purpose of reactor regulation and protection, cobalt SPNDs that have a prompt response for changes in power is used for in-core flux mapping, vanadium SPNDs that provide accurate measure of neutron flux, even though having slow response is used In core SPNDs are placed in Vertical Flux Units (VFU) and Horizontal Flux Units (HFUs). These SPNDs were to be replaced at regular intervals to meet the design intent. Cobalt SPNDs have dose rates of the order of 1000 Gy/h and the Mineral Insulated (MI) cables of Vanadium SPNDs have dose rates of the order of 100 Gy/h. So far 3 Cobalt SPNDs were removed from HFUs and are being stored in lead shielding inside spent fuel storage facility. These high active components were handled with meticulous planning with lowest exposures to the maintainers. Radiological safety aspects of handling and storage of SPNDs are discussed in this report. (author)

In-core neutron flux measurements at PARR using self powered neutron detector

International Nuclear Information System (INIS)

Hussain, A.; Ansari, S.A.

1989-10-01

This report describes experimental reactor physics measure ments at PARR using the in-core neutron detectors. Rhodium self powered neutron detectors (SPND) were used in the PARR core and several measurements were made aimed at detector calibration, response time determination and neutron flux measurements. The detectors were calibrated at low power using gold foils and full power by the thermal channel. Based on this calibration it was observed that the detector response remains almost linear throughout the power range. The self powered detectors were used for on-line determination of absolute neutron flux in the core as well as the spatial distribution of neutron flux or reactor power. The experimental, axial and horizontal flux mapping results at certain locations in the core are presented. The total response time of rhodium detector was experimentally determined to be about 5 minutes, which agree well with the theoretical results. Because of longer response time of SPND of the detectors it is not possible to use them in the reactor protection system. (author). 10 figs

Self-powered neutron detector

International Nuclear Information System (INIS)

Goldstein, N.P.; Todt, W.H.

1974-01-01

The invention relates a self-powered neutron detector comprising an emitting body, an insulating material surrounding said body, and a conducting outer cover, a power conductor connected to the emitting body and passing through the insulating material permitting to insert an ammeter between said emitting body and said cover. The invention is characterized in that said emitting body is surrounded by a thin conducting layer of small cross section for neutrons made of high density material said material being capable of absorbing the beta-radiations due to the degradation of the emitting body activating product, while transmitting the fast electrons of high average energy emitted by said emitting body. This can be applied to safety control devices required to provide a quick answer [fr

Self-powered detectors with thulium emitter

International Nuclear Information System (INIS)

Haller, P.; Klar, E.

1978-01-01

In addition to fission chambers, prompt-indicating self-powered (SPN) detectors are used for measuring the neutron flux density in the core of power reactors. Although current SPN detectors with a cobalt emitter give satisfactora results, detectors with other emitter materials have been analyzed and tested. The author describes the properties and decay pattern of the nuclide thulium and presents the results of measurements made while testing thulium detectors. (orig.) [de

Self-powered neutron flux detector

International Nuclear Information System (INIS)

Kroon, J.

1979-01-01

A self-powered neutron flux detector having an emitter electrode, at least a major portion of which is, 95 Mo encased in a tubular collector electrode and separated therefrom by dielectric material. The 95 Mo emitter electrode has experimentally shown a 98% prompt response, is primarily sensitive to neutron flux, has adequate sensitivity and has low burn up. Preferably the emitter electrode is molybdenum which has been enriched 75% to 99% by weight with 95 Mo

Response characteristics of self-powered flux detectors in CANDU reactors

International Nuclear Information System (INIS)

Allan, C.J.

1978-05-01

As part of the development of a new flux-detector assembly for future CANDU reactors, the sensitivities of a variety of vanadium, cobalt and platinum self-powered detectors have been determined in a simulated CANDU core installed in the ZED-2 test reactor at CRNL. While the vanadium and cobalt detectors had solid emitters, the platinum detectors were of two types, having either solid platinum emitters, or emitters consisting of a platinum sheath over an Inconel core. Almost all of the signal from the cobalt and vanadium detectors is due to neutron events in the emitters. For these detectors we have measured the total sensitivities per unit length. For the platinum detectors, reactor γ-rays and neutrons both contribute appreciably to the output signal, and in addition to the total sensitivity, we have determined the individual neutron and γ-ray sensitivities for these detectors. It was found that the detector sensitivities depend primarily on emitter diameter and that the observed variations can be fitted by means of power laws. (author)

Characteristics of self-powered neutron detectors used in power reactors

International Nuclear Information System (INIS)

Todt, William H. Sr.

1998-01-01

Self-powered neutron detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors worldwide. This paper describes the basic properties of these radiation sensors including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs, which are being effectively, used in in-core instrumentation systems for pressurized water, heavy water and graphite moderated light water reactors. Also examples are shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurized water and heavy water reactors worldwide. (author)

Sensitivity Calculation of Vanadium Self-Powered Neutron Detector

International Nuclear Information System (INIS)

Cha, Kyoon Ho

2011-01-01

Self-powered neutron detector (SPND) is being widely used to monitor the reactor core of the nuclear power plants. The SPND contains a neutron-sensitive metallic emitter surrounded by a ceramic insulator. Currently, the rhodium SPND has been used in many nuclear power plants. The lifetime of rhodium is too short (about 3∼5 years) to operate the nuclear power plant economically. The vanadium (V) SPND is also primarily sensitive to neutrons like rhodium, but is a somewhat slower reaction time as that of a rhodium SPND. The benefit of vanadium over rhodium is its low depletion rate, which is a factor of 7 times less than that of rhodium. For this reason, a vanadium SPND has been being developed to replace the rhodium SPND which is used in OPR1000. Some Monte Carlo simulations were accomplished to calculate the initial sensitivity of vanadium emitter material and alumina (Al 2 O 3 ) insulator with a cylindrical geometry. An MCNP-X code was used to simulate some factors (neutron self shielding factor and electron escape probability from the emitter) necessary to calculate the sensitivity of vanadium detector. The simulation results were compared with some theoretical and experimental values. The method presented here can be used to analyze the optimum design of the vanadium SPND

Characteristics of self-powered neutron detectors used in power reactors

International Nuclear Information System (INIS)

Todt, W.H.

1997-01-01

Self-Powered Neutron Detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors world-wide. The basic properties of these radiation sensors are described including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs which are being effectively used in in-core instrumentation systems for pressurised water, heavy water and graphite moderated light water reactors. Examples are also shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurised water and heavy water reactors worldwide. This paper is a summary of a new IEC standard to be issued in 1996 describing the characteristics and test methods of self-powered detectors used in nuclear power reactors. (author)

Self-powered neutron flux detector assembly

International Nuclear Information System (INIS)

Allan, C.J.; McIntyre, I.L.

1980-01-01

A self-powered neutron flux detector has both the central emitter electrode and its surrounding collector electrode made of inconel 600. The lead cables may also be made of inconel. Other nickel alloys, or iron, nickel, titamium, chromium, zirconium or their alloys may also be used for the electrodes

Environmental effects on the response of self-powered flux detectors in CANDU reactors

International Nuclear Information System (INIS)

Lynch, G.F.; Shields, R.B.; Joslin, C.W.

1976-01-01

Self-powered flux detectors are playing an increasingly important role in the control and safety systems of CANDU-type reactors. In this paper we report on recent experiments to determine how local reactor conditions affect the output signals from self-powered detectors with vanadium, platinum and cobalt emitters. The results are interpreted in terms of variations in the local neutron, γ-ray and electron fluxes. (author)

Self-powered in-core neutron detector assembly with uniform perturbation characteristics

International Nuclear Information System (INIS)

Todt, W.H.; Playfoot, K.C.

1979-01-01

Disclosed is a self-powered in-core neutron detector assembly in which a plurality of longitudinally extending self-powered detectors have neutron responsive active portions spaced along a longitudinal path. A low neutron absorptive extension extends from the active portions of the spaced active portions of the detectors in symmetrical longitudinal relationship with the spaced active detector portions of each succeeding detector. The detector extension terminates with the detector assembly to provide a uniform perturbation characteristic over the entire assembly length

Large-area self-powered neutron-detectors for neutron-flux measurements in HTRs. Status of developmental work

International Nuclear Information System (INIS)

Brixy, H.; Hecker, R.; Serpekian, T.; Benninghofen, G.; Serafin, N.; Spillekothen, H.G.

1982-06-01

The development is described of the large-area SPN-detector as an out of core power monitoring system. Gadolinium or cobalt was used as the emitter. Response functions of the gadolinium SPN-detector were found with regard to the reactor power, the effect of the gamma field, its short-term behaviour following reactor shutdown and long-term behaviour during reactor operation. It was shown that a detector of 0.1 mm emitter thickness can withstand an integral thermal neutron flux of 2.10 20 nvt almost without efficiency loss thus indicating that the large-area gadolinium SPN-detector is a suitable means for power monitoring in large HTGR's

Characterization of hybrid self-powered neutron detector under neutron irradiation

CERN Document Server

Nakamichi, M; Yamamura, C; Nakazawa, M; Kawamura, H

2000-01-01

To evaluate the irradiation behaviour of a blanket mock-up on in-pile functional test, it is necessary to measure the neutron flux change in the in-pile mock-up by a neutron detector, such as the self-powered neutron detector (SPND). With its small-sized emitter, which has high sensitivity and fast response time, SPND is an indispensable tool in order to measure the local neutron flux change. In the case of an in-pile functional test, it is necessary that response time is less than 1s and ratio of SPND output current is more than 0.3 of output current of SPND with Rh emitter. Therefore, a hybrid SPND with high sensitivity and fast response time was developed. This hybrid SPND used a hybrid emitter, i.e. Co cladded Pt-13%Rh.

Investigation of the response of improved self-powered neutron detectors

International Nuclear Information System (INIS)

Erk, S.

1982-01-01

The self-powered neutron detectors have been successfully employed for the most important parameters both for neutron flux and flux fluence determination. Their preference for such measurements due to their simplicity, convenience in use, rigidity, voluminal smallness and low price. However, self-powered neutron detectors depend on the type used, can only follow the neutron flux changes with a certain delay when they are compared to fission chambers which are thought to be the best detectors. In this thesis, a system has been proposed and considered carefully in order to speed up the response time, in another word, to correct the detector response to a level very near to fission chamber performance, a circuitry has been realized in the frame of principles so forth and applied to the experiments carried out in the TR-1 Reactor. Their positive results are presented. (author)

Characterization of hybrid self-powered neutron detector under neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Nakamichi, M. E-mail: masaru@oarai.jaeri.go.jp; Nagao, Y.; Yamamura, C.; Nakazawa, M.; Kawamura, H

2000-11-01

To evaluate the irradiation behaviour of a blanket mock-up on in-pile functional test, it is necessary to measure the neutron flux change in the in-pile mock-up by a neutron detector, such as the self-powered neutron detector (SPND). With its small-sized emitter, which has high sensitivity and fast response time, SPND is an indispensable tool in order to measure the local neutron flux change. In the case of an in-pile functional test, it is necessary that response time is less than 1s and ratio of SPND output current is more than 0.3 of output current of SPND with Rh emitter. Therefore, a hybrid SPND with high sensitivity and fast response time was developed. This hybrid SPND used a hybrid emitter, i.e. Co cladded Pt-13%R000.

Self-powered neutron detector of high sensitivity

International Nuclear Information System (INIS)

Brixy, H.; Spillekothen, H.G.; Benninghofen, G.; Serafin, N.

1983-01-01

A self-powered neutron detector is proposed, consisting of three concentrically arranged electrically conducting tubes; where the central one forms the emitter and the inner and outer ones form the collector and where the tubes are electrically insulated from each other by insulating material. The emitter consists of a material with a high absorption cross-section for thermal neutrons, particularly of gadolinium, and is provided with an auxiliary emitter layer on the inside or the outside. With suitable dimensions and material, the auxiliary emitter layer increases the yield of electrons. (orig./HP) [de

Initial absolute calibration factors for some neutron sensitive self-powered detectors

International Nuclear Information System (INIS)

Kroon, J.

1975-01-01

Self-powered flux detectors have found extensive use as monitoring devices in PWR (Pressurized Water Reactor) cores and CANDU (Canada Deuterium Uranium) type power reactors. The detectors measure fuel power distributions and indicate trip parameters for reactor control and safety requirements. Both applications demand accurate absolute initial calibration factors. Experimental results obtained in calibrating some neutron sensitive self-powered detectors is presented. (author)

Real time neutron flux monitoring using Rh self powered neutron detector

Energy Technology Data Exchange (ETDEWEB)

Juna, Byung Jin; Lee, Byung Chul; Park, Sang Jun; Jung, Hoan Sung [KAERI, Daejeon (Korea, Republic of)

2012-10-15

Rhodium (Rh) self powered neutron detectors (SPNDs) are widely used for on line monitoring of local neutron flux. Its signal is slower than the actual variation of neutron flux owing to a delayed {beta} decay of the Rh activation product, but real time monitoring is possible by solving equations between the neutron reaction rate in the detector and its signal. While the measuring system is highly reliable, the accuracy depends on the method solving the equations and accuracy of the parameters in the equations. The uncertain parameters are the contribution of gamma rays to the signal, and the branching ratios of Rh 104 and Rh 104m after the neutron absorption of Rh 103. Real time neutron flux monitoring using Rh SPNDs has been quite successful for neutron transmutation doping (NTD) at HANARO. We revisited the initial data used for the verification of a real time monitoring system, to refine algorithm for a better solution and to check the parameters for correctness. As a result, we suggest an effective way to determine the prompt parameter.

Real time neutron flux monitoring using Rh self powered neutron detector

International Nuclear Information System (INIS)

Juna, Byung Jin; Lee, Byung Chul; Park, Sang Jun; Jung, Hoan Sung

2012-01-01

Rhodium (Rh) self powered neutron detectors (SPNDs) are widely used for on line monitoring of local neutron flux. Its signal is slower than the actual variation of neutron flux owing to a delayed β decay of the Rh activation product, but real time monitoring is possible by solving equations between the neutron reaction rate in the detector and its signal. While the measuring system is highly reliable, the accuracy depends on the method solving the equations and accuracy of the parameters in the equations. The uncertain parameters are the contribution of gamma rays to the signal, and the branching ratios of Rh 104 and Rh 104m after the neutron absorption of Rh 103. Real time neutron flux monitoring using Rh SPNDs has been quite successful for neutron transmutation doping (NTD) at HANARO. We revisited the initial data used for the verification of a real time monitoring system, to refine algorithm for a better solution and to check the parameters for correctness. As a result, we suggest an effective way to determine the prompt parameter

Robust filtering for dynamic compensation of self-powered neutron detectors

International Nuclear Information System (INIS)

Peng, Xingjie; Li, Qing; Zhao, Wenbo; Gong, Helin; Wang, Kan

2014-01-01

Highlights: • Three dynamic compensation methods based on robust filtering theory are proposed. • Filter design problems are converted into linear matrix inequality problems. • Rhodium and Vanadium self-powered neutron detectors are used to validate the use of these three dynamic compensation methods. • The numerical simulation results show that all three methods can provide a reasonable balance between response speed and noise suppression. - Abstract: Self-powered neutron detectors (SPNDs), which are widely used in nuclear reactors to obtain core neutron flux distribution, are accurate at steady state but respond slowly to changes in neutron flux. Dynamic compensation methods are required to improve the response speed of the SPNDs and make it possible to apply the SPNDs for core monitoring and surveillance. In this paper, three digital dynamic compensation methods are proposed. All the three methods are based on the convex optimization framework using linear matrix inequalities (LMIs). The simulation results show that all three methods can provide a reasonable balance between response speed and noise suppression

Development of self-powered neutron detectors for neutron flux monitoring in HCLL and HCPB ITER-TBM

International Nuclear Information System (INIS)

Angelone, M.; Klix, A.; Pillon, M.; Batistoni, P.; Fischer, U.; Santagata, A.

2014-01-01

Highlights: •Self powered neutron detector (SPND) is attractive neutron monitor for TBM in ITER. •In hard neutron spectra (e.g. TBM) there is the need to optimize their response. •Three state-of-the-art SPNDs were tested using fast and 14 MeV neutrons. •The response of SPNDs is much lower than in thermal neutron flux. •FISPACT calculations performed to find out candidate materials in hard spectra. -- Abstract: Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum. This work represents the first attempt to study SPNDs as neutron flux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14 MeV neutrons at the Frascati neutron generator (FNG). The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutron flux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra

Development of self-powered neutron detectors for neutron flux monitoring in HCLL and HCPB ITER-TBM

Energy Technology Data Exchange (ETDEWEB)

Angelone, M., E-mail: maurizio.angelone@enea.it [Associazione ENEA-EURATOM sulla FusioneENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); Klix, A. [Association KIT-EURATOM, Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pillon, M.; Batistoni, P. [Associazione ENEA-EURATOM sulla FusioneENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); Fischer, U. [Association KIT-EURATOM, Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Santagata, A. [ENEA C.R. Casaccia, via Anguillarese Km. 1,300, 00100 Roma (Italy)

2014-10-15

Highlights: •Self powered neutron detector (SPND) is attractive neutron monitor for TBM in ITER. •In hard neutron spectra (e.g. TBM) there is the need to optimize their response. •Three state-of-the-art SPNDs were tested using fast and 14 MeV neutrons. •The response of SPNDs is much lower than in thermal neutron flux. •FISPACT calculations performed to find out candidate materials in hard spectra. -- Abstract: Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum. This work represents the first attempt to study SPNDs as neutron flux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14 MeV neutrons at the Frascati neutron generator (FNG). The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutron flux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra.

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Slack, J.; Norton, J.L.; Malkoske, G.R.

2003-01-01

MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments. The major application is in the health care industry where irradiators are used to sterilize single use medical products. These irradiators are designed and built by MDS Nordion and are used by manufacturers of surgical kits, gloves, gowns, drapes and other medical products. The irradiator is a large shielded room with a storage pool for the cobalt-60 sources. The medical products are circulated through the shielded room and exposed to the cobalt-60 sources. This treatment sterilizes the medical products which can then be shipped to hospitals for immediate use. Other applications for this irradiation technology include sanitisation of cosmetics, microbial reduction of pharmaceutical raw materials and food irradiation. The cobalt-60 sources are manufactured by MDS Nordion in their Cobalt Operations Facility in Kanata. More than 75,000 cobalt-60 sources for use in irradiators have been manufactured by MDS Nordion. The cobalt-60 sources are double encapsulated in stainless steel capsules, seal welded and helium leak tested. Each source may contain up to 14,000 curies. These sources are shipped to over 170 industrial irradiators around the world. This paper will focus on the MDS Nordion proprietary technology used to produce the cobalt-60 isotope in CANDU reactors. Almost 55 years ago MDS Nordion and Atomic Energy of Canada developed the process for manufacturing cobalt-60 at the Chalk River Labs, in Ontario, Canada. A cobalt-59 target was introduced into a research reactor where the cobalt-59 atom absorbed one neutron to become cobalt-60. Once the cobalt-60 material was removed from the research reactor it was encapsulated in stainless steel and seal welded using a Tungsten Inert Gas weld. The first cobalt-60 sources manufactured using material from the Chalk River Labs were used in cancer

Speciation analysis of cobalt in foods by high-performance liquid chromatography and neutron activation analysis

International Nuclear Information System (INIS)

Muto, Toshio; Koyama, Motoko

1994-01-01

A combined method by coupling high-performance liquid chromatography (HPLC, as a separation method) with neutron activation analysis (as a detection method) have been applied to the speciation analysis of cobalt in daily foods (e.g. egg, fish and milk). Cobalt species including free cobalt, vitamin B 12 and protein-bound cobalt were separated with a preparative HPLC and a centrifuge. Subsequently, the determination of cobalt in the separated species was made by neutron activation analysis. The results showed that the content of the total cobalt in the foods was found to lie in the range 0.4-11ng/g(0.4-11ppb) based on wet weight. The compositions of free cobalt, vitamin B 12 and protein-bound cobalt were ranged 16-43%, 55-73%, 2.3-17%, respectively. These experimental evidences suggest that the combination of HPLC and neutron activation analysis is expected to be a useful tool for speciation analysis of trace elements in biological as well as environmental materials. (author)

Analysis of the sensitivity concept of self-powered neutron detector (SPND)

International Nuclear Information System (INIS)

Moreira, O.; Lescano, H.

2012-01-01

Self powered neutron detectors (SPND) are widely used to monitor the neutron flux, either in nuclear as in irradiation facilities and medical treatments. However, the physical meaning of the parameter that is used to relate the detector signal (an electrical current) with the neutron flux, i.e., the sensitivity of the detector, has not been sufficiently analyzed. Since the definition of sensitivity, ε=i/φ is calculated for particular reactor conditions, i.e., for thermal neutrons at room temperature, it does not take into account the deviation originated from other conditions of temperature (above ambient), as found for example in nuclear power plants. In this work we calculated the microscopic cross section weighted with the neutron flux, defined in the usual way. This weighted microscopic cross section reveals the no proportionality between the absorption rate and the neutron flux, exhibiting the problem that the SPND current signal has to properly represent the neutron flux (author)

Self-powered neutron and gamma-ray flux detector

International Nuclear Information System (INIS)

Allan, C.J.; Shields, R.B.; Lynch, G.F.; Cuttler, J.M.

1980-01-01

A new type of self-powered neutron detector was developed which is sensitive to both the neutron and gamma-ray fluxes. The emitter comprises two parts. The central emitter core is made of materials that generate high-energy electrons on exposure to neutrons. The outer layer acts as a gamma-ray/electron converter, and since it has a higher atomic number and higher back-scattering coefficient than the collector, increases the net outflow or emmission of electrons. The collector, which is around the emitter outer layer, is insulated from the outer layer electrically with dielectric insulation formed from compressed metal-oxide powder. The fraction of electrons given off by the emitter that is reflected back by the collector is less than the fraction of electrons emitted by the collector that is reflected back by the emitter. The thickness of the outer layer needed to achieve this result is very small. A detector of this design responds to external reactor gamma-rays as well as to neutron capture gamma-rays from the collector. The emitter core is either nickel, iron or titanium, or alloys based on these metals. The outer layer is made of platinum, tantalum, osmium, molybdenum or cerium. The detector is particularly useful for monitoring neutron and gamma ray flux intensities in nuclear reactor cores in which the neutron and gamma ray flux intensities are closely proportional, are unltimately related to the fission rate, and are used as measurements of nuclear reactor power. (DN)

Comparison of dynamic compensation methods for delayed self-powered neutron detector

International Nuclear Information System (INIS)

In, Wang Kee; Kim, Joon Sung; Auh, Geun Sun; Yoon, Tae Young

1993-01-01

Dynamic compensation methods for rhodium self-powered neutron detector have been developed by Banda and Hoppe to compensate for the time delay associated with detector signals. The time delay is due to the decay of the neutron-activated rhodium and results in delayed detector response. Two digital dynamic compensation methods, were compared for step change of neutron flux in this paper. The inverse kinetics method gave slightly better response time and noise gain. However, the inverse kinetics method also showed overshooting of neutron flux for the step change. (Author)

Neutron diffraction studies on cobalt substituted BiFeO3

Science.gov (United States)

Ray, J.; Biswal, A. K.; Acharya, S.; Babu, P. D.; Siruguri, V.; Vishwakarma, P. N.

2013-02-01

A dilute concentration of single phase Cobalt substituted Bismuth ferrite, BiFe1-XCoXO3; (x=0, 0.02) is prepared by sol-gel auto combustion method. Room temperature neutron diffraction patterns show no change in the crystal and magnetic structure upon cobalt doping. The calculation of magnetic moments shows 3.848 μB for Fe+ and 2.85 μB for Co3+. The cobalt is found to be in intermediate spin state.

A study on the sensitivity depletion laws for rhodium self-powered neutron detectors

International Nuclear Information System (INIS)

Kim, Gil Gon

1999-02-01

The rhodium self-powered neutron detectors (SPND) in a reactor core provide the operator with the on-line 3-dimensional nuclear power distribution. The signal produced by rhodium SPND is interpreted into the local neutron flux by using a sensitivity depletion law and the local neutron flux is interpreted into the local power by using a power conversion factor. This work on the sensitivity depletion laws for rhodium self-powered neutron detectors (SPND) is performed to improve the uncertainty of the sensitivity depletion law used in ABB-CE reactors employing a rhodium SPND and to develop a calculational tool for providing the sensitivity depletion laws to interpret the signal of the newly designed rhodium SPND into the local neutron flux. The calculational tools for a time dependent neutron flux distribution in the rhodium emitter during depletion and for a time dependent beta escape probability that a beta generated in the emitter is escaped into the collector were developed. Due to the cost, the exposure to the radiation, and the longer fuel cycle, there is a strong incentive that the loading density of an in-core instrumentation is reduced and the lifetime of the detector is lengthened. These objectives can be achieved by reducing the uncertainty which is amplified as it depletes. The calculational tools above provide the sensitivity depletion law and show the reduction of the uncertainty to about 1 % in interpreting the signal into the local neutron flux compared to the method employed by ABB-CE. The reduction in the uncertainty of 1 % in interpreting the signal into the local neutron flux is equivalent to the reduction in the uncertainty of 1 % or more in interpreting the signal into the local power and to the extension of the lifetime of rhodium SPND to about 10 % as reported by ABB-CE

Dynamical model of computation of the rhodium self-powered neutron detector current

International Nuclear Information System (INIS)

Erben, O.; Slovacek, M.; Zerola, L.

1992-01-01

A model is presented for the calculation of the rhodium self-powered neutron detector current in dependence on the neutron flux density during reactor core transients. The total signal consists of a beta emission, prompt, and gamma component and a background signal. The model has been verified by means of experimental data obtained during measurements on the LVR-15 research reactor and at the Dukovany nuclear power plant. (author) 9 figs., 21 refs

Use of self powered neutron detectors in the IEA-R1 reactor

International Nuclear Information System (INIS)

Galo Rocha, F. del.

1989-01-01

A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

Kalman filtering for rhodium self-powered neutron detectors

International Nuclear Information System (INIS)

Kantrowitz, M.L.

1988-01-01

Rhodium self-powered neutron detectors are utilized in many pressurized water reactors to determine the neutronic behavior within the core. In order to compensate for the inherent time delay associated with the response of these detectors, a dynamic compensation algorithm is currently used in Combustion Engineering plants to reconstruct the dynamic flux signal which is being sensed by the rhodium detectors. This paper describes a new dynamic compensation algorithm, based on Kalman filtering, which improves on the noise gain and response time characteristics of the algorithm currently used, and offers the possibility of utilizing the proven rhodium detector based fixed in-core detector system as an integral part of advanced core control and/or protection systems

Kalman filtering of self-powered neutron detectors

International Nuclear Information System (INIS)

Kantrowitz, M.L.

1992-01-01

Pressurized water reactors employ a wide variety of in-core detectors to determine the neutronic behavior within the core. Among the detectors used are rhodium and vanadium self-powered detectors (SPDs), which are very accurate, but respond slowly to changes in neutron flux. This paper describes a new dynamic compensation algorithm, based on Kalman filtering, which converts delayed-responding rhodium and vanadium SPDs into prompt-responding detectors by reconstructing the dynamic flux signal sensed by the detectors from the prompt and delayed components. This conversion offers the possibility of utilizing current fixed in-core detector systems based on these delayed-responding detectors for core control and/or core protection functions without the need for fixed in-core detectors which are prompt-responding. As a result, the capabilities of current fixed in-core detector systems could be expanded significantly without a major hardware investment

Self-powered radiation detector

International Nuclear Information System (INIS)

Playfoot, K.C.; Bauer, R.F.; Goldstein, N.P.

1980-01-01

This invention relates to a self powered radiation detector requiring no excitation potential to generate a signal indicating a radiation flux. Such detectors comprise two electrically insulated electrodes, at a distance from each other. These electrodes are made of conducting materials having a different response for neutron and/or gamma ray radiation flux levels, as in nuclear power stations. This elongated detector generates an electric signal in terms of an incident flux of radiations cooperating with coaxial conductors insulated from each other and with different radiation reaction characteristics. The conductor with the greatest reaction to the radiations forms the central emitting electrode and the conductor with the least reaction to the radiations forms a tubular coaxial collecting electrode. The rhodium or cobalt tubular emitting electrode contains a ductile central conducting cable placed along the longitudinal axis of the detector. The latter is in high nickel steel with a low reaction to radiation [fr

Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

International Nuclear Information System (INIS)

Park, Donghwan; Kim, Youngae; Kim, Sungmin

2016-01-01

CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future

Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

Energy Technology Data Exchange (ETDEWEB)

Park, Donghwan; Kim, Youngae; Kim, Sungmin [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2016-10-15

CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future.

Design, construction and testing of a self-powered neutron detector

International Nuclear Information System (INIS)

Correa, R.F.

1987-01-01

The design, construction and testing of a self-powered neutron detector (SPN) and associated electronics are described. Several tests were performed giving information about dielectrical properties of detector and cable, gamma spectra induced in the detector through reactor irradiation, detector response as a function of neutron flux, current stability and reproductibility with the neutron flux. The gamma and neutron sensitivities were also evaluated, by means of thermoluminescent dosimeters and gold foils as references. The test results are presented and show that the detector response is reliable. The gamma and neutron sensitivities are in agreement with those found in the available literature. Nevertheless, a ceramic insulated cable should be employed for permanent use in a reactor. The tests were performed in a 100 KW TRIGA Mark I reactor at the Centro de Desenvolvimento da Tecnologia Nuclear of NUCLEBRAS, in Belo Horizonte, Brazil. (author) [pt

Design constrution and testing of a self-powered neutron detector

International Nuclear Information System (INIS)

Correa, R.F.

1987-01-01

The design, contruction and testing of a self-powered neutron detector (SPN) and associated electronics are described. Several tests were performed giving information about dielectrical properties od detector and cable, gamma spectra induced in the detector through reactor irradiation, detector response as a function of neutron flux, current stability and reproductibility with the neutron flux. The gamma and neutron sensitivities were also evaluated, by means of thermoluminescent dosimeters and gold foils as references. The test results are presented and show that the detector response is reliable. The gamma and neutron sensitivities are in agreement with those found in the available literature. Neverthe less, a ceramic insulated cable should be employed for permanent use in a reactor. The tests were perfomance in a 100 kW TRIGA Mark I reactor at the Centro de Desenvolvimento da Tecnologia Nuclear of NUCLEBRAS,in Belo Horizonte, Brazil. (Author) [pt

Self-powered in-core neutron detector assembly with uniform perturbation characteristics

International Nuclear Information System (INIS)

1981-01-01

An in-core neutron detector assembly consisting of a number of longitudinally extending self-powered detectors is described. The uniform mechanical structures and materials are placed symmetrically at each active detector portion thus ensuring that local perturbation factors are uniform. (U.K.)

Development of measuring system with self-powered neutron detectors for the LR-0 reactor

International Nuclear Information System (INIS)

Erben, O.; Horinek, K.; Szasz, Z.

1989-01-01

A measuring channel with self-powered detectors was developed for measuring neutron fluxs density in the reactor core. The measuring channel consists of a measuring probe with standard self-powered detectors of Soviet make, a signal pathway, a current/voltage converter and a measuring and recording unit. Neutron flux density in the LR-0 reactor core reaches a maximum of 10 13 m -2 s -1 . Experiments using the channel were carried out both in steady-state operation and after emergency shutdown of the reactor, this from power levels of 2,096 W and 1,830 W. The results of the experiments are tabulated and briefly analyzed. (Z.M.). 4 figs., 3 tabs., 5 refs

Application of nuclear pumped laser to an optical self-powered neutron detector

Science.gov (United States)

Yamanaka, N.; Takahashi, H.; Iguchi, T.; Nakazawa, M.; Kakuta, T.; Yamagishi, H.; Katagiri, M.

1996-05-01

A Nuclear Pumped Laser (NPL) using 3He/Ne/Ar gas mixture is investigated for a purpose of applying to an optical self-powered neutron detector. Reactor experiments and simulations on lasing mechanism have been made to estimate the best gas pressure and mixture ratios on the threshold input power density (or thermal neutron flux) in 3He/Ne/Ar mixture. Calculational results show that the best mixture pressure is 3He/Ne/Ar=2280/60/100 Torr and thermal neutron flux threshold 5×1012 n/cm2 sec, while the reactor experiments made in the research reactor ``YAYOI'' of the University of Tokyo and ``JRR-4'' of JAERI also demonstrate that excitational efficiency is maximized in a similar gas mixture predicted by the calculation.

An evaluated neutronic data file for elemental cobalt

Energy Technology Data Exchange (ETDEWEB)

Guenther, P.; Lawson, R.; Meadows, J.; Sugimoto, M.; Smith, A.; Smith, D.; Howerton, R.

1988-08-01

A comprehensive evaluated neutronic data file for elemental cobalt is described. The experimental data base, the calculational methods, the evaluation techniques and judgments, and the physical content are outlined. The file contains: neutron total and scattering cross sections and associated properties, (n,2n) and (n,3n) processes, neutron radiative capture processes, charged-particle-emission processes, and photon-production processes. The file extends from 10/sup /minus/5/ eV to 20 MeV, and is presented in the ENDF/B-VI format. Detailed attention is given to the uncertainties and correlations associated with the prominent neutron-induced processes. The numerical contents of the file have been transmitted to the National Nuclear Data Center, Brookhaven National Laboratory. 143 refs., 16 figs., 5 tabs.

Performance of self-powered neutron detectors in pressurized water reactors

International Nuclear Information System (INIS)

Warren, H.D.; Bozarch, D.P.

1977-01-01

A typical Babcock and Wilcox pressurized water reactor (PWR) contains 364 rhodium self-powered neutron detectors (SPNDs) and 52 background detectors. The detectors are inserted into the reactor core in 52 dry, multidetector assemblies. Each assembly contains seven SPNDs and one background detector. By mid-1977, eight B and W PWRs, each fitted with SPNDs, were in operation. Many of the SPNDs have operated successfully for more than four years. This paper describes the operational performance of the SPNDs and special tests conducted to improve that performance. Topics included are (1) insulation performance versus neutron dose to the SPND, (2) background signals in the leadwire region of the SPND, and (3) depletion of the SPND emitter versus absorbed neutron dose

Self-Powered Neutron Detector Calibration Using a Large Vertical Irradiation Hole of HANARO

OpenAIRE

Kim Myong-Seop; Park Byung-Gun; Kang Gi-Doo

2018-01-01

A calibration technology of the self-powered neutron detectors (SPNDs) using a large vertical irradiation hole of HANARO is developed. The 40 Rh-SPNDs are installed on the polycarbonate plastic support, and the gold wires with the same length as the effective length of the rhodium emitter of the SPND are also installed to measure the neutron flux on the SPND. They are irradiated at a low reactor power, and the SPND current is measured using the pico-ammeter. The external gamma-rays which affe...

The use of self-powered neutron detectors (SPN) to measure reactor power

International Nuclear Information System (INIS)

Hamidian, M.R.

1974-01-01

Due to the complicated design and technical difficulties of the conventional chamber method, neutron flux measurement between fuel rods is not possible. SPN detectors have a very simple design and a small size; therefore, they are suitable for the use inside the space available in the reactor core. The SPN detector consists of an emitter which emits particles by neutron absorption. The particles penetrate in an insulator and reach an outer collector. Among fission products, electrons might be produced by photoeffect or compton effect, internal conversion or pair production which, will also reach the collector. Cobalt and rhodium emitters have found practical application because of their fast response and sensitivity. The (n, γ) reaction in 103 Rh and 60 Co yields several isotopes and isomers which are discussed in the present report

Uncertainties in measuring trace amounts of cobalt and europium with low-flux neutron activation analysis

Directory of Open Access Journals (Sweden)

Burnham Steven

2017-01-01

Full Text Available Neutron activation analysis is widely used for identification of elements and their quantities even in trace amounts in the samples of almost any type. The challenges in detecting trace amounts of particular elements are often associated with the neutron flux produced at the research reactors. Low-flux neutron activation analysis usually presents the biggest challenge when analyzing trace quantities of elements with lower magnitude of radiative capture cross-sections. In this paper, we present the methodology and the quantified uncertainties associated with the detection of trace amounts of cobalt and europium, using as an example concrete aggregates. Recent growing interest is in improving structural concrete (increasing its strength but reducing its activation in nuclear power plant environments. Aside from buildings, structural concrete is also used as a biological shield in nuclear power plant that become radioactive after exposure to neutron flux. Due to radiative capture interactions, artificial radionuclides are generated to high enough concentrations that classify concrete as low-level radioactive waste at the time of the plant's decommissioning. Disposal of this concrete adds to the expense of nuclear power plant financing and its construction. Three radionuclides, 60Co, 152Eu, and 154Eu, account for 99 % of total residual radioactivity of nuclear power plant decommissioned concrete. IAEA document RS-G-1.7, Application of the Concepts of Exclusion, Exemption, and Clearance, specifies clearance levels of radionuclides specific activities: a specific activity lower than 0.1 Bqg-1 for 60Co and 152Eu, and 154Eu allows for a concrete to be recycled after decommissioning of the nuclear power plant. Therefore, low-flux neutron activation analysis is used to test the detection limits of trace elements in samples of cement, coarse, and fine concrete aggregates. These samples are irradiated at the University of Utah's 100 kW TRIGA Reactor at

Burnup Estimation of Rhodium Self-Powered Neutron Detector Emitter in VVER Reactor Core Using Monte Carlo Simulations

OpenAIRE

Khrutchinsky, А. А.; Kuten, S. A.; Babichev, L. F.

2011-01-01

Estimation of burn-up in a rhodium-103 emitter of self-powered neutron detector in VVER-1000 reactor core has been performed using Monte Carlo simulations within approximation of a constant neutron flux.

Activation of cobalt by neutrons from the Hiroshima bomb

International Nuclear Information System (INIS)

Kerr, G.D.; Dyer, F.F.; Emery, J.F.; Pace, J.V. III; Brodzinski, R.L.; Marcum, J.

1990-02-01

A study has been completed of cobalt activation in samples from two new locations in Hiroshima. The samples consisted of a piece of steel from a bridge located at a distance of about 1300 m from the hypocenter and pieces of both steel and concrete from a building located at approximately 700 m. The concrete was analyzed to obtain information needed to calculate the cobalt activation in the two steel samples. Close agreement was found between calculated and measured values for cobalt activation of the steel sample from the building at 700 m. It was found, however, that the measured values for the bridge sample at 1300 m were approximately twice the calculated values. Thus, the new results confirm the existence of a systematic error in the transport calculations for neutrons from the Hiroshima bomb. 52 refs., 32 figs., 16 tabs

Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

Energy Technology Data Exchange (ETDEWEB)

Vermeeren, L. [SCK/CEN, B-2400 Mol (Belgium)

2001-07-01

Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

International Nuclear Information System (INIS)

Vermeeren, L.

2001-01-01

Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

Thermal neutron flux measurement using self-powered neutron detector (SPND) at out-core locations of TRIGA PUSPATI Reactor (RTP)

Science.gov (United States)

Ali, Nur Syazwani Mohd; Hamzah, Khaidzir; Mohamad Idris, Faridah; Hairie Rabir, Mohamad

2018-01-01

The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND’s signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

Self-Powered Neutron Detector Qualification for Absolute On-Line In-Pile Neutron Flux Measurements in BR2

Science.gov (United States)

Vermeeren, L.; Wéber, M.

2003-06-01

A set of ten Self-Powered Neutron Detectors with Co, Rh and Ag emitters has been irradiated in several channels of the BR2 research reactor at SCK•CEN aiming at a comparison of their performance as thermal neutron flux detectors under various conditions. To allow for a correct interpretation of their signals, all detector sensitivity contributions (prompt and delayed) were calculated using a dedicated Monte Carlo model. The various contributions were also measured separately; the agreement between calculated and experimental data, including data from activation dosimetry, was excellent. Detailed neutron flux profiles were obtained from the SPND data, after correction for the finite detector lengths and for the slow response of delayed SPNDs.

Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

International Nuclear Information System (INIS)

Jammes, Christian; Filliatre, Philippe; Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan

2015-01-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

Energy Technology Data Exchange (ETDEWEB)

Jammes, Christian; Filliatre, Philippe [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan [Division of Applied Nuclear Physics, Uppsala University, SE-75120 Uppsala, (Sweden)

2015-07-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

Self-powered detectors for power reactors: an overview

International Nuclear Information System (INIS)

Ma, J.

2006-01-01

In this paper, Self-Powered Detectors (SPDs) for applications in nuclear power reactors have been reviewed. Based on their responses to radiation, these detectors can be divided into delayed response Self-Powered Neutron Detector (SPND), prompt response SPND and Self-Powered Gamma Detector (SPGD). The operational principles of these detectors are presented and their distinctive characteristics are examined accordingly. The analytical models and Monte Carlo method to calculate the responses of these detectors to neutron flux and external gamma rays are reviewed. The paper has also considered some related signal processing techniques, such as detector calibrations and detector signal compensations. Furthermore, a couple of failure modes have also been analyzed. Finally, applications of SPD in nuclear power reactors are summarized. (author)

Self-powered detectors for power reactors: an overview

Energy Technology Data Exchange (ETDEWEB)

Ma, J. [Univ. of Western Ontario, Dept. of Mechanical and Materials Engineering, London, Ontario (Canada)]. E-mail: jma64@uwo.ca

2006-07-01

In this paper, Self-Powered Detectors (SPDs) for applications in nuclear power reactors have been reviewed. Based on their responses to radiation, these detectors can be divided into delayed response Self-Powered Neutron Detector (SPND), prompt response SPND and Self-Powered Gamma Detector (SPGD). The operational principles of these detectors are presented and their distinctive characteristics are examined accordingly. The analytical models and Monte Carlo method to calculate the responses of these detectors to neutron flux and external gamma rays are reviewed. The paper has also considered some related signal processing techniques, such as detector calibrations and detector signal compensations. Furthermore, a couple of failure modes have also been analyzed. Finally, applications of SPD in nuclear power reactors are summarized. (author)

Neutron and gamma sensitivities of self-powered detectors: Monte Carlo modelling

Energy Technology Data Exchange (ETDEWEB)

Vermeeren, Ludo [SCK-CEN, Nuclear Research Centre, Boeretang 200, B-2400 Mol, (Belgium)

2015-07-01

This paper deals with the development of a detailed Monte Carlo approach for the calculation of the absolute neutron sensitivity of SPNDs, which makes use of the MCNP code. We will explain the calculation approach, including the activation and beta emission steps, the gamma-electron interactions, the charge deposition in the various detector parts and the effect of the space charge field in the insulator. The model can also be applied for the calculation of the gamma sensitivity of self-powered detectors and for the radiation-induced currents in signal cables. The model yields detailed information on the various contributions to the sensor currents, with distinct response times. Results for the neutron sensitivity of various types of SPNDs are in excellent agreement with experimental data obtained at the BR2 research reactor. For typical neutron to gamma flux ratios, the calculated gamma induced SPND currents are significantly lower than the neutron induced currents. The gamma sensitivity depends very strongly upon the immediate detector surroundings and on the gamma spectrum. Our calculation method opens the way to a reliable on-line determination of the absolute in-pile thermal neutron flux. (authors)

Responses of platinum, vanadium and cobalt self-powered flux detectors near simulated booster rods in a ZED-2 mockup of a Bruce reactor core

International Nuclear Information System (INIS)

French, P.M.; Shields, R.B.; Kroon, J.C.

1978-02-01

The static responses of Pt, V and Co self-powered detectors have been compared with copper-foil neutron activation profiles in reference and perturbed Bruce reactor core mockups assembled in the ZED-2 test reactor at Chalk River Nuclear Laboratories. The results indicate that the normalized response of each self-powered detector is an accurate measure of the thermal-neutron flux at locations greater than one lattice pitch from either a booster rod or the core boundary. They indicate that, in the Bruce booster/detector configuration, the normalized static Pt response overestimates the neutron flux by less than 3.5% upon full booster-rod insertion. (author)

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

Science.gov (United States)

Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

2017-07-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

Comparison study on in-core neutron detector for online neutron flux mapping of research and power reactor

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Mohd Idris Taib; Izhar Abu Husin; Nurfarhana Ayuni

2010-01-01

This paper presents the comparison study on In-Core neutron detector using for online flux mapping of Research and Power reactor. Technical description of in-core neutron also taken into consideration to identify the different characterization of neutron detector and describe on Self Power neutron detector (SPND) for online neutron flux mapping. Able to provide information on the neutron flux distribution and understand how in-core neutron detector are being used in nuclear power plant including to enable to state the principles of neutron detector. (author)

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Verma, V., E-mail: vasudha.verma@physics.uu.se [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Barbot, L.; Filliatre, P. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Hellesen, C. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Jammes, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Svärd, S. Jacobsson [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden)

2017-07-11

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment. - Highlights: • Studied possibility of using SPNDs as in-core detectors in SFRs. • Study done to detect local power profile changes when reactor is at nominal power. • SPND with a Pt-emitter gives measurable prompt current of the order of 600 nA/m. • Dominant proportion of prompt response is maintained throughout the operation. • Detector signal gives dynamic information on the power fluctuations.

Long-term performance of the CANDU-type of vanadium self-powered neutron detectors in NRU

Energy Technology Data Exchange (ETDEWEB)

Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)]. E-mail: leungt@aecl.ca

2007-07-01

The CANDU-type of in-core vanadium self-powered neutron flux detectors have been installed in NRU to monitor the axial neutron flux distributions adjacent to the loop fuel test sites since 1996. This paper describes how the thermal neutron fluxes were measured at two monitoring sites, and presents a method of correcting the vanadium burn-up effect, which can be up to 2 to 3% per year, depending on the detector locations in the reactor. It also presents the results of measurements from neutron flux detectors that have operated for over eight-years in NRU. There is good agreement between the measured and simulated neutron fluxes, to within {+-} 6.5%, and the long-term performance of the CANDU-type of vanadium neutron flux detectors in NRU is satisfactory. (author)

Cobalt

International Nuclear Information System (INIS)

Stolyarova, I.A.; Bunakova, N.Yu.

1983-01-01

The neutron-activation method for determining cobalt in rocks, polymetallic and iron ores and rockforming minerals at 2x10 -6 -5x10 -3 % content is developed. Cobalt determination is based on the formation under the effect of thermal neutrons of nuclear reactor of the 60 Co radioactive isotope by the 59 Co (n, γ) 60 Co reaction with radiation energy of the most intensive line of 1333 keV. Cobalt can be determined by the scheme of the multicomponent analysis from the sample with other elements. Co is determined in the solution after separation of all determinable by the scheme elements. The 60 Co intensity is measured by the mUltichannel gamma-spectrometer with Ge(Li)-detector

EPRTM Reactor neutron instrumentation

International Nuclear Information System (INIS)

Pfeiffer, Maxime; SALA, Stephanie

2013-06-01

The core safety during operation is linked, in particular, to the respect of criteria related to the heat generated in fuel rods and to the heat exchange between the rods and the coolant. This local power information is linked to the power distribution in the core. In order to evaluate the core power distribution, the EPR TM reactor relies on several types of neutron detectors: - ionization chambers located outside the vessel and used for protection and monitoring - a fixed in-core instrumentation based on Cobalt Self Powered Neutron Detectors used for protection and monitoring - a mobile reference in-core instrumentation based on Vanadium aero-balls This document provides a description of this instrumentation and its use in core protection, limitation, monitoring and control functions. In particular, a description of the detectors and the principles of their signal generation is supplied as well as the description of the treatments related to these detectors in the EPR TM reactor I and C systems (including periodical calibration). (authors)

Self-Powered Neutron Detector Calibration Using a Large Vertical Irradiation Hole of HANARO

Directory of Open Access Journals (Sweden)

Kim Myong-Seop

2018-01-01

Full Text Available A calibration technology of the self-powered neutron detectors (SPNDs using a large vertical irradiation hole of HANARO is developed. The 40 Rh-SPNDs are installed on the polycarbonate plastic support, and the gold wires with the same length as the effective length of the rhodium emitter of the SPND are also installed to measure the neutron flux on the SPND. They are irradiated at a low reactor power, and the SPND current is measured using the pico-ammeter. The external gamma-rays which affect the SPND current response are analyzed using the Monte Carlo simulation for various irradiation conditions in HANARO. It is confirmed that the effect of the external gamma-rays to the SPND current is dependent on the reactor characteristics, and that it is affected by materials around the detector. The current signals due to the external gamma-rays can be either positive or negative, in that the net flow of the current may be either in the same or the opposite direction as the neutron-induced current by the rhodium emitter. From the above procedure, the effective calibration methodology of multiple SPNDs using the large hole of HANARO is developed. It could be useful for the calibration experiment of the neutron detectors in the research reactors.

Self-Powered Neutron Detector Calibration Using a Large Vertical Irradiation Hole of HANARO

Science.gov (United States)

Kim, Myong-Seop; Park, Byung-Gun; Kang, Gi-Doo

2018-01-01

A calibration technology of the self-powered neutron detectors (SPNDs) using a large vertical irradiation hole of HANARO is developed. The 40 Rh-SPNDs are installed on the polycarbonate plastic support, and the gold wires with the same length as the effective length of the rhodium emitter of the SPND are also installed to measure the neutron flux on the SPND. They are irradiated at a low reactor power, and the SPND current is measured using the pico-ammeter. The external gamma-rays which affect the SPND current response are analyzed using the Monte Carlo simulation for various irradiation conditions in HANARO. It is confirmed that the effect of the external gamma-rays to the SPND current is dependent on the reactor characteristics, and that it is affected by materials around the detector. The current signals due to the external gamma-rays can be either positive or negative, in that the net flow of the current may be either in the same or the opposite direction as the neutron-induced current by the rhodium emitter. From the above procedure, the effective calibration methodology of multiple SPNDs using the large hole of HANARO is developed. It could be useful for the calibration experiment of the neutron detectors in the research reactors.

An investigation of models of rhodium emitter used in self-powered neutron detector

International Nuclear Information System (INIS)

Borisenko, V.I.; Piontkovskij, Yu.F.; Goranchuk, V.V.

2017-01-01

he paper presents the results of MCNP simulation of the self-powered neutron detector (SPND) signal formation as a result of emitter nuclei activation under the irradiation with neutrons generated in the fuel assemblies. To account for the non-uniformity of emitter burnup along the radius, its model was divided radially into 10 layers of equal thickness. It has been shown that the main contribution of about 88 % of SPND signal is provided by the four peripheral emitter layers. The contribution of different parts of emitter to the SPND signal formation throughout the lifetime of the SPND in the In-Core Monitoring System was found. Simulation results allow us to determine the SPND signal when the spectral characteristics of the neutron flux at the detector location change during the fuel campaign. The study has investigated and proposed a SPND model with the higher neutron sensitivity even though a smaller amount of expensive rhodium is used.

Industrial development of neutron detectors, fission chambers, self powered detectors, ionization chambers

International Nuclear Information System (INIS)

Constans, H.; Coville, P.; Guerre, J.

1975-01-01

Reactor control requires the determination of neutron flux at all times. The needed characteristics lead to use of several types of detectors: boron lined counters, boron lined ionization chambers, fission ionization chambers and self powered detectors. The principle of the reaction involved the fabrication requirements, the different modes of utilization and the characteristics obtained are examined for each detector. The problem of electric connections in the active area has been solved by developing ''integrated cables'' [fr

Self-subunit swapping occurs in another gene type of cobalt nitrile hydratase.

Directory of Open Access Journals (Sweden)

Yi Liu

Full Text Available Self-subunit swapping is one of the post-translational maturation of the cobalt-containing nitrile hydratase (Co-NHase family of enzymes. All of these NHases possess a gene organization of , which allows the activator protein to easily form a mediatory complex with the α-subunit of the NHase after translation. Here, we discovered that the incorporation of cobalt into another type of Co-NHase, with a gene organization of , was also dependent on self-subunit swapping. We successfully isolated a recombinant NHase activator protein (P14K of Pseudomonas putida NRRL-18668 by adding a Strep-tag N-terminal to the P14K gene. P14K was found to form a complex [α(StrepP14K(2] with the α-subunit of the NHase. The incorporation of cobalt into the NHase of P. putida was confirmed to be dependent on the α-subunit substitution between the cobalt-containing α(StrepP14K(2 and the cobalt-free NHase. Cobalt was inserted into cobalt-free α(StrepP14K(2 but not into cobalt-free NHase, suggesting that P14K functions not only as a self-subunit swapping chaperone but also as a metallochaperone. In addition, NHase from P. putida was also expressed by a mutant gene that was designed with a order. Our findings expand the general features of self-subunit swapping maturation.

Use of FET in automatic scanning of measurements using thermocouples and self-powered neutron detectors

International Nuclear Information System (INIS)

Plaige, Yves.

1977-01-01

Advantages lying in using FET switches in the relays of multiplexing systems are shown with two examples of application. Their performance as regard fast reliable operation are used in temperature measurement scanning inside nuclear reactors. As for current measurements using self-powered neutron detectors, the weak leakage currents of said switches ( [fr

Development of high sensitivity transimpedance amplifier module for self powered neutron detectors

International Nuclear Information System (INIS)

Khan, T.K.; Tamboli, P.K.; Antony, J.; Balasubramanian, R.; Agilandaeswari, K.; Pramanik, M.

2010-01-01

This paper describes design and development of a Transimpedance Amplifier for amplification of very low current from in core Self Powered Neutron Detectors (SPND). Measurement of neutron flux is very important for operation, control and protection of Nuclear Power Plant (NPP). SPND is used to measure Reactor incore flux/power. Based on sensitivity of emitter material used in SPND, pitch length and neutron flux (power level); the current output from SPND varies from few pA to few μA. The described amplifier is suitable to use for this current range. The amplifier provides a very high gain using a resistive T network feedback topology. The amplifier is designed in two stages using ultra low bias current FET OPAMPs. Design of Transimpedance amplifier is carefully done to include ultra low input bias current, low offset voltage and noise. The amplifier has in built test facility for calibration and on line test facility for measurement of insulation resistance (IR). The amplifier module has on board isolated DC-DC converter circuit complying MIL/STD/461C/D which generate isolated +/-15V and +12V supply to provide parameter to parameter ground isolation and independence among each module/signal.The output from the amplifier is 0V to 6V for 0 to 150%FP. The design is simulated in computer and amplifier used at TAPS-3 was modified as per new design and has been tested at TAPS-3 site. The amplifier performed satisfactorily. The results showed that the IR measurement technique adopted in the design can tolerate lower IR of SPND in existing design. (author)

Comparison of Thermal Neutron Flux Measured by Uranium 235 Fission Chamber and Rhodium Self-Powered Neutron Detector in MTR

International Nuclear Information System (INIS)

Fourmentel, D.; Filliatre, P.; Barbot, L.; Villard, J.-F.; Lyoussi, A.; Geslot, B.; Malo, J.-Y.; Carcreff, H.; Reynard-Carette, C.

2013-06-01

Thermal neutron flux is one of the most important nuclear parameter to be measured on-line in Material Testing Reactors (MTRs). In particular two types of sensors with different physical operating principles are commonly used: self-powered neutron detectors (SPND) and fission chambers with uranium 235 coating. This work aims to compare on one hand the thermal neutron flux evaluation given by these two types of sensors and on the other hand to compare these evaluations with activation dosimeter measurements, which are considered as the reference for absolute neutron flux assessment. This study was conducted in an irradiation experiment, called CARMEN-1, performed during 2012 in OSIRIS reactor (CEA Saclay - France). The CARMEN-1 experiment aims to improve the neutron and photon flux and nuclear heating measurements in MTRs. In this paper we focus on the thermal neutron flux measurements performed in CARMEN-1 experiment. The use of fission chambers to measure the absolute thermal neutron flux in MTRs is not very usual. An innovative calibration method for fission chambers operated in Campbell mode has been developed at the CEA Cadarache (France) and tested for the first time in the CARMEN-1 experiment. The results of these measurements are discussed, with the objective to measure with the best accuracy the thermal neutron flux in the future Jules Horowitz Reactor. (authors)

Device for neutron flux monitoring in IEA-R1 reactor using rhodium self powered neutron detector; Dispositivo de mapeamento de fluxo de neutron atraves do SPN/Rodio no IEA-R1

Energy Technology Data Exchange (ETDEWEB)

Ricci Filho, Walter; Fernando, Alberto de Jesus; Jerez, Rogerio; Tondin, Julio B.M.; Pasqualetto, Hertz [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

2000-07-01

The IEA-R1 reactor has undergone a modernization tio increase its operating power to 5 MW, in order to allow a more efficient production of radioisotopes. The objective of this work is to provide the reactor with flux monitoring device using a rhodium self powered neutron detector. Self powered detectors are rugged miniature devices with are increasingly being used for fixed in core reactor monitoring both for safety purposes and flux mapping. The work presents the results obtained with Rhodium-SPND in several irradiation position inside the reactor core. (author)

New avenues in cobalt-60 production at Ontario Hydro

International Nuclear Information System (INIS)

Mylvaganam, C.K.; Ronchka, R.A.

1990-01-01

Ontario Hydro produces cobalt-60 in the control rods of twelve power reactors. These reactors have a typical flux of 2 x 10 14 neutrons/cm 2 /s, making them efficient producers of cobalt-60. Current annual production is 45 million curies. Since the primary function of these reactors is the production of electricity, their flexibility to meet the needs of commercial cobalt production by the control rod route is limited. Ontario Hydro is therefore developing innovative production techniques, making use of the CANDU reactor's unique ability to be fuelled on-power. These techniques will enable production to better respond to the market's requirements for quantity and specific activity. As it is supplementary to control rod production, annual supply could potentially reach 165 million curies. (author)

A study on the sensitivity of self-powered neutron detectors(SPNDs)

International Nuclear Information System (INIS)

Lee, Wan No

1997-02-01

Self-Powered Neutron Detectors(SPND) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. While they have several advantages such as small size, low cost, and relatively simple electronics required in conjunction with those usage, they have some intrinsic problems of the low level of output current, a slow response time, the rapid change of sensitivity which makes it difficult to use for a long term. In this paper, Monte Carlo simulation is accomplished to calculate the escape probability as a function of the birth position of emitted beta particle for a geometry of rhodium-based SPNDs. A simple numerical method calculates the initial generation rate of beta particles and the change of generation rate due to rhodium burn-up. Using the simulation result, the burn-up profile of rhodium number density and the neutron sensitivity are calculated as a function of burn-up time in reactors. The sensitivity of the SPND decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. In addition, for improvement of some properties of rhodium-based SPNDs which are currently used, this paper presents a new material and modified geometry. From searching nuclear data, Ag 109 is chosen as a replacing material for rhodium. Silver has a low neutron absorption cross-section and a high beta energy and a low density when it is compared with rhodium. The sensitivity and the density change of silver as a function of burn-up are calculated using this method. Also, this paper compares the initial sensitivity of a solid type with its of a tube type. The initial sensitivity is increased with the new material and the tube type. Silver is also found to be used for longer time than rhodium. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long term usage

Cadmium-emitter self-powered thermal neutron detector performance characterization & reactor power tracking capability experiments performed in ZED-2

Energy Technology Data Exchange (ETDEWEB)

LaFontaine, M.W., E-mail: physics@execulink.com [LaFontaine Consulting, Kitchener, Ontario (Canada); Zeller, M.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Nielsen, K. [Royal Military College of Canada, SLOWPOKE-2 Reactor, Kingston, Ontario (Canada)

2014-07-01

Cadmium-emitter self-powered thermal neutron flux detectors (SPDs), are typically used for flux monitoring and control applications in low temperature, test reactors such as the SLOWPOKE-2. A collaborative program between Atomic Energy of Canada, academia (Royal Military College of Canada (RMCC)) and industry (LaFontaine Consulting) was initiated to characterize the incore performance of a typical Cd-emitter SPD; and to obtain a definitive measure of the capability of the detector to track changes in reactor power in real time. Prior to starting the experiment proper, Chalk River Laboratories' ZED-2 was operated at low power (5 watts nominal) to verify the predicted moderator critical height. Test measurements were then performed with the vertical center of the SPD emitter positioned at the vertical mid-plane of the ZED-2 reactor core. Measurements were taken with the SPD located at lattice position L0 (near center), and repeated at lattice position P0 (in D{sub 2}O reflector). An ionization chamber (part of the ZED-2 control instrumentation) monitored reactor power at a position located on the south side of the outside wall of the reactor's calandria. These experiments facilitated measurement of the absolute thermal neutron sensitivity of the subject Cd-emitter SPD, and validated the power tracking capability of said SPD. Procedural details of the experiments, data, calculations and associated graphs, are presented and discussed. (author)

Implantable self-powered detector for on-line determination of neutron flux in patients during NCT treatment.

Science.gov (United States)

Miller, M E; Mariani, L E; Gonçalves-Carralves, M L Sztejnberg; Skumanic, M; Thorp, S I

2004-11-01

A novel system to determine thermal neutron flux in real time during NCT treatments was developed in the National Atomic Energy Commission of Argentina. The system is based on a special self-powered detector that can be implanted in patients owing to its small size and biocompatibility. High voltage is not required to operate this kind of detectors, which is a considerable advantage in terms of medical uses. By choosing the appropriate materials, it was possible to obtain a prototype with thermal neutron sensitivity providing for an adequate signal level in typical NCT thermal fluxes. It was also possible to minimize gamma response in order to neglect its contribution.

Implantable self-powered detector for on-line determination of neutron flux in patients during NCT treatment

Energy Technology Data Exchange (ETDEWEB)

Miller, M.E. E-mail: miller@cae.cnea.gov.ar; Mariani, L.E.; Sztejnberg Goncalves-Carralves, M.L.; Skumanic, M.; Thorp, S.I

2004-11-01

A novel system to determine thermal neutron flux in real time during NCT treatments was developed in the National Atomic Energy Commission of Argentina. The system is based on a special self-powered detector that can be implanted in patients owing to its small size and biocompatibility. High voltage is not required to operate this kind of detectors, which is a considerable advantage in terms of medical uses. By choosing the appropriate materials, it was possible to obtain a prototype with thermal neutron sensitivity providing for an adequate signal level in typical NCT thermal fluxes. It was also possible to minimize gamma response in order to neglect its contribution.

An analysis of cobalt irradiation in CANDU 6 reactor core

International Nuclear Information System (INIS)

Gugiu, E.D.; Dumitrache, I.

2003-01-01

In CANDU reactors, one has the ability to replace the stainless steel adjuster rods with neutronically equivalent Co assemblies with a minimum impact on the power plant safety and efficiency. The 60 Co produced by 59 Co irradiation is used extensively in medicine and industry. The paper mainly describes some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronically equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and the heating of the irradiated cobalt rods are performed using the Monte Carlo codes MCNP5 and MONTEBURNS2.1. The 60 Co activity and heating evaluations are closely related to the neutronics computations and to the density evolution of cobalt isotopes during assumed in-core irradiation period. Unfortunately, the activities of these isotopes could not be evaluated directly using the burn-up capabilities of the MONTEBURNS code because of the lack of their neutron cross-section from the MCNP5 code library. Additional MCNP5 runs for all the cobalt assemblies have been done in order to compute the flux-spectrum, the 59 Co and the 60 Co radiative capture reaction rates in the adjusters. The 60m Co cross-section was estimated using the flux-spectrum and the ORIGEN2.1 code capabilities THERM and RES. These computational steps allowed the evaluation of the one-group cross-section for the radiative capture reactions of cobalt isotopes. The values obtained replaced the corresponding ones from the ORIGEN library, which have been estimated using the flux-spectrum specific to the fuel. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters. (authors)

Cold neutron source with self-regulation

International Nuclear Information System (INIS)

Kawai, T.

2003-01-01

A way to increase the cold neutron flux is to cool moderator from where cold neutrons are extracted. Although various kinds of cooling system are considered, the closed thermo-siphon cooling system is adopted in many institutes. The notable feature of this system is to be able to keep the liquid level stable in the moderator cell against thermal disturbances, by using self-regulation, which allows a stable supply of cold neutrons. The main part of the closed thermo-siphon consists of a condenser, a moderator transfer tube and moderator cell, which is called the hydrogen cold system. When an extra heat load is applied to the hydrogen cold system having no flow resistance in a moderator transfer tube, the system pressure rises by evaporation of liquid hydrogen. Then the boiling point of hydrogen rises. The liquefaction capacity of the condenser is increasing with a rise of temperature, because a refrigerating power of the helium refrigerator increases linearly with temperature rise of the system. Therefore, the effect of thermal heat load increase is compensated and cancelled out. The closed thermo-siphon has this feature generally, when the moderator transfer tube is designed to be no flow resistance. The report reviews the concept of self-regulation, and how to design and construct the cold neutron source with self-regulation. (author)

A study on the sensitivity of self-powered neutron detectors (SPNDs) and a new proposal

International Nuclear Information System (INIS)

Lee, Wan No; Cho, Gyu Seong

1997-01-01

Self-Powered Neutron Detectors (SPNDs) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. In this paper, Monte Carlo simulation is accomplished to calculate the escape probability of beta particle as a function of their birth position for the typical geometry of rhodium-based SPNDs. Also, a simple numerical method calculates the initial generation rate of beta particles and the change of generation rate due to rhodium burn-up. Using the simulation and the numerical method, the burn-up profile of rhodium density and the neutron sensitivity are calculated as a function of burn-up time in the reactor. The sensitivity of the SPNDs decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. In addition, for improvement of some properties of rhodium-based SPNDs which are currently used, this paper presents a new material. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long term usage

Determination of Cobalt in Seawater Using Neutron Activation Analysis after Preconcentration by Adsorption onto γ-MnO2 Nanomaterial

Directory of Open Access Journals (Sweden)

Van-Phuc Dinh

2018-01-01

Full Text Available The γ-MnO2 nanomaterial has been used to adsorb cobalt in the seawater at Phan Thiet City, Binh Thuan Province, Vietnam. Its concentration is determined by using the neutron activation analysis (NAA method at the Dalat nuclear research reactor. Factors affecting the uptake of cobalt on the γ-MnO2 material such as the pH, adsorption time, and initial cobalt(II concentration are investigated. The irradiated experiment data are calculated using the K0-Dalat program. The results obtained show that the trace dissolved cobalt in Phan Thiet seawater is found equal to 0.25 ± 0.04 μg/L (n=5, P=95% with the adsorption efficiency being higher than 95% (n=4, P=95%.

A study on the sensitivity of self-powered neutron detectors (SPNDs)

Science.gov (United States)

Lee, Wanno; Cho, Gyuseong; Kim, Kwanghyun; Kim, Hee Joon; choi, Yuseon; Park, Moon Chu; Kim, Soongpyung

2001-08-01

Self-powered neutron detectors (SPNDs) are widely used in reactors to monitor neutron flux, while they have several advantages such as small size, and relatively simple electronics required in conjunction with those usages, they have some intrinsic problems of the low level of output current-a slow response time and the rapid change of sensitivity-that make it difficult to use for a long term. Monte Carlo simulation was used to calculate the escape probability as a function of the birth position of emitted beta particle for geometry of rhodium-based SPNDs. A simple numerical method calculated the initial generation rate of beta particles and the change of generation rate due to rhodium burnup. Using results of the simulation and the simple numerical method, the burnup profile of rhodium number density and the neutron sensitivity were calculated as a function of burnup time in reactors. This method was verified by the comparison of this and other papers, and data of YGN3.4 (Young Gwang Nuclear plant 3, 4) about the initial sensitivity. In addition, for improvement of some properties of rhodium-based SPNDs, which are currently used, a modified geometry is proposed. The proposed geometry, which is tube-type, is able to increase the initial sensitivity due to increase of the escape probability. The escape probability was calculated by changing the thickness of the insulator and compared solid-type with tube-type about each insulator thickness. The method used here can be applied to the analysis and design of other types of SPNDs.

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments.

Science.gov (United States)

Miller, Marcelo E; Sztejnberg, Manuel L; González, Sara J; Thorp, Silvia I; Longhino, Juan M; Estryk, Guillermo

2011-12-01

A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comisión Nacional de Energía Atómica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Local mixed-field thermal neutron sensitivities and global thermal and mixed

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments

Energy Technology Data Exchange (ETDEWEB)

Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.; Thorp, Silvia I.; Longhino, Juan M.; Estryk, Guillermo [Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429, Argentina and CONICET, Av. Rivadavia 1917, Ciudad de Buenos Aires 1033 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina)

2011-12-15

Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments

International Nuclear Information System (INIS)

Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.; Thorp, Silvia I.; Longhino, Juan M.; Estryk, Guillermo

2011-01-01

Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and global

EFTF cobalt test assembly results

International Nuclear Information System (INIS)

Rawlins, J.A.; Wootan, D.W.; Carter, L.L.; Brager, H.R.; Schenter, R.E.

1988-01-01

A cobalt test assembly containing yttrium hydride pins for neutron moderation was irradiated in the Fast Flux Test Facility during Cycle 9A for 137.7 equivalent full power days at a power level fo 291 MW. The 36 test pins consisted of a batch of 32 pins containing cobalt metal to produce Co-60, and a set of 4 pins with europium oxide to produce Gd-153, a radioisotope used in detection of the bone disease Osteoporosis. Post-irradiation examination of the cobalt pins determined the Co-60 produced with an accuracy of about 5 %. The measured Co-60 spatially distributed concentrations were within 20 % of the calculated concentrations. The assembly average Co-60 measured activity was 4 % less than the calculated value. The europium oxide pins were gamma scanned for the europium isotopes Eu-152 and Eu-154 to an absolute accuracy of about 10 %. The measured europium radioisotpe anc Gd-153 concentrations were within 20 % of calculated values. In conclusion, the hydride assembly performed well and is an excellent vehicle for many Fast Flux Test Facility isotope production applications. The results also demonstrate that the calculational methods developed by the Westinghouse Hanford Company are very accurate. (author)

Small angle neutron and x-ray scattering studies of self-assembled nano structured materials

International Nuclear Information System (INIS)

Choi, Sung Min

2009-01-01

Full text: Small angle neutron and x-ray scattering are very powerful techniques to investigate nano structured materials. In this presentation, examples of nano structured materials investigated by neutron and x-ray scattering will be presented. Part I: The unique anisotropic physical properties of columnar discotic liquid crystals (DLCs) have attracted considerable interest for their potential applications as electronic devices. For many practical applications, however, it is crucial to obtain uniaxially oriented and highly ordered columnar superstructures of DLC molecules covering macroscopic area. Here, we present a simple and straight-forward approach to fabricate uniaxially oriented and highly ordered columnar superstructures of cobalt octa(n-decylthio) porphyrazine (CoS 1 0), a discotic supra-molecule, in bulk and on substrates [1] over a macroscopic length scale, utilizing an applied magnetic field and the interaction of CoS 1 0 with an OTS-functionalized substrate. The details of the oriented and ordered columnar nano-structures are investigated by SANS and GISAXS. Part II: Self-assembly of one-dimensional (1D) nanoparticles with metallic or semiconducting properties into highly ordered superstructures using various interactions has been of great interest as a route towards materials with new functionalities. Here, we report a new phase diagram of negatively charged 1D nanoparticle (cROD) and cationic liposome (CL) complexes in water which exhibit three different highly ordered phases [2]. Small angle neutron and x-ray scattering measurements show that the cROD-CL complexes exhibit three different highly ordered phases, intercalated lamellar, doubly intercalated lamellar and centered rectangular phases, depending on particle curvature and electrostatic interactions. The new phase diagram can be used to understand and design new highly ordered self-assemblies of 1D nanoparticles in soft matter which provide new functionalities. (author)

Sulfur, cobalt and europium activation from the A-bombs in Japan

International Nuclear Information System (INIS)

Gritzner, M.L.; Egbert, S.D.; Woolson, W.A.

1987-01-01

Comparisons are made between calculations and measurements of in-situ sulfur, cobalt, and europium activation by neutrons at Hiroshima and Nagasaki to check the validity of the neutron sources and transport air-over-ground used in the new Dosimetry System 1986 (DS86). Europium and cobalt are activated by epithermal/thermal neutrons. The sulfur activation has an effective threshold around 2 MeV. Comparisons of sulfur activation at Hiroshima, which necessitate considering the tilt and heading of the cylindrically shaped Little Boy bomb, confirm the high energy neutron portion of the source relevant to sulfur activation. The europium and cobalt activation calculations tend to be higher than the measurements for ground ranges within 300 m and lower than the measurements beyond 700 m. Thus, the cobalt and europium measurements do not validate the source or the transport, embodied in DS87, for those neutrons relevant to europium and cobalt activation. Presently, it is unresolved as to whether the discrepancy can be explained as a problem in the measurements, in their interpretation, or as a problem in the calculations due to the source neutrons and/or their transport from source to the sample site

Design and fabrication of self-powered in-core neutron flux monitor assembly

International Nuclear Information System (INIS)

Chung, M.K.; Cho, S.W.; Kang, H.D.; Cho, K.K.; Cho, B.S.; Kang, S.S.

1980-01-01

This is the final report on the prototypical fabrication of an in-core neutron flux monitor detector assembly for a specific power reactor conducted by KAERI from July 1, 1978 to December 31, 1979. It is well known that power reactors require a large number of in-core neutron flux detector for reactor regulation and the structures of detector assemblies are different from reactor to reactor. Therefore, from the nature of this project, it should be noted here that the target model of the prototypical farbrication of an in-core neutron flux monitor detector assembly is a VFD-2 System for Wolsung CANDU. It is concluded that fabrication of in-core neutron flux monitor detector assembly for CANDU reactor is technically feasible and will bring economical benefit as much as 50 % of the unit price if they are fabricated in Korea by using partially materials which are available from local market. (author)

Self-powered neutron and γ-ray flux detector

International Nuclear Information System (INIS)

Allan, C.J.

1983-01-01

According to the invention there is provided a self-powered neutron and γ-ray flux detector, comprising: a) an emitter core wire; b) an emitter outer layer around the core wire and of different metal thereto; c) a metal collector around the emitter core wire and the emitter outer layer; and d) dielectric insulation electrically insulating the emitter core wire and the emitter outer layer from the metal collector. The improvement comprises: a) the overall diameter of the emitter core wire and the emitter outer layer is at least of the order of 0.4 mm in diameter; b) the emitter outer layer covers only of the order of l0 percent of the order of 90 percent of the emitter core wire surface area and comprises at least one band around the emitter core wire and is of a thickness in the range of the order 0.02 mm to of the order of 0.07 mm; and c) the metal of the emitter core wire, the metal of the emitter outer layer, the metal of the metal collector, the overall diameter of the emitter core wire and the emitter outer layer and the surface area of the emitter core wire that is covered by the emitter outer layer are selected so that the detector has a prompt fraction in the range of the order of 90 percent to of the order of 96 percent and has a dynamic response which substantially matches the dynamic response of the power in the fuel of the nuclear reactor in which the detector is to be used

Application of the neutron noise analysis technique in nuclear power plants

International Nuclear Information System (INIS)

Lescano, Victor H.; Wentzeis, Luis M.

1999-01-01

Using the neutron noise analysis in nuclear power plants, and without producing any perturbation in the normal operation of the plant, information of the vibration state of the reactor internals and the behavior of the operating conditions of the reactor primary circuit can be obtained. In Argentina, the neutron noise analysis technique is applied in customary way in the nuclear power plants Atucha I and Embalse. A database was constructed and vibration frequencies corresponding to different reactor internals were characterized. Reactor internals with particular mechanical vibrations have been detected and localized. In the framing of a cooperation project between Argentina and Germany, we participated in the measurements, analysis and modelisation, using the neutron noise technique, in the Obrigheim and Gundremmingen nuclear power plants. In the nuclear power plant Obrigheim (PWR, 350 M We), correlations between the signals measured from self-power neutron detectors and accelerometers located inside the reactor core, were made. In the nuclear power plant Gundremmingen (BWR, 1200 M We) we participated in the study of a particular mechanical vibration detected in one of the instrumentation tube. (author)

Rhodium self-powered detector for monitoring neutron fluence, energy production, and isotopic composition of fuel

International Nuclear Information System (INIS)

Sokolov, A.P.; Pochivalin, G.P.; Shipovskikh, Yu.M.; Garusov, Yu.V.; Chernikov, O.G.; Shevchenko, V.G.

1993-01-01

The use of self-powered detectors (SPDs) with a rhodium emitter customarily involves monitoring of neutron fields in the core of a nuclear reactor. Since current in an SPD is generated primarily because of the neutron flux, which is responsible for the dynamics of particular nuclear transformations, including fission reactions of heavy isotopes, the detector signal can be attributed unambiguously to energy release at the location of the detector. Computation modeling performed with the KOMDPS package of programs of the current formation in a rhodium SPD along with the neutron-physical processes that occur in the reactor core makes it possible to take account of the effect of the principal factors characterizing the operating conditions and the design features of the fuel channel and the detector, reveal quantitative relations between the generated signal and individual physical parameters, and determine the metrological parameters of the detector. The formation and transport of changed particles in the sensitive part of the SPC is calculated by the Monte Carlo method. The emitter activation, neutron transport, and dynamics of the isotopic composition in the fuel channel containing the SPD are determined by solving the kinetic equation in the multigroup representation of the neutron spectrum, using the discrete ordinate method. In this work the authors consider the operation of a rhodium SPD in a bundle of 49 fuel channels of the RBMK-1000 reactor with a fuel enrichment of 2.4% from the time it is inserted into a fresh channel

Study on the sensitivity of Self-Powered Neutron Detectors (SPND) and its change due to burn-up

International Nuclear Information System (INIS)

Cho, Gyuseong; Lee, Wanno; Yoon, Jeong-Hyoun.

1996-01-01

Self-Powered Neutron Detectors (SPND) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. While they have several advantages such as small size, low cost, and relatively simple electronics required in conjunction with its usage, it has some intrinsic problems of the low level of output current, a slow response time, the rapid change of sensitivity which makes it difficult to use for a long term. In this paper, Monte Carlo simulation was accomplished to calculate the escape probability as a function of the birth position for the typical geometry of rhodium-based SPNDs. Using the simulation result, the burn-up profile of rhodium number density and the neutron sensitivity is calculated as a function of burn-up time in the reactor. The sensitivity of the SPND decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long-term usage. (author)

Numerische und experimentelle Untersuchungen zur Beschreibung der Wirkungsweise und Effektivität von grossflächigen hochsensitiven SPNDs : Self-Powered-Neutron-Detectors

OpenAIRE

Al-Dabagh, D.

1981-01-01

In the present work the neutron sensitivity of large surface "Self-Powered- Neutron-Detectors«SPNDs) is determined numericaily and experimentally. First a computer model for the parametric caIculation of the (n,$\\gamma$)-reaction rates in the emitter of the SPND was developed under use of the Monte-Carlo-Program KENO 11 ‘To determine the escape-probability of the electrons produced via Comptonor photoelectric effect by the emitter as weil as fron the surrounding isolator, a new Monte-Carlo-Pr...

Self-powered flux detectors

International Nuclear Information System (INIS)

Shields, R.B.

1983-02-01

This bibliography attempts to cover the published literature on the class of radiation detectors most often referred to as 'self-powered'. For this purpose, self-powered detectors are defined as those that have two or more conducting electrodes separated by solid insulation and that generate a signal current without an external power source. Primary sensitivity is unrestricted, but it is usually to neutrons or gamma-rays. The main application is in the core of a nuclear reactor. All relevant facets of the subject are covered including: theory, experiment, development, design, manufacture, instrumentation and application. In addition to the usual reference information, various other designations are included where available, such as CONF-and abstract serial numbers. Where possible, a summary of the content is given with emphasis on specific results and conclusions. Indexing is by author and subject

Dynamic compensation of the Silver self-powered neutron detector in the ramp program at the OSIRIS reactor

International Nuclear Information System (INIS)

Moulin, D.J.

1994-01-01

The Silver self-powered neutron detector (SPND) is a common detector used in the ramp program at the OSIRIS reactor. The Silver SPND signal is a reference during steady states, but its response is too slow for monitoring transient tests. In order to compensate for the inherent time delay a mathematical processing method of the Silver SPND signal was developed. Based on a convolution-type resolution of the kinetics equations, a dynamic compensation algorithm can be used for transient conditions as well as steady state conditions. A computer program reconstructs, in real-time, the dynamic neutron flux sensed by the Silver detector from the current measured between the emitter and the collector of the SPND. Although this method decreases slightly the signal-to-noise ratio, it maintains the SPND's characteristics and reduces the response time from about 10 minutes to less than 4 seconds for a step change in flux. This provides for prompt and accurate measurement of fuel rod power during ramp experiments in the OSIRIS reactor. This development makes the Silver SPND very suitable for many on-line monitoring applications

Measuring delayed part of the current of a self powered neutron detector and comparison with calculations

International Nuclear Information System (INIS)

Kophazi, J.; Czifrus, Sz.; Feher, S.; Por, G.

2001-01-01

The paper describes the measurement of the delayed signal of a Rh emitter Self Powered Neutron Detector (SPND) separately from other signal components originating from (n-gamma-e), (background gamma-e) and other effects. In order to separate the delayed signal, the detector was removed from the reactor core and placed to an adequately distant location during the measurement, where the radiation from the core was negligible. The experiment was carried out on the 100kW light water tank-type reactor of Technical University of Budapest and the results of the measurement were compared with the results of Monte Carlo calculations.(author)

Cobalt-60 production in CANDU reactors

International Nuclear Information System (INIS)

Ross, Michel; Lemire, Christian

2002-01-01

CANDU reactors can produce cobalt-60 very efficiently and with an interesting return on investment. This paper discusses what is needed to convert a CANDU reactor into a cobalt-60 producer: what are the different phases, the safety studies required, the physical modifications needed, and what is the minimum involvement of the utility owning the plant. The past ten years of experience of Hydro-Quebec as a cobalt-60 producer will be reviewed, including the management of the risk of both incident and electricity generation loss, and including the benefits for the utility and its personnel. Originally a simple metal used for centuries as a pigment, cobalt-59 today is transformed into cobalt-60, a radioactive element of unprecedented value. Well known in medicine for cancer treatment, cobalt-60 is also used to sterilize a wide range of disposable medical products used in hospitals and to sanitize pharmaceutical and cosmetic products. Cobalt-60 is proving to be a new and effective solution, in the food sector, for preserving harvests and controlling food-borne diseases, or to advantageously replace certain gases and chemical products which are suspected of being harmful or carcinogenic. There are also other applications, such as: hardening of some plastics, treatment of sewage sludge and elimination of harmful insect populations. With a half-life of 5,3 years, cobalt-60 is a metal not found in nature. It is a radioactive isotope produced by exposing stable nuclei of cobalt-59 to neutrons. One of the best places to find such an important neutron source is a nuclear reactor. High energy gamma rays are then emitted during the process of radioactive decay, where cobalt-60 seeks again its stable state

Determination of concentration factors for Chromium cesium, iron and cobalt in corvine and shrimp in the vicinities of Almirante Alvaro Alberto Nuclear Power Plant

International Nuclear Information System (INIS)

Kamel, L.N.

1988-01-01

In order to verify if the concentration factors for dose calculations in critical population recommended by the International Atomic Energy (IAEA), on the Safety Series 57, are in agreement with local factors, studies were carried out at Piraquara de Dentro bay, a region in the vicinities of the Almirante Alvaro Alberto Nuclear Power Plant (CNAAA). Chromium, caesium, iron and cobalt concentration factors for corvine and shrimp were determined using the activation analysis method by neutrons, taking into account the same behaviour between radioactive and stable from the same physicochemical form. This study has evidenced that: The local values of cesium, iron and cobalt concentration factors for corvine (fish) are in the same order of magnitude of the IAEA recommended values; The chromium, caesium and cobalt concentration factor values determined for shrimp for Piraquara de Dentro bay are in the same order of magnitude or smaller than those proposed by IAEA, while the concentration factor value for ion is one order of magnitude higher than the IAEA recommended value. (author) [pt

Separation and purification of carrier-free cobalt-58 from neutron irradiated nickel foil for electrochemical studies

International Nuclear Information System (INIS)

Egamediev, S.; Nurbaeva, D.; Rakhmanov, A.

2004-01-01

Full text: Cobalt-58 will be used for tracer studies of the behaviour of cobalt radionuclides in no- carrier-added form during electrochemical deposition on metal backing. The 58 Co can be produced by using 58 Ni(n,p) 58 Co nuclear reaction in nuclear reactor. 58 Co (T 1/2 =71 days) decays by positron emitting (15%) and electron capture (85%) with simultaneous γ -irradiation. In this study, we have developed the simple method for separation and purification of 58 Co in no- carrier-added form from neutron irradiated nickel foil. Previously, we have studied the dissolution of nickel foil in various media to find best conditions for rapid dissolution of nickel target. It was found that nickel foil dissolved completely without heating in 6.3 M hydrobromic acid with addition a few drops of hydrogen peroxide. After dissolution of the target material, the cobalt-58 is separated from nickel, copper, iron and other elements by extraction chromatography. The solution in 6.3 M hydrobromic acid is passed through a column containing suspension of polytetrafluoroethylene powder with 0.5 M trioctylamine in xylene, equilibrated with the same acid. Nickel is not extracted and passed through column. Cobalt is retained and finally eluted with 3 M HBr in the one free column volume. The cobalt fraction is percolated through a column filled with suspension of pure polytetrafluoroethylene powder to purify from the admixture of extractant. The obtained solution is evaporated to dryness and the dry residue is treated by evaporation with aqua regia. After treatment the damp residue is dissolved in electrolyte and the obtained solution is used to study of 58 Co electrochemical deposition procedure. The yield of cobalt-58 was higher than 93% and the radiochemical purity was more than 99%. This method will be used for separation and purification of cobalt-57 to make of sealed sources for X-ray fluorescence analysis

Determination of selenium in samples of concentrated of nickel plus cobalt using the analysis through activation with epi thermic neutron

International Nuclear Information System (INIS)

Capote Rodriguez, G.; Perez Zayas, G.; Hernandez Rivero, A.; Moreno Bermudez, G.; Ribeiro Guevara, S.; Arribere, M.A.; Molina Insfran, J.

1997-01-01

In this work, the concentration of selenium and other elements was determined. Seventeen samples of sulfides plus cobalt were used. Among the employed methods for such purpose, the analysis was made through neutronic activation, in its instrumental variants, and by means of the application of radiochemical separation. (author) [es

Rhodium self-powered neutron detector's lifetime for korean standard nuclear power plants

International Nuclear Information System (INIS)

Yoo, Choon Sung; Kim, Byoung Chul; Park, Jong Ho; Fero, Arnold H.; Anderson, S. L.

2005-01-01

A method to estimate the relative sensitivity of a self-powered rhodium detector for an upcoming cycle is developed by combining the rhodium depletion data from a nuclear design with the site measurement data. This method can be used both by nuclear power plant designers and by site staffs of Korean standard nuclear power plants for determining which rhodium detectors should be replaced during overhauls

Cobalt, fast neutrons and physical models: Nuclear data and measurements series

International Nuclear Information System (INIS)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.; Lawson, R.D.

1987-07-01

Energy-averaged neutron total cross sections of cobalt were measured from ≅0.5 to 12.0 MeV. Differential elastic- and inelastic-scattering cross sections were measured from ≅1.5 to 10.0 MeV over the scattering-angle range ≅18 0 to 160 0 , with sufficient detail to define the energy-averaged behavior. Inelastic neutron groups were observed corresponding to ''levels'' at: 1115 +- 29, 1212 +- 24, 1307 +- 24, 1503 +- 33, 1778 +- 40, 2112 +- 40, 2224 +- 35, 2423 +- 39, 2593 +- 41 and 2810 keV. The experimental results were interpreted in terms of the spherical optical-statistical and coupled-channels models. An unusually successful description of observables was achieved over a wide energy range ( 20.0 MeV) with a spherical model having energy-dependent strengths and geometries. The energy dependencies are large below ≅7.0 MeV (i.e., ≅19.0 MeV above the Fermi energy), but become smaller and similar to those reported for ''global'' potentials at higher energies. The imaginary strength is large and decreases with energy. These imaginary-potential characteristics are attributed to neutron shell closure and collective-vibrational processes. The weak-coupling model also offers an explanation of the unusual negative energy slope and relatively small radius of the imaginary potential. The spherical optical model derived from the neutron-scattering results was extrapolated to bound energies using the dispersion relationship and the method of moments. The resulting real-potential strength and radius peak at ≅-10.0 MeV, while concurrently the real diffuseness is at a minimum. The extrapolated potential is ≅8% larger than that implied by reported particle-state energies, and ≅13% smaller than indicated by hole-state energies. 68 refs., 15 figs., 1 tab

Results and interpretation of noise measurements using in-core self powered neutron detector strings at Unit 2 of the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Gloeckler, O.; Por, G.; Valko, J.

1986-11-01

In-core neutron noise and fuel assembly outlet temperature noise measurements were performed at Unit 2 of Paks Nuclear Power Plant. Characteristics of the reactor and the noise measuring equipment are briefly described. The in-core Rhodium emitter selfpowered neutron detector strings positioned axially above the other show high coherence and linear phase at low frequencies indicating a marked transport effect, not regularly measured in PWRs. The coherence between horizontally placed neutron detectors is small and the phase is zero. A transport effect of different nature is obtained between neutron detectors (in-core and ex-core) and fuel assembly outlet thermocouples. The observed characteristics depend on reactor and fuel assembly power in a way supporting interpretation in terms of coolant density and void content changes and power feedback effects. During routine analysis vibration of 1.1 Hz appeared as a strong peak in the power spectra. The control assembly that was responsible for the observed behaviour could be localized with high certainty. (author)

The influence of rhodium burn-up on the sensitivity of rhodium self-powered neutron detectors

International Nuclear Information System (INIS)

Erben, O.

1980-01-01

Depression and self-shielding coefficients are presented for thermal and epithermal neutron flux densities. Functions are shown describing the distribution of beta particle sources on the emitter cross section for 0 to 50% rhodium burnup. The values are calculated of detector sensitivity to thermal and epithermal neutron flux densities for the said burnup for main types of rhodium SPN detectors made by SODERN. (J.B.)

Analysis and solution of current spike occurred in dynamic compensation of self-powered neutron detectors

International Nuclear Information System (INIS)

Peng, Xingjie; Li, Qing; Wang, Kan

2017-01-01

Highlights: • The current spike problem is observed in the dynamic compensation process of SPNDs. • The current spike is caused by unphysical current change due to range switching. • Modification on the compensation algorithm is introduced to deal with current spike. - Abstract: Dynamic compensation methods are required to improve the response speed of the Self-Powered Neutron Detectors (SPNDs) and make it possible to apply the SPNDs for core monitoring and surveillance. During the experimental test of the compensation method based on linear matrix inequality (LMI), spikes are observed in the compensated SPND current. After analyzing the measurement data, the cause is fixed on the unphysical change of the uncompensated SPND current due to range switching. Then some modifications on the dynamic compensation algorithms are proposed to solve the current spike problem.

Prototypes of Self-Powered Radiation Detectors Employing Intrinsic High-Energy Current (HEC) (POSTPRINT)

Science.gov (United States)

2016-01-01

neutron sensi- tivities of a Pt self - powered detector ,” IEEE Trans. Nucl. Sci. 25, 292–295 (1978). 6T. A. Dellin, R. E. Huddleston, and C. J...Gamma-sensitive self - powered detectors and their use for in-core flux -mapping,” IEEE Trans. Nucl. Sci. 28, 752–757 (1981). 9E. A. Burke and J. Wall...AFCEC-CX-TY-TP-2016-0006 PROTOTYPES OF SELF - POWERED RADIATION DETECTORS EMPLOYING INTRINSIC HIGH-ENERGY CURRENT (HEC) (POSTPRINT) Piotr

A new self-powered flux detector

International Nuclear Information System (INIS)

Allan, C.J.

1979-11-01

It has been found that an Inconel-Inconel coaxial cable can be used as a fast-responding, neutron, self-powered flux detector if the core wire is sufficiently large. Test results obtained with such a detector, having a core wire approximately 1.5 mm in diameter, are presented. Other materials suitable for use as an emitter material, in such a relatively large diameter detector, also are included. (auth)

Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

International Nuclear Information System (INIS)

Tzika, F.; Stamatelatos, I.E.

2004-01-01

Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample

Code of practice for in-core instrumentation for neutron fluence rate (flux) measurements in power reactors

International Nuclear Information System (INIS)

Anon.

1982-01-01

This standard applies to in-core (on-line) neutron detectors and instrumentation which is designed for safety, information or control purposes. It also applies to components in so far as these components are contained within the primary envelope of the reactor. The detector types usually used are dc ionization chambers and self-powered neutron detectors

Sensitivity change of rhodium self -powered detectors with burn-up

International Nuclear Information System (INIS)

Girgis, R.; Akimov, I.S.; Hamouda, I.

1976-01-01

The scope of the present paper is to obtain the calculation formulae to evaluate the rate of sensitivity change of the neutron self-powered detectors with burn-up. A code written in FORTRAN 4 was developed to be operational on the IBM-1130 computer. It has been established in the case of rhodium detectors that neglecting the β-particle absorption in the calculations leads to the underestimation of the detector sensitivity decrease up to 40%. The derived formulae can be used for other self-powered detectors. (author)

Development of signal processing electronics for self powered neutron detector signal with built-in on-line insulation monitoring [Paper No.:E3

International Nuclear Information System (INIS)

Das, Amitabha; Chaganty, S.P.

1993-01-01

Self powered neutron detectors (SPNDs) are employed to monitor in-core neutron flux in nuclear reactors for control, safety and mapping of in-core neutron flux. The d.c. current produced by SPND is converted into a proportional d.c. voltage, which in turn is used for various purposes stated above. This paper describes various features of the SPND amplifier developed in the Electronics Division of Bhabha Atomic Research Centre (BARC). It also outlines the principle of working of on-line monitoring of insulation resistance (IR) of the detector and associated mineral insulated (MI) and soft cables. The amplifier generates an alarm in case of the IR of the detector and the cable assembly falls below an accepted value or the cable is not connected to the amplifier and relieves the operator from periodic and manual checking of each of the individual detectors and ensures the validity of the signal for further processing. (author). 3 figs

Feasibility studies on large sample neutron activation analysis using a low power research reactor

International Nuclear Information System (INIS)

Gyampo, O.

2008-06-01

Instrumental neutron activation analysis (INAA) using Ghana Research Reactor-1 (GHARR-1) can be directly applied to samples with masses in grams. Samples weights were in the range of 0.5g to 5g. Therefore, the representativity of the sample is improved as well as sensitivity. Irradiation of samples was done using a low power research reactor. The correction for the neutron self-shielding within the sample is determined from measurement of the neutron flux depression just outside the sample. Correction for gamma ray self-attenuation in the sample was performed via linear attenuation coefficients derived from transmission measurements. Quantitative and qualitative analysis of data were done using gamma ray spectrometry (HPGe detector). The results of this study on the possibilities of large sample NAA using a miniature neutron source reactor (MNSR) show clearly that the Ghana Research Reactor-1 (GHARR-1) at the National Nuclear Research Institute (NNRI) can be used for sample analyses up to 5 grams (5g) using the pneumatic transfer systems.

Detailed evaluation of ET-RR-2 neutronic design: approach to criticality

International Nuclear Information System (INIS)

Ashoub, N.; Amin, E.

2000-01-01

The ET-RR-2 safety analysis evaluation effort included the assessment of the reactor core nuclear design parameters. The present work complements the previous effort of neutronic calculations using NCNSRC computer code packages. This paper is the first in comprehensive neutronic calculation assessment up to the equilibrium core. It deals with evaluation of neutronic calculation submitted during the early phase of the licensing process; namely the proposal to assemble the first critical core, calculation of power peaking factor and determination of detailed energy group flux distribution in the core particularly in the cobalt irradiation position. The neutronic design criteria are examined qualitatively and quantitatively in the present paper

Self-energy dispersion effects on neutron matter superfluidity

International Nuclear Information System (INIS)

Zuo Wei

2001-01-01

The effects of the dispersion and ground state correlation of the single particle self-energy on neutron matter superfluidity have been investigated in the framework of the Extended Brueckner-Hartree-Fock and the generalized BCS approaches. A sizable reduction of the energy gap is found due to the energy dependence of the self-energy. And the inclusion of the ground state correlations in the self-energy suppresses further the neutron matter superfluidity

Free neutron-proton analyzing power at medium energies

International Nuclear Information System (INIS)

Newsom, C.R.

1980-01-01

In recent years, increasing efforts have been made to measure the nucleon-nucleon polarization parameters. To date, no free neutron-proton spin correlated parameters have been published in the energy range 500 to 800 MeV. Existing analyzing power data is of low precision and in most cases was obtained by quasi-free proton scattering. As a first step in determining the neutron-proton scattering matrix, the free neutron-proton analyzing power has been measured at the Los Alamos Physics Facility as a function of energy and angle. The experiment was performed by scattering a neutron beam from a polarized proton target. The neutron beam was generated by scattering 800 MeV protons from a Beryllium target and using the neutrons produced at 0 degrees. The incident energy ranged from 300 MeV to 800 MeV. The energy spread of the neutron beam made it possible to measure the analyzing power at different energies simultaneously. Angular distributions were taken from 60 to 170 degrees in the center of mass system (c.m.)

Substrate dependent self-organization of mesoporous cobalt oxide nanowires with remarkable pseudocapacitance

KAUST Repository

Baby, Rakhi Raghavan

2012-05-09

A scheme of current collector dependent self-organization of mesoporous cobalt oxide nanowires has been used to create unique supercapacitor electrodes, with each nanowire making direct contact with the current collector. The fabricated electrodes offer the desired properties of macroporosity to allow facile electrolyte flow, thereby reducing device resistance and nanoporosity with large surface area to allow faster reaction kinetics. Co 3O 4 nanowires grown on carbon fiber paper collectors self-organize into a brush-like morphology with the nanowires completely surrounding the carbon microfiber cores. In comparison, Co 3O 4 nanowires grown on planar graphitized carbon paper collectors self-organize into a flower-like morphology. In three electrode configuration, brush-like and flower-like morphologies exhibited specific capacitance values of 1525 and 1199 F/g, respectively, at a constant current density of 1 A/g. In two electrode configuration, the brush-like nanowire morphology resulted in a superior supercapacitor performance with high specific capacitances of 911 F/g at 0.25 A/g and 784 F/g at 40 A/g. In comparison, the flower-like morphology exhibited lower specific capacitance values of 620 F/g at 0.25 A/g and 423 F/g at 40 A/g. The Co 3O 4 nanowires with brush-like morphology exhibited high values of specific power (71 kW/kg) and specific energy (81 Wh/kg). Maximum energy and power densities calculated for Co 3O 4 nanowires with flower-like morphology were 55 Wh/kg and 37 kW/kg respectively. Both electrode designs exhibited excellent cycling stability by retaining ∼91-94% of their maximum capacitance after 5000 cycles of continuous charge-discharge. © 2012 American Chemical Society.

H{infinity} Filtering for Dynamic Compensation of Self-Powered Neutron Detectors - A Linear Matrix Inequality Based Method -

Energy Technology Data Exchange (ETDEWEB)

Park, M.G.; Kim, Y.H.; Cha, K.H.; Kim, M.K. [Korea Electric Power Research Institute, Taejon (Korea)

1999-07-01

A method is described to develop and H{infinity} filtering method for the dynamic compensation of self-powered neutron detectors normally used for fixed incore instruments. An H{infinity} norm of the filter transfer matrix is used as the optimization criteria in the worst-case estimation error sense. Filter modeling is performed for both continuous- and discrete-time models. The filter gains are optimized in the sense of noise attenuation level of H{infinity} setting. By introducing Bounded Real Lemma, the conventional algebraic Riccati inequalities are converted into Linear Matrix Inequalities (LMIs). Finally, the filter design problem is solved via the convex optimization framework using LMIs. The simulation results show that remarkable improvements are achieved in view of the filter response time and the filter design efficiency. (author). 15 refs., 4 figs., 3 tabs.

Neutronics requirements for a DEMO fusion power plant

Energy Technology Data Exchange (ETDEWEB)

Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion Consortium , Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA UT-FUS C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy)

2015-10-15

Highlights: • Discussion and specification of neutronic requirements for a DEMO power plant. • TBR uncertainties are reviewed/discussed and design margins are elaborated. • Limits are given for radiation loads to super-conducting magnets and steel structural components. • Available DEMO results are compared to recommended limits and TBR design target. - Abstract: This paper addresses the neutronic requirements a DEMO fusion power plant needs to fulfil for a reliable and safe operation. The major requirement is to ensure Tritium self-sufficiency taking into account the various uncertainties and plant-internal losses that occur during DEMO operation. A further major requirement is to ensure sufficient protection of the superconducting magnets against the radiation penetrating in-vessel components and vessel. Reliable criteria for the radiation loads need to be defined and verified to ensure the reliable operation of the magnets over the lifetime of DEMO. Other issues include radiation induced effects on structural materials such as the accumulated displacement damage, the generation of gases such as helium which may deteriorate the material performance. The paper discusses these issues and their impact on design options for DEMO taking into account results obtained in the frame of European Power Plant Physics and Technology (PPPT) 2013 programme activities with DEMO models employing the helium cooled pebble bed (HCPB), the helium cooled lithium lead (HCLL), and the water-cooled (WCLL) blanket concepts.

Phase-transformation and subgrain-deformation characteristics in a cobalt-based superalloy

International Nuclear Information System (INIS)

Benson, M.L.; Reetz, B.; Liaw, P.K.; Reimers, W.; Choo, H.; Brown, D.W.; Saleh, T.A.; Klarstrom, D.L.

2011-01-01

Research highlights: → The mechanical behavior of a cobalt-based superalloy was investigated. → Two diffraction techniques were used to study deformation mechanisms of materials. → In-situ neutron diffraction provides the volume-averaged information. → The peak-profile analysis reveals the information on a subgrain level. → The material exhibited a transformation texture for the HCP phase under loading. - Abstract: A complimentary set of experiments, in situ neutron diffraction and ex situ synchrotron X-ray diffraction, were used to study the phase-transformation and subgrain-deformation characteristics of a cobalt-based superalloy. The neutron diffraction indicated a strain-induced phase transformation in the cobalt-based superalloy under uniaxial tension and compression. The synchrotron X-ray diffraction revealed stacking-fault accumulation and twinning under the same loading conditions. The extent of transformation was found to be greater under tension than under compression. Tensile plastic strains below 2% were accommodated by the stacking-fault creation, while those greater than 2% were accommodated by the phase transformation. Twinning was found to be more active under compressive loading than under tensile loading.

Experimental measurements and theoretical simulations for neutron flux in self-serve facility of Dhruva reactor

International Nuclear Information System (INIS)

Rana, Y.S.; Mishra, Abhishek; Singh, Tej

2016-06-01

Dhruva is a 100 MW th tank type research reactor with natural metallic uranium as fuel and heavy water as coolant, moderator and reflector. The reactor is utilized for production of a large variety of radioisotopes for fulfilling growing demands of various applications in industrial, agricultural and medicinal sectors, and neutron beam research in condensed matter physics. The core consists of two on-power tray rods for radioisotope production and fifteen experimental beam holes for neutron beam research. Recently, a self-serve facility has also been commissioned in one of the through tubes in the reactor for carrying out short term irradiations. To get accurate information about neutron flux spectrum, measurements have been carried out in self-serve facility of Dhruva reactor. The present report describes measurement method, analysis technique and results. Theoretical estimations for neutron flux were also carried out and a comparison between theoretical and experimental results is made. (author)

National Low-Level Waste Management Program Radionuclide Report Series: Volume 12, Cobalt-60

International Nuclear Information System (INIS)

Adams, J.P.

1995-06-01

This report outlines the basic radiological and chemical characteristics of cobalt-60 ( 60 Co) and examines how these characteristics affect the behavior of 60 Co in various environmental media, such as soils, groundwater, plants, animals, the atmosphere, and the human body. Discussions also include methods of 60 Co production, waste types, and waste forms that contain 60 Co. All cobalt atoms contain 27 protons (Z = 27) and various numbers of neutrons (typically N = 27 to 37 neutrons) within the atom's nucleus. There is only one stable isotope of cobalt, namely 59 Co. All other cobalt isotopes, including 60 Co, are radioactive. The radioactive isotopes of cobalt have half-lives ranging from less than a second ( 54 Co-0.19 s) to 5.2 years ( 60 Co). The radioactive isotopes of cobalt are not a normal constituent of the natural environment and are generated as a result of human activities. The primary source of 60 Co in the environment has been low-level radioactive waste material generated as a result of neutron activation of stable 59 Co that is present in the structural components of nuclear reactor vessels. This isotope is also intentionally produced, usually in reactors but also to some degree in accelerators for industrial and medical uses, such as for radiation sources for cancer treatment and nondestructive testing of metals and welds. 60 Co may enter the environment as a result of the activities associated with nuclear reactor operations and decommissioning and when industrial and medical sources are being used, manufactured, or disposed

Point kernel technique for calculating dose rates due to cobalt-60 hot particles

International Nuclear Information System (INIS)

Thornhill, M.J.; McCarthy, J.T.; Morrissette, R.R.; Leach, B.N.

1989-01-01

This paper reports on a computer code called BETA that has been developed by health physicists at the Vermont Yankee Nuclear Power Station which accounts for the mass and size of hot particles of Cobalt-60, and therefore corrects the Loevinger-based dose calculation for self-absorption

Introduction to the neutron kinetics of nuclear power reactors

CERN Document Server

Tyror, J G; Grant, P J

2013-01-01

An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

2008-10-29

The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

Probing fine magnetic particles with neutron scattering

International Nuclear Information System (INIS)

Pynn, R.

1991-01-01

Because thermal neutrons are scattered both by nuclei and by unpaired electrons, they provide an ideal probe for studying the atomic and magnetic structures of fine-grained magnetic materials, including nanocrystalline solids, thin epitaxial layers, and colloidal suspensions of magnetic particles, known as ferrofluids. Diffraction, surface reflection, and small angle neutron scattering (SANS) are the techniques used. With the exception of surface reflection, these methods are described in this article. The combination of SANS with refractive-index matching and neutron polarisation analysis is particularly powerful because it allows the magnetic and atomic structures to be determined independently. This technique has been used to study both dilute and concentrated ferrofluid suspensions of relatively monodisperse cobalt particles, subjected to a series of applied magnetic fields. The size of the cobalt particle core and the surrounding surfactant layer were determined. The measured interparticle structure factor agrees well with a recent theory that allows correlations in binary mixtures of magnetic particles to be calculated in the case of complete magnetic alignment. When one of the species in such a binary mixture is a nonmagnetic, cyclindrical macromolecule, application of a magnetic field leads to some degree of alignment of the nonmagnetic species. This result has been demonstrated with tobacco mosaic virus suspended in a water-based ferrofluid

Neutron stars with spin polarized self-interacting dark matter

OpenAIRE

Rezaei, Zeinab

2018-01-01

Dark matter, one of the important portion of the universe, could affect the visible matter in neutron stars. An important physical feature of dark matter is due to the spin of dark matter particles. Here, applying the piecewise polytropic equation of state for the neutron star matter and the equation of state of spin polarized self-interacting dark matter, we investigate the structure of neutron stars which are influenced by the spin polarized self-interacting dark matter. The behavior of the...

Enhancement of thermal neutron self-shielding in materials surrounded by reflectors

International Nuclear Information System (INIS)

Cornelia Chilian; Gregory Kennedy

2012-01-01

Materials containing from 41 to 1124 mg chlorine and surrounded by polyethylene containers of various thicknesses, from 0.01 to 5.6 mm, were irradiated in a research reactor neutron spectrum and the 38 Cl activity produced was measured as a function of polyethylene reflector thickness. For the material containing the higher amount of chlorine, the 38 Cl specific activity decreased with increasing reflector thickness, indicating increased neutron self-shielding. It was found that the amount of neutron self-shielding increased by as much as 52% with increasing reflector thickness. This is explained by neutrons which have exited the material subsequently reflecting back into it and thus increasing the total mean path length in the material. All physical and empirical models currently used to predict neutron self-shielding have ignored this effect and need to be modified. A method is given for measuring the adjustable parameter of a self-shielding model for a particular sample size and combination of neutron reflectors. (author)

Concentration differences between serum and plasma of the elements cobalt, iron, mercury, rubidium, selenium and zinc determined by neutron activation analysis

International Nuclear Information System (INIS)

Kasperek, K.; Kiem, J.; Iyengar, G.V.; Feinendegen, L.E.

1981-01-01

The differences in concentrations of cesium, cobalt, iron, mercury, rubidium, selenium and zinc between serum and plasma were examined with the aid of instrumental neutron activation analysis. Eighty serum and plasma samples obtained from 13 donors were compared. Serum was prepared in plastic tubes immediately after clotting, and plasma was separated with heparin as anticoagulant. No significant differences in the concentrations of cesium, cobalt, mercury and selenium were observed. However, the concentrations of iron, rubidium and zinc were significantly higher in serum than in plasma. The average differences were 322, 12 and 20 ng/ml for iron, rubidium and zinc, respectively. The average differences found for cesium, rubidium and zinc were far below that which can be expected from a complete, or considerable release of these elements from platelets which aggregate or disintegrate during the clotting process in preparing serum. (orig.)

Thermal neutron measurement using the instrumented test bundle and assessment of maximum linear power in HANARO

Energy Technology Data Exchange (ETDEWEB)

Lee, C. S.; Seo, C. K.; Lee, B. C.; Kim, H. N.; Kang, B. W. [KAERI, Taejon (Korea, Republic of)

2000-10-01

The HANARO fuel, U{sub 3}Si-Al, has been developed by AECL and tested in NRU reactor. Due to the lack of the data performed under the high power, the repetitive conduct of the irradiation test was required under the power greater than 108kW/m, which is the estimated maximum linear power in the design stage. Accordingly, the instrumented test bundle with SPND(Self Powered Neutron Detector) was fabricated and its irradiation test was performed in IR2 of HANARO. The measured thermal neutron flux with SPND is compared with calculation results by HANAFMS(HANARO Fuel Management System). The difference in the measured and calculated thermal flux values are below {+-}11% and the accuracy of the linear power predicted by HANAFMS is consequently accompanied. Therefore, it is believed that the maximum linear power above 120kW/m is achieved during the irradiation test of the test bundle.

National Low-Level Waste Management Program Radionuclide Report Series: Volume 12, Cobalt-60

Energy Technology Data Exchange (ETDEWEB)

Adams, J.P.

1995-06-01

This report outlines the basic radiological and chemical characteristics of cobalt-60 ({sup 60}Co) and examines how these characteristics affect the behavior of {sup 60}Co in various environmental media, such as soils, groundwater, plants, animals, the atmosphere, and the human body. Discussions also include methods of {sup 60}Co production, waste types, and waste forms that contain {sup 60}Co. All cobalt atoms contain 27 protons (Z = 27) and various numbers of neutrons (typically N = 27 to 37 neutrons) within the atom`s nucleus. There is only one stable isotope of cobalt, namely {sup 59}Co. All other cobalt isotopes, including {sup 60}Co, are radioactive. The radioactive isotopes of cobalt have half-lives ranging from less than a second ({sup 54}Co-0.19 s) to 5.2 years ({sup 60}Co). The radioactive isotopes of cobalt are not a normal constituent of the natural environment and are generated as a result of human activities. The primary source of {sup 60}Co in the environment has been low-level radioactive waste material generated as a result of neutron activation of stable {sup 59}Co that is present in the structural components of nuclear reactor vessels. This isotope is also intentionally produced, usually in reactors but also to some degree in accelerators for industrial and medical uses, such as for radiation sources for cancer treatment and nondestructive testing of metals and welds. {sup 60}Co may enter the environment as a result of the activities associated with nuclear reactor operations and decommissioning and when industrial and medical sources are being used, manufactured, or disposed.

Method and apparatus for determination of temperature, neutron absorption cross section and neutron moderating power

Science.gov (United States)

Vagelatos, Nicholas; Steinman, Donald K.; John, Joseph; Young, Jack C.

1981-01-01

A nuclear method and apparatus determines the temperature of a medium by injecting fast neutrons into the medium and detecting returning slow neutrons in three first energy ranges by producing three respective detection signals. The detection signals are combined to produce three derived indicia each systematically related to the population of slow neutrons returning from the medium in a respective one of three second energy ranges, specifically exclusively epithermal neutrons, exclusively substantially all thermal neutrons and exclusively a portion of the thermal neutron spectrum. The derived indicia are compared with calibration indicia similarly systematically related to the population of slow neutrons in the same three second energy ranges returning from similarly irradiated calibration media for which the relationships temperature, neutron absorption cross section and neutron moderating power to such calibration indicia are known. The comparison indicates the temperature at which the calibration indicia correspond to the derived indicia and consequently the temperature of the medium. The neutron absorption cross section and moderating power of the medium can be identified at the same time.

Ion exchange of Cobalt and Cadmium in Zeolite X

International Nuclear Information System (INIS)

Nava M, I.

1994-01-01

The growing development in the industry has an important contribution to the environmental damage, where the natural effluents are each day more contaminated by toxic elements, such as: mercury, chromium, lead and cadmium. So as to separate such elements it has sorbent must have enough stability, and have a sharp capacity of sorption. In this work it was studied the sorption behavior of cobalt and on the other hand, cadmium in aqueous solutions, which along with sodic form of the Zeolite X, undergoes a phenomenon of ionic interchange. Such interchange was verify to different concentration of cadmium, cobalt and hydronium ion. The content of cobalt and sodium in the interchanged samples was detected through the neutronic activation analysis. The results disclose a higher selectivity for cadmium than cobalt. (Author)

Neutron and gamma irradiation effects on power semiconductor switches

Science.gov (United States)

Schwarze, G. E.; Frasca, A. J.

1990-01-01

The performance characteristics of high power semiconductor switches subjected to high levels of neutron fluence and gamma dose must be known by the designer of the power conditioning, control and transmission subsystem of space nuclear power systems. Location and the allowable shielding mass budget will determine the level of radiation tolerance required by the switches to meet performance and reliability requirements. Neutron and gamma ray interactions with semiconductor materials and how these interactions affect the electrical and switching characteristics of solid state power switches is discussed. The experimental measurement system and radiation facilities are described. Experimental data showing the effects of neutron and gamma irradiation on the performance characteristics are given for power-type NPN Bipolar Junction Transistors (BJTs), and Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs). BJTs show a rapid decrease in gain, blocking voltage, and storage time for neutron irradiation, and MOSFETs show a rapid decrease in the gate threshold voltage for gamma irradiation.

Recent advances in self-powered flux detector development for CANDU reactors

International Nuclear Information System (INIS)

Allan, C.J.; Drewell, N.H.; Hall, D.S.

1983-01-01

The characteristics of self-powered flux detectors used in CANDU reactors are reviewed. Detectors with emitters of vanadium, platinum, platinum-clad Inconel and Inconel are used. Data on dynamic response, relative neutron and gamma-ray sensitivities, and burnout, obtained both from experiments and from the Monte Carlo code ICARES, are presented. Since the response of a detector depends on the relative magnitudes of the various current-producing mechanisms, the operating principles of self-powered detectors are briefly reviewed. Current research programmes are discussed. These include modifying the design of the platinum-clad Inconel detector in order to match its dynamic response to that of the fuel power and developing a prompt-responding flux-mapping detector. (author)

Limits on Self-Interacting Dark Matter from Neutron Stars

DEFF Research Database (Denmark)

Kouvaris, C.

2012-01-01

We impose new severe constraints on the self-interactions of fermionic asymmetric dark matter based on observations of nearby old neutron stars. Weakly interacting massive particle (WIMP) self-interactions mediated by Yukawa-type interactions can lower significantly the number of WIMPs necessary...... for gravitational collapse of the WIMP population accumulated in a neutron star. Even nearby neutron stars located at regions of low dark matter density can accrete a sufficient number of WIMPs that can potentially collapse, form a mini black hole, and destroy the host star. Based on this, we derive constraints...

Neutron and gamma irradiation effects on power semiconductor switches

International Nuclear Information System (INIS)

Schwarze, G.E.; Frasca, A.J.

1990-01-01

The performance characteristics of high power semiconductor switches subjected to high levels of neutron fluence and gamma dose must be known by the designer of the power conditioning, control and transmission subsystem of space nuclear power systems. Location and the allowable shielding mass budget will determine the level of radiation tolerance required by the switches to meet performance and reliability requirements. Neutron and gamma ray interactions with semiconductor materials and how these interactions affect the electrical and switching characteristics of solid state power switches is discussed. The experimental measurement system and radiation facilities are described. Experimental data showing the effects of neutron and gamma irradiation on the performance characteristics are given for power-type NPN bipolar junction transistors (BJTs), and metal-oxide-semiconductor field effect transistors (MOSFETs)

Some properties of zero power neutron noise in a time-varying medium with delayed neutrons

International Nuclear Information System (INIS)

Kitamura, Y.; Pal, L.; Pazsit, I.; Yamamoto, A.; Yamane, Y.

2008-01-01

The temporal evolution of the distribution of the number of neutrons in a time-varying multiplying system, producing only prompt neutrons, was treated recently with the master equation technique by some of the present authors. Such a treatment gives account of both the so-called zero power reactor noise and the power reactor noise simultaneously. In particular, the first two moments of the neutron number, as well as the concept of criticality for time-varying systems, were investigated and discussed. The present paper extends these investigations to the case when delayed neutrons are also taken into account. Due to the complexity of the description, only the expectation of the neutron number is calculated. The concept of criticality of a time-varying system is also generalized to systems with delayed neutrons. The temporal behaviour of the expectation of the number of neutrons and its asymptotic properties are displayed and discussed

Cold neutron radiography using low power accelerator

International Nuclear Information System (INIS)

Kiyanagi, Yoshiaki; Iwasa, Hirokatu

1993-01-01

A cold neutron source which can be adopted at a low power accelerator was studied. Time-of-flight radiography using the cold neutron source was performed. It is suggested that time-of-flight cold neutron radiography has possibility to distinguish the materials more clearly than the traditional film method since large contrast differences can be obtained by using digital data of the neutron intensity at different energies from thermal to cold region. Material will be identified at the same time by this method. (author)

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

Indian Academy of Sciences (India)

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the ...

Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

International Nuclear Information System (INIS)

Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.

2007-01-01

In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required

Cobalt oxide polymorph growth on electrostatic self-assembled nanoparticle arrays for dually tunable nano-textures

International Nuclear Information System (INIS)

Bulliard, Xavier; Benayad, Anass; Lee, Kwang-Hee; Choi, Yun-Hyuk; Lee, Jae Cheol; Park, Jong-Jin; Kim, Jong-Min

2011-01-01

We report on a method for surface nano-texturing on a plastic substrate. Nano-objects with a silica nanoparticle core and a textured cobalt oxide crown are created with selectable density on the plastic substrate. The resulting dual morphology is easily tuned over large areas, either by changing the parameters directing nanoparticle deposition through electrostatic self-arrangement for nano-object density control, or the parameter directing cobalt oxide deposition for shape control. The entire process takes place at room temperature, with no chemicals harmful to the plastic substrate. The ready modulation of the dual morphology is used to control the wettability properties of the plastic film, which is covered by nano-objects.

A Self-Powered Thin-Film Radiation Detector Using Intrinsic High-Energy Current (HEC) (Author’s Final Version)

Science.gov (United States)

2016-09-08

of electromagnetic 85 pulse effects on cables and electrical devices4 and as a self - powered detector for in-core neutron flux measurement in nuclear...AFCEC-CX-TY-TP-2016-0003 A SELF - POWERED THIN-FILM RADIATION DETECTOR USING INTRINSIC HIGH-ENERGY CURRENT (HEC) (AUTHOR’S FINAL VERSION...14 -- 5 Oct 15 A self - powered thin-film radiation detector using intrinsic high-energy current (HEC) (Author’s Final Version) FA8051-15-P-0010

Neutron measurements at nuclear power reactors [55

CERN Document Server

Scherpelz, R I

2002-01-01

Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D sub 2 O-moderated sup 2 sup 5 sup 2 Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields. Battelle has participate...

Self-shielding for thick slabs in a converging neutron beam

CERN Document Server

Mildner, D F R

1999-01-01

We have previously given a correction to the neutron self-shielding for a thin slab to account for the increased average path length through the slab when irradiated in a converging neutron beam. This expression overstates the case for the self-shielding for a thick (or highly absorbing) slab. We give a better approximation to the increase in effective shielding correction for a slab placed in a converging neutron beam. It is negligible at large absorption mean free paths. (author)

Fusion neutronics plan in the development of fusion reactor. With the aim of realizing electric power

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Hiroo; Morimoto, Yuichi; Ochiai, Kentarou; Sugimoto, Masayoshi; Nishitani, Takeo; Takeuchi, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-10-01

On June 1992, Atomic Energy Commission in Japan has settled Third Phase Program of Fusion Research and Development to achieve self-ignition condition, to realize long pulse burning plasma and to establish basis of fusion engineering for demonstration reactor. This report describes research plan of Fusion Neutron Laboratory in JAERI toward a development of fusion reactor with an aim of realizing electric power. The fusion neutron laboratory has a fusion neutronics facility (FNS), intense fusion neutron source. The plan includes research items in the FNS; characteristics of shielding and breeding materials, nuclear characteristics of materials, fundamental irradiation process of insulator, diagnostics materials and structural materials, and development of in-vessel diagnostic technology. Upgrade of the FNS is also described. Also, the International Fusion Material Irradiation Facility (IFMIF) for intense neutron source to develop fusion materials is described. (author)

Transport properties of cobalt at low temperatures

DEFF Research Database (Denmark)

Radharkishna, P.; Nielsen, Mourits

1965-01-01

Measurements are made of electrical resistivity, absolute thermoelectric power, and thermal conductivity of polycrystalline cobalt between 1.2 and 6 K; results are discussed on basis of inter-electronic scattering.......Measurements are made of electrical resistivity, absolute thermoelectric power, and thermal conductivity of polycrystalline cobalt between 1.2 and 6 K; results are discussed on basis of inter-electronic scattering....

The problem of resonance self-shielding effect in neutron multigroup calculations

International Nuclear Information System (INIS)

Wang Qingming; Huang Jinghua

1991-01-01

It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations

The Self-Powered Detector Simulation `MATiSSe' Toolbox applied to SPNDs for severe accident monitoring in PWRs

Science.gov (United States)

Barbot, Loïc; Villard, Jean-François; Fourrez, Stéphane; Pichon, Laurent; Makil, Hamid

2018-01-01

In the framework of the French National Research Agency program on nuclear safety and radioprotection, the `DIstributed Sensing for COrium Monitoring and Safety' project aims at developing innovative instrumentation for corium monitoring in case of severe accident in a Pressurized Water nuclear Reactor. Among others, a new under-vessel instrumentation based on Self-Powered Neutron Detectors is developed using a numerical simulation toolbox, named `MATiSSe'. The CEA Instrumentation Sensors and Dosimetry Lab developed MATiSSe since 2010 for Self-Powered Neutron Detectors material selection and geometry design, as well as for their respective partial neutron and gamma sensitivity calculations. MATiSSe is based on a comprehensive model of neutron and gamma interactions which take place in Selfpowered neutron detector components using the MCNP6 Monte Carlo code. As member of the project consortium, the THERMOCOAX SAS Company is currently manufacturing some instrumented pole prototypes to be tested in 2017. The full severe accident monitoring equipment, including the standalone low current acquisition system, will be tested during a joined CEA-THERMOCOAX experimental campaign in some realistic irradiation conditions, in the Slovenian TRIGA Mark II research reactor.

Heating analysis of cobalt adjusters in reactor core

International Nuclear Information System (INIS)

Mei Qiliang; Li Kang; Fu Yaru

2011-01-01

In order to produce 60 Co source for industry and medicine applications in CANDU-6 reactor, the stainless steel adjusters were replaced with the cobalt adjusters. The cobalt rod will generate the heat when it is irradiated by neutron and γ ray. In addition, 59 Co will be activated and become 60 Co, the ray released due to 60 Co decay will be absorbed by adjusters, and then the adjusters will also generate the heat. So the heating rate of adjusters to be changed during normal operation must be studied, which will be provided as the input data for analyzing the temperature field of cobalt adjusters and the relative heat load of moderator. MCNP code was used to simulate whole core geometric configuration in detail, including reactor fuel, control rod, adjuster, coolant and moderator, and to analyze the heating rate of the stainless steel adjusters and the cobalt adjusters. The maximum heating rate of different cobalt adjuster based on above results will be provided for the steady thermal hydraulic and accident analysis, and make sure that the reactor is safe on the thermal hydraulic. (authors)

Research advances in adsorption treatment of cesium and cobalt in waste liquid of nuclear power plants by ion exchange resin

International Nuclear Information System (INIS)

Liu Dan; Wang Xin; Liu Jiean; Zhu Laiye; Chen Bin; Weng Minghui

2014-01-01

Liquid radwaste is unavoidably produced during the operation of nuclear power plants (NPP). With the development of China's nuclear power, much more attention will be paid to the impact of NPP construction on water environment. Radionuclides are the main targets for adsorption treatment of NPP liquid radwaste. Thereinto, cesium and cobalt are typical cationic ones. Currently, ion exchange process is widely used in liquid radwaste treatment system (WLS) of NPP, wherein resin plays a key role. The paper reviewed the research progress on adsorption of cesium and cobalt existed in NPP liquid radwaste using ion exchange resin. (authors)

A study on the linearity characteristics of neutron power measurement system for Hanaro

International Nuclear Information System (INIS)

Kang, Tai Ki; Kim, Young Ki; Lee, Byung Chul; Park, Sang Jun

1999-06-01

It is briefly described the general principles of neutron detection and the method of neutron measurement in the nuclear reactor which neutron flux varies widely and gamma radiation also exists. Wide-range Fission Chamber System which is excellent in electrical and mechanical performances has been selected for neutron power measurement system for Hanaro. The linearity characteristics of neutron power signals is a critical factor of the reliability in reactor power control. In particular , the linearity of the log power signal, which covers 10 decade form 10 -8 %FP to 200 %FP was a matter of primary concern during commissioning. In case of the linear power signal for reactor control at high power condition, the output signals were additionally analyzed in connection with the reactor thermal power and the delayed neutron signal from the primary pipe as well as the output signal from the compensated ion chamber as a reference signal. (author). 13 refs., 7 tabs., 33 figs

A study on the linearity characteristics of neutron power measurement system for Hanaro

Energy Technology Data Exchange (ETDEWEB)

Kang, Tai Ki; Kim, Young Ki; Lee, Byung Chul; Park, Sang Jun

1999-06-01

It is briefly described the general principles of neutron detection and the method of neutron measurement in the nuclear reactor which neutron flux varies widely and gamma radiation also exists. Wide-range Fission Chamber System which is excellent in electrical and mechanical performances has been selected for neutron power measurement system for Hanaro. The linearity characteristics of neutron power signals is a critical factor of the reliability in reactor power control. In particular , the linearity of the log power signal, which covers 10 decade form 10 {sup -8} %FP to 200 %FP was a matter of primary concern during commissioning. In case of the linear power signal for reactor control at high power condition, the output signals were additionally analyzed in connection with the reactor thermal power and the delayed neutron signal from the primary pipe as well asthe output signal from the compensated ion chamber as a reference signal. (author). 13 refs., 7 tabs., 33 figs.

Calculation of the power factor using the neutron diffusion hybrid equation

International Nuclear Information System (INIS)

Costa da Silva, Adilson; Carvalho da Silva, Fernando; Senra Martinez, Aquilino

2013-01-01

Highlights: ► A neutron diffusion hybrid equation with an external neutron source was used. ► Nodal expansion method to obtain the neutron flux was used. ► Nuclear power factors in each fuel element in the reactor core were calculated. ► The results obtained were very accurate. -- Abstract: In this paper, we used a neutron diffusion hybrid equation with an external neutron source to calculate nuclear power factors in each fuel element in the reactor core. We used the nodal expansion method to obtain the neutron flux for a given control rods bank position. The results were compared with results obtained for eigenvalue problem near criticality condition and fixed source problem during the start-up of the reactor, where external neutron sources are extremely important for the stabilization of external neutron detectors.

Analysis of Opportunity to Create Self-Regulating Reactor Facility of Extra-Low Power

International Nuclear Information System (INIS)

Kazansky, Y.A.; Levtchenko, V.A.; Yuriev, Y.S.

2002-01-01

This paper deals with fundamental possibilities (economy, safety, self-regulation) of creating an extra-low power reactor facility for heat supply. It contains the results of calculations for thermal and fast neutron reactors. The concept of this type of a reactor had been developed by the contributors earlier

Boiling detection using signals of self-powered neutron detectors and thermocouples

International Nuclear Information System (INIS)

Kozma, R.

1989-01-01

A specially-equipped simulated fuel assembly has been placed into the core of the 2 MW research reactor of the IRI, Delft. In this paper the recent results concerning the detection of coolant boiling in the simulated fuel assembly are introduced. Applying the theory of boiling temperature noise, different stages of boiling, i.e. one-phase flow, subcooled boiling, volume boiling, were identified in the measurements using the low-frequency noise components of the thermocouple signals. It has been ascertained that neutron noise spectra remained unchanged when subcooled boiling appeared, and that they changed reasonably only when developed volume boiling took place in the channels. At certain neutron detector positions neutron spectra did not vary at all, although developed volume boiling occurred at a distance of 3-4 cm from these neutron detectors. This phenomenon was applied in studying the field-of-view of neutron detectors

Applications of neutron radiography for the nuclear power industry

Energy Technology Data Exchange (ETDEWEB)

Craft, Aaron E.; Barton, John P.

2016-11-01

The World Conference on Neutron Radiography (WCNR) and International Topical Meeting on Neutron Radiography (ITMNR) series have been running over 35 years. The most recent event, ITMNR-8, focused on industrial applications and was the first time this series was hosted in China. In China, more than twenty new nuclear power plants are in construction and plans have been announced to increase the nuclear capacity further by a factor of three within fifteen years. There are additional prospects in many other nations. Neutron tests were vital during previous developments of materials and components for nuclear power applications, as reported in this conference series. For example a majority of the 140 papers in the Proceedings of the First WCNR are for the benefit of the nuclear power industry. Included are reviews of the diverse techniques being applied in Europe, Japan, the United States, and at many other centers. Many of those techniques are being utilized and advanced to the present time. Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Applications include examination of nuclear waste, nuclear fuels, cladding, control elements, and other critical components. In this paper, the techniques developed and applied internationally for the nuclear power industry since the earliest years are reviewed, and the question is asked whether neutron test techniques can be of value in development of the present and future generations of nuclear power plants world-wide.

Neutron self-shielding with k0-NAA irradiations

International Nuclear Information System (INIS)

Chilian, C.; Chambon, R.; Kennedy, G.

2010-01-01

A sample of SMELS Type II reference material was mixed with powdered Cd-nitrate neutron absorber and analysed by k 0 NAA for 10 elements. The thermal neutron self-shielding effect was found to be 34.8%. When flux monitors were irradiated sufficiently far from the absorbing sample, it was found that the self-shielding could be corrected accurately using an analytical formula and an iterative calculation. When the flux monitors were irradiated 2 mm from the absorbing sample, the calculations over-corrected the concentrations by as much as 30%. It is recommended to irradiate flux monitors at least 14 mm from a 10 mm diameter absorbing sample.

Research on self-powered detectors for gamma-ray monitoring

International Nuclear Information System (INIS)

Cho, S.W.; Lee, Y.J.

1984-01-01

Self-powered neutron detectors are used extensively in power reactors both for flux mapping and for power control and over-power protection, because of their small size, ruggedness and simplicity. But they have a few disadvantages such as high burn-up rate and background signal produced by the gamma-rays from the reactor itself. In order to overcome these disadvantages and to achieve a better understanding of gamma-ray effects of self-power detectors, a new type of self-powered detectors was designed and fabricated by the author,and experiments have carried out in the 10kCi sup(60)Co gamma irradiation facility in Korea Advanced Energy Research Institute. The configuration of the new type detectors is not of coaxial type but of paralled plate in order to obtain directional effects of gamma-ray incidence. Detector materials and dimensions are so chosen that the output current signal is large enough to be detected using some commercial measuring divice even at low dose rate and the contribution of the lead cable to the total signal is negligibly small. The results are 1)sensitivity is depended primarily on the materials of the insulator, 2)output signal has a good linearity to gamma dose rate, 3) response of detectors is prompt, but not perfect, 4) critical thickness for satusation of the output current is thinner than the range of photoelectron in the materials. (Author)

Research on self-powered detectors for gamma-ray monitoring

International Nuclear Information System (INIS)

Cho, S.W.

1983-01-01

Self-powered neutron detectors are used extensively in power reactors both for flux mapping and for power control and over-power protection, because of their small size, ruggedness and simplicity. But they have a few disadvantages such as high burn-up rate and background signal produced by the gamma-rays from the reactor itself. In order to overcome these disadvantages and to achieve a better understanding of gamma-ray effects of self-powered detectors, a new type of self-powered detectors was designed and fabricated by the author, and experiments have been carried out in the 10kCi sup(60)Co gamma irradiation facility in Korea Advanced Energy Research Institute. The configuration of the new type detectors is not of coaxial type but of paralled plate in order to obtain directional effects of gamma-ray incidence. Detector materials and dimensions are so chosen that the output current signal is large enough to be detected using some commercial measuring divice even at low dose rate and the contribution of the lead cable to the totel signal is negligibly small. The results are 1) sensitivity is depended primarily on the materials of the insulator, 2) output signal has a good linearity to gamma dose rate, 3) response of detectors is prompt, but not perfect, 4) critical thickness for satusation of the output current is thinner than the range of photoelectron in the materials. (Author) πT

Design and fabrication of a novel self-powered solid-state neutron detector

Science.gov (United States)

LiCausi, Nicholas

key advantages of the novel devices are theoretical neutron detection efficiency of ˜48%, a self-passivating structure that reduces leakage current and detector operation with no bias resulting in extremely low device noise. Processes required to fabricate the 3D type detector were explored and developed in this thesis. The detector capacitance and processing steps have been simulated with MEDICI and TSuprem-4, respectively. Lithography masks were then designed using Cadence. The fabrication process development was conducted in line with standard CMOS grade integrated circuit processing to allow for simple integration with existing fabrication facilities. A number of new processes were developed including the low pressure chemical vapor deposition of conformal boron films using diborane on very high aspect-ratio trenches and holes. Development also included methods for "wet" chemical etching and "dry" reactive ion etching of the deposited boron films. Fabricated detectors were characterized with the transmission line method, 4-point probe, I-V measurements and C-V measurements. Finally the detector response to thermal neutrons was studied. Characterization has shown significant reduction in reverse leakage current density to ˜8x10-8 A/cm2 (nearly 4 orders of magnitude over the previously published data). Results show that the fabrication process developed is capable of producing efficient (˜22.5%) solid-state thermal neutron detectors.

Standard Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation Techniques

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 The purpose of this test method is to define a general procedure for determining an unknown thermal-neutron fluence rate by neutron activation techniques. It is not practicable to describe completely a technique applicable to the large number of experimental situations that require the measurement of a thermal-neutron fluence rate. Therefore, this method is presented so that the user may adapt to his particular situation the fundamental procedures of the following techniques. 1.1.1 Radiometric counting technique using pure cobalt, pure gold, pure indium, cobalt-aluminum, alloy, gold-aluminum alloy, or indium-aluminum alloy. 1.1.2 Standard comparison technique using pure gold, or gold-aluminum alloy, and 1.1.3 Secondary standard comparison techniques using pure indium, indium-aluminum alloy, pure dysprosium, or dysprosium-aluminum alloy. 1.2 The techniques presented are limited to measurements at room temperatures. However, special problems when making thermal-neutron fluence rate measurements in high-...

Neutron excess generation by fusion neutron source for self-consistency of nuclear energy system

International Nuclear Information System (INIS)

Saito, Masaki; Artisyuk, V.; Chmelev, A.

1999-01-01

The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities. (author)

Self-powered detector probes for electron and gamma-ray beam monitoring in high-power industrial accelerators

International Nuclear Information System (INIS)

Lone, M.A.

1992-08-01

A self-powered detector (SPD) is a simple passive device that consists of a coaxial probe with a metallic outer sleeve, a mineral oxide insulating layer, and a metallic inner core. SPDs are used in nuclear reactors for monitoring neutron and gamma ray fields. Responses of various SPDs to electron and gamma ray beams from industrial accelerators were investigated with Monte Carlo simulations. By judicious choice of transmission filters, threshold SPD probes were investigated for on-line monitoring of the beam energy spectrum of the high-power IMPELA industrial electron accelerator. (Author) (14 figs, 16 refs.)

Neutron flux monitoring device

International Nuclear Information System (INIS)

Shimazu, Yoichiro.

1995-01-01

In a neutron flux monitoring device, there are disposed a neutron flux measuring means for outputting signals in accordance with the intensity of neutron fluxes, a calculation means for calculating a self power density spectrum at a frequency band suitable to an object to be measured based on the output of the neutron flux measuring means, an alarm set value generation means for outputting an alarm set value as a comparative reference, and an alarm judging means for comparing the alarm set value with the outputted value of the calculation means to judge requirement of generating an alarm and generate an alarm in accordance with the result of the judgement. Namely, the time-series of neutron flux signals is put to fourier transformation for a predetermined period of time by the calculation means, and from each of square sums for real number component and imaginary number component for each of the frequencies, a self power density spectrum in the frequency band suitable to the object to be measured is calculated. Then, when the set reference value is exceeded, an alarm is generated. This can reliably prevent generation of erroneous alarm due to neutron flux noises and can accurately generate an alarm at an appropriate time. (N.H.)

Tris(trimethylsilyl)phosphate as electrolyte additive for self-discharge suppression of layered nickel cobalt manganese oxide

International Nuclear Information System (INIS)

Liao, Xiaolin; Zheng, Xiongwen; Chen, Jiawei; Huang, Ziyu; Xu, Mengqing; Xing, Lidan; Liao, Youhao; Lu, Qilun; Li, Xiangfeng; Li, Weishan

2016-01-01

Highlights: • TMSP is effective for self-discharge suppression of the charged NCM under 4.5 V. • TMSP oxidizes preferentially forming protective cathode interface film on NCM. • The film suppresses electrolyte decomposition and prevents NCM destruction. - Abstract: Application of layered nickel cobalt manganese oxide as cathode under higher potential than conventional 4.2 V yields a significant improvement in energy density of lithium ion battery. However, the cathode fully charged under high potential suffers serious self-discharge, in which the interaction between the cathode and electrolyte proceeds without potential limitation. In this work, we use tris(trimethylsilyl)phosphate (TMSP) as an electrolyte additive to solve this problem. A representative layered nickel cobalt manganese oxide, LiNi 1/3 Co 1/3 Mn 1/3 O 2 , is considered. The effect of TMSP on self-discharge behavior of LiNi 1/3 Co 1/3 Mn 1/3 O 2 is evaluated by physical and electrochemical methods. It is found that the self-discharge of charged LiNi 1/3 Co 1/3 Mn 1/3 O 2 can be suppressed significantly by using TMSP. TMSP is oxidized preferentially in comparison with the standard electrolyte during initial charging process forming a protective cathode interface film, which avoids the interaction between cathode and electrolyte at any potential and thus prevents electrolyte decomposition and protects LiNi 1/3 Co 1/3 Mn 1/3 O 2 from structure destruction.

Performance of SPNDs used in control and safety systems

International Nuclear Information System (INIS)

Fernando, M.P.S.; Raj, Manish; Kumar, A.N.

2006-01-01

Large sized reactor such as 540 MWe Pressurised Heavy Water Reactor (PHWR) requires continuous in core monitoring of local flux in order to provide effective control and protection. About 198 self powered neutron detectors (SPNDs) of the straight individually replaceable type are distributed in the reactor core. For purposes of reactor regulation, 42 prompt responding cobalt SPNDs called zone control detectors (ZCDs) are housed in vertical flux units (VFUs) and these are uniformly distributed in 14 power zones. The in core detectors used for spatial control by ZCCs do not accurately represent average zone power as they sense the flux over a small volume. Flux mapping system (FMS) comprising of 102 vanadium SPNDs in 26 VFUs, provide accurate measure of neutron flux, even though they have slow response to change in neutron flux levels. For reactor protection system-1 (RPS-1), 36 cobalt SPNDs are placed in VFUs and become part of core overpower protection system-1 (COPPS-1). Similarly, for RPS-2, 18 cobalt SPNDs are placed in horizontal flux units (HFUs) and become part of the COPPS-2. The present study discusses the performance of in core SPNDs used in TAPP-4 by comparing the measured fluxes with detailed simulations. The performances of SPNDs are evaluated at different power levels and several full power day of reactor operation. (author)

Note: Self-biased voltage to suppress secondary electrons by a ZnO varistor in a compact pulsed neutron generator

Science.gov (United States)

Yang, Z.; Li, X.; Li, J.; Long, J. D.; Lan, C. H.; Wang, T.; Dong, P.; He, J. L.

2017-03-01

A large amount of back streaming electrons will bring about a part of current drain on power supply, cause sparking or high-voltage breakdowns, and affect the neutron yield and waveform for a compact sealed-tube pulsed neutron generator. A novel idea which uses a ZnO varistor to provide a constant self-biased voltage to suppress the secondary electrons is introduced. The I-V curve for the ZnO varistor was measured in the experiment. The effects of suppressing the secondary electrons were investigated using a ZnO varistor, linear resistors, and an independent power supply, respectively. The results show that the secondary electrons are suppressed effectively by the compact ZnO varistor, while not increasing the size and the component of the device. It is a promising design for compact sealed-tube neutron generators.

Cobalt

Science.gov (United States)

Slack, John F.; Kimball, Bryn E.; Shedd, Kim B.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

2017-12-19

Cobalt is a silvery gray metal that has diverse uses based on certain key properties, including ferromagnetism, hardness and wear-resistance when alloyed with other metals, low thermal and electrical conductivity, high melting point, multiple valences, and production of intense blue colors when combined with silica. Cobalt is used mostly in cathodes in rechargeable batteries and in superalloys for turbine engines in jet aircraft. Annual global cobalt consumption was approximately 75,000 metric tons in 2011; China, Japan, and the United States (in order of consumption amount) were the top three cobalt-consuming countries. In 2011, approximately 109,000 metric tons of recoverable cobalt was produced in ores, concentrates, and intermediate products from cobalt, copper, nickel, platinum-group-element (PGE), and zinc operations. The Democratic Republic of the Congo (Congo [Kinshasa]) was the principal source of mined cobalt globally (55 percent). The United States produced a negligible amount of byproduct cobalt as an intermediate product from a PGE mining and refining operation in southeastern Montana; no U.S. production was from mines in which cobalt was the principal commodity. China was the leading refiner of cobalt, and much of its production came from cobalt ores, concentrates, and partially refined materials imported from Congo (Kinshasa).The mineralogy of cobalt deposits is diverse and includes both primary (hypogene) and secondary (supergene) phases. Principal terrestrial (land-based) deposit types, which represent most of world’s cobalt mine production, include primary magmatic Ni-Cu(-Co-PGE) sulfides, primary and secondary stratiform sediment-hosted Cu-Co sulfides and oxides, and secondary Ni-Co laterites. Seven additional terrestrial deposit types are described in this chapter. The total terrestrial cobalt resource (reserves plus other resources) plus past production, where available, is calculated to be 25.5 million metric tons. Additional resources of

Neutron reflectivity of electrodeposited thin magnetic films

International Nuclear Information System (INIS)

Cooper, Joshaniel F.K.; Vyas, Kunal N.; Steinke, Nina-J.; Love, David M.; Kinane, Christian J.; Barnes, Crispin H.W.

2014-01-01

Highlights: • Electrodeposited magnetic bi-layers were measured by polarised neutron reflectivity. • When growing a CoNiCu alloy from a single bath a Cu rich region is initially formed. • This Cu rich region is formed in the first layer but not subsequent ones. • Ni deposition is inhibited in thin film growth and Co deposits anomalously. • Alloy magnetism and neutron scattering length give a self-consistent model. - Abstract: We present a polarised neutron reflectivity (PNR) study of magnetic/non-magnetic (CoNiCu/Cu) thin films grown by single bath electrodeposition. We find that the composition is neither homogeneous with time, nor consistent with bulk values. Instead an initial, non-magnetic copper rich layer is formed, around 2 nm thick. This layer is formed by the deposition of the dilute, but rapidly diffusing, Cu 2+ ions near the electrode surface at the start of growth, before the region is depleted and the deposition becomes mass transport limited. After the region has been depleted, by growth etc., this layer does not form and thus may be prevented by growing a copper buffer layer immediately preceding the magnetic layer growth. As has been previously found, cobalt deposits anomalously compared to nickel, and even inhibits Ni deposition in thin films. The layer magnetisation and average neutron scattering length are fitted independently but both depend upon the alloy composition. Thus these parameters can be used to check for model self-consistency, increasing confidence in the derived composition

Improvement of core monitoring code cecor by the virtual segmentation of the self powered neutron detector loaded at Korean Standard Nuclear Plant

International Nuclear Information System (INIS)

Choi, T.; Jung, Y.S.

2006-01-01

Full text: Full text: Korean Standard Nuclear Plant uses Self Powered Neutron Detectors (SPNDs) to measure the neutron flux in the reactor core. The SPND's height is 40 cm and is located axially at the five different positions and 45 radial places. The design code simulated a reactor core is calculated by segmentation of the core. The segmentation is called as 'node', of which size is normally 20 cm. The axial height of the detector is larger than that of the node, and the larger detector's height maybe product some error on the axially complex shape. The analysis with the detector's signals showed some errors at the non-cosine axial flux shape. In order to reduce the errors for the shape, we tried to divide the detector by introducing the virtual boundary in the detector. Then, each axially 5 detectors had two virtual segmentations respectively and the detector's signal was divided by the inputs. So the more virtual detector's signals were gotten, the more accurate axial shape was produced. The result with virtual segmentations in a detector gave less deviation than the case without virtual segmentation (the current model). After the middle of cycle at the initial core specially, the axial neutron flux shape is changed to the saddle type one. The current model gave some error in Root Mean Square (RMS) between the measured value and the calculated one. The virtual segmentation model gave the better agreement at that time

Operational experiences of the spallation neutron source superconducting linac and power ramp-up

International Nuclear Information System (INIS)

Kim, Sang-Ho

2009-01-01

The spallation neutron source (SNS) is a second generation pulsed neutron source and designed to provide a 1-GeV, 1.44-MW proton beam to a mercury target for neutron production. Since the commissioning of the accelerator complex in 2006, the SNS has started its operation for neutron production and beam power ramp-up has been in progress toward the design goal. All subsystems of the SNS were designed and developed for substantial improvements compared to existing accelerators because the design beam power is almost an order of magnitude higher compared to existing neutron facilities and the achievable neutron scattering performance will exceed present sources by more than a factor of 20 to 100. In this paper, the operational experiences with the SNS Superconducting Linac (SCL), Power Ramp-up Plan to reach the design goal and the Power Upgrade Plan (PUP) will be presented including machine, subsystem and beam related issues.

Non-enzymatic glucose sensing platform using self assembled cobalt oxide/graphene nanocomposites immobilized graphite modified electrode

DEFF Research Database (Denmark)

Vivekananth, R.; Babu, R. Suresh; Prasanna, K.

2018-01-01

A new strategy to prepare the densely packed cobalt oxide (Co3O4)/graphene nanocomposites by a self-assembly method were adopted in this work. A new non-enzymatic glucose determination has been fabricated by using Co3O4/graphene nanocomposites modified electrode as a sensing material. The nanocom...... of the modified electrode for glucose determination has been evaluated in urine samples....

Cobalt production in RAPS-1

International Nuclear Information System (INIS)

Krishnan, P.D.; Purandare, H.D.

1978-01-01

At present in RAPS-1 radioisotope Co 60 is produced by irradiating Co 59 in the adjusters which perform the function of regulation of reactivity, power and xenon override. But the manrem expenditure of the crew handling the charge and discharge of the adjusters is going to be prohibitively high. It is therefore proposed to irradiate Co 59 in the fuel channel positions. The physics optimisation study for such irradiation is presented. The burnup penalty and loss of power are estimated to produce the required quantity of Co 60 after optimising the number of cobalt pencils in a bundle and the positions of the cobalt producing channels in the reactor core. (author)

Seminar Neutronika-2012. Neutron-physical problems of nuclear-power engineering. Program and abstracts

International Nuclear Information System (INIS)

2012-01-01

On October, 30 - November, 2 in State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky a seminar Neutron-physical problems of nuclear power engineering - Neutronika-2012 took place. On the seminar the following problems were discussed: justification of neutron-physical characteristics of reactor facilities and innovation projects; constant support of neutron-physical calculations of nuclear power installations; numerical simulation during solving reactor physics problems; simulation of neutron-physical processes in reactor facilities by Monte Carlo method; development and verification of programs for reactor facilities neutron-physical calculations; algorithms and programs for solving nonstationary problems of neutron-physical calculation of nuclear reactors; analysis of integral and reactor experiments, experimental database; justification of nuclear and radiation safety of fuel cycle [ru

Intense neutron source: high-voltage power supply specifications

International Nuclear Information System (INIS)

Riedel, A.A.

1980-08-01

This report explains the need for and sets forth the electrical, mechanical and safety specifications for a high-voltage power supply to be used with the intense neutron source. It contains sufficient information for a supplier to bid on such a power supply

Contribution to development of SPNDs for instantaneous and selective measurement of different radiation fields in nuclear reactors; Contribution au developpement de collectrons pour la mesure instantanee et selective des differents champs de rayonnements en reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Blandin, Christophe [Institut National Polytechnique, 38 - Grenoble (France)

1998-02-20

The objective of this work was conceiving and experimentally optimizing the SPNDs (Self-Powdered Neutron Detector) able to control fast power transients in test reactors and also to cope with requirements of surveillanceand protection of EDF reactors. Thus, different SPND emitters of platinum, gadolinium, hafnium and cobalt were provided according to their nature with sheathing and stainless steel plugs as well as with zirconium over-sheathing in order to render them faster, more selective and adapted for wear checking. Special experimental devices were designed for measuring inside the Siloe reactor the promptness of the signals from SPND, on one hand, and their sensitivity to thermal and epithermal neutrons as well as to gamma rays, on the other hand. The follow-up of power transients in test reactors is ensured by the instantaneous measurement of thermal and epithermal neutron flux as well as of gamma field by means of three special SPND with gadolinium, hafnium and platinum. Also, we have defined the characteristics of a new SPND with cobalt, that delivers a current of unique neutronic origin, able to ensure the surveillance and protection of a power reactor over a period of at least six years.

Contribution to development of SPNDs for instantaneous and selective measurement of different radiation fields in nuclear reactors

International Nuclear Information System (INIS)

Blandin, Christophe

1998-01-01

The objective of this work was conceiving and experimentally optimizing the SPNDs (Self-Powdered Neutron Detector) able to control fast power transients in test reactors and also to cope with requirements of surveillance and protection of EDF reactors. Thus, different SPND emitters of platinum, gadolinium, hafnium and cobalt were provided according to their nature with sheathing and stainless steel plugs as well as with zirconium over-sheathing in order to render them faster, more selective and adapted for wear checking. Special experimental devices were designed for measuring inside the Siloe reactor the promptness of the signals from SPND, on one hand, and their sensitivity to thermal and epithermal neutrons as well as to gamma rays, on the other hand. The follow-up of power transients in test reactors is ensured by the instantaneous measurement of thermal and epithermal neutron flux as well as of gamma field by means of three special SPND with gadolinium, hafnium and platinum. Also, we have defined the characteristics of a new SPND with cobalt, that delivers a current of unique neutronic origin, able to ensure the surveillance and protection of a power reactor over a period of at least six years

Calculating the Responses of Self-Powered Radiation Detectors.

Science.gov (United States)

Thornton, D. A.

Available from UMI in association with The British Library. The aim of this research is to review and develop the theoretical understanding of the responses of Self -Powered Radiation Detectors (SPDs) in Pressurized Water Reactors (PWRs). Two very different models are considered. A simple analytic model of the responses of SPDs to neutrons and gamma radiation is presented. It is a development of the work of several previous authors and has been incorporated into a computer program (called GENSPD), the predictions of which have been compared with experimental and theoretical results reported in the literature. Generally, the comparisons show reasonable consistency; where there is poor agreement explanations have been sought and presented. Two major limitations of analytic models have been identified; neglect of current generation in insulators and over-simplified electron transport treatments. Both of these are developed in the current work. A second model based on the Explicit Representation of Radiation Sources and Transport (ERRST) is presented and evaluated for several SPDs in a PWR at beginning of life. The model incorporates simulation of the production and subsequent transport of neutrons, gamma rays and electrons, both internal and external to the detector. Neutron fluxes and fuel power ratings have been evaluated with core physics calculations. Neutron interaction rates in assembly and detector materials have been evaluated in lattice calculations employing deterministic transport and diffusion methods. The transport of the reactor gamma radiation has been calculated with Monte Carlo, adjusted diffusion and point-kernel methods. The electron flux associated with the reactor gamma field as well as the internal charge deposition effects of the transport of photons and electrons have been calculated with coupled Monte Carlo calculations of photon and electron transport. The predicted response of a SPD is evaluated as the sum of contributions from individual

The effects of impurity composition and concentration in reactor structure material on neutron activation inventory in pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Cha, Gil Yong; Kim, Soon Young [RADCORE, Daejeon (Korea, Republic of); Lee, Jae Min [TUV Rheinland Korea, Seoul (Korea, Republic of); Kim, Yong Soo [Hanyang University, Seoul (Korea, Republic of)

2016-06-15

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

The cobalt radioactive isotopes in environment; Les isotopes radioactifs du cobalt dans l'environnement

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

For the year 1993 the total activity released in cobalt is 69 GBq for the whole of nuclear power plants. The part of activity in cobalt for La Hague in 1993 is 8 GBq of {sup 58}Co and 2 GBq of {sup 60}Co. The radioactive isotopes released by nuclear power plants or the reprocessing plant of La Hague under liquid effluents are shared by half between {sup 58}Co and {sup 60}Co. The exposure to sealed sources is the most important risk for the cobalt. The risk of acute exposure can associate a local irradiation of several decades of grays inducing a radiological burns, deep burn to treat in surgery by resection or graft even amputation. A global irradiation of organism for several grays induces an acute irradiation syndrome, often serious. At long term the stochastic effects are represented by leukemia and radio-induced cancers. The increase of probability of their occurrence is 1% by sievert. We must remind that the natural spontaneous probability is 25%. (N.C.)

Variation of Neutron Moderating Power on HDPE by Gamma Radiation

International Nuclear Information System (INIS)

Park, Kwang June; Ju, June Sik; Kang, Hee Young; Shin, Hee Sung; Kim, Ho Dong

2009-01-01

High density polyethylene (HDPE) is degraded due to a radiation-induced oxidation when it is used as a neutron moderator in a neutron counter for a nuclear material accounting of spent fuels. The HDPE exposed to the gamma-ray emitted from the fission products in a spent nuclear fuel results in a radiation-induced degradation which changes its original molecular structure to others. So a neutron moderating power variation of HDPE, irradiated by a gamma radiation, was investigated in this work. Five HDPE moderator structures were exposed to the gamma radiation emitted from a 60 Co source to a level of 10 5 -10 9 rad to compare their post-irradiation properties. As a result of the neutron measurement test with 5 irradiated HDPE structures and a neutron measuring system, it was confirmed that the neutron moderating power for the 105 rad irradiated HDPE moderator revealed the largest decrease when the un-irradiated pure one was used as a reference. It implies that a neutron moderating power variation of HDPE is not directly proportional to the integrated gamma dose rate. To clarify the cause of these changes, some techniques such as a FTIR, an element analysis and a densitometry were employed. As a result of these analyses, it was confirmed that the molecular structure of the gamma irradiated HDPEs had partially changed to others, and the contents of hydrogen and oxygen had varied during the process of a radiation-induced degradation. The mechanism of these changes cannot be explained in detail at present, and thus need further study

An Optical Model Study of Neutrons Elastically Scattered by Iron, Nickel, Cobalt, Copper, and Indium in the Energy Region 1.5 to 7.0 MeV

Energy Technology Data Exchange (ETDEWEB)

Holmqvist, B; Wiedling, T

1967-03-15

Angular distributions of elastically scattered neutrons have been measured for cobalt and copper at nine energies between 1.5 and 7.0 MeV, for natural iron at 4.6 MeV, for natural nickel and indium at four energies between 3.0 and 4.6 MeV, by using time-of-flight technique. The observed angular distributions were corrected for neutron flux attenuation, multiple elastic scattering, and the finite geometry of the source-sample-detector system by using a Monte Carlo program. Theoretical angular distributions have been fitted to the experimental angular distributions by using an optical model potential with Saxon-Woods form factors. A computer program was used to find parameter values of the potential giving the best fittings to the experimental angular distributions.

The cobalt radioactive isotopes in environment

International Nuclear Information System (INIS)

2007-01-01

For the year 1993 the total activity released in cobalt is 69 GBq for the whole of nuclear power plants. The part of activity in cobalt for La Hague in 1993 is 8 GBq of 58 Co and 2 GBq of 60 Co. The radioactive isotopes released by nuclear power plants or the reprocessing plant of La Hague under liquid effluents are shared by half between 58 Co and 60 Co. The exposure to sealed sources is the most important risk for the cobalt. The risk of acute exposure can associate a local irradiation of several decades of grays inducing a radiological burns, deep burn to treat in surgery by resection or graft even amputation. A global irradiation of organism for several grays induces an acute irradiation syndrome, often serious. At long term the stochastic effects are represented by leukemia and radio-induced cancers. The increase of probability of their occurrence is 1% by sievert. We must remind that the natural spontaneous probability is 25%. (N.C.)

Absolute measurement of the critical scattering cross section in cobalt

International Nuclear Information System (INIS)

Glinka, C.J.; Minkiewicz, V.J.; Passell, L.

1975-01-01

Small-angle neutron scattering techniques have been used to study the angular distribution of the critical scattering from cobalt above T/sub c/. These measurements have been put on an absolute scale by calibrating the critical scattering directly against the nuclear incoherent scattering from cobalt. In this way the interaction range r 1 , which appears in the classical and modified Ornstein--Zernike expressions for the asymptotic form of the spin pair correlation function and is related to the strength of the spin correlations, has been determined. We obtain r 1 /a = 0.46 +- 0.03 for the ratio of the interaction range to the nearest-neighbor distance in cobalt. This result is in good agreement with theoretical predictions. Lack of agreement among previous determinations of the ratio r 1 /a made in iron failed to provide a definitive comparison with theory

Molecular dynamics simulation of thermophysical properties of undercooled liquid cobalt

International Nuclear Information System (INIS)

Han, X J; Wang, J Z; Chen, M; Guo, Z Y

2004-01-01

Molecular dynamics simulations with two different embedded-atom-method (EAM) potentials are applied to calculate the density, specific heat and self-diffusion coefficient of liquid cobalt at temperatures above and below the melting temperature. Simulation shows that Pasianot's EAM model of cobalt constructed on the basis of a hcp structure is more successful than Stoop's EAM model in the framework of a fcc structure in predicting the thermophysical properties of liquid cobalt. Simulations with Pasianot's EAM model indicate that the density fits into ρ = 7.49-9.17 x 10 -4 (T- T m ) g cm -3 , and the self-diffusion coefficient is given by D = 1.291 x 10 -7 exp(-48 795.71/RT) m 2 s -1 . Dissimilar to the linear dependence of the density and the Arrhenius dependence of the self-diffusion coefficient on temperature, the specific heat shows almost a constant value of 38.595 ± 0.084 J mol -1 K -1 within the temperature range of simulation. The simulated properties of liquid cobalt are compared with experimental data available. Comparisons show reasonable agreements between the simulated results from Pasianot's EAM model and experimental data

Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

Energy Technology Data Exchange (ETDEWEB)

Palermo, Iole, E-mail: iole.palermo@ciemat.es; Fernández, Iván; Rapisarda, David; Ibarra, Angel

2016-11-01

Highlights: • We perform neutronic calculations for the preliminary DCLL Blanket design. • We study the tritium breeding capability of the reactor. • We determine the nuclear heating in the main components. • We verify if the shielding of the TF coil is maintained. - Abstract: In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

High-power liquid-lithium jet target for neutron production

Science.gov (United States)

Halfon, S.; Arenshtam, A.; Kijel, D.; Paul, M.; Berkovits, D.; Eliyahu, I.; Feinberg, G.; Friedman, M.; Hazenshprung, N.; Mardor, I.; Nagler, A.; Shimel, G.; Tessler, M.; Silverman, I.

2013-12-01

A compact liquid-lithium target (LiLiT) was built and tested with a high-power electron gun at the Soreq Nuclear Research Center. The lithium target, to be bombarded by the high-intensity proton beam of the Soreq Applied Research Accelerator Facility (SARAF), will constitute an intense source of neutrons produced by the 7Li(p,n)7Be reaction for nuclear astrophysics research and as a pilot setup for accelerator-based Boron Neutron Capture Therapy. The liquid-lithium jet target acts both as neutron-producing target and beam dump by removing the beam thermal power (>5 kW, >1 MW/cm3) with fast transport. The target was designed based on a thermal model, accompanied by a detailed calculation of the 7Li(p,n) neutron yield, energy distribution, and angular distribution. Liquid lithium is circulated through the target loop at ˜200 °C and generates a stable 1.5 mm-thick film flowing at a velocity up to 7 m/s onto a concave supporting wall. Electron beam irradiation demonstrated that the liquid-lithium target can dissipate electron power areal densities of >4 kW/cm2 and volume power density of ˜2 MW/cm3 at a lithium flow of ˜4 m/s while maintaining stable temperature and vacuum conditions. The LiLiT setup is presently in online commissioning stage for high-intensity proton beam irradiation (1.91-2.5 MeV, 1-2 mA) at SARAF.

Self-absorption of neutron capture gamma-rays in gold samples

International Nuclear Information System (INIS)

Wisshak, K.; Walter, G.; Kaeppeler, F.

1983-06-01

The self absorption of neutron capture gamma rays in gold samples has been determined experimentally for two standard setups used in measurements of neutron capture cross sections. One makes use of an artificially collimated neutron beam and two C 6 D 6 detectors, the other of kinematically collimated neutrons and three Moxon-Rae detectors. Correction factors for an actual measurement of a neutron capture cross section using a gold standard of 1 mm thickness up to 12% were found for the first setup while they are only 4% for the second setup. The present data allow to determine the correction in an actual measurement with an accuracy of 0.5-1%. (orig.) [de

Effect of neutron irradiation on the breakdown voltage of power MOSFET's

International Nuclear Information System (INIS)

Hasan, S.M.Y.; Kosier, S.L.; Schrimpf, R.D.; Galloway, K.F.

1994-01-01

The effect of neutron irradiation on power metal-oxide-semiconductor field effect transistor (power MOSFET) breakdown voltage has been investigated. Transistors with various breakdown voltage ratings were irradiated in a TRIGA nuclear reactor with cumulative fluence levels up to 5 x 10 14 neutrons/cm 2 (1 MeV equivalent). Noticeable increases in the breakdown voltages are observed in n-type MOSFET's after 10 13 neutrons/cm 2 and in p-type MOSFETs after 10 12 neutrons/cm 2 . An increase in breakdown voltage of as much as 30% is observed after 5 x 10 14 neutrons/cm 2 . The increase in breakdown voltage is attributed to the neutron-irradiation-induced defects which decrease the mean free path and trap majority carriers in the space charge region. The effect of positive trapped oxide charge due to concomitant gamma radiation and the effect of the termination structure on the increase in breakdown voltage are considered. An empirical model is presented to predict the value of the breakdown voltage as a function of neutron fluence

Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

Science.gov (United States)

Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

2016-01-01

Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

Structure of a carbon monoxide adduct of cobalt-exchanged zeolite A (Co/sub 5.25/Na/sub 1.5/-A x 1.5CO) by neutron profile refinement

International Nuclear Information System (INIS)

Adams, J.M.; Haselden, D.A.

1984-01-01

The structure of a carbon monoxide adduct of cobalt-exchanged zeolite A, Co/sub 5.25/Na/sub 1.5/-A x 1.5CO, has been determined by Rietveld neutron profile refinement. All exchangeable cations are located in sites adjacent to the 6-ring; 3.75 of the cobalt cations sit 0.4 A inside the β-cage (Co(2)) and are arranged tetrahedrally about the eight 6-rings sites in the β-cage. The sodium cations (Na(1)) reside just inside the α-cage in sites similar to those found previously for zeolites 3A, 4A, and 5A. The remaining 1.5 cobalt cations (Co(1)) are located in sites similar to those for sodium, but they are also coordinated to the carbon monoxide molecules, which lie on, or close-to, the threefold axis which passes through the 6-ring. Inside the β-cage there is a tetrahedral aluminum complex of AlO 4 type, the oxygen atoms (O*) of which point toward six rings not occupied by cobalt cations, Co(2). Each of the oxygen atoms of this complex is involved in a hydrogen bond (2.83 A) to the 6-ring oxygen O(3). Approximately 2/3 of these bonds are of type O*-H ... O(3) and 1/3 of type O* ...H-O(3)

Studies on Flat Sandwich-type Self-Powered Detectors for Flux Measurements in ITER Test Blanket Modules

Science.gov (United States)

Raj, Prasoon; Angelone, Maurizio; Döring, Toralf; Eberhardt, Klaus; Fischer, Ulrich; Klix, Axel; Schwengner, Ronald

2018-01-01

Neutron and gamma flux measurements in designated positions in the test blanket modules (TBM) of ITER will be important tasks during ITER's campaigns. As part of the ongoing task on development of nuclear instrumentation for application in European ITER TBMs, experimental investigations on self-powered detectors (SPD) are undertaken. This paper reports the findings of neutron and photon irradiation tests performed with a test SPD in flat sandwich-like geometry. Whereas both neutrons and gammas can be detected with appropriate optimization of geometries, materials and sizes of the components, the present sandwich-like design is more sensitive to gammas than 14 MeV neutrons. Range of SPD current signals achievable under TBM conditions are predicted based on the SPD sensitivities measured in this work.

Experimental study of the IPR-R1 TRIGA reactor power channels responses

International Nuclear Information System (INIS)

Mesquita, Henrique F.A.; Ferreira, Andrea V.

2015-01-01

The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

Neutron capture cross section standards for BNL-325

International Nuclear Information System (INIS)

Holden, N.E.

1980-01-01

The most common cross section standards for capture reactions in the thermal neutron energy region are gold, cobalt, and manganese. In preparation for the fourth edition of BNL-325, data on the thermal cross section and resonance integral were evaluated for these three standards. For gold, only measurements below the Bragg scattering cutoff were used and extrapolated to a neutron velocity of 2200 meters/second. A non 1/v correction due to the 4.9 eV resonance was made. The resonance integral is based on Jirlow's integral measurement and Tellier's parameters. The resonance integrals for cobalt and manganese are based solely on integral measurements because the capture widths of the first major resonance either vary by 20% in various measurements (cobalt), or have never been measured (manganese). Recommended thermal cross sections and resonance integrals are respectively gold: 98.65/plus or minus/0.9 barns, 1550/plus or minus/28 barns; cobalt: 37.18/plus or minus/0.06 barns, 74.2/plus or minus/2.0 barns and manganese: 13.3/plus or minus/0.2 barns, and 14.0/plus or minus/0.3 barns. 72 refs

Fast neutron sensitivity of polymer dosemeters

International Nuclear Information System (INIS)

Harper, M.W.; Pearson, D.W.; Moran, P.R.

1975-01-01

The responses of polymer thermocurrent dosemeters to fission spectrum and 14 MeV neutrons were measured. The dosemeters are in the form of disks 1 cm diam by 0.5 mm thick. Relative to Cobalt 60 gamma responses, teflon PTFE dosemeters show a 6 percent response to 14 MeV neutrons and a 5 percent response to fission neutrons on a tissue rad basis. Polymethylpentene dosemeters show a 49 percent response to 14 MeV neutrons and a 40 percent response to fission neutrons on a tissue rad basis when provided with adequate recoil proton buildup. The sensitivity of these dosemeters is limited to neutron doses greater than 10 rads by spurious background currents

Characteristics of Fabricated SiC Neutron Detectors for Neutron Flux Monitoring

Energy Technology Data Exchange (ETDEWEB)

Kim, Han Soo; Ha, Jang Ho; Park, Se Hwan; Lee, Kyu Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Ho [Hanyang University, Seoul (Korea, Republic of)

2011-05-15

An SPND (Self-powered Neutron Detector) is commonly used for neutron detection in NPP (Nuclear Power Plant) by virtue of un-reactivity for gamma-rays. But it has a drawback, which is that it cannot detect neutrons in real time due to beta emissions (about > 48 s) after reactions between neutrons and {sup 103}Rh in an SPND. And Generation IV reactors such as MSR (Molten-salt reactor), SFR (Sodium-cooled fast reactor), and GFR (Gas-cooled fast reactor) are designed to compact size and integration type. For GEN IV reactor, neutron monitor also must be compact-sized to apply such reactor easily and much more reliable. The wide band-gap semiconductors such as SiC, AlN, and diamond make them an attractive alternative in applications in harsh environments by virtue of the lower operating voltage, faster charge-collection times compared with gas-filled detectors, and compact size.1) In this study, two PIN-type SiC semiconductor neutron detectors, which are for fast neutron detection by elastic and inelastic scattering SiC atoms and for thermal neutron detection by charged particle emissions of 6LiF reaction, were designed and fabricated for NPP-related applications. Preliminary tests such as I-V and alpha response were performed and neutron responses at ENF in HANARO research reactor were also addressed. The application feasibility of the fabricated SiC neutron detector as an in-core neutron monitor was discussed

Neutron energy spectra calculations in the low power research reactor

International Nuclear Information System (INIS)

Omar, H.; Khattab, K.; Ghazi, N.

2011-01-01

The neutron energy spectra have been calculated in the fuel region, inner and outer irradiation sites of the zero power research reactor using the MCNP-4C code and the combination of the WIMS-D/4 transport code for generation of group constants and the three-dimensional CITATION diffusion code for core analysis calculations. The neutron energy spectrum has been divided into three regions and compared with the proposed empirical correlations. The calculated thermal and fast neutron fluxes in the low power research reactor MNSR inner and outer irradiation sites have been compared with the measured results. Better agreements have been noticed between the calculated and measured results using the MCNP code than those obtained by the CITATION code. (author)

Neutron, gamma ray and post-irradiation thermal annealing effects on power semiconductor switches

Science.gov (United States)

Schwarze, G. E.; Frasca, A. J.

1991-01-01

The effects of neutron and gamma rays on the electrical and switching characteristics of power semiconductor switches must be known and understood by the designer of the power conditioning, control, and transmission subsystem of space nuclear power systems. The SP-100 radiation requirements at 25 m from the nuclear source are a neutron fluence of 10(exp 13) n/sq cm and a gamma dose of 0.5 Mrads. Experimental data showing the effects of neutrons and gamma rays on the performance characteristics of power-type NPN Bipolar Junction Transistors (BJTs), Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs), and Static Induction Transistors (SITs) are presented. These three types of devices were tested at radiation levels which met or exceeded the SP-100 requirements. For the SP-100 radiation requirements, the BJTs were found to be most sensitive to neutrons, the MOSFETs were most sensitive to gamma rays, and the SITs were only slightly sensitive to neutrons. Post-irradiation thermal anneals at 300 K and up to 425 K were done on these devices and the effectiveness of these anneals are also discussed.

Progress made by Framatome in neutron flux mapping technology

International Nuclear Information System (INIS)

Mourlevat, J.L.

1999-01-01

A fixed in-core system based on self-powered neutron detectors (SPNDs) could improve the power distribution measurement accuracy and hence improve core operating margin knowledge. For that purpose, Framatome has tested rhodium neutron detectors as fixed in-core sensors. Two such in-core strings were installed during tie summer of 1997 at Golfech Unit 2, for a three-year experiment. (author)

Real time neutronic evolution CNE (Embalse nuclear power plant)

International Nuclear Information System (INIS)

Notari, C.; Waldman, R.M.

1993-01-01

The simulator of the Embalse nuclear power plant uses a Point Reactor Model(PRM) for the neutronic evolution calculation. As this model is not conservative for transients produced by the sudden or localized reactivity insertion in big cores, it is convenient to use spatial models in these cases. In this report we show the results obtained using a nodal model (codes NODOS-TIEMPO). This model has been fitted against a more exact solution for the neutron flux and delayed neutron precursors. This has been done for the reactor at full power with nominal values for the reactivity control devices (liquid zones and adjusters rods). Transients corresponding to the global variation of the liquid zones and to the insertion of fresh fuel in some channels are shown. The results are compared with calculations made with the quasi-static model of the PUMA code. (author). 1 ref

Self-powered radiation detectors

International Nuclear Information System (INIS)

Gillies, Wallace.

1980-01-01

This invention aims to create a self fed radiation detector comprising a long central emitter-conductor absorbing the neutrons, wrapped in an insulating material, and a thin collector-conductor placed coaxially around the emitter and the insulation, the emitter being constructed of several stranded cables in a given conducting material so that the detector is flexible enough [fr

Neutronics investigation of advanced self-cooled liquid blanket systems in helical reactor

International Nuclear Information System (INIS)

Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M.Z.

2006-10-01

Neutronics performances of advanced self-cooled liquid blanket systems have been investigated in design activity of the helical-type reactor FFHR2. In the present study, a new three-dimensional (3-D) neutronics calculation system has been developed for the helical-type reactor to enhance quick feedback between neutronics evaluation and design modification. Using this new calculation system, advanced Flibe-cooled and Li-cooled liquid blanket systems proposed for FFHR2 have been evaluated to make clear design issues to enhance neutronics performance. Based on calculated results, modification of the blanket dimensions and configuration have been attempted to achieve the adequate tritium breeding ability and neutron shielding performance in the helical reactor. The total tritium breeding ratios (TBRs) obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. Issues in neutron shielding performance have been investigated quantitatively using 3-D geometry of the helical blanket system, support structures, poloidal coils etc. Shielding performance of the helical coils against direct neutrons from core plasma would achieve design target by further optimization of shielding materials. However, suppression of the neutron streaming and reflection through the divertor pumping areas in the original design is important issue to protect the poloidal coils and helical coils, respectively. Investigation of the neutron wall loading indicated that the peaking factor of the neutron wall load distribution would be moderated by the toroidal and helical effect of the plasma distribution in the helical reactor. (author)

Change of cobalt magnetic anisotropy and spin polarization with alkanethiolates self-assembled monolayers

International Nuclear Information System (INIS)

Campiglio, Paolo; Breitwieser, Romain; Repain, Vincent; Guitteny, Solène; Chacon, Cyril; Bellec, Amandine; Lagoute, Jérôme; Girard, Yann; Rousset, Sylvie; Sassella, Adele; Imam, Mighfar; Narasimhan, Shobhana

2015-01-01

We demonstrate that the deposition of a self-assembled monolayer of alkanethiolates on a 1 nm thick cobalt ultrathin film grown on Au(111) induces a spin reorientation transition from in-plane to out-of-plane magnetization. Using ab initio calculations, we show that a methanethiolate layer changes slightly both the magnetocrystalline and shape anisotropy, both effects almost cancelling each other out for a 1 nm Co film. Finally, the change in hysteresis cycles upon alkanethiolate adsorption could be assigned to a molecular-induced roughening of the Co layer, as shown by STM. In addition, we calculate how a methanethiolate layer modifies the spin density of states of the Co layer and we show that the spin polarization at the Fermi level through the organic layer is reversed as compared to the uncovered Co. These results give new theoretical and experimental insights for the use of thiol-based self-assembled monolayers in spintronic devices. (paper)

Sensitivity Measurement of a long SPND by a Split Neutron Irradiation

International Nuclear Information System (INIS)

Choi, Young San; Park, Sang Jun; Jung, Hwan Sung

2006-01-01

A Self Powered neutron Detector (SPND) has been used effectively as an in-core neutron monitor for a long time both in nuclear research reactors and power reactors due to its very simple dimensions and structures. For power reactor applications, Rhodium, Vanadium, Platinum, and Silver are usually used as the typical emitter materials for the SPNDs. The SPNDs with a Rhodium emitter (Rh-SPNDs) are most frequently used at Korean Standard Nuclear Power stations (KSNP) as a fixed incore neutron detector. This paper describes a new method to measure the neutron sensitivity of the Rh-SPNDs with a long sensitive length at a horizontal beam tube of HANARO by a split irradiation of them

Neutron, gamma ray and post-irradiation thermal annealing effects on power semiconductor switches

International Nuclear Information System (INIS)

Schwarze, G.E.; Frasca, A.J.

1994-01-01

The effects of neutrons and gamma rays on the electrical and switching characteristics of power semiconductor switches must be known and understood by the designer of the power conditioning, control, and transmission subsystem of space nuclear power systems. The SP-100 radiation requirements at 25 m from the nuclear source are a neutron fluence of 10 13 n/cm 2 and a gamma dose of 0.5 Mrads. Experimental data showing the effects of neutrons and gamma rays on the performance characteristics of power-type NPN Bipolar Junction Transistors (BJTs), Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs), and Static Induction Transistors (SITs) are given in this paper. These three types of devices were tested at radiation levels which met or exceeded the SP-100 requirements. For the SP-100 radiation requirements, the BJTs were found to be most sensitive to neutrons, the MOSFETs were most sensitive to gamma rays, and the SITs were only slightly sensitive to neutrons. Post-irradiation thermal anneals at 300 K and up to 425 K were done on these devices and the effectiveness of these anneals are also discussed

SELF-TRAPPING OF DISKOSEISMIC CORRUGATION MODES IN NEUTRON STAR SPACETIMES

Energy Technology Data Exchange (ETDEWEB)

Tsang, David [Center for Theory and Computation, Department of Astronomy, University of Maryland, College Park, MD 20742 (United States); Pappas, George [Department of Physics and Astronomy, The University of Mississippi, University, MS 38677 (United States)

2016-02-10

We examine the effects of higher-order multipole contributions of rotating neutron star (NS) spacetimes on the propagation of corrugation (c-)modes within a thin accretion disk. We find that the Lense–Thirring precession frequency, which determines the propagation region of the low-frequency fundamental corrugation modes, can experience a turnover allowing for c-modes to become self-trapped for sufficiently high dimensionless spin j and quadrupole rotational deformability α. If such self-trapping c-modes can be detected, e.g., through phase-resolved spectroscopy of the iron line for a high-spin low-mass accreting neutron star, this could potentially constrain the spin-induced NS quadrupole and the NS equation of state.

Self-Trapping of Diskoseismic Corrugation Modes in Neutron Star Spacetimes

Science.gov (United States)

Tsang, David; Pappas, George

2016-02-01

We examine the effects of higher-order multipole contributions of rotating neutron star (NS) spacetimes on the propagation of corrugation (c-)modes within a thin accretion disk. We find that the Lense-Thirring precession frequency, which determines the propagation region of the low-frequency fundamental corrugation modes, can experience a turnover allowing for c-modes to become self-trapped for sufficiently high dimensionless spin j and quadrupole rotational deformability α. If such self-trapping c-modes can be detected, e.g., through phase-resolved spectroscopy of the iron line for a high-spin low-mass accreting neutron star, this could potentially constrain the spin-induced NS quadrupole and the NS equation of state.

Sorption of cobalt in zeolites and natural clays of the clinoptilolite and kaolinite type

International Nuclear Information System (INIS)

Davila R, J.I.; Solache R, M.

2006-01-01

In this work the sorption of cobalt of aqueous solutions in two natural zeolites (clinoptilolite) and a clay (kaolinite) of origin in the center-north region of Mexico is evaluated. The effect of the pH and the time of contact in the process of sorption were evaluated. The cobalt retained in the aluminosilicates was determined by neutron activation analysis. The cobalt sorption in the materials in a range of pH from 4 to 7 does not present significant variations. The studies of reaction kinetics show a very fast sorption in the first 5 hours of contact, reaching the equilibrium in approximately 24 hours. The kinetics of sorption of the cobalt ions was represented better by the Ritchie reaction model modified of second order. The experimental data for the zeolites obtained at ambient temperature and varying the concentration were adjusted to the models of Freundlich, Langmuir and Freundlich-Langmuir isotherms and it was observed that the cobalt sorption it behaves according to the Freundlich isotherm model. (Author)

Neutron dosimetry in EDF experimental surveillance programme for VVER-440 nuclear power plants

International Nuclear Information System (INIS)

Brumovsky, M.; Erben, O.; Novosad, P.; Zerola, L.; Hogel, J.; Trollat, C.

2001-01-01

Fourteen chains containing experimental surveillance material specimens of the VVER 440/213 nuclear power reactor pressure vessels were irradiated in the surveillance channels of the Nuclear Power Plant Dukovany in the Czech Republic. The irradiation periods were one, two or three cycles. The chains contained different number and types of containers, the omitted ones were replaced by chain elements. All of the containers were instrumented with wire neutron fluence detectors, some of the containers in the chain had spectrometric sets of neutron fluence monitors. For the absolute fluence values evaluation it was taken into account time history of the reactor power and local changes of the neutron flux along the reactor core height, correction factors due to the orientation of monitors with respect to the reactor core centre. Unfolding programs SAND-II or BASA-CF were used. The relative axial fluence distribution was obtained from the O-wire measurements. Neutron fluence values above 0.5 MeV energy and above 1.0 MeV energy in the container axis on the axial positions of the sample centres and fluence values in the geometric centre of the samples was calculated making use the exponential attenuation model of the incident neutron beam. Received fast neutron fluence values can be used as reference values to all VVER-440 type 213 nuclear power plant reactors. (author)

Self-Assembled Hierarchical Formation of Conjugated 3D Cobalt Oxide Nanobead-CNT-Graphene Nanostructure Using Microwaves for High-Performance Supercapacitor Electrode.

Science.gov (United States)

Kumar, Rajesh; Singh, Rajesh Kumar; Dubey, Pawan Kumar; Singh, Dinesh Pratap; Yadav, Ram Manohar

2015-07-15

Here we report the electrochemical performance of a interesting three-dimensional (3D) structures comprised of zero-dimensional (0D) cobalt oxide nanobeads, one-dimensional (1D) carbon nanotubes and two-dimensional (2D) graphene, stacked hierarchically. We have synthesized 3D self-assembled hierarchical nanostructure comprised of cobalt oxide nanobeads (Co-nb), carbon nanotubes (CNTs), and graphene nanosheets (GNSs) for high-performance supercapacitor electrode application. This 3D self-assembled hierarchical nanostructure Co3O4 nanobeads-CNTs-GNSs (3D:Co-nb@CG) is grown at a large scale (gram) through simple, facile, and ultrafast microwave irradiation (MWI). In 3D:Co-nb@CG nanostructure, Co3O4 nanobeads are attached to the CNT surfaces grown on GNSs. Our ultrafast, one-step approach not only renders simultaneous growth of cobalt oxide and CNTs on graphene nanosheets but also institutes the intrinsic dispersion of carbon nanotubes and cobalt oxide within a highly conductive scaffold. The 3D:Co-nb@CG electrode shows better electrochemical performance with a maximum specific capacitance of 600 F/g at the charge/discharge current density of 0.7A/g in KOH electrolyte, which is 1.56 times higher than that of Co3O4-decorated graphene (Co-np@G) nanostructure. This electrode also shows a long cyclic life, excellent rate capability, and high specific capacitance. It also shows high stability after few cycles (550 cycles) and exhibits high capacitance retention behavior. It was observed that the supercapacitor retained 94.5% of its initial capacitance even after 5000 cycles, indicating its excellent cyclic stability. The synergistic effect of the 3D:Co-nb@CG appears to contribute to the enhanced electrochemical performances.

Generation of Transparent Oxygen Evolution Electrode Consisting of Regularly Ordered Nanoparticles from Self-Assembly Cobalt Phthalocyanine as a Template

KAUST Repository

Ziani, Ahmed

2016-11-04

The decoration of (photo)electrodes for efficient photoresponse requires the use of electrocatalysts with good dispersion and high transparency for efficient light absorption by the photoelectrode. As a result of the ease of thermal evaporation and particulate self-assembly growth, the phthalocyanine molecular species can be uniformly deposited layer-by-layer on the surface of substrates. This structure can be used as a template to achieve a tunable amount of catalysts, high dispersion of the nanoparticles, and transparency of the catalysts. In this study, we present a systematic study of the structural and optical properties, surface morphologies, and electrochemical oxygen evolution reaction (OER) performance of cobalt oxide prepared from a phthalocyanine metal precursor. Cobalt phthalocyanine (CoPc) films with different thicknesses were deposited by thermal evaporation on different substrates. The films were annealed at 400 °C in air to form a material with the cobalt oxide phase. The final Co oxide catalysts exhibit high transparency after thermal treatment. Their OER measurements demonstrate well expected mass activity for OER. Thermally evaporated and treated transition metal oxide nanoparticles are attractive for the functionalization of (photo)anodes for water oxidation.

Mathematical model of rhodium self-powered detectors and algorithms for correction of their time delay

International Nuclear Information System (INIS)

Bur'yan, V.I.; Kozlova, L.V.; Kuzhil', A.S.; Shikalov, V.F.

2005-01-01

The development of algorithms for correction of self-powered neutron detector (SPND) inertial is caused by necessity to increase the fast response of the in-core instrumentation systems (ICIS). The increase of ICIS fast response will permit to monitor in real time fast transient processes in the core, and in perspective - to use the signals of rhodium SPND for functions of emergency protection by local parameters. In this paper it is proposed to use mathematical model of neutron flux measurements by means of SPND in integral form for creation of correction algorithms. This approach, in the case, is the most convenient for creation of recurrent algorithms for flux estimation. The results of comparison for estimation of neutron flux and reactivity by readings of ionization chambers and SPND signals, corrected by proposed algorithms, are presented [ru

Target development for the SINQ high-power neutron spallation source

International Nuclear Information System (INIS)

Wagner, Werner

2002-01-01

SINQ is a 1 MW class research spallation neutron source, driven by the PSI proton accelerator system. In terms of beam power, it is, by a large margin, the most powerful spallation neutron source currently in operation worldwide. As a consequence, target load levels prevail in SINQ which are beyond the realm of existing experience. Therefore, an extensive materials irradiation program (STIP) is currently underway which will help to select the proper structural material and make dependable life time estimates accounting for the real operating conditions that prevail in the facility. In parallel, both theoretical and experimental work is going on within the MEGAPIE (MEGAwatt Pilot Experiment) project, to develop a liquid lead-bismuth spallation target for a beam power level of 1MW

Self-shielding coefficient and thermal flux depression factor of voluminous sample in neutron activation analysis

International Nuclear Information System (INIS)

Noorddin Ibrahim; Rosnie Akang

2009-01-01

Full text: One of the major problems encountered during the irradiation of large inhomogeneous samples in performing activation analysis using neutron is the perturbation of the neutron field due to absorption and scattering of neutron within the sample as well as along the neutron guide in the case of prompt gamma activation analysis. The magnitude of this perturbation shown by self-shielding coefficient and flux depression depend on several factors including the average neutron energy, the size and shape of the sample, as well as the macroscopic absorption cross section of the sample. In this study, we use Monte Carlo N-Particle codes to simulate the variation of neutron self-shielding coefficient and thermal flux depression factor as a function of the macroscopic thermal absorption cross section. The simulation works was carried out using the high performance computing facility available at UTM while the experimental work was performed at the tangential beam port of Reactor TRIGA PUSPATI, Malaysia Nuclear Agency. The neutron flux measured along the beam port is found to be in good agreement with the simulated data. Our simulation results also reveal that total flux perturbation factor decreases as the value of absorption increases. This factor is close to unity for low absorbing sample and tends towards zero for strong absorber. In addition, sample with long mean chord length produces smaller flux perturbation than the shorter mean chord length. When comparing both the graphs of self-shielding factor and total disturbance, we can conclude that the total disturbance of the thermal neutron flux on the large samples is dominated by the self-shielding effect. (Author)

Dose Evaluation of Neutron within Containment Building of a CE type Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Wook; Han, Jae Mun; Kim, Kyung Doek; Yun, Cheol Whan; Suh, Jang Soo; Kim, Young Jae [Nuclear Environment Technology Institute, Daejeon (Korea, Republic of)

2005-03-15

From measured results of the neutron fields at some principal places within the containment building in a CE type nuclear power plant in operation, the radiation exposure of a worker to the neutron at there was evaluated and the equivalent dose reflecting new recommendation (ICRP 60) was compared with that doing the old one (ICRP 26). The measured neutron field was also compared with calibration neutron field. From the analysis, the following conclusion was obtained: the average neutron radiation weighting factor according to new recommendation is 2.41 to 2.71 times higher than the old one. The average neutron radiation weighting factor at the measured place was similar to that at calibration neutron field. The average neutron energy at measured place was between 42 and 158 keV and higher than that of calibration field of 500 keV. So, the measured equivalent dose in nuclear power plant could be overestimated compared to the real equivalent dose.

β-Cobalt sulfide nanoparticles decorated graphene composite electrodes for high capacity and power supercapacitors

Science.gov (United States)

Qu, Baihua; Chen, Yuejiao; Zhang, Ming; Hu, Lingling; Lei, Danni; Lu, Bingan; Li, Qiuhong; Wang, Yanguo; Chen, Libao; Wang, Taihong

2012-11-01

Electrochemical supercapacitors have drawn much attention because of their high power and reasonably high energy densities. However, their performances still do not reach the demand of energy storage. In this paper β-cobalt sulfide nanoparticles were homogeneously distributed on a highly conductive graphene (CS-G) nanocomposite, which was confirmed by transmission electron microscopy analysis, and exhibit excellent electrochemical performances including extremely high values of specific capacitance (~1535 F g-1) at a current density of 2 A g-1, high-power density (11.98 kW kg-1) at a discharge current density of 40 A g-1 and excellent cyclic stability. The excellent electrochemical performances could be attributed to the graphene nanosheets (GNSs) which could maintain the mechanical integrity. Also the CS-G nanocomposite electrodes have high electrical conductivity. These results indicate that high electronic conductivity of graphene nanocomposite materials is crucial to achieving high power and energy density for supercapacitors.

β-Cobalt sulfide nanoparticles decorated graphene composite electrodes for high capacity and power supercapacitors.

Science.gov (United States)

Qu, Baihua; Chen, Yuejiao; Zhang, Ming; Hu, Lingling; Lei, Danni; Lu, Bingan; Li, Qiuhong; Wang, Yanguo; Chen, Libao; Wang, Taihong

2012-12-21

Electrochemical supercapacitors have drawn much attention because of their high power and reasonably high energy densities. However, their performances still do not reach the demand of energy storage. In this paper β-cobalt sulfide nanoparticles were homogeneously distributed on a highly conductive graphene (CS-G) nanocomposite, which was confirmed by transmission electron microscopy analysis, and exhibit excellent electrochemical performances including extremely high values of specific capacitance (~1535 F g(-1)) at a current density of 2 A g(-1), high-power density (11.98 kW kg(-1)) at a discharge current density of 40 A g(-1) and excellent cyclic stability. The excellent electrochemical performances could be attributed to the graphene nanosheets (GNSs) which could maintain the mechanical integrity. Also the CS-G nanocomposite electrodes have high electrical conductivity. These results indicate that high electronic conductivity of graphene nanocomposite materials is crucial to achieving high power and energy density for supercapacitors.

Self-ion emulation of high dose neutron irradiated microstructure in stainless steels

Science.gov (United States)

Jiao, Z.; Michalicka, J.; Was, G. S.

2018-04-01

Solution-annealed 304L stainless steel (SS) was irradiated to 130 dpa at 380 °C, and to 15 dpa at 500 °C and 600 °C, and cold-worked 316 SS (CW 316 SS) was irradiated to 130 dpa at 380 °C using 5 MeV Fe++/Ni++ to produce microstructures and radiation-induced segregation (RIS) for comparison with that from neutron irradiation at 320 °C to 46 dpa in the BOR60 reactor. For the 304L SS alloy, self-ion irradiation at 380 °C produced a dislocation loop microstructure that was comparable to that by neutron irradiation. No voids were observed in either the 380 °C self-ion irradiation or the neutron irradiation conditions. Irradiation at 600 °C produced the best match to radiation-induced segregation of Cr and Ni with the neutron irradiation, consistent with the prediction of a large temperature shift by Mansur's invariant relations for RIS. For the CW 316 SS alloy irradiated to 130 dpa at 380 °C, both the irradiated microstructure (dislocation loops, precipitates and voids) and RIS reasonably matched the neutron-irradiated sample. The smaller temperature shift for RIS in CW 316 SS was likely due to the high sink (dislocation) density induced by the cold work. A single self-ion irradiation condition at a dose rate ∼1000× that in reactor does not match both dislocation loops and RIS in solution-annealed 304L SS. However, a single irradiation temperature produced a reasonable match with both the dislocation/precipitate microstructure and RIS in CW 316 SS, indicating that sink density is a critical factor in determining the temperature shift for self-ion irradiations.

Reactor power measuring device

International Nuclear Information System (INIS)

Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

1993-01-01

The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

Tidal Love numbers of neutron and self-bound quark stars

International Nuclear Information System (INIS)

Postnikov, Sergey; Prakash, Madappa; Lattimer, James M.

2010-01-01

Gravitational waves from the final stages of inspiraling binary neutron stars are expected to be one of the most important sources for ground-based gravitational wave detectors. The masses of the components are determinable from the orbital and chirp frequencies during the early part of the evolution, and large finite-size (tidal) effects are measurable toward the end of inspiral, but the gravitational wave signal is expected to be very complex at this time. Tidal effects during the early part of the evolution will form a very small correction, but during this phase the signal is relatively clean. The accumulated phase shift due to tidal corrections is characterized by a single quantity related to a star's tidal Love number. The Love number is sensitive, in particular, to the compactness parameter M/R and the star's internal structure, and its determination could provide an important constraint to the neutron star radius. We show that Love numbers of self-bound strange quark matter stars are qualitatively different from those of normal neutron stars. Observations of the tidal signature from coalescing compact binaries could therefore provide an important, and possibly unique, way to distinguish self-bound strange quark stars from normal neutron stars. Tidal signatures from self-bound strange quark stars with masses smaller than 1M · are substantially smaller than those of normal stars owing to their smaller radii. Thus tidal signatures of stars less massive than 1M · are probably not detectable with Advanced LIGO. For stars with masses in the range 1-2M · , the anticipated efficiency of the proposed Einstein telescope would be required for the detection of tidal signatures.

Determination of the Jet Neutron Rate and Fusion Power using the Magnetic Proton Recoil Neutron Spectrometer

Energy Technology Data Exchange (ETDEWEB)

Sjoestrand, Henrik

2003-01-01

In this thesis a new independent method has been developed to enable precise measurements of neutron yields and rates from fusion plasmas and thereby determining the fusion power and fusion energy. The new method, together with the associated diagnostics, can provide information of great importance to present and future high fusion yield experiments, such as the Joint European Torus (JET) tokamak and the International Thermonuclear Experiment Reactor (ITER). The method has been applied to data from high fusion rate experiments from the tritium campaign at JET. By using the count-rate from the Magnetic Proton Recoil (MPR) neutron spectrometer the number of neutrons in the spectrometer's line of sight has been calculated. To be able to do this, all relevant factors between the plasma and the instrument have been evaluated. The number of neutrons in the MPR line of sight has been related to the total number of produced neutrons in the plasma by using information on the neutron emission profile. The achieved results have been compared with other JET neutron diagnostic data and the agreement is shown to be very good.

Prompt neutron decay constant estimation of RSG-GAS at high power noise experiment

International Nuclear Information System (INIS)

Jujuratisbela, U.; Kristedjo; Tukiran; Pinem, S.; Iman, J.; Puryono; Sanjaya, A.; Suwarno

1998-01-01

The determination of prompt neutron decay constant (α) of RGS-GAS by using low power noise experiment method at the equilibrium core indicated that the result is not good. The bad result was due to the small ratio of the noise signal to background which was caused by low detector efficiency or contaminated core after long time operation. To solve the problem is tried by using noise experiment technique at high power. The voltage output of neutron detectors at power of 5, 12, and 23 MW were connected to preamplifier and filter then to the Dynamic Signal Analyzer Version-2 and then the power spectral density of each channel of JKT04 and JKT03, the cut off frequency of each channel can be determined by using linear regression technique such that the prompt neutron decay constant can be estimated

Neutrons and fusion nuclear technology

International Nuclear Information System (INIS)

Hirayama, Shoichi

1991-01-01

The strategy of the devolopment of the fusion reactor has been compared with the history of the development of the fission reactor. More than 50 neutron reactors (neutron sources for research and development of reactor components and materials, and for Pu production) have been constructed and operated before the introduction of demonstration power reactors. This fact suggests us to introduce a new path of neutron reactor in the strategy of the development of fusion power reactor in addition to the orthodox approach which goes through the break-even, self-ignition, ETR, and DEMO. One of the benefits of the introduction of such neutron reactor or into the strategy of the fusion reactor development has been studied numerically. The results demonstrate that the introduction of fission-fusion hybrid reactor in 2030, can save ∝20% of natural uranium by 2100 in Japan, in comparison with the case when the fast breeder reactor is introduced in 2030. This saving is recognized large enough to justify earlier construction of the fusion neutron reactor. (orig.)

Sensitivity of self-powered detector probes to electron and gamma-ray fields

International Nuclear Information System (INIS)

Lone, M.A.; Wong, P.Y.

1995-01-01

A self-powered detector (SPD) is a simple, passive device that consists of a coaxial probe with a metallic outer sleeve, a mineral oxide insulating layer, and a metallic inner core. SPD's are used in nuclear reactors to monitor neutron and gamma fields. Responses of SPD's to electrons and γ-rays of various energies were investigated with Monte Carlo simulations. Transmission filters were studied for the design of threshold SPD probes used for online monitoring of the energy spectrum of high-power industrial electron accelerator beams. Filters were also investigated for the enhancement of γ-ray sensitivity of an SPD placed in a mixed electron and γ-ray field. (author). 30 refs., 1 tab., 8 figs

Sensitivity of self-powered detector probes to electron and gamma-ray fields

Energy Technology Data Exchange (ETDEWEB)

Lone, M A; Wong, P Y [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

1996-12-31

A self-powered detector (SPD) is a simple, passive device that consists of a coaxial probe with a metallic outer sleeve, a mineral oxide insulating layer, and a metallic inner core. SPD`s are used in nuclear reactors to monitor neutron and gamma fields. Responses of SPD`s to electrons and {gamma}-rays of various energies were investigated with Monte Carlo simulations. Transmission filters were studied for the design of threshold SPD probes used for online monitoring of the energy spectrum of high-power industrial electron accelerator beams. Filters were also investigated for the enhancement of {gamma}-ray sensitivity of an SPD placed in a mixed electron and {gamma}-ray field. (author). 30 refs., 1 tab., 8 figs.

NOKIA - nuclear power plant monitoring system

International Nuclear Information System (INIS)

Anon.

The monitoring system is described developed specially for the LOVIISA-1 and -2 nuclear power plants with two WWER-440 units. The multiprocessor system of the WWER-440 contains 3 identical main computers. The in core instrumentation is based on stationary self-powered neutron detectors and on thermocouples for measuring the coolant temperature. The system has equipment for the automatic control of the insulation resistance of the self-powered detectors. It is also equipped with a wide range of standard and special programmes. The standard programmes permit the recording of analog and digital data at different frequencies depending on the pre-set requirements. These data are processed and form data files which are accessible from all programmes. The heart of the special programme is a code for the determination of the power distribution in the core of the WWER-440 reactor. The main part of the programme is the algorithm for computing measured neutron fluxes derived from the signals of the self-powered detectors and the algorithm for deriving the global distribution of the neutron flux in the core. The computed power distribution is used for the determination of instantaneous thermal loads and the distribution of burnup in the core. The production programme of the FINNATOM company for nuclear power plants is listed. (B.S.)

Power Burst Reactor Facility as an epithermal neutron source for brain cancer therapy

International Nuclear Information System (INIS)

Wheeler, F.J.

1986-01-01

The Power Burst Facility (PBF) reactor is considered for modification to provide an intense, clean source of intermediate-energy (epithermal) neutrons desirable for clinical studies of neutron capture therapy (NCT) for malignant tumors. The modifications include partial replacement of the reflector, installation of a neutron-moderating, shifting region, additional shielding, and penetration of the present concrete shield with a collimating (and optionally) filtering region. The studies have indicated that the reactor, after these modifications, will be safely operable at full power (28 MW) within the acceptable limits of the plant protection systems. The neutron beam exiting from the collimator port is predicted to be of sufficient intensity (approx.10 10 neutrons/cm 2 -s) to provide therapeutic doses in very short irradiation times. The beam would be relatively free of undesirable fast neutrons, thermal neutrons and gamma rays. The calculated neutron energy spectrum and associated gamma rays in the beam were provided as input in simulation studies that used a computer model of a patient with a brain tumor to determine predicted dose rates to the tumor and healthy tissue. The results of this conceptual study indicate an intense, clean beam of epithermal neutrons for NCT clinical trials is attainable in the PBF facility with properly engineered design modifications. 9 refs., 11 figs., 3 tabs

Calculation support for industrial production of cobalt-60 at Leningrad NPP

International Nuclear Information System (INIS)

Artemov, Vladimir; Elshin, Alexander; Ivanov, Alexander; Gorbunov, Evgeny; Ikonnikov, Roman; Pimenov, Alexander

2008-01-01

Cobalt-60 is industrially produced at the Leningrad NPP by irradiation of cobalt-59 in special-purpose facilities loaded into the RBMK reactor core (all 4 units). The paper describes calculation methods used to determine the current activity of cobalt in irradiation assemblies for their timely unloading. The described peculiarities of core calculation model account for continuous refueling, overloading of irradiation assemblies and individual thermohydraulics in each channel under variation of reactor power. Fuel burnup in the core is calculated with a time step of about 24 hours. The resulting values for cobalt activity and uncertainties are presented in the paper as well. Deviation of calculated cobalt activity from measured activity is within the experimental accuracy of 10% (at confidence probability of 0.95). (authors)

Self-shielding and burn-out effects in the irradiation of strongly-neutron-absorbing material

International Nuclear Information System (INIS)

Sekine, T.; Baba, H.

1978-01-01

Self-shielding and burn-out effects are discussed in the evaluation of radioisotopes formed by neutron irradiation of a strongly-neutron-absorbing material. A method of the evaluation of such effects is developed both for thermal and epithermal neutrons. Gadolinium oxide uniformly mixed with graphite powder was irradiated by reactor-neutrons together with pieces of a Co-Al alloy wire (the content of Co being 0.475%) as the neutron flux monitor. The configuration of the samples and flux monitors in each of two irradiations is illustrated. The yields of activities produced in the irradiated samples were determined by the γ-spectrometry with a Ge(Li) detector of a relative detection efficiency of 8%. Activities at the end of irradiation were estimated by corrections due to pile-up, self-absorption, detection efficiency, branching ratio, and decay of the activity. Results of the calculation are discussed in comparison with the observed yields of 153 Gd, 160 Tb, and 161 Tb for the case of neutron irradiation of disc-shaped targets of gadolinium oxide. (T.G.)

Competitive Adsorption-Assisted Formation of One-Dimensional Cobalt Nanochains with High CO Hydrogenation Activity

Energy Technology Data Exchange (ETDEWEB)

Liang, Xin [State; Ren, Zhibo [State; Institute; Zhu, Xiaolin [State; Zhang, Qinwei [State; Mei, Donghai [Institute; Chen, Biaohua [State

2017-10-31

In the present work, cobalt nanochains have been successfully synthesized by a novel co assisted self-assembling formation strategy. A dramatic morphology transformation from cobalt nanoparticles to nanochains are observed when co molecules were introduced into the synthetic system. DFT calculations further confirm that competitive co-adsorbed co and oleylamine over the cobalt nanoparticles facilitates the formation of cobalt nanochains, which show higher co hydrogenation performance. The present work provides a new strategic and promising method for controllable synthesis of catalyst nanomaterials with the preferred surface structure and morphology.

Hard hardfacing by welding in the manufacture of valves; Problem Cobalt, alternatives, advantages, disadvantages

International Nuclear Information System (INIS)

Piquer Caballero, J.

2014-01-01

Alloys of recharge usually used in the field of the valves are base alloys cobalt (stellite), but in the field of nuclear power plants, due to radioactive activation of the cobalt, there is a growing trend to replace these alloys with other calls cobalt free . In this paper we will explore the most frequent and will be deducted the relevant advantages and disadvantages of these, in comparison with base alloys cobalt. (Author)

Self-assembled monolayers of a disulphide-derivatised cobalt-porphyrin on gold

International Nuclear Information System (INIS)

Viana, A.S.; Leupold, S.; Montforts, F.-P.; Abrantes, L.M.

2005-01-01

A self-assembled monolayer (SAM) of a novel cobalt(II)porphyrin disulphide derivative was prepared on flat gold(1 1 1) electrode. Evidence for surface modification was provided by electrochemical reductive desorption of the monolayer and ellipsometry, consistent with a coverage of 2.5 x 10 -10 mol cm -2 and a thickness of 13 A, respectively. Both results support the presence of SAMs where the molecules share an intermediate position between perpendicular and flat orientation. Scanning tunnelling microscopy have also proven the formation of CoPSS SAMs, however high-resolution images could only be obtained when the CoPSS molecules were diluted in an hexanethiol SAM. The electrocatalytic activity of the surface confined Co-porphyrin was evaluated for the oxygen reduction. Voltammetric data indicate that reaction involves two electrons consistent with the formation of hydrogen peroxide. Under similar experimental conditions the data obtained for an iron-porphyrin analogue points for a full reduction of dioxygen to water

MCNP and visualization of neutron flux and power distributions

International Nuclear Information System (INIS)

Snoj, L.; Lengar, I.; Zerovnik, G.; Ravnik, M.

2009-01-01

The visualization of neutron flux and power distributions in two nuclear reactors (TRIG A type research reactor and typical PWR) and one thermonuclear reactor (tokamak type) are treated in the paper. The distributions are calculated with MCNP computer code and presented using Amira and Voxler software. The results in the form of figures are presented in the paper together with comments qualitatively explaining the figures. The remembrance of most of the people is better, if they visualize a process. Therefore a representation of the reactor and neutron transport parameters is a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for core and irradiation planning. (authors)

Neutron dosimetry inside the containment building of Spanish nuclear power plants with PADC based dosemeters

Energy Technology Data Exchange (ETDEWEB)

Garcia-Fuste, M.J. [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain); Domingo, C., E-mail: carles.domingo@uab.ca [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain); Amgarou, K.; Bouassoule, T.; Castelo, J. [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain)

2009-10-15

The Spanish Nuclear Safety Council (Consejo de Seguridad Nuclear, CSN) recommends performing neutron individual dose assignments at workplaces based on ambient dose equivalent measurements using area monitors and by estimating the amount of time that workers spend in the different monitored environments. In addition, some Spanish nuclear power plants estimate the neutron dose equivalent using albedo thermoluminescence dosemeters (TLD). In the period 2004-2006, our group, together with other research centers, participated in a project, funded by the CSN, with the support of the Spanish Nuclear Power Plants Association (UNESA), to investigate in situ which could be the best practical procedure for individual neutron dose monitoring in nuclear power plants. As part of this survey, several units of the UAB PADC based neutron dosemeter were exposed, on a methacrylate phantom simulating a human body, at four different places inside the containment building of the Asco I nuclear power plant. The influence of different types of calibration neutron fields is analysed and the dose equivalent for each point is estimated.

Recovery of Cobalt as Cobalt Oxalate from Cobalt Tailings Using Moderately Thermophilic Bioleaching Technology and Selective Sequential Extraction

Directory of Open Access Journals (Sweden)

Guobao Chen

2016-07-01

Full Text Available Cobalt is a very important metal which is widely applied in various critical areas, however, it is difficult to recover cobalt from minerals since there is a lack of independent cobalt deposits in nature. This work is to provide a complete process to recover cobalt from cobalt tailings using the moderately thermophilic bioleaching technology and selective sequential extraction. It is found that 96.51% Co and 26.32% Cu were extracted after bioleaching for four days at 10% pulp density. The mean compositions of the leach solutions contain 0.98 g·L−1 of Co, 6.52 g·L−1 of Cu, and 24.57 g·L−1 of Fe (III. The copper ion was then recovered by a solvent extraction process and the ferric ions were selectively removed by applying a goethite deironization process. The technological conditions of the above purification procedures were deliberately discussed. Over 98.6% of copper and 99.9% of ferric ions were eliminated from the leaching liquor. Cobalt was finally produced as cobalt oxalate and its overall recovery during the whole process was greater than 95%. The present bioleaching process of cobalt is worth using for reference to deal with low-grade cobalt ores.

Polytypic transformations during the thermal decomposition of cobalt hydroxide and cobalt hydroxynitrate

International Nuclear Information System (INIS)

Ramesh, Thimmasandra Narayan

2010-01-01

The isothermal decomposition of cobalt hydroxide and cobalt hydroxynitrate at different intervals of temperature leads to the formation of Co 3 O 4 . The phase evolution during the decomposition process was monitored using powder X-ray diffraction. The transformation of cobalt hydroxide to cobalt oxide occurs via three phase mixture while cobalt hydroxynitrate to cobalt oxide occurs through a two phase mixture. The nature of the sample and its preparation method controls the decomposition mechanism. The comparison of topotactical relationship between the precursors to the decomposed product has been reported in relation to polytypism. - Graphical abstract: Isothermal thermal decomposition studies of cobalt hydroxide and cobalt hydroxynitrate at different intervals of temperature show the metastable phase formed prior to Co 3 O 4 phase.

Certification of a nickel metal reference material for neutron dosimetry (EC Nuclear Reference Material 521)

International Nuclear Information System (INIS)

Pauwels, J.

1988-01-01

Nickel metal, of 99.99 % nominal purity and natural isotopic composition, in the form of 0.1 mm thick foil and 0.5 mm diameter wire has been certified for its cobalt mass fraction. The certified value of cobalt (<0.1μg.g-1) is based on 38 results obtained by neutron activation analysis, emission spectrometry with inductively coupled plasma excitation and atomic absorption spectrometry, whereas the isotopic composition of the nickel was verified by thermal ionization mass spectrometry. The material is intended to be used as a reference material in neutron metrology

Notes on neutron flux measurement

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1984-01-01

The main purpose of this work is to get an useful guide to carry out topical neutron flux measurements. Although the foil activation technique is used in the majority of the cases, other techniques, such as those based on fission chambers and self-powered neutron detectors, are also shown. Special interest is given to the description and application of corrections on the measurement of relative and absolute induced activities by several types of detectors (scintillators, G-M and gas proportional counters). The thermal arid epithermal neutron fluxes, as determined in this work, are conventional or effective (West cots fluxes), which are extensively used by the reactor experimentalists; however, we also give some expressions where they are related to the integrated neutron fluxes, which are used in neutron calculations. (Author) 16 refs

High-power liquid-lithium jet target for neutron production

OpenAIRE

Halfon, S.; Arenshtam, A.; Kijel, D.; Paul, M.; Berkovits, D.; Eliyahu, I.; Feinberg, G.; Friedman, M.; Hazenshprung, N.; Mardor, I.; Nagler, A.; Shimel, G.; Tessler, M.; Silverman, I.

2013-01-01

A compact Liquid-Lithium Target (LiLiT) was built and tested with a high-power electron gun at Soreq Nuclear Research Center. The lithium target, to be bombarded by the high-intensity proton beam of the Soreq Applied Research Accelerator Facility (SARAF), will constitute an intense source of neutrons produced by the 7Li(p,n)7Be reaction for nuclear astrophysics research and as a pilot setup for accelerator-based Boron Neutron Capture Therapy (BNCT). The liquid-lithium jet target acts both as ...

Major activated corrosion products cobalt, silver and antimony in the primary coolant of PWR power plants

International Nuclear Information System (INIS)

Xu Mingxia

2012-01-01

The production of the major activated corrosion products such as cobalt, silver and antimony in the primary coolant of PWR power plants and the impacts on the increase of the dose rates caused by these corrosion products during the shutdown are described in the paper. Investigating the corrosion product behavior during the operation and shutdown periods aims at detecting the appearance of these radiological pollutants in the early time and searching relevant solutions that may enable eventually to decrease the dose rate. The solutions may include: Replacing critical material in the primary system's equipment and components, which contact with primary coolant circuit to possibly limit the source term, Elaborating strictly the specific chemical and shutdown procedure to optimize the purification capacity and to minimize the over-contaminations; Improving purification techniques according to the real operation circumstance, and limiting the impacts of these pollutants. It is obvious in the real practices that implementing appropriate solution will be benefit to decrease or limit the pollutants species like cobalt, silver and antimony. (author)

Evaluation of power history during power burst experiments in TRACY by combination of gamma-ray and thermal neutron detectors

International Nuclear Information System (INIS)

Yanagisawa, Hiroshi; Ohno, Akio

2002-01-01

A combination method using γ-ray and thermal neutron detectors was newly applied to the accurate evaluation of power histories during reactivity-initiated power burst experiments in the Transient Experiment Critical Facility (TRACY). During an initial power burst, the power history was determined using a fast response γ-ray ionization chamber, which was used because of its ability to exactly trace the power history within a short duration of the initial burst. After the initial burst, a micro fission chamber containing highly enriched uranium was used for the determination of the power history because the γ-ray ionization chamber could not be applied due to the contribution of delayed γ-rays from fission products. By the present method, the power histories were evaluated for the experiments in the range of 1.50 to 2.93$ of the reactivity insertion. It was found that the peak power and integrated power as determined by the previous method using only the micro fission chamber were underestimated to be 40% and 30% in maximum, respectively, in comparison with the results from the present evaluation. The numerical simulation performed by using the Monte Carlo method indicated that the underestimation could be comprehended by considering the time delay of thermal neutron detection of the fission chamber, which arose from the flight-time of neutrons from the TRACY core to the fission chamber. (author)

Transport of Cobalt 60 from the Argentine Nuclear Power Plant in Embalse

International Nuclear Information System (INIS)

Lopez Vietri, J.R.

2011-01-01

Full text: The purpose of this presentation is to point out some relevant issues related to the transport of cobalt 60 for domestic uses and export from the Argentine Nuclear Power Plant located in Embalse (ENPP). It is remarked that in last 22 years, ENPP produced about 2600 PBq (70 MCi) of Co-60, and that Argentina is the third Co-60 producer, after Canada and Russia. It is mentioned that in Argentina there are facilities which manufacture Zircaloy seamless tubes and reactivity control bars containing cobalt 59 to be used in ENPP, the operation of Candu reactor at ENPP including cobalt-related activities, and the manufacture and commercialization of Co-60 sealed sources. The Argentine Nuclear Regulatory Authority verifies the compliance of nuclear activities with regulatory standards, including those related to the transport of radioactive material. Then, it is described the process for obtaining Co-60 at ENPP with an activity concentration to allow its use in medicine and industry. After that, is considered in detail the type of package used for the transport of Co-60 from ENPP: it is to say, Type B(U) package designs that must be approved by the Competent Authority origin of the design, and they are able to withstand routine, normal and accident conditions of transport, including severe impacts, fires and immersion in water. The overall dimensions of packages are about 1.0 m to 1.5 m diameter by 1.2 m to 1.8 m high, as well as the gross mass could vary from 6500 kg to 9500 kg. Radioactive contents consist of Co-60 sealed sources (called pencils), with an activity range from 8 PBq to 16 PBq. Then, it is considered activities prior to dispatch of Type B(U) packages at ENPP, including the preparation of radioactive contents, packaging and package (labelling and marking) as well as necessary tests and documentation for transport in order to fulfil pertinent requirements of package design operation manual of the package and related standards. Finally, it is pointed

Dosimetry Characterization of the Neutron Fields of the Intermediate Temporary Storage of the Trillo Nuclear Power Plant

International Nuclear Information System (INIS)

Campo Blanco, X.

2015-01-01

The Neutron Standards Laboratory of CIEMAT, in collaboration with the Trillo Nuclear Power Plant, has conducted a detailed dosimetric and spectrometric characterization of the neutron fields at the Intermediate Temporary Storage of the Trillo Nuclear Power Plant, as well as the neutron fields of ENSA-DPT spent fuel casks. For neutron measurements, neutron monitors and a Bonner spheres spectrometry system have been used. In addition, a Monte Carlo model of the installation and the cask has been developed and validated.

Beta-delayed gamma and neutron emission near the double shell closure at 78Ni

International Nuclear Information System (INIS)

Rykaczewski, Krzysztof Piotr; Mazzocchi, C.; Grzywacz, R.; Batchelder, J. C.; Bingham, C.R.; Fong, D.; Hamilton, J.H.; Hwang, J.K.; Karny, M.; Krolas, W.; Liddick, S. N.; Morton, A. C.; Mantica, P. F.; Mueller, W. F.; Steiner, M.; Stolz, A.; Winger, J.A.

2005-01-01

An experiment was performed at the National Superconducting Cyclotron Laboratory at Michigan State University to investigate β decay of very neutron-rich cobalt isotopes. Beta-delayed neutron emission from 71-74 Co has been observed for the first time. Preliminary results are reported

Blood doping by cobalt. Should we measure cobalt in athletes?

Directory of Open Access Journals (Sweden)

Guidi Gian

2006-07-01

Full Text Available Abstract Background Blood doping is commonplace in competitive athletes who seek to enhance their aerobic performances through illicit techniques. Presentation of the hypothesis Cobalt, a naturally-occurring element with properties similar to those of iron and nickel, induces a marked and stable polycythemic response through a more efficient transcription of the erythropoietin gene. Testing the hypothesis Although little information is available so far on cobalt metabolism, reference value ranges or supplementation in athletes, there is emerging evidence that cobalt is used as a supplement and increased serum concentrations are occasionally observed in athletes. Therefore, given the athlete's connatural inclination to experiment with innovative, unfair and potentially unhealthy doping techniques, cobalt administration might soon become the most suited complement or surrogate for erythropoiesis-stimulating substances. Nevertheless, cobalt administration is not free from unsafe consequences, which involve toxic effects on heart, liver, kidney, thyroid and cancer promotion. Implications of the hypothesis Cobalt is easily purchasable, inexpensive and not currently comprehended within the World Anti-Doping Agency prohibited list. Moreover, available techniques for measuring whole blood, serum, plasma or urinary cobalt involve analytic approaches which are currently not practical for antidoping laboratories. Thus more research on cobalt metabolism in athletes is compelling, along with implementation of effective strategies to unmask this potentially deleterious doping practice

Dosimetry of the Embalse nuclear power plant neutron/gamma mixed fields

International Nuclear Information System (INIS)

Salas, C.A.

1990-01-01

The aim of this work is to describe the method used at the Embalse nuclear power plant for carrying out personal dosimetry of the agents affected to the tasks on the Embalse nuclear power plant neutron-gamma mixed fields. (Author) [es

Feasibility study for the investigation of Nitinol self-expanding stents by neutron techniques

International Nuclear Information System (INIS)

Rogante, M.; Pasquini, U.; Rosta, L.; Lebedev, V.

2011-01-01

In this paper, neutron techniques - in particular, small angle neutron scattering (SANS) and neutron diffraction (ND) - are considered for the non-destructive characterization of Nitinol artery stents. This roughly equiatomic (50Ni-50Ti at%) shape memory alloy (SMA) exhibits significant properties of superelasticity and biocompatibility that make it suitable to be typically used as smart material for medical implants and devices. Nitinol self-expanding artery stents, as permanent vascular support structures, supply an ideal option to bypass surgery, but they are submitted for the whole of patient's life to the dynamical stress of the artery pulsation and the aggression from the biological environment. These stents, consequently, can suffer from wear and fracture occurrence likely due to a variety of cyclic fatigue, overload conditions and residual stresses. Neutrons have recently become a progressively more important probe for various materials and components and they allow achieving information complementary to those obtained from the traditional microstructural analyses. The outputs from the preliminary works already carried out in this field consent to consider neutron techniques capable to contribute to the development of these crucial medical implants. The achievable results can yield trends adoptable in monitoring of the stent features. -- Research Highlights: → Neutron techniques can contribute to develop Nitinol self-expanding artery stents. → Neutrons investigations can help avoiding wear and fracture events in Nitinol stents. → Neutron techniques can yield trends adoptable in monitoring of Nitinol stent features. → SANS is able to perform a micro- and nano-scale characterization of Nitinol stents. → Neutron Diffraction helps assessing stresses due to the exercise in Nitinol stents.

Visualization of neutron flux and power distributions in TRIGA Mark II reactor as an educational tool

International Nuclear Information System (INIS)

Snoj, Luka; Ravnik, Matjaz; Lengar, Igor

2008-01-01

Modern Monte Carlo computer codes (e.g. MCNP) for neutron transport allow calculation of detailed neutron flux and power distribution in complex geometries with resolution of ∼1 mm. Moreover they enable the calculation of individual particle tracks, scattering and absorption events. With the use of advanced software for 3D visualization (e.g. Amira, Voxler, etc.) one can create and present neutron flux and power distribution in a 'user friendly' way convenient for educational purposes. One can view axial, radial or any other spatial distribution of the neutron flux and power distribution in a nuclear reactor from various perspectives and in various modalities of presentation. By visualizing the distribution of scattering and absorption events and individual particle tracks one can visualize neutron transport parameters (mean free path, diffusion length, macroscopic cross section, up-scattering, thermalization, etc.) from elementary point of view. Most of the people remember better, if they visualize the processes. Therefore the representation of the reactor and neutron transport parameters is a convenient modern educational tool for the (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. The visualization of neutron flux and power distributions in Jozef Stefan Institute TRIGA Mark II research reactor is treated in the paper. The distributions are calculated with MCNP computer code and presented using Amira and Voxler software. The results in the form of figures are presented in the paper together with comments qualitatively explaining the figures. (authors)

Resonance self-shielding effect in uncertainty quantification of fission reactor neutronics parameters

International Nuclear Information System (INIS)

Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

2014-01-01

In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

Formation and electrochemical investigation of ordered cobalt coordinated peptide monolayers on gold substrates

International Nuclear Information System (INIS)

Wang Xinxin; Nagata, Kenji; Higuchi, Masahiro

2012-01-01

The monolayers composed of cobalt coordinated peptides were prepared on gold substrates by two different approaches. One was the self-assembly method, which was used to prepare a peptide monolayer on the gold substrate via the spontaneous attachment of peptides owing to the interaction between gold and sulfur at the N-terminal of the peptide. The other one was the stepwise polymerization method that was utilized to fabricate the unidirectionally arranged peptide monolayer by the stepwise condensation of amino acids from the initiator fixed on the gold substrate. Leu 2 Ala(4-Pyri)Leu 6 Ala(4-Pyri)Leu 6 sequence was chosen as the cobalt coordinated peptide. The 4-pyridyl alanines, Ala(4-Pyri)s, were introduced as ligands for cobalt to the leucine-rich sequential peptide. The complexation between cobalt and pyridyl groups of the peptide induced the formation of a stable α-helical bundle, which oriented perpendicularly to the substrate surface. In the case of the monolayer fabricated by the stepwise polymerization method, the direction of the peptide macro-dipole moment aligned unidirectionally, and the cobalt complexes were fixed in the monolayer to form the ordered arrangement. On the other hand, the peptides prepared by the self-assembly method formed the mixture of parallel and antiparallel packing owing to the dipole-dipole interaction. The spatial location of the cobalt complexes in the monolayer prepared by the self-assembly method was distorted, compared with that in the monolayer fabricated by the stepwise polymerization method. The vectorial electron flow through the peptide monolayer was achieved by the regular alignment of the peptide macro-dipole moment and the cobalt complexes in the monolayer fabricated by the stepwise polymerization method. - Highlights: ► We fabricated ordered Co coordinated peptide monolayers on the gold substrates. ► The Co complexes in peptide monolayer formed an ordered arrangement of the peptide. ► The peptide macro

Feasibility study of 24-month cycle using enriched Gadolinium as burnable poison for OPR1000

Energy Technology Data Exchange (ETDEWEB)

Moon, Sang-Rae; Shin, Ho-Cheol [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2016-10-15

All of cobalt-60 sources are imported to Korea. CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor's full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly.

Dense granular Flows: a conceptual design of high-power neutron source

Directory of Open Access Journals (Sweden)

Yang Lei

2017-01-01

Full Text Available A high-power neutron source system is very useful for multifunctional applications, such as material facilities for advanced nuclear power, space radiation studies, radiography and tomography. Here the idea of inclined dense granular flow is utilized and developed in a new conceptual design of a compact high-power target to produce a high-energy and high-flux neutron irradiation (the flux is up to 1015 n/cm2/s or even 1016. Comparing to the traditional solid and liquid heavy metal targets, this design has advantages in material choice, fluid stability, heat removal, etc. In this paper the natures of the granular flows in an inclined chute are investigated and preliminary experimental and numerical results are reported. Then the feasibility of this design is discussed.

Thermofluid-neutronic stability of the rotating, fluidized bed, space-power reactor

International Nuclear Information System (INIS)

Lee, C.C.; Jones, O.C.; Becker, M.

1993-01-01

A rotating fluidized bed nuclear reactor has the potential of being a vary attractive option for ultra-high power space systems, especially for propulsion. Research has already examined fuel bed expansion due to variations in state variables, propellant flow rate, and rotational speed, and has also considered problems related to thermal stress. This paper describes the results of a coupled thermofluid-neutronic analysis where perturbations in fuel bed height caused by maneuvering changes in operating conditions alter power levels due to varying absorption of neutrons which would otherwise leak from the system, mainly through the nozzle. This first analysis was not a detailed stability analysis. Rather, it utilized simplified neutronic methods, and was intended to provide an order-of-magnitude assessment of the stability of the reactor with the intention to determine whether or not stability might be a 'concept killer'. Stability was compared with a fixed-fuel-bed reactor of identical geometry for three different cases comprising a set of small, medium and large sizes/powers from 250 MW to 5 GW. It was found that power fluctuations in the fluidized bed reactor were larger by 100 db or more than expected in a packed bed reactor of the same geometry, but never resulted in power excursions. Margins to unit gain in some cases, however, were sufficiently small that the approximations in this quasi-2-dimensional model may not be sufficiently accurate to preclude significant excursions. (orig.)

Neutron-deuteron analyzing power data at En=22.5 MeV

Science.gov (United States)

Weisel, G. J.; Tornow, W.; Crowell, A. S.; Esterline, J. H.; Hale, G. M.; Howell, C. R.; O'Malley, P. D.; Tompkins, J. R.; Witała, H.

2014-05-01

We present measurements of n-d analyzing power, Ay(θ), at En=22.5 MeV. The experiment uses a shielded neutron source which produced polarized neutrons via the 2H(d⃗,n⃗)3He reaction. It also uses a deuterated liquid-scintillator center detector and six pairs of liquid-scintillator neutron side detectors. Elastic neutron scattering events are identified by using time-of-flight techniques and by setting a window in the center detector pulse-height spectrum. The beam polarization is monitored by using a high-pressure helium gas cell and an additional pair of liquid-scintillator side detectors. The n-d Ay(θ) data were corrected for finite-geometry and multiple-scattering effects using a Monte Carlo simulation of the experiment. The 22.5-MeV data demonstrate that the three-nucleon analyzing power puzzle also exists at this energy. They show a significant discrepancy with predictions of high-precision nucleon-nucleon potentials alone or combined with Tucscon-Melbourne or Urbana IX three-nucleon forces, as well as currently available effective-field theory based potentials of next-to-next-to-next-to-leading order.

Neutron fluctuations in accelerator driven and power reactors via backward master equations

International Nuclear Information System (INIS)

Zhifeng Kuang

2000-05-01

The transport of neutrons in a reactor is a random process, and thus the number of neutrons in a reactor is a random variable. Fluctuations in the number of neutrons in a reactor can be divided into two categories, namely zero noise and power reactor noise. As the name indicates, they dominate (i.e. are observable) at different power levels. The reasons for their occurrences and utilization are also different. In addition, they are described via different mathematical tools, namely master equations and the Langevin equation, respectively. Zero noise carries information about some nuclear properties such as reactor reactivity. Hence methods such as Feynman- and Rossi-alpha methods have been established to determine the subcritical reactivity of a subcritical system. Such methods received a renewed interest recently with the advent of the so-called accelerator driven systems (ADS). Such systems, intended to be used either for energy production or transuranium transmutation, will use a subcritical core with a strong spallation source. A spallation source has statistical properties that are different from those of the traditionally used radioactive sources which were also assumed in the derivation of the Feynman- and Rossi-alpha formulae. Therefore it is necessary to re-derive the Feynman- and Rossi-alpha formulae. Such formulae for ADS have been derived recently but in simpler neutronic models. One subject of this thesis is the extension of such formulae to a more general case in which six groups of delayed neutron precursors are taken into account, and the full joint statistics of the prompt and all delayed groups is included. The involved complexity problems are solved with a combination of effective analytical techniques and symbolic algebra codes. Power reactor noise carries information about parametric perturbation of the system. Langevin technique has been used to extract such information. In such a treatment, zero noise has been neglected. This is a pragmatic

A two-dimensional simulator of the neutronic behaviour of low power fast reactors

International Nuclear Information System (INIS)

Penha, M.A.V.R. da.

1984-01-01

A model to simulate the temporal neutronic behaviour of fast breeder reactors was developed. The effective cross-sections are corrected, whenever the reactor state change; by using linear correlations and interpolation schemes with data contained in a library previously compiled. This methodology was coupled with a simplified spatial neutronic calculation to investigate the temporal behaviour of neutronic parameters such as breeding gain, flux and power. (Author) [pt

Cobalt release from inexpensive jewellery

DEFF Research Database (Denmark)

Thyssen, Jacob Pontoppidan; Jellesen, Morten Stendahl; Menné, Torkil

2010-01-01

. Conclusions: This study showed that only a minority of inexpensive jewellery purchased in Denmark released cobalt when analysed with the cobalt spot test. As fashion trends fluctuate and we found cobalt release from dark appearing jewellery, cobalt release from consumer items should be monitored in the future......Objectives: The aim was to study 354 consumer items using the cobalt spot test. Cobalt release was assessed to obtain a risk estimate of cobalt allergy and dermatitis in consumers who would wear the jewellery. Methods: The cobalt spot test was used to assess cobalt release from all items...

Self-biased cobalt ferrite nanocomposites for microwave applications

International Nuclear Information System (INIS)

Hannour, Abdelkrim; Vincent, Didier; Kahlouche, Faouzi; Tchangoulian, Ardaches; Neveu, Sophie; Dupuis, Vincent

2014-01-01

Oriented CoFe 2 O 4 nanoparticles, dispersed in polymethyl methacrylate (PMMA) matrix, were fabricated by magnetophoretic deposition of functionalized nanocolloidal cobalt ferrite particles into porous alumina membrane. Their magnetic behavior exhibits an out-of-plane easy axis with a large remanent magnetization and coercitivity. This orientation allows high effective internal magnetic anisotropy that contributes to the permanent bias along the wire axis. The microwave studies reveal a ferromagnetic resonance at 46.5 and 49.5 GHz, depending on the filling ratio of the membrane. Ansoft High Frequency Structure Simulator (Ansoft HFSS) simulations are in good agreement with experimental results. Such nanocomposite is presented as one of the promising candidates for microwave devices (circulators, isolators, noise suppressors etc.). - Highlights: • Oriented magnetic CoFe 2 O 4 nanoparticles were fabricated by magnetophoretic deposition of functionalized cobalt ferrite particles into porous alumina membrane. • The nanocomposite obtained presents an out-of-plane easy axis with a large remanent magnetization and coercitivity. • The high effective internal magnetic anisotropy contributes to the permanent bias along the wire axis. • The frequency ferromagnetic resonance ranges from 46.5 to 49.5 GHz, depending on the filling ratio of the membrane. • We have obtained a good agreement between Ansoft High Frequency Structure Simulator simulations and experimental results

RESONANCE SELF-SHIELDING EFFECT IN UNCERTAINTY QUANTIFICATION OF FISSION REACTOR NEUTRONICS PARAMETERS

Directory of Open Access Journals (Sweden)

GO CHIBA

2014-06-01

Full Text Available In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

Neutron generator power supply modeling in EMMA

International Nuclear Information System (INIS)

Robinson, A.C.; Farnsworth, A.V.; Montgomery, S.T.; Peery, J.S.; Merewether, K.O.

1996-01-01

Sandia National Laboratories has prime responsibility for neutron generator design and manufacturing, and is committed to developing predictive tools for modeling neutron generator performance. An important aspect of understanding component performance is explosively driven ferroelectric power supply modeling. EMMA (ElectroMechanical Modeling in ALEGRA) is a three dimensional compile time version of Sandia's ALEGRA code. The code is built on top of the general ALEGRA framework for parallel shock-physics computations but also includes additional capability for modeling the electric potential field in dielectrics. The overall package includes shock propagation due to explosive detonation, depoling of ferroelectric ceramics, electric field calculation and coupling with a general lumped element circuit equation system. The AZTEC parallel iterative solver is used to solve for the electric potential. The DASPK differential algebraic equation package is used to solve the circuit equation system. Sample calculations are described

First observations of power MOSFET burnout with high energy neutrons

International Nuclear Information System (INIS)

Oberg, D.L.; Wert, J.L.; Normand, E.; Majewski, P.P.; Wender, S.A.

1996-01-01

Single event burnout was seen in power MOSFETs exposed to high energy neutrons. Devices with rated voltage ≥400 volts exhibited burnout at substantially less than the rated voltage. Tests with high energy protons gave similar results. Burnout was also seen in limited tests with lower energy protons and neutrons. Correlations with heavy-ion data are discussed. Accelerator proton data gave favorable comparisons with burnout rates measured on the APEX spacecraft. Implications for burnout at lower altitudes are also discussed

Neutron flux measurements in C-9 capsule pressure tube

International Nuclear Information System (INIS)

Barbos, D.; Roth, C. S.; Gugiu, D.; Preda, M.

2001-01-01

C-9 capsule is a fuel testing facility in which the testing consists of a daily cycle ranging between the limits 100% power to 50% power. C-9 in-pile section with sample holder an instrumentation are introduced in G-9 and G-10 experimental channels. The experimental fuel channel has a maximum value when the in-pile section (pressure tube) is in G-9 channel and minimum value in G-10 channel. In this paper the main goals are determination or measurements of: - axial thermal neutron flux distribution in C-9 pressure tube both in G-9 and G-10 channel; - ratio of maximum neutron flux value in G-9 and the same value in G-9 channel and the same value in G-10 channel; - neutron flux-spectrum. On the basis of axial neutron flux distribution measurements, the experimental fuel element in sample holder position in set. Both axial neutron flux distribution of thermal neutrons and neutron flux-spectrum were performed using multi- foil activation technique. Activation rates were obtained by absolute measurements of the induced activity using gamma spectroscopy methods. To determine the axial thermal neutron flux distribution in G-9 and G-10, Cu 100% wire was irradiated at the reactor power of 2 MW. Ratio between the two maximum values, in G-9 and G-10 channels, is 2.55. Multi-foil activation method was used for neutron flux spectrum measurements. The neutron spectra and flux were obtained from reaction rate measurements by means of SAND 2 code. To obtain gamma-ray spectra, a HPGe detector connected to a multichannel analyzer was used. The spectrometer is absolute efficiency calibrated. The foils were irradiated at 2 MW reactor power in previously determined maximum flux position resulted from wire measurements. This reaction rates were normalized for 10 MW reactor power. Neutron self shielding corrections for the activation foils were applied. The self-shielding corrections are computed using Monte Carlo simulation methods. The measured integral flux is 1.1·10 14 n/cm 2 s

Neutron-nucleus interactions and fission. Chapter 1

International Nuclear Information System (INIS)

1998-01-01

The central problem in nuclear-reactor kinetics is to predict the evolution in time of the neutron population in a multiplying medium. Point kinetics allows study of the global behaviour of the neutron population from the average properties of the medium. Before tackling, in the following chapters, the equations governing the time variation of the reactor power (proportional to the total neutron population), the properties of a neutron-multiplying medium shall be discussed briefly. After recalling a number of definitions, a qualitative description shall be given of the principal nuclear reactions at play in a self-sustaining chain reaction, with emphasis on the source of fission neutrons. Since delayed neutrons play a crucial role in reactor kinetics, their production in a reactor shall be described in greater detail. (author)

Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry

International Nuclear Information System (INIS)

Rossa, Riccardo; Borella, Alessandro; Labeau, Pierre-Etienne; Pauly, Nicolas; Meer, Klaas van der

2015-01-01

The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive non-destructive assay (NDA) technique that is proposed for the direct measurement of 239 Pu in a spent fuel assembly. The insertion of neutron detectors wrapped with different neutron absorbing materials, or neutron filters, in the central guide tube of a PWR fuel assembly is envisaged to measure the neutron flux in the energy region close to the 0.3 eV resonance of 239 Pu. In addition, the measurement of the fast neutron flux is foreseen. This paper is focused on the determination of the Gd and Cd neutron filters thickness to maximize the detection of neutrons within the resonance region. Moreover, several detector types are compared to identify the optimal condition and to assess the expected total neutron counts that can be obtained with the SINRD measurements. Results from Monte Carlo simulations showed that ranges between 0.1–0.3 mm and 0.5–1.0 mm ensure the optimal conditions for the Gd and Cd filters, respectively. Moreover, a 239 Pu fission chamber is better suited to measure neutrons close to the 0.3 eV resonance and it has the highest sensitivity to 239 Pu, in comparison with a 235 U fission chamber, with a 3 He proportional counter, and with a 10 B proportional counter. The use of a thin Gd filter and a thick Cd filter is suggested for the 239 Pu and 235 U fission chambers to increase the total counts achieved in a measurement, while a thick Gd filter and a thin Cd filter are envisaged for the 3 He and 10 B proportional counters to increase the sensitivity to 239 Pu. We concluded that an optimization process that takes into account measurement time, filters thickness, and detector size is needed to develop a SINRD detector that can meet the requirement for an efficient verification of spent fuel assemblies

Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry

Energy Technology Data Exchange (ETDEWEB)

Rossa, Riccardo, E-mail: rrossa@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium); Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Borella, Alessandro, E-mail: aborella@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium); Labeau, Pierre-Etienne, E-mail: pelabeau@ulb.ac.be [Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Pauly, Nicolas, E-mail: nipauly@ulb.ac.be [Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels (Belgium); Meer, Klaas van der, E-mail: kvdmeer@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol (Belgium)

2015-08-11

The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive non-destructive assay (NDA) technique that is proposed for the direct measurement of {sup 239}Pu in a spent fuel assembly. The insertion of neutron detectors wrapped with different neutron absorbing materials, or neutron filters, in the central guide tube of a PWR fuel assembly is envisaged to measure the neutron flux in the energy region close to the 0.3 eV resonance of {sup 239}Pu. In addition, the measurement of the fast neutron flux is foreseen. This paper is focused on the determination of the Gd and Cd neutron filters thickness to maximize the detection of neutrons within the resonance region. Moreover, several detector types are compared to identify the optimal condition and to assess the expected total neutron counts that can be obtained with the SINRD measurements. Results from Monte Carlo simulations showed that ranges between 0.1–0.3 mm and 0.5–1.0 mm ensure the optimal conditions for the Gd and Cd filters, respectively. Moreover, a {sup 239}Pu fission chamber is better suited to measure neutrons close to the 0.3 eV resonance and it has the highest sensitivity to {sup 239}Pu, in comparison with a {sup 235}U fission chamber, with a {sup 3}He proportional counter, and with a {sup 10}B proportional counter. The use of a thin Gd filter and a thick Cd filter is suggested for the {sup 239}Pu and {sup 235}U fission chambers to increase the total counts achieved in a measurement, while a thick Gd filter and a thin Cd filter are envisaged for the {sup 3}He and {sup 10}B proportional counters to increase the sensitivity to {sup 239}Pu. We concluded that an optimization process that takes into account measurement time, filters thickness, and detector size is needed to develop a SINRD detector that can meet the requirement for an efficient verification of spent fuel assemblies.

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

Science.gov (United States)

Bowman, Charles D.

1992-01-01

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

Science.gov (United States)

Bowman, C.D.

1992-11-03

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

Self-induced inverse spin-Hall effect in an iron and a cobalt single-layer films themselves under the ferromagnetic resonance

Science.gov (United States)

Kanagawa, Kazunari; Teki, Yoshio; Shikoh, Eiji

2018-05-01

The inverse spin-Hall effect (ISHE) is produced even in a "single-layer" ferromagnetic material film. Previously, the self-induced ISHE in a Ni80Fe20 film under the ferromagnetic resonance (FMR) was discovered. In this study, we observed an electromotive force (EMF) in an iron (Fe) and a cobalt (Co) single-layer films themselves under the FMR. As origins of the EMFs in the films themselves, the ISHE was main for Fe and dominant for Co, respectively 2 and 18 times larger than the anomalous Hall effect. Thus, we demonstrated the self-induced ISHE in an Fe and a Co single-layer films themselves under the FMR.

Evaluation of the neutron self-interrogation approach for assay of plutonium in high materials

International Nuclear Information System (INIS)

Russo, P.A.; Menlove, H.O.; Fife, K.W.; West, M.H.

1987-01-01

The pyrochemical scrap recovery processes, designed to extract impurities from plutonium metal and compounds, generate a variety of plutonium-laden residues consisting of high (α,n) matrices of varying chemical composition, and often containing grams to tens of grams of americium. For such materials, multiplication corrections based on real neutron coincidence count rate, R, and total neutron count rate, T, measurements cannot be applied because of the large, unknown, and variable (α,n) component in the total neutron emission rate. A study of the prototype self-interrogation assay method is in progress at the Los Alamos plutonium facility. In the self-interrogation approach, the assay signature R(IF)/T is a function of effective fissile plutonium content, where R(IF) is the induced fission component of the measured reals rate, and T is the measured, (α,n)-dominated totals rate. The present study includes a calibration effort using standards consisting of mixtures of PuO 2 and PuF 4 in a salt-strip matrix. The neutron measurements of the standards and the process materials have been performed at the Los Alamos Plutonium Facility. The precision and accuracy of the self-interrogation method applied to pyrochemical residues is examined in this study

COBALT: Development of a Platform to Flight Test Lander GN&C Technologies on Suborbital Rockets

Science.gov (United States)

Carson, John M., III; Seubert, Carl R.; Amzajerdian, Farzin; Bergh, Chuck; Kourchians, Ara; Restrepo, Carolina I.; Villapando, Carlos Y.; O'Neal, Travis V.; Robertson, Edward A.; Pierrottet, Diego;

Detection of low caloric power of coal by pulse fast-thermal neutron analysis

International Nuclear Information System (INIS)

Gu De-shan; Sang Hai-feng; Qiao Shuang; Liu Yu-ren, Liu Lin-mao; Jing Shi-wei; Chinese Academy of Sciences, Changchun

2004-01-01

Analysis method and principle of pulse fast-thermal neutron analysis (PFTNA) are introduced. A system for the measurement of low caloric power of coal by PFTNA is also presented. The 14 MeV pulse neutron generator and BGO detector and 4096 MCA were applied in this system. A multiple linear regression method applied to the data solved the interferential problem of multiple elements. The error of low caloric power between chemical analysis and experiment was less than 0.4 MJ/kg. (author)

Theory of neutron resonance cross sections for safety applications

International Nuclear Information System (INIS)

Froehner, F.H.

1992-09-01

Neutron resonances exert a strong influence on the behaviour of nuclear reactors, especially on their response to the temperature changes accompanying power excursions, and also on the efficiency of shielding materials. The relevant theory of neutron resonance cross sections including the practically important approximations is reviewed, both for the resolved and the unresolved resonance region. Numerical techniques for Doppler broadening of resonances are presented, and the construction of group constants and especially of self-shielding factors for neutronics calculations is outlined. (orig.) [de

Nanogenerators for Self-Powered Gas Sensing

Science.gov (United States)

Wen, Zhen; Shen, Qingqing; Sun, Xuhui

2017-10-01

Looking toward world technology trends over the next few decades, self-powered sensing networks are a key field of technological and economic driver for global industries. Since 2006, Zhong Lin Wang's group has proposed a novel concept of nanogenerators (NGs), including piezoelectric nanogenerator and triboelectric nanogenerator, which could convert a mechanical trigger into an electric output. Considering motion ubiquitously exists in the surrounding environment and for any most common materials used every day, NGs could be inherently served as an energy source for our daily increasing requirements or as one of self-powered environmental sensors. In this regard, by coupling the piezoelectric or triboelectric properties with semiconducting gas sensing characterization, a new research field of self-powered gas sensing has been proposed. Recent works have shown promising concept to realize NG-based self-powered gas sensors that are capable of detecting gas environment without the need of external power sources to activate the gas sensors or to actively generate a readout signal. Compared with conventional sensors, these self-powered gas sensors keep the approximate performance. Meanwhile, these sensors drastically reduce power consumption and additionally reduce the required space for integration, which are significantly suitable for the wearable devices. This paper gives a brief summary about the establishment and latest progress in the fundamental principle, updated progress and potential applications of NG-based self-powered gas sensing system. The development trend in this field is envisaged, and the basic configurations are also introduced.

Neutronic scoping studies for the tokamak experimental power reactor

International Nuclear Information System (INIS)

Santoro, R.T.; Bettis, E.S.; McAlees, D.G.; Watts, H.L.; Williams, M.L.

1976-02-01

One-dimensional neutron and photon radiation transport methods have been used to investigate candidate blanket configurations and compositions for use in the Tokamak Experimental Power Reactor. Seven blanket designs are compared in terms of energy recovery, radiation attenuation, potential radiation damage, and, where applicable, tritium breeding

Cobalt release from inexpensive jewellery: has the use of cobalt replaced nickel following regulatory intervention?

Science.gov (United States)

Thyssen, Jacob Pontoppidan; Jellesen, Morten S; Menné, Torkil; Lidén, Carola; Julander, Anneli; Møller, Per; Johansen, Jeanne Duus

2010-08-01

Before the introduction of the EU Nickel Directive, concern was raised that manufacturers of jewellery might turn from the use of nickel to cobalt following the regulatory intervention on nickel exposure. The aim was to study 354 consumer items using the cobalt spot test. Cobalt release was assessed to obtain a risk estimate of cobalt allergy and dermatitis in consumers who would wear the jewellery. The cobalt spot test was used to assess cobalt release from all items. Microstructural characterization was made using scanning electron microscope (SEM) and energy-dispersive spectroscopy (EDS). Cobalt release was found in 4 (1.1%) of 354 items. All these had a dark appearance. SEM/EDS was performed on the four dark appearing items which showed tin-cobalt plating on these. This study showed that only a minority of inexpensive jewellery purchased in Denmark released cobalt when analysed with the cobalt spot test. As fashion trends fluctuate and we found cobalt release from dark appearing jewellery, cobalt release from consumer items should be monitored in the future. Industries may not be fully aware of the potential cobalt allergy problem.

Fusion Power Measurement Using a Combined Neutron Spectrometer-Camera System at ITER

International Nuclear Information System (INIS)

Sjoestrand, Henrik; Sunden, E. Andersson; Conroy, S.; Ericsson, G.; Johnson, M. Gatu; Giacomelli, L.; Hellesen, C.; Hjalmarsson, A.; Ronchi, E.; Weiszflog, M.; Kaellne, J.

2008-01-01

A central task for fusion plasma diagnostics is to measure the 2.5 and 14 MeV neutron emission rate in order to determine the fusion power. A new method for determining the neutron yield has been developed at JET. It makes use of the magnetic proton recoil neutron spectrometer and a neutron camera and provides the neutron yield with small systematic errors. At ITER a similar system could operate if a high-resolution, high-performance neutron spectrometer similar to the MPR was installed. In this paper, we present how such system could be implemented and how well it would perform under different assumption of plasma scenarios and diagnostic capabilities. It is found that the systematic uncertainty for using such a system as an absolute calibration reference is as low as 3% and hence it would be an excellent candidate for the calibration of neutron monitors such as fission chambers. It is also shown that the system could provide a 1 ms time resolved estimation of the neutron rate with a total uncertainty of 5%

Experiments at the GELINA facility for the validation of the self-indication neutron resonance densitometry technique

Directory of Open Access Journals (Sweden)

Rossa Riccardo

2017-01-01

Full Text Available Self-Indication Neutron Resonance Densitometry (SINRD is a passive non-destructive method that is being investigated to quantify the 239Pu content in a spent fuel assembly. The technique relies on the energy dependence of total cross sections for neutron induced reaction. The cross sections show resonance structures that can be used to quantify the presence of materials in objects, e.g. the total cross-section of 239Pu shows a strong resonance close to 0.3 eV. This resonance will cause a reduction of the number of neutrons emitted from spent fuel when 239Pu is present. Hence such a reduction can be used to quantify the amount of 239Pu present in the fuel. A neutron detector with a high sensitivity to neutrons in this energy region is used to enhance the sensitivity to 239Pu. This principle is similar to self-indication cross section measurements. An appropriate detector can be realized by surrounding a 239Pu-loaded fission chamber with appropriate neutron absorbing material. In this contribution experiments performed at the GELINA time-of-flight facility of the JRC at Geel (Belgium to validate the simulations are discussed. The results confirm that the strongest sensitivity to the target material was achieved with the self-indication technique, highlighting the importance of using a 239Pu fission chamber for the SINRD measurements.

Cobalt release from implants and consumer items and characteristics of cobalt sensitized patients with dermatitis

DEFF Research Database (Denmark)

Thyssen, Jacob Pontoppidan; Menne, Torkil; Liden, Carola

2012-01-01

-containing dental alloys and revised hip implant components.Results. Six of eight dental alloys and 10 of 98 revised hip implant components released cobalt in the cobalt spot test, whereas none of 50 mobile phones gave positive reactions. The clinical relevance of positive cobalt test reactions was difficult......-tested dermatitis patients in an attempt to better understand cobalt allergy.Materials and methods. 19 780 dermatitis patients aged 4-99 years were patch tested with nickel, chromium or cobalt between 1985 and 2010. The cobalt spot test was used to test for cobalt ion release from mobile phones as well as cobalt...

Self-powered radiation detectors

International Nuclear Information System (INIS)

Goldstein, N.P.; Todt, W.H.

1982-01-01

A self-powered nuclear radiation detector has an emitter electrode of an alloy of a first major constituent metal having a desired high radiation response, and a second minor constituent which imparts to the alloy a desired thermal or mechanical characteristic without diminishing the desired high radiation response. A gamma responsive self-powered detector is detailed which has an emitter with lead as the major constituent, with the minor constituent selected from aluminum, copper, nickel, platinum, or zinc. (author)

Neutron dosimetry at nuclear power plants with light water reactors (LWR)

International Nuclear Information System (INIS)

Hofmann, B.; Schwarz, W.; Burgkhardt, B.; Piesch, E.

1989-02-01

During nuclear start-up of the Muelheim-Kaerlich nuclear power plant in 1986 the neutron radiation fields in the primary and auxiliary component rooms of the containment were investigated using the Single Sphere Albedo Technique and additional measurement techniques. For personnel monitoring albedo neutron dosemeters were used consisting of thermoluminescent detectors and track etch detectors combined with boron converters. Results: (1) The neutron radiation fields reach dose rate values up to 1000 mSv/h at the sleeves of the reactor coolant pipes, in the refuelling pool and the reactor cavity sump. The neutron component varies between 10% in the steam generator rooms up to 92% in the refuelling pool. (2) The mean value of the effective neutron energy at the different locations was found to be about 100 keV. Thermal neutrons contribute with about 10% to the area dose. (3) By direct intercomparisons and different evaluation methods of the Single Sphere Albedo Dosemeter it was shown, that rem-counters used within routine monitoring in the mixed radiation fields of the LWR overestimate the neutron dose rate only insignificantly (+20%) and are therefore usable for practical radiation protection work. (4) The sensitivity of albedo neutron dosemeters allows the detection of neutrons above 10 μSv. The contribution of neutrons to the total personnel dose was 25% in maximum. For the evaluation of albedo detectors a constant calibration factor can be applied. (orig./HP) [de

Self-biased cobalt ferrite nanocomposites for microwave applications

Energy Technology Data Exchange (ETDEWEB)

Hannour, Abdelkrim, E-mail: abdelkrim.hannour@hotmail.com [LT2C Laboratory, Jean-Monnet University, 25 rue Dr. Rémy Annino, F-42000, Saint-Etienne (France); Vincent, Didier; Kahlouche, Faouzi; Tchangoulian, Ardaches [LT2C Laboratory, Jean-Monnet University, 25 rue Dr. Rémy Annino, F-42000, Saint-Etienne (France); Neveu, Sophie; Dupuis, Vincent [UPMC Univ Paris 06, UMR 7195, PECSA, F-75005, Paris (France)

2014-03-15

Oriented CoFe{sub 2}O{sub 4} nanoparticles, dispersed in polymethyl methacrylate (PMMA) matrix, were fabricated by magnetophoretic deposition of functionalized nanocolloidal cobalt ferrite particles into porous alumina membrane. Their magnetic behavior exhibits an out-of-plane easy axis with a large remanent magnetization and coercitivity. This orientation allows high effective internal magnetic anisotropy that contributes to the permanent bias along the wire axis. The microwave studies reveal a ferromagnetic resonance at 46.5 and 49.5 GHz, depending on the filling ratio of the membrane. Ansoft High Frequency Structure Simulator (Ansoft HFSS) simulations are in good agreement with experimental results. Such nanocomposite is presented as one of the promising candidates for microwave devices (circulators, isolators, noise suppressors etc.). - Highlights: • Oriented magnetic CoFe{sub 2}O{sub 4} nanoparticles were fabricated by magnetophoretic deposition of functionalized cobalt ferrite particles into porous alumina membrane. • The nanocomposite obtained presents an out-of-plane easy axis with a large remanent magnetization and coercitivity. • The high effective internal magnetic anisotropy contributes to the permanent bias along the wire axis. • The frequency ferromagnetic resonance ranges from 46.5 to 49.5 GHz, depending on the filling ratio of the membrane. • We have obtained a good agreement between Ansoft High Frequency Structure Simulator simulations and experimental results.

Monitoring of the Irradiated Neutron Fluence in the Neutron Transmutation Doping Process of Hanaro

Science.gov (United States)

Kim, Myong-Seop; Park, Sang-Jun

2009-08-01

Neutron transmutation doping (NTD) for silicon is a process of the creation of phosphorus impurities in intrinsic or extrinsic silicon by neutron irradiation to obtain silicon semiconductors with extremely uniform dopant distribution. HANARO has two vertical holes for the NTD, and the irradiation for 5 and 6 inch silicon ingots has been going on at one hole. In order to achieve the accurate neutron fluence corresponding to the target resistivity, the real time neutron flux is monitored by self-powered neutron detectors. After irradiation, the total irradiation fluence is confirmed by measuring the absolute activity of activation detectors. In this work, a neutron fluence monitoring method using zirconium foils with the mass of 10 ~ 50 mg was applied to the NTD process of HANARO. We determined the proportional constant of the relationship between the resistivity of the irradiated silicon and the neutron fluence determined by using zirconium foils. The determined constant for the initially n-type silicon was 3.126 × 1019 n·Ω/cm. It was confirmed that the difference between this empirical value and the theoretical one was only 0.5%. Conclusively, the practical methodology to perform the neutron transmutation doping of silicon was established.

Design innovations in neutron and gamma detectors

International Nuclear Information System (INIS)

Prasad, K.R.

2003-01-01

Neutron and gamma radiation needs to be monitored in most nuclear installations since it is highly penetrating. On-line monitoring of these radiations is very important for the safe and controlled operation of nuclear reactors, accelerators etc. Several design innovations have been carried out on gas ionisation detectors such as boron-lined proportional counters and ion chambers, fission detectors, gamma ion chambers as well as self-powered detectors. The use of additional structures within boron-lined detectors has enhanced their neutron sensitivity without a corresponding increase in the unwanted gamma sensitivity. The neutron sensitivity of fission counters can be enhanced by designing them as transmission line devices. Ion chambers with two and six pairs of electrodes have been developed for monitoring pulsed x-ray background at accelerator areas. Ion chambers have been employed at gamma fields up to 80 kR/h by deriving the exposure levels on-line using microcontroller devices programmed on the basis of theoretical and empirical formulas. The use of gas electron multiplier foils is proposed for charge multiplication in ion chambers. Self-powered detectors with new emitter materials like Hi, Ni and Inconel have been developed. (author)

Use of the Power Burst Facility for boron neutron capture therapy

International Nuclear Information System (INIS)

Crocker, J.G.; Griebenow, M.L.; Leatham, J.

1990-01-01

A program is under development at the Idaho National Engineering Laboratory (INEL) that involves using the Power Burst Facility (PBF) for research into boron neutron capture therapy (BNCT). BNCT utilizes the ionizing energy from boron-neutron capture to stop reproduction of or destroy cells in cancerous tissue in a two-step process. The first step is to selectively concentrate a boron isotope within the tumor cell, that when activated by neutron capture emits highly ionizing, short range particles. The second step involves activation of the isotope only in the vicinity of the tumor with a narrow neutron beam. The ( 10 B[n, 4 He] 7 Li) reaction with thermal neutrons produces fission products with track lengths approximately equal to a cell diameter. The INEL program includes the modification of the PBF by the addition of a filter and treatment area. The filter will down-scatter high energy neutrons into the epithermal range and remove thermal neutrons and excessively damaging gamma components. The intense source of epithermal neutrons from PBF is considered necessary to achieve optimum therapy for deep-seated tumors with minimum damage to surface tissue. THe neutron filter conceptualized for PBF utilizes aluminum and heavy water to down-scatter neutrons into the proper energy range. Bismuth will be used for gamma shielding and cadmium will remove the thermal neutron contaminant from the beam. The INEL program leads to human clinical trials at PBF which are intended to prove that brain tumors can be successfully treated through noninvasive techniques. Further research into BNCT at PBF for other cancer types is also anticipated

Solution of the multigroup neutron diffusion Eigenvalue problem in slab geometry by modified power method

Energy Technology Data Exchange (ETDEWEB)

Zanette, Rodrigo [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pós-Graduação em Matemática Aplicada; Petersen, Claudio Z.; Tavares, Matheus G., E-mail: rodrigozanette@hotmail.com, E-mail: claudiopetersen@yahoo.com.br, E-mail: matheus.gulartetavares@gmail.com [Universidade Federal de Pelotas (UFPEL), RS (Brazil). Programa de Pós-Graduação em Modelagem Matemática

2017-07-01

We describe in this work the application of the modified power method for solve the multigroup neutron diffusion eigenvalue problem in slab geometry considering two-dimensions for nuclear reactor global calculations. It is well known that criticality calculations can often be best approached by solving eigenvalue problems. The criticality in nuclear reactors physics plays a relevant role since establishes the ratio between the numbers of neutrons generated in successive fission reactions. In order to solve the eigenvalue problem, a modified power method is used to obtain the dominant eigenvalue (effective multiplication factor (K{sub eff})) and its corresponding eigenfunction (scalar neutron flux), which is non-negative in every domain, that is, physically relevant. The innovation of this work is solving the neutron diffusion equation in analytical form for each new iteration of the power method. For solve this problem we propose to apply the Finite Fourier Sine Transform on one of the spatial variables obtaining a transformed problem which is resolved by well-established methods for ordinary differential equations. The inverse Fourier transform is used to reconstruct the solution for the original problem. It is known that the power method is an iterative source method in which is updated by the neutron flux expression of previous iteration. Thus, for each new iteration, the neutron flux expression becomes larger and more complex due to analytical solution what makes propose that it be reconstructed through an polynomial interpolation. The methodology is implemented to solve a homogeneous problem and the results are compared with works presents in the literature. (author)

Neutron-deuteron analyzing power data at 19.0 MeV

International Nuclear Information System (INIS)

Weisel, G. J.; Tornow, W.; Crowe, B. J. III; Crowell, A. S.; Esterline, J. H.; Howell, C. R.; Kelley, J. H.; Macri, R. A.; Pedroni, R. S.; Walter, R. L.; Witala, H.

2010-01-01

Measurements of neutron-deuteron (n-d) analyzing power A y (θ) at E n =19.0 MeV are reported at 16 angles from θ c.m. =46.7 to 152.0 deg. The objective of the experiment is to better characterize the discrepancies between n-d data and the predictions of three-nucleon calculations for neutron energies above 16.0 MeV. The experiment used a shielded neutron source, which produced polarized neutrons via the 2 H(d-vector,n-vector) 3 He reaction, a deuterated liquid scintillator center detector (CD) and liquid-scintillator neutron side detectors. A coincidence between the CD and the side detectors isolated the elastic-scattering events. The CD pulse height spectrum associated with each side detector was sorted by using pulse-shape discrimination, time-of-flight techniques, and by removing accidental coincidences. A Monte Carlo computer simulation of the experiment accounted for effects due to finite geometry, multiple scattering, and CD edge effects. The resulting high-precision data (with absolute uncertainties ranging from 0.0022 to 0.0132) have a somewhat lower discrepancy with the predictions of three-body calculations, as compared to those found at lower energies.

NEUTRON-STAR MERGER EJECTA AS OBSTACLES TO NEUTRINO-POWERED JETS OF GAMMA-RAY BURSTS

Energy Technology Data Exchange (ETDEWEB)

Just, O.; Janka, H.-T.; Schwarz, N. [Max-Planck-Institut für Astrophysik, Karl-Schwarzschild-Str. 1, D-85748 Garching (Germany); Obergaulinger, M. [Departament d´Astronomia i Astrofísica, Universitat de València, Edifici d´Investigació Jeroni Muñoz, C/ Dr. Moliner, 50, E-46100 Burjassot (València) (Spain); Bauswein, A., E-mail: ojust@mpa-garching.mpg.de, E-mail: thj@mpa-garching.mpg.de [Department of Physics, Aristotle University of Thessaloniki, 54124 Thessaloniki (Greece)

2016-01-10

We present the first special relativistic, axisymmetric hydrodynamic simulations of black hole-torus systems (approximating general relativistic gravity) as remnants of binary-neutron star (NS–NS) and neutron star–black hole (NS–BH) mergers, in which the viscously driven evolution of the accretion torus is followed with self-consistent energy-dependent neutrino transport and the interaction with the cloud of dynamical ejecta expelled during the NS–NS merging is taken into account. The modeled torus masses, BH masses and spins, and the ejecta masses, velocities, and spatial distributions are adopted from relativistic merger simulations. We find that energy deposition by neutrino annihilation can accelerate outflows with initially high Lorentz factors along polar low-density funnels, but only in mergers with extremely low baryon pollution in the polar regions. NS–BH mergers, where polar mass ejection during the merging phase is absent, provide sufficiently baryon-poor environments to enable neutrino-powered, ultrarelativistic jets with terminal Lorentz factors above 100 and considerable dynamical collimation, favoring short gamma-ray bursts (sGRBs), although their typical energies and durations might be too small to explain the majority of events. In the case of NS–NS mergers, however, neutrino emission of the accreting and viscously spreading torus is too short and too weak to yield enough energy for the outflows to break out from the surrounding ejecta shell as highly relativistic jets. We conclude that neutrino annihilation alone cannot power sGRBs from NS–NS mergers.

NEUTRON-STAR MERGER EJECTA AS OBSTACLES TO NEUTRINO-POWERED JETS OF GAMMA-RAY BURSTS

International Nuclear Information System (INIS)

Just, O.; Janka, H.-T.; Schwarz, N.; Obergaulinger, M.; Bauswein, A.

2016-01-01

We present the first special relativistic, axisymmetric hydrodynamic simulations of black hole-torus systems (approximating general relativistic gravity) as remnants of binary-neutron star (NS–NS) and neutron star–black hole (NS–BH) mergers, in which the viscously driven evolution of the accretion torus is followed with self-consistent energy-dependent neutrino transport and the interaction with the cloud of dynamical ejecta expelled during the NS–NS merging is taken into account. The modeled torus masses, BH masses and spins, and the ejecta masses, velocities, and spatial distributions are adopted from relativistic merger simulations. We find that energy deposition by neutrino annihilation can accelerate outflows with initially high Lorentz factors along polar low-density funnels, but only in mergers with extremely low baryon pollution in the polar regions. NS–BH mergers, where polar mass ejection during the merging phase is absent, provide sufficiently baryon-poor environments to enable neutrino-powered, ultrarelativistic jets with terminal Lorentz factors above 100 and considerable dynamical collimation, favoring short gamma-ray bursts (sGRBs), although their typical energies and durations might be too small to explain the majority of events. In the case of NS–NS mergers, however, neutrino emission of the accreting and viscously spreading torus is too short and too weak to yield enough energy for the outflows to break out from the surrounding ejecta shell as highly relativistic jets. We conclude that neutrino annihilation alone cannot power sGRBs from NS–NS mergers

Thermoelectric self-cooling for power electronics: Increasing the cooling power

International Nuclear Information System (INIS)

Martinez, Alvaro; Astrain, David; Aranguren, Patricia

2016-01-01

Thermoelectric self-cooling was firstly conceived to increase, without electricity consumption, the cooling power of passive cooling systems. This paper studies the combination of heat pipe exchangers and thermoelectric self-cooling, and demonstrates its applicability to the cooling of power electronics. Experimental tests indicate that source-to-ambient thermal resistance reduces by around 30% when thermoelectric self-cooling system is installed, compared to that of the heat pipe exchanger under natural convection. Neither additional electric power nor cooling fluids are required. This thermal resistance reaches 0.346 K/W for a heat flux of 24.1 kW/m"2, being one order of magnitude lower than that obtained in previous designs. In addition, the system adapts to the cooling demand, reducing this thermal resistance for increasing heat. Simulation tests have indicated that simple system modifications allow relevant improvements in the cooling power. Replacement of a thermoelectric module with a thermal bridge leads to 33.54 kW/m"2 of top cooling power. Likewise, thermoelectric modules with shorter legs and higher number of pairs lead to a top cooling power of 44.17 kW/m"2. These results demonstrate the applicability of thermoelectric self-cooling to power electronics. - Highlights: • Cooling power of passive systems increased. • No electric power consumption. • Applicable for the cooling of power electronics. • Up to 44.17 kW/m"2 of cooling power, one order of magnitude higher. • Source-to-ambient thermal resistance reduces by 30%.

Neutron Exposures in Human Cells: Bystander Effect and Relative Biological Effectiveness

Science.gov (United States)

Seth, Isheeta; Schwartz, Jeffrey L.; Stewart, Robert D.; Emery, Robert; Joiner, Michael C.; Tucker, James D.

2014-01-01

Bystander effects have been observed repeatedly in mammalian cells following photon and alpha particle irradiation. However, few studies have been performed to investigate bystander effects arising from neutron irradiation. Here we asked whether neutrons also induce a bystander effect in two normal human lymphoblastoid cell lines. These cells were exposed to fast neutrons produced by targeting a near-monoenergetic 50.5 MeV proton beam at a Be target (17 MeV average neutron energy), and irradiated-cell conditioned media (ICCM) was transferred to unirradiated cells. The cytokinesis-block micronucleus assay was used to quantify genetic damage in radiation-naïve cells exposed to ICCM from cultures that received 0 (control), 0.5, 1, 1.5, 2, 3 or 4 Gy neutrons. Cells grown in ICCM from irradiated cells showed no significant increase in the frequencies of micronuclei or nucleoplasmic bridges compared to cells grown in ICCM from sham irradiated cells for either cell line. However, the neutron beam has a photon dose-contamination of 5%, which may modulate a neutron-induced bystander effect. To determine whether these low doses of contaminating photons can induce a bystander effect, cells were irradiated with cobalt-60 at doses equivalent to the percent contamination for each neutron dose. No significant increase in the frequencies of micronuclei or bridges was observed at these doses of photons for either cell line when cultured in ICCM. As expected, high doses of photons induced a clear bystander effect in both cell lines for micronuclei and bridges (pbystander effect in these cells. Finally, neutrons had a relative biological effectiveness of 2.0±0.13 for micronuclei and 5.8±2.9 for bridges compared to cobalt-60. These results may be relevant to radiation therapy with fast neutrons and for regulatory agencies setting standards for neutron radiation protection and safety. PMID:24896095

Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor

International Nuclear Information System (INIS)

Rojas R, E.; Benitez R, J. S.; Segovia de los Rios, J. A.; Rivero G, T.

2009-10-01

In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

Measurements and Monte-Carlo simulations of the particle self-shielding effect of B4C grains in neutron shielding concrete

Science.gov (United States)

DiJulio, D. D.; Cooper-Jensen, C. P.; Llamas-Jansa, I.; Kazi, S.; Bentley, P. M.

2018-06-01

A combined measurement and Monte-Carlo simulation study was carried out in order to characterize the particle self-shielding effect of B4C grains in neutron shielding concrete. Several batches of a specialized neutron shielding concrete, with varying B4C grain sizes, were exposed to a 2 Å neutron beam at the R2D2 test beamline at the Institute for Energy Technology located in Kjeller, Norway. The direct and scattered neutrons were detected with a neutron detector placed behind the concrete blocks and the results were compared to Geant4 simulations. The particle self-shielding effect was included in the Geant4 simulations by calculating effective neutron cross-sections during the Monte-Carlo simulation process. It is shown that this method well reproduces the measured results. Our results show that shielding calculations for low-energy neutrons using such materials would lead to an underestimate of the shielding required for a certain design scenario if the particle self-shielding effect is not included in the calculations.

Neutron Activation analysis of waste water

International Nuclear Information System (INIS)

Hernandez H, V.

1997-01-01

An instrumental neutron activation analysis for the simultaneous determination of chlorine, bromine, sodium, manganese, cobalt, copper, chromium, zinc, nickel, antimony and iron in waste water is described. They were determined in waste water samples under normal conditions by non-destructive neutron activation simultaneously using a suitable monostandard method. Standardized water samples were used and irradiated in polyethylene ampoules at a neutron flux of 10 13 cm -2 s -1 for periods of 1 minute, 1 and 10 hours. A Ge hyperpure detector was used for your activity determination, with count times of 60, 180, 300 and 600 seconds. The obtained results show than the method can be utilized for the determination of this elements without realize anything previous treatment of the samples. (Author)

Radioactive and stable cobalt concentrations in mussel in Kyushu island, Japan

International Nuclear Information System (INIS)

Momoshima, Noriyuki; Shiki, Atsushi; Takashima, Yoshimasa; Maki, Takao; Koriyama, Munehiro; Shimozono, Seika; Imamura, Hiroka; Nakamata, Kojiro.

1985-01-01

Two kinds of mussel, Septifer virgatus and Mytilus edulis, were collected from Kyushu island, Japan, in order to elucidate a background level of 60 Co, which is one of the most significant radionuclide for environmental monitoring around a nuclear power plant. The mussels were collected from 7 locations in 1983 and classified 2 or 3 groups depending on their shell size at each location. Activities of 60 Co were measured by a low-background β counter after purified by means of chemical separation and electrodeposition. Stable cobalt concentrations were determined by colorimetric method. The concentrations of cobalt in Septifer virgatus are one order higher level than that in Mytilus edulis. There are not so large difference in cobalt content depending on shell size so long as comparing them at the same location. The radioactivities in mussels show the same trend as stable cobalt. It has become apparent that Septifer virgatus has a tendency to concentrate cobalt with growing but Mytilus edulis is opposite. The cobalt-60 introduced to sea from nuclear explosions seems to be relatively constant in coastal seawater since specific activities are distributed in a narrow range in spite of kind, shell size and location. (author)

Self-sustainability of a research reactor facility with neutron activation analysis

International Nuclear Information System (INIS)

Chilian, C.; Kennedy, G.

2010-01-01

Long-term self-sustainability of a small reactor facility is possible because there is a large demand for non-destructive chemical analysis of bulk materials that can only be achieved with neutron activation analysis (NAA). The Ecole Polytechnique Montreal SLOWPOKE Reactor Facility has achieved self-sustainability for over twenty years, benefiting from the extreme reliability, ease of use and stable neutron flux of the SLOWPOKE reactor. The industrial clientele developed slowly over the years, mainly because of research users of the facility. A reliable NAA service with flexibility, high accuracy and fast turn-around time was achieved by developing an efficient NAA system, using a combination of the relative and k0 standardisation methods. The techniques were optimized to meet the specific needs of the client, such as low detection limit or high accuracy at high concentration. New marketing strategies are presented, which aim at a more rapid expansion. (author)

A neutron amplifier: prospects for reactor-based waste transmutation

International Nuclear Information System (INIS)

Blanovsky, A.

2004-01-01

A design concept and characteristics for an epithermal breeder controlled by variable feedback and external neutron source intensity are presented. By replacing the control rods with neutron sources, we could maintain good power distribution and perform radioactive waste burning in high flux subcritical reactors (HFSR) that have primary system size, power density and cost comparable to a pressurized water reactor (PWR). Another approach for actinide transmutation is a molten salt subcritical reactor proposed by Russian scientists. To increase neutron source intensity the HFSR is divided into two zones: a booster and a blanket with solid and liquid fuels. A neutron gate (absorber and moderator) imposed between two zones permits fast neutrons from the booster to flow to the blanket. Neutrons moving in the reverse direction are moderated and absorbed in the absorber zone. In the HFSR, neptunium-plutonium fuel is circulated in the booster and blanket, and americium-curium in the absorber zone and outer reflector. Use of a liquid actinide fuel permits transport of the delayed-neutron emitters from the blanket to the booster, where they can provide additional neutrons (source-dominated mode) or all the necessary excitation without an external neutron source (self-amplifying mode). With a blanket neutron multiplication gain of 20 and a booster gain of 50, an external neutron source rate of at least 10 15 n/s (0.7 MW D-T or 2.5 MW electron beam power) is needed to control the HFSR that produces 300 MWt. Most of the power could be generated in the blanket that burns about 100 kg of actinides a year. The analysis takes into consideration a wide range of HFSR design aspects including the wave model of observed relativistic phenomena, plant seismic diagnostics, fission electric cells (FEC) with a multistage collector (anode) and layered cathode. (author)

Reaction-in-flight neutrons as a test of stopping power in degenerate plasmas

Science.gov (United States)

Hayes, A. C.; Jungman, Gerard; Schulz, A. E.; Boswell, M.; Fowler, M. M.; Grim, G.; Klein, A.; Rundberg, R. S.; Wilhelmy, J. B.; Wilson, D.; Cerjan, C.; Schneider, D.; Sepke, S. M.; Tonchev, A.; Yeamans, C.

2015-08-01

We present the first measurements of reaction-in-flight (RIF) neutrons in an inertial confinement fusion system. The experiments were carried out at the National Ignition Facility, using both Low Foot and High Foot drives and cryogenic plastic capsules. In both cases, the high-energy RIF ( En> 15 MeV) component of the neutron spectrum was found to be about 10-4 of the total. The majority of the RIF neutrons were produced in the dense cold fuel surrounding the burning hotspot of the capsule, and the data are consistent with a compressed cold fuel that is moderately to strongly coupled (Γ˜ 0.6) and electron degenerate (θFermi/θe˜ 4). The production of RIF neutrons is controlled by the stopping power in the plasma. Thus, the current RIF measurements provide a unique test of stopping power models in an experimentally unexplored plasma regime. We find that the measured RIF data strongly constrain stopping models in warm dense plasma conditions, and some models are ruled out by our analysis of these experiments.

The basic concepts of a fuel-power detector for nuclear power reactors

International Nuclear Information System (INIS)

Lynch, G.F.

1979-01-01

Fuel power is proposed as an alternative to neutron or gamma-ray flux for control and safety functions in CANDU power reactors. To satisfy in-core power monitoring requirements, a detector whose dynamic response corresponds to the heat production rate in the fuel is needed. This report explores the concept of tailoring the response characteristics of a mixed-response self-powered flux detector to match the requirements of an ideal fuel-power detector. (author)

Brightness coatings of zinc-cobalt alloys by electrolytic way

International Nuclear Information System (INIS)

Julve, E.

1993-01-01

Zinc-cobalt alloys provide corrosion resistance for the ferrous based metals. An acidic electrolyte for zinc-cobalt electrodeposition is examined in the present work. The effects of variations in electrolyte composition, in electrolyte temperature, pH and agitation on electrodeposit composition have been studied, as well as the current density influence. It was found that the following electrolyte gave the optimum results: 79 g.1''-1 ZnCl 2 , 15.3 g.1''-1 CoCl 2 .6H 2 O, 160 g.1''-1 KCl, 25 g.1''-1 H 3 BO 3 and 5-10 cm''3.1''-1 of an organic additive (caffeine, coumarin and sodium lauryl-sulphonate). The operating conditions were: pH=5,6 temperature: 30 degree centigree, current density: 0,025-0,035 A. cm''2, anode: pure zinc, agitation: slowly with air and filtration: continuous. The throwing power and cathode current efficiency of the electrolyte were also studied. This electrolyte yielded zinc-cobalt alloys white and lustrous and had a cobalt content of 0,5-0,8% (Author) 3 refs. 5 fig

Research on rejection performance of reverse osmosis to cobalt in simulated radioactive wastewater

International Nuclear Information System (INIS)

Kong Jinsong; Tian Yanjie

2012-01-01

In order to reveal the rejection performance of the reverse osmosis applied in the radioactive wastewater treatment, treatment experiments were carried out on a pilot reverse osmosis equipment using wastewater containing cobalt nuclide. Results showed that the rejection ratio of reverse osmosis to cobalt was almost not affected by the operation pressure and the ratio of reclaiming, and had no direct relation with the salt rejection ratio. The ratio of cobalt rejection reached 90% in the experiment condition and could meet the requirement on the disposal of radioactive wastewater produced by nuclear powered installations. (authors)

Cobalt-60 production in the BN-350 fast power reactor

International Nuclear Information System (INIS)

Zvonarev, A.V.; Korobejnikov, V.V.; Matveenko, I.P.

1994-01-01

A possibility of Co-60 isotope production in the BN-350 fast reactor was considered. A special irradiating device, which is an assembly with a central hole, where a container containing cobalt and zirconium hydride is placed. The irradiating device tested permits generating 60 Co with specific activity of 100 Ci/g

Workshop: Research and development plans for high power spallation neutron testing at BNL

International Nuclear Information System (INIS)

1996-01-01

This report consists of vugraphs from presentations at the meeting. The papers covered the following topics: (1) APS as a proton source; (2) target status for NSNS (National Spallation Neutron Source); (3) spallation neutron source in Japan; (4) liquid LiBi flow loop; and (5) research and development plans for high power tests at the AGS

Self-Powered Functional Device Using On-Chip Power Generation

KAUST Repository

Hussain, Muhammad Mustafa

2012-01-26

An apparatus, system, and method for a self-powered device using on-chip power generation. In some embodiments, the apparatus includes a substrate, a power generation module on the substrate, and a power storage module on the substrate. The power generation module may include a thermoelectric generator made of bismuth telluride.

Self-Powered Functional Device Using On-Chip Power Generation

KAUST Repository

Hussain, Muhammad Mustafa

2012-01-01

An apparatus, system, and method for a self-powered device using on-chip power generation. In some embodiments, the apparatus includes a substrate, a power generation module on the substrate, and a power storage module on the substrate. The power generation module may include a thermoelectric generator made of bismuth telluride.

Power measurement in the boiling capsules in R2 using delayed neutron detector

International Nuclear Information System (INIS)

Roennberg, G.

1979-03-01

LWR fuel testing is performed in the R2 reactor by irradiation in both loops and so-called boiling capsules. The loops have forced cooling, and the power can be measured calorimetrically by conventional instrumentation. The boiling capsules have convection cooling, and it has therefore been necessary to develop a special technique for power measurement, the delayed neutron detector (DND). The DND is a pneumatic rabbit system, which activates small uranium samples in the boiling capsules and counts the delayed neutrons for determination of the fission rate. This report describes the equipment used, the procedure of measurement, and the method of evaluation. (atuhor)

Discrimination of neutrons and gamma quanta with the aid of their power density spectra

International Nuclear Information System (INIS)

Buchmueller, R.

1977-01-01

The paper introduces a method of using only one fission chamber to discriminate the neutron flux against the gamma flux. The gamma chamber current may be several orders of magnitude higher than the neutron chamber current. In specially dimensioned fission chambers the neutrons and gamma quanta are made to generate different current pulses. Discrimination becomes possible by recording the power density spectrum of the mixture of pulses over a broad frequency range ( [de

Passivation and corrosion behaviours of cobalt and cobalt-chromium-molybdenum alloy

International Nuclear Information System (INIS)

Metikos-Hukovic, M.; Babic, R.

2007-01-01

Passivation and corrosion behaviour of the cobalt and cobalt-base alloy Co30Cr6Mo was studied in a simulated physiological solution containing chloride and bicarbonate ions and with pH of 6.8. The oxido-reduction processes included solid state transformations occurring at the cobalt/electrolyte interface are interpreted using theories of surface electrochemistry. The dissolution of cobalt is significantly suppressed by alloying it with chromium and molybdenum, since the alloy exhibited 'chromium like' passivity. The structural and protective properties of passive oxide films formed spontaneously at the open circuit potential or during the anodic polarization were studied using electrochemical impedance spectroscopy in the wide frequency range

Optical studies of cobalt implanted rutile TiO2 (110) surfaces

International Nuclear Information System (INIS)

Joshi, Shalik Ram; Padmanabhan, B.; Chanda, Anupama; Mishra, Indrani; Malik, V.K.; Mishra, N.C.; Kanjilal, D.; Varma, Shikha

2016-01-01

Highlights: • The present study displays formation of nanostructures after Co implantation on TiO 2 surfaces. • Preferential sputtering leads to the creation of oxygen vacancies on the surface. • A large enhancement in visible light absorbance (nearly 5 times compared to pristine) is observed. • Creation of self-organized nanostructures and Ti 3+ oxygen vacancies promote photoabsorption. • Formation of Co-nanoclusters and Co–Ti–O phase play concerted role in enhancing photo-absorption. - Abstract: Present study investigates the photoabsorption properties of single crystal rutile TiO 2 (110) surfaces after they have been implanted with low fluences of cobalt ions. The surfaces, after implantation, demonstrate fabrication of nanostructures and anisotropic nano-ripple patterns. Creation of oxygen vacancies (Ti 3+ states), development of cobalt nano-clusters as well as band gap modifications have also been observed. Results presented here demonstrate that fabrication of self organized nanostructures, upon implantation, along with the development of oxygen vacancies and ligand field transitions of cobalt ion promote the enhancement of photo-absorbance in both UV (∼2 times) and visible (∼5 times) regimes. These investigations on nanostructured TiO 2 surfaces can be important for photo-catalysis.

Qualification of the neutronic evolution of LWR fuels in MELUSINE

International Nuclear Information System (INIS)

Beretz, D.; Garcin, J.; Ducros, G.; Vanhumbeeck, D.; Chaucheprat, P.

1984-09-01

MELUSINE, a swimming pool type reactor, in Grenoble, for research and technological irradiations is well fitted to the neutronic evolution qualification of the LWR fuel. Thus, with an adjustment of the lattice pitch, representative neutron spectrum locations are available. The re-leading management and the regulation mode flexibility of MELUSINE lead to reproductible neutronic parameters configurations without restricting the reactor to this purpose only. Under these conditions, simple calculations can be carried out for interpretation, without taking into account the whole core. An instrumentation by Self Power Neutron Detectors (collectrons) gives on-line information on the fluxes at the periphery of the device. When required by the neutronicians, experimental pins can be unloaded during the irradiation process and scanned on a gammametry bench immersed in the reactor-pool itself, before their isotopic composition analysis. Thus, within the framework of neutronic evolution qualification, are studied fuel pins for advanced assemblies for the light water reactors or their derivatives, with large advantages over irradiations in power reactors [fr

Activation analysis of stainless steel flux monitors using 252Cf neutron sources

International Nuclear Information System (INIS)

Williams, J.G.; Newton, T.H. Jr.; Cogburn, C.O.

1984-01-01

Activation analysis was performed on stainless steel beads from a chain which is used in reactor pressure vessel surveillance experiments at the Arkansas Power and Light Company reactors. The beads allow monitoring of two fast and three thermal neutron induced reactions: 58 Ni(n,p) 58 Co, 54 Fe(n,p) 54 Mn, 58 Fe(n,γ) 59 Fe, 59 Co(n,γ) 60 Co and 50 Cr(n,γ) 51 Cr. The analysis was performed using 12 beads from various positions along 5 different batches of chain and standard materials in an H 2 O moderator tank using two intense californium sources which had a total neutron emission rate of 3.97 x 10 10 /s. Semiconductor gamma spectrometers were used to count the products of the above reactions in the specimens. The percentage by weight of the iron, chromium and cobalt in the beads were found to be 62.1%, 20.2% and 0.120%, respectively. The excellent uniformity found in the bead compositions demonstrates the reproducibility of the experimental techniques and enhances considerably the value of the beads as neutron flux montitors

An application of CCD read-out technique to neutron distribution measurement using the self-activation method with a CsI scintillator plate

International Nuclear Information System (INIS)

Nohtomi, Akihiro; Kurihara, Ryosuke; Kinoshita, Hiroyuki; Honda, Soichiro; Tokunaga, Masaaki; Uno, Heita; Shinsho, Kiyomitsu; Wakabayashi, Genichiro; Koba, Yusuke; Fukunaga, Junichi; Umezu, Yoshiyuki; Nakamura, Yasuhiko; Ohga, Saiji

2016-01-01

In our previous paper, the self-activation of an NaI scintillator had been successfully utilized for detecting photo-neutrons around a high-energy X-ray radiotherapy machine; individual optical pulses from the self-activated scintillator are read-out by photo sensors such as a photomultiplier tube (PMT). In the present work, preliminary observations have been performed in order to apply a direct CCD read-out technique to the self-activation method with a CsI scintillator plate using a Pu-Be source and a 10-MV linac. In conclusion, it has been revealed that the CCD read-out technique is applicable to neutron measurement around a high-energy X-ray radiotherapy machine with the self-activation of a CsI plate. Such application may provide a possibility of novel method for simple neutron dose-distribution measurement. - Highlights: • Preliminary observations have been performed by a CCD for the CsI self-activation method. • It has been revealed that the CCD read-out technique is applicable to neutron measurement. • Such application may provide a novel method for simple neutron distribution measurement.

An application of CCD read-out technique to neutron distribution measurement using the self-activation method with a CsI scintillator plate

Energy Technology Data Exchange (ETDEWEB)

Nohtomi, Akihiro, E-mail: nohtomi@hs.med.kyushu-u.ac.jp [Graduate School of Medical Sciences, Kyushu University, 3-1-1 Maidashi, Higashi-ku, Fukuoka 812-8582 (Japan); Kurihara, Ryosuke; Kinoshita, Hiroyuki; Honda, Soichiro; Tokunaga, Masaaki; Uno, Heita [Graduate School of Medical Sciences, Kyushu University, 3-1-1 Maidashi, Higashi-ku, Fukuoka 812-8582 (Japan); Shinsho, Kiyomitsu [Graduate School of Human Sciences, Tokyo Metropolitan University, 7-2-10 Higashi-oku, Arakawa-ku, Tokyo 116-8551 (Japan); Wakabayashi, Genichiro [Atomic Energy Research Institute, Kinki University, 3-4-1 Kowakae, Higashiosaka-shi, Osaka 577-8502 (Japan); Koba, Yusuke [National Institute of Radiological Sciences, 4-9-1 Anagawa, Inage-ku, Chiba-shi, Chiba 263-8555 (Japan); Fukunaga, Junichi; Umezu, Yoshiyuki; Nakamura, Yasuhiko [Department of Radiology, Kyushu University Hospital, 3-1-1 Maidashi, Higashi-ku, Fukuoka 812-8582 (Japan); Ohga, Saiji [Department of Clinical Radiology, Graduate School of Medical Sciences, Kyushu University, 3-1-1 Maidashi, Higashi-ku, Fukuoka 812-8582 (Japan)

2016-10-01

In our previous paper, the self-activation of an NaI scintillator had been successfully utilized for detecting photo-neutrons around a high-energy X-ray radiotherapy machine; individual optical pulses from the self-activated scintillator are read-out by photo sensors such as a photomultiplier tube (PMT). In the present work, preliminary observations have been performed in order to apply a direct CCD read-out technique to the self-activation method with a CsI scintillator plate using a Pu-Be source and a 10-MV linac. In conclusion, it has been revealed that the CCD read-out technique is applicable to neutron measurement around a high-energy X-ray radiotherapy machine with the self-activation of a CsI plate. Such application may provide a possibility of novel method for simple neutron dose-distribution measurement. - Highlights: • Preliminary observations have been performed by a CCD for the CsI self-activation method. • It has been revealed that the CCD read-out technique is applicable to neutron measurement. • Such application may provide a novel method for simple neutron distribution measurement.

Cobalt as a gastric juice volume marker: Comparison of two methods of estimation

International Nuclear Information System (INIS)

Gana, T.J.; MacPherson, B.R.; Ng, D.; Koo, J.

1990-01-01

We investigated the use of cobalt-EDTA, a novel, nonabsorbable liquid phase marker, in the estimation of secretory volumes during topical misoprostol (synthetic PGE, analog) administration in the canine chambered gastric segment. We compared atomic absorption spectrophotometry (AAS) and instrumental neutron activation analysis (INAA) in the estimation of [Co]. Mucosal bathing solutions containing cobalt-EDTA were instilled into and recovered from the chamber by gravity every 15-min period as follows: (i) basal--60 min; (ii) misoprostol periods--150 min (plus 0.1-, 1-, 10-, 100-, and 1000-micrograms doses of misoprostol for two periods per dose). The recovered solutions were analyzed for [Co] by AAS and INAA. Total cobalt recovery by AAS after chamber washout was 102.97 +/- 0.98%. Mean +/- SE volumes (12.14 +/- 0.33 and 13.24 +/- 0.60 ml/15 min) obtained respectively from AAS and INAA were significantly higher (P less than 0.001) than the recovered mean volumes (10.51 +/- 0.17 ml/15 min). The percentage error in volume collection increased (range: 9.3-52.7%) with the volume of secretion. Values of [Co] obtained by the two techniques were comparable and not significantly different from each other (P greater than 0.05). INAA-estimated mean +/- SE [Co] showed consistently higher coefficients of variation. Spectra obtained for all samples during INAA measurements showed significant Compton background activity from 24Na and 38Cl. Cobalt-EDTA did not grossly or histologically damage the gastric mucosa. We conclude that cobalt is not adsorbed, absorbed, or metabolized, and is a suitable and reliable volume marker in this model

A continuously self regenerating high-flux neutron-generator facility

Science.gov (United States)

Rogers, A. M.; Becker, T. A.; Bernstein, L. A.; van Bibber, K.; Bleuel, D. L.; Chen, A. X.; Daub, B. H.; Goldblum, B. L.; Firestone, R. B.; Leung, K.-N.; Renne, P. R.; Waltz, C.

2013-10-01

A facility based on a next-generation, high-flux D-D neutron generator (HFNG) is being constructed at UC Berkeley. The current generator, designed around two RF-driven multicusp deuterium ion sources, is capable of producing a neutron output of >1011 n/s. A specially designed titanium-coated copper target located between the ion sources accelerates D+ ions up to 150 keV, generating 2.45 MeV neutrons through the d(d,3He)n fusion reaction. Deuterium in the target is self loaded and regenerating through ion implantation, enabling stable and continuous long-term operation. The proposed science program is focused on pioneering advances in the 40Ar/39Ar dating technique for geochronology, new nuclear data measurements, basic nuclear science research including statistical model studies of radiative-strength functions and level densities, and education. An overview of the facility and its unique capabilities as well as first measurements from the HFNG commissioning will be presented. Work supported by NSF Grant No. EAR-0960138, U.S. DOE LBL Contract No. DE-AC02-05CH11231, and U.S. DOE LLNL Contract No. DE-AC52-07NA27344.

Quality Management for Neutron Transmutation Doping of Silicon Ingot in HANARO

Energy Technology Data Exchange (ETDEWEB)

Kang, Ki-Doo; Kim, Ji-Uk; Yun, Hwa-Kyung; Lim, Chul-Hong; Kim, Young-Chil; Kim, Myong-Seop; Park, Sang-Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2007-10-15

By using this doping method, silicon semiconductors with extremely uniform dopant distributions can be produced, and this is the dominant advantage of NTD compared with a conventional chemical doping. Good uniformity of a dopant concentration is usually required for high power applications such as thyristor (SCR), IGBT, IGCT and GTO and for special sensors. Achieving an accurate neutron fluence corresponding to a target resistivity as well as a uniform irradiation is the prime target of a neutron irradiation for NTD. Generally, in order to reach an accurate neutron fluence, a real time neutron flux is monitored by a neutron detector such as a Self-powered Neutron Detector(SPND). And, after an irradiation, the total irradiation fluence is confirmed by measuring the absolute activity of a neutron activation sample that has been irradiated with a silicon ingot, and thus the SPND can be properly calibrated. Excellent irradiation uniformity and a high accuracy for a target neutron dose have been achieved from the early works of NTD. However, to maintain this excellent quality, the neutron irradiation fluence should be continuously modified and controlled. So, in this work, an activity to maintain the irradiation quality is introduced.

Elicitation threshold of cobalt chloride

DEFF Research Database (Denmark)

Fischer, Louise A; Johansen, Jeanne D; Voelund, Aage

2016-01-01

: On the basis of five included studies, the ED10 values of aqueous cobalt chloride ranged between 0.0663 and 1.95 µg cobalt/cm(2), corresponding to 30.8-259 ppm. CONCLUSIONS: Our analysis provides an overview of the doses of cobalt that are required to elicit allergic cobalt contactdermatitis in sensitized...

Neutron activation analysis in minerals prospecting

International Nuclear Information System (INIS)

Gomez, H.; Duque O, J.

1988-01-01

One method multielemental analysis in geological samples has been developed by neutron activation analysis without using standards and by eliminating many of the error sources of the absolute method. It uses the ratio of the activities induced by mass unit, between the element in the sample and one cobalt monitor. The detection limits are good for more than thirty elements in many prospecting programs, with a standard deviation less than 7%. The neutron flux used is 2x10 11 nxcm -2 .S -1 and the HPGE detector has a relative efficiency of 20% and an energy resolution of 1.9 KeV in 1332 KeV photopeak

Online In-Core Thermal Neutron Flux Measurement for the Validation of Computational Methods

International Nuclear Information System (INIS)

Mohamad Hairie Rabir; Muhammad Rawi Mohamed Zin; Yahya Ismail

2016-01-01

In order to verify and validate the computational methods for neutron flux calculation in RTP calculations, a series of thermal neutron flux measurement has been performed. The Self Powered Neutron Detector (SPND) was used to measure thermal neutron flux to verify the calculated neutron flux distribution in the TRIGA reactor. Measurements results obtained online for different power level of the reactor. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and measured thermal neutron flux in the core are in very good agreement indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux distribution in the reactor core. Since the computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of RTP utilization. (author)

High-power liquid-lithium target prototype for accelerator-based boron neutron capture therapy.

Science.gov (United States)

Halfon, S; Paul, M; Arenshtam, A; Berkovits, D; Bisyakoev, M; Eliyahu, I; Feinberg, G; Hazenshprung, N; Kijel, D; Nagler, A; Silverman, I

2011-12-01

A prototype of a compact Liquid-Lithium Target (LiLiT), which will possibly constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals, was built. The LiLiT setup is presently being commissioned at Soreq Nuclear Research Center (SNRC). The liquid-lithium target will produce neutrons through the (7)Li(p,n)(7)Be reaction and it will overcome the major problem of removing the thermal power generated using a high-intensity proton beam (>10 kW), necessary for sufficient neutron flux. In off-line circulation tests, the liquid-lithium loop generated a stable lithium jet at high velocity, on a concave supporting wall; the concept will first be tested using a high-power electron beam impinging on the lithium jet. High intensity proton beam irradiation (1.91-2.5 MeV, 2-4 mA) will take place at Soreq Applied Research Accelerator Facility (SARAF) superconducting linear accelerator currently in construction at SNRC. Radiological risks due to the (7)Be produced in the reaction were studied and will be handled through a proper design, including a cold trap and appropriate shielding. A moderator/reflector assembly is planned according to a Monte Carlo simulation, to create a neutron spectrum and intensity maximally effective to the treatment and to reduce prompt gamma radiation dose risks. Copyright © 2011 Elsevier Ltd. All rights reserved.

Neutron radiation shielding properties of polymer incorporated self compacting concrete mixes.

Science.gov (United States)

Malkapur, Santhosh M; Divakar, L; Narasimhan, Mattur C; Karkera, Narayana B; Goverdhan, P; Sathian, V; Prasad, N K

2017-07-01

In this work, the neutron radiation shielding characteristics of a class of novel polymer-incorporated self-compacting concrete (PISCC) mixes are evaluated. Pulverized high density polyethylene (HDPE) material was used, at three different reference volumes, as a partial replacement to river sand in conventional concrete mixes. By such partial replacement of sand with polymer, additional hydrogen contents are incorporated in these concrete mixes and their effect on the neutron radiation shielding properties are studied. It has been observed from the initial set of experiments that there is a definite trend of reductions in the neutron flux and dose transmission factor values in these PISCC mixes vis-à-vis ordinary concrete mix. Also, the fact that quite similar enhanced shielding results are recorded even when reprocessed HDPE material is used in lieu of the virgin HDPE attracts further attention. Copyright © 2017 Elsevier Ltd. All rights reserved.

Radiation detectors for the control of PWR nuclear boilers

International Nuclear Information System (INIS)

Duchene, J.

1977-01-01

The neutronic control in French PWR is effected by: 2 channels of measurement of intermediate power using γ'-compensated boron-coated ionization chambers 4 channels of measurement of high power with 'long' boron chambers also used in axial off-set measurement. A movable in-core measuring system is used for the fuel management and the power distribution monitoring. The instrumentation of start-up and intermediate power is conventional; the chambers of the axial off-set measurement and the in-core system are special for this type of power plant, they are discussed in details. The essential properties of the various types of detector, their major advantages or drawbacks, their comparative adaptation to the functions to be performed in the plant are summarized in a table. The 'long chambers' (on use in Fessenheim I and II, and soon in Bugey II) are boron coated current ionization chambers, without γ compensation, intended for power measurement. In-core measurements first involved activation methods - movable wires giving flux profiles, -or activable nuts (the Aeroball System at Trino Vercellese, Chooz...). In on-line neutron detectors, used at fixed positions, the electric signal is generated from: ionization the gas filling fission ionization chambers and γ ionization chambers; direct collection of the charged particles emitted from the convertor element in self-powered neutron detectors (rhodium, silver or vanadium) or self-powered γ detectors (cobalt); or thermoelectric effect in neutron and γ thermometers. The in-core measurement unit developped by Framatome is a movable miniaturized fission chamber system (at Tihange), every French exported power plant being now equipped with it [fr

Does power corrupt or enable? When and why power facilitates self-interested behavior.

Science.gov (United States)

DeCelles, Katherine A; DeRue, D Scott; Margolis, Joshua D; Ceranic, Tara L

2012-05-01

Does power corrupt a moral identity, or does it enable a moral identity to emerge? Drawing from the power literature, we propose that the psychological experience of power, although often associated with promoting self-interest, is associated with greater self-interest only in the presence of a weak moral identity. Furthermore, we propose that the psychological experience of power is associated with less self-interest in the presence of a strong moral identity. Across a field survey of working adults and in a lab experiment, individuals with a strong moral identity were less likely to act in self-interest, yet individuals with a weak moral identity were more likely to act in self-interest, when subjectively experiencing power. Finally, we predict and demonstrate an explanatory mechanism behind this effect: The psychological experience of power enhances moral awareness among those with a strong moral identity, yet decreases the moral awareness among those with a weak moral identity. In turn, individuals' moral awareness affects how they behave in relation to their self-interest. (PsycINFO Database Record (c) 2012 APA, all rights reserved).

Evidence of interface exchange magnetism in self-assembled cobalt-fullerene nanocomposites exposed to air

Czech Academy of Sciences Publication Activity Database

Lavrentiev, Vasyl; Stupakov, Alexandr; Lavrentieva, Inna; Motylenko, M.; Barchuk, M.; Rafaja, D.

2017-01-01

Roč. 28, č. 12 (2017), č. článku 125704. ISSN 0957-4484 R&D Projects: GA ČR(CZ) GBP108/12/G108; GA MŠk LM2015056 Institutional support: RVO:68378271 ; RVO:61389005 Keywords : fullerene * cobalt clusters * cobalt oxide * nanocomposite * interface exchange magnetism Subject RIV: BM - Solid Matter Physics ; Magnetism; JB - Sensors, Measurment, Regulation (FZU-D) OBOR OECD: Condensed matter physics (including formerly solid state physics, supercond.); Electrical and electronic engineering (FZU-D) Impact factor: 3.440, year: 2016

Synthesis of novel complexing macromolecular surfactants and study of their interactions with cobalt for the development of a decontamination process of textiles in dense CO2 medium

International Nuclear Information System (INIS)

Chirat, M.

2012-01-01

This study is about textile decontamination in dense CO 2 (liquid CO 2 or supercritical CO 2 ). The study is carried out in the framework of decontamination of textile used in the nuclear industry. The dense CO 2 offers an alternative to aqueous medium used in the current process which generates a huge quantity of contaminated aqueous effluent requiring a post-treatment. Cobalt is the targeted contamination and can be found as ionic species or particles. The cobalt extraction in dense CO 2 is achieved with an additive: a complexing CO 2 -philic/CO 2 -phobic macromolecular surfactant. Several types of additives were synthesized by controlled free radical polymerization: gradient copolymers made with CO 2 -philic groups (silicone-based or fluorinated moieties) and CO 2 -phobic complexing groups (aceto acetoxy, di-ethylphosphonate or phosphonic acid moieties). The copolymer behavior in dense CO 2 was determined by phase diagram measurements (cloud point method) and their self-assembly in dense CO 2 was investigated by small angle neutron scattering. The fluorinated copolymers were found advantageous in terms of solubility. Nevertheless, the silicone-based copolymers showed solubilities which are compatible with the process, therefore they are a good alternative to avoid fluorinated compounds which are unwanted in the conditioning of nuclear wastes. The study of cobalt complexation by the copolymers (UV-vis spectroscopy and inductively coupled plasma-mass spectroscopy) established relations between the type of complexing group and the affinity with the cobalt. The solubility of copolymer-cobalt complexes in dense CO 2 is similar to those of copolymers. Moreover, the self-assembly study of the complex revealed a low aggregation. Finally, the synthesized copolymers were used in particle or ionic decontamination processes. In the case of ionic decontamination process, a rate of 70% of decontamination was reached with the use of gradient copolymer poly(1

Neutron-deuteron analyzing power data at En = 21 MeV and the energy dependence of the three-nucleon analyzing power puzzle

Science.gov (United States)

Weisel, G. J.; Tornow, W.; Esterline, J. H.

2015-08-01

We present measurements of n-d analyzing power, {A}y(θ ), at En = 21.0 MeV. The experiment produces neutrons via the 2H(d, n)3He reaction and uses a deuterated liquid-scintillator center detector and six pairs of liquid-scintillator neutron side detectors. Elastic neutron scattering events are identified by using time-of-flight techniques and by setting a gate in the center-detector pulse-height spectrum. Beam polarization is monitored by using a high-pressure helium gas scintillator. The n-d {A}y(θ ) data at 21.0 MeV show a significant discrepancy with the results of rigorous three-body calculations and are consistent with data taken previously by us at 19.0 and 22.5 MeV. We review the overall energy dependence of the three-nucleon analyzing power puzzle in neutron-deuteron elastic scattering, using the best data available. We find that the relative difference between calculations and data is nearly constant at 25% up to En = 22.5 MeV.

Measurement of the thermal neutron self shielding coefficient in the Syrian miniature neutron source reactor inner irradiation site using the dy soils

International Nuclear Information System (INIS)

Khattab, K.; Khamis, I.

2007-01-01

Measurement of the thermal self shielding coefficient ( Gth ) in the Syrian Miniature Neutron Source Reactor (MNSR) inner irradiation site using Dy foils is presented in this paper. The thermal self shielding coefficient is measured as a function of the foil thickness or numbers. The mathematical equation which calculates the average relative radioactivity (Bq/g) versus the foil number is found as well.

Cobalt toxicity: Chemical and radiological combined effects on HaCaT keratinocyte cell line

International Nuclear Information System (INIS)

Sandre, C.; Moulin, C.; Bresson, C.; Gault, N.; Poncy, J. L.; Lefaix, J. L.

2010-01-01

Cobalt (Co) is an essential trace element well known as a constituent of vitamin B 12 , but different compounds of Co are also described as highly toxic and/or radio-toxic for individuals or the environment. In nuclear power plants, 58 Co and 60 Co are radioactive isotopes of cobalt present as activation products of stable Co and Ni used in alloys. Skin exposure is a current occupational risk in the hard metal and nuclear industries. As biochemical and molecular cobalt-induced toxicological mechanisms are not fully identified, we investigated cobalt toxicity in a model human keratinocyte cell line, HaCaT. In this study, we propose a model to determine the in vitro chemical impact on cell viability of a soluble form of cobalt (CoCl 2 ) with or without gamma-ray doses to mimic contamination by 60 Co, to elucidate the mechanisms of cobalt intracellular chemical and radiological toxicity. Intracellular cobalt concentration was determined after HaCaT cell contamination and chemical toxicity was evaluated in terms of cellular viability and clonogenic survival. We investigated damage to DNA in HaCaT cells by combined treatment with chemical cobalt and a moderate gamma-ray dose. Additive effects of cobalt and irradiation were demonstrated. The underlying mechanism of cobalt toxicity is not clearly established, but our results seem to indicate that the toxicity of Co(II) and of irradiation arises from production of reactive oxygen species. (authors)

Cobalt toxicity: Chemical and radiological combined effects on HaCaT keratinocyte cell line

Energy Technology Data Exchange (ETDEWEB)

Gault, N. [CEA Fontenay aux Roses, DSV/IRCM/SCSR/LRTS, 92265 Fontenay aux Rose (France); Sandre, C.; Moulin, B.; Bresson, C. [CEA, DEN, SECR, Laboratoire de Speciation des Radionucleides et des Molecules, F-91191 Gif-sur-Yvette (France); Poncy, J.L. [CEA Bruyeres Le Chatel, DSV/IRCM/SREIT/LRT, 91680 Bruyeres Le Chatel (France); Lefaix, J.L. [CEA Caen, DSV/IRCM/SRO/LARIA, 14070 Caen (France)

2010-07-01

Cobalt (Co) is an essential trace element well known as a constituent of vitamin B12, but different compounds of Co are also described as highly toxic and/or radio-toxic for individuals or the environment. In nuclear power plants, {sup 58}Co and {sup 60}Co are radioactive isotopes of cobalt present as activation products of stable Co and Ni used in alloys. Skin exposure is a current occupational risk in the hard metal and nuclear industries. As biochemical and molecular cobalt-induced toxicological mechanisms are not fully identified, we investigated cobalt toxicity in a model human keratinocyte cell line, HaCaT. In this study, we propose a model to determine the in vitro chemical impact on cell viability of a soluble form of cobalt (CoCl{sub 2}) with or without {gamma}-ray doses to mimic contamination by {sup 60}Co, to elucidate the mechanisms of cobalt intracellular chemical and radiological toxicity. Intracellular cobalt concentration was determined after HaCaT cell contamination and chemical toxicity was evaluated in terms of cellular viability and clonogenic survival. We investigated damage to DNA in HaCaT cells by combined treatment with chemical cobalt and a moderate {gamma}-ray dose. Additive effects of cobalt and irradiation were demonstrated. The underlying mechanism of cobalt toxicity is not clearly established, but our results seem to indicate that the toxicity of Co(II) and of irradiation arises from production of reactive oxygen species. (authors)

Cobalt toxicity: Chemical and radiological combined effects on HaCaT keratinocyte cell line

Energy Technology Data Exchange (ETDEWEB)

Sandre, C.; Moulin, C.; Bresson, C. [CEA Saclay, DEN, SECR, Lab Speciat Radionucleides and Mol, F-91191 Gif Sur Yvette (France); Gault, N. [CEA Fontenay Roses, DSV IRCM SCSR LRTS, F-92265 Fontenay Aux Roses (France); Poncy, J. L. [CEA Bruyeres Le Chatel, DSV IRCM SREIT LRT, F-91680 Bruyeres Le Chatel (France); Lefaix, J. L. [CEA Caen, DSV IRCM SRO LARIA, F-14070 Caen (France)

2010-07-01

Cobalt (Co) is an essential trace element well known as a constituent of vitamin B{sub 12}, but different compounds of Co are also described as highly toxic and/or radio-toxic for individuals or the environment. In nuclear power plants, {sup 58}Co and {sup 60}Co are radioactive isotopes of cobalt present as activation products of stable Co and Ni used in alloys. Skin exposure is a current occupational risk in the hard metal and nuclear industries. As biochemical and molecular cobalt-induced toxicological mechanisms are not fully identified, we investigated cobalt toxicity in a model human keratinocyte cell line, HaCaT. In this study, we propose a model to determine the in vitro chemical impact on cell viability of a soluble form of cobalt (CoCl{sub 2}) with or without gamma-ray doses to mimic contamination by {sup 60}Co, to elucidate the mechanisms of cobalt intracellular chemical and radiological toxicity. Intracellular cobalt concentration was determined after HaCaT cell contamination and chemical toxicity was evaluated in terms of cellular viability and clonogenic survival. We investigated damage to DNA in HaCaT cells by combined treatment with chemical cobalt and a moderate gamma-ray dose. Additive effects of cobalt and irradiation were demonstrated. The underlying mechanism of cobalt toxicity is not clearly established, but our results seem to indicate that the toxicity of Co(II) and of irradiation arises from production of reactive oxygen species. (authors)

Monte Carlo estimation of the dose and heating of cobalt adjuster rods irradiated in the CANDU 6 reactor core

International Nuclear Information System (INIS)

Gugiu, D.; Dumitrache, I.

2005-01-01

The present work is a part of a more complex project related to the replacement of the original stainless steel adjuster rods with cobalt assemblies in the CANDU 6 reactor core. The 60 Co produced by 59 Co irradiation could be used extensively in medicine and industry. The paper will mainly describe some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronic equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and heating of the irradiated cobalt rods, are performed using the Monte Carlo codes MCNP5 and MONTEBURNS 2.1. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters. (authors)

Biological fate of cobalt-60 released during the corrosion of neutron-activated stanless steel in seawater

International Nuclear Information System (INIS)

Young, J.S.

1982-03-01

Passing seawater over radioactive Type 347 stainless steel in a sediment/seawater laboratory system and exposing marine animals to this environment provided information on the bioaccumulation of 60 Co from radioactive structural material. Exposure of marine organisms to radioactive corrosion products and directly to radioactive stainless steel in seawater simulated some of the possible conditions which could arise from the deposition of radioactive stainless steel on the ocean floor. Detectable levels of 60 Co in marine animals were not observed on a short term basis (5 weeks). Longterm (13 months) exposure of marine animals in a sediment/seawater system resulted in 60 Co bioaccumulation. The specific activity of 60 Co in the organisms was as much as one million times less than that initially present in the radioactive stainless steel. This was due to the dilution of 60 Co by stable cobalt in the seawater, sediments and organisms. As expected the 60 Co specific activity of the organisms never increased above that of the radioactive source. This is because 60 Co is chemicaly indistinguishable from stable Co. Increasing 60 Co concentration factors with decreasing 60 Co concentrations in the seawater and sediment media coupled with relatively constant 60 Co specific activities suggest a possible homeostatic control of cobalt concentrations in certain marine organisms. The evidence indicates that the marine animals derived more of the accumulated 60 Co from the sediments and interstitial water than from seawater. Cobalt-60 concentration factors were generally found to be lower than published cobalt concentration factors due to the predominantly insoluble nature of the corrosion products. Baseline information is provided on trace element concentrations in deep-sea organisms. Stable Co and twenty other elements were measured in abyssal invertebrates and a fish

Non-equilibrium thermodynamics and energy distribution function of neutron gas in constant power reactor under coupling of neutrons and medium

International Nuclear Information System (INIS)

Hayasaka, Hideo

1983-01-01

The thermodynamics and the energy distribution function of the neutron gas in a constant power reactor are considered, taking into account the burn-up of fuel. To separate the secular motion of neutrons owing to fuel burn-up and the microscopic fluctuations of neutrons around this motion, a long time of the order of several months is divided into m equal intervals, and the respective states corresponding to m small time intervals are treated as quasi-stationary states. The local energy distribution function of the neutron gas in the quasi-stationary state is given by a generalized Boltzmann distribution specified by the respective generalized activity coefficient for each subsystem. The effects of fuel burn-up on the respective distribution functions for successive small time intervals are taken into account through various quantities relating to reactor physics, depending upon the fuel burn-up, by successive approximation. (author)

Facile fabrication of cobalt oxalate nanostructures with superior specific capacitance and super-long cycling stability

Science.gov (United States)

Cheng, Guanhua; Si, Conghui; Zhang, Jie; Wang, Ying; Yang, Wanfeng; Dong, Chaoqun; Zhang, Zhonghua

2016-04-01

Transition metal oxalate materials have shown huge competitive advantages for applications in supercapacitors. Herein, nanostructured cobalt oxalate supported on cobalt foils has been facilely fabricated by anodization, and could directly serve as additive/binder-free electrodes for supercapacitors. The as-prepared cobalt oxalate electrodes present superior specific capacitance of 1269 F g-1 at the current density of 6 A g-1 in the galvanostatic charge/discharge test. Moreover, the retained capacitance is as high as 87.2% as the current density increases from 6 A g-1 to 30 A g-1. More importantly, the specific capacitance of cobalt oxalate retains 91.9% even after super-long cycling of 100,000 cycles. In addition, an asymmetric supercapacitor assembled with cobalt oxalate (positive electrode) and activated carbon (negative electrode) demonstrates excellent capacitive performance with high energy density and power density.

The mathematical description of the electrosynthesis of composites of oxy-hydroxycompounds cobalt with polypyrrole overooxidazed

Directory of Open Access Journals (Sweden)

V. V. Tkach

2016-03-01

Full Text Available The electrosynthesis of pereoxidized polypyrrole composite with oxy-hydroxy compounds cobalt in a strongly acidic environment has been described mathematically, using linear stability theory and bifurcation analysis. The conditions of stability of stationary states and self-oscillatory and monotonic instability have been described also. The system behavior was compared with behavior of other systems with pereoxidation, electropolymerization of heterocyclic compounds and electrosynthesis of the oxy-hydroxy compounds cobalt.

A Compact Self-Driven Liquid Lithium Loop for Industrial Neutron Generation

Science.gov (United States)

Stemmley, Steven; Szott, Matt; Kalathiparambil, Kishor; Ahn, Chisung; Jurczyk, Brian; Ruzic, David

2017-10-01

A compact, closed liquid lithium loop has been developed at the University of Illinois to test and utilize the Li-7(d,n) reaction. The liquid metal loop is housed in a stainless steel trench module with embedded heating and cooling. The system was designed to handle large heat and particle fluxes for use in neutron generators as well as fusion devices, solely operating via thermo-electric MHD. The objectives of this project are two-fold, 1) produce a high energy, MeV-level, neutron source and 2) provide a self-healing, low Z, low recycling plasma facing component. The flowing volume will keep a fresh, clean, lithium surface allowing Li-7(d,n) reactions to occur as well as deuterium adsorption in the fluid, increasing the overall neutron output. Expected yields of this system are 107 n/s for 13.5 MeV neutrons and 108 n/s for 2.45 MeV neutrons. Previous work has shown that using a tapered trench design prevents dry out and allows for an increase in velocity of the fluid at the particle strike point. For heat fluxes on the order of 10's MW/m2, COMSOL models have shown that high enough velocities ( 70 cm/s) are attainable to prevent significant lithium evaporation. Future work will be aimed at addressing wettability issues of lithium in the trenches, experimentally determine the velocities required to prevent dry out, and determine the neutron output of the system. The preliminary results and discussion will be presented. DOE SBIR project DE-SC0013861.

In core system mapping reactor power distribution

International Nuclear Information System (INIS)

Yoriyaz, H.; Moreira, J.M.L.

1989-01-01

Based on the signals of SPND'S (Self Powered Neutron Detectors) distributed inside of a core, the spatial power distribution is obtained using the MAP program, developed in this work. The methodology applied in MAP program uses a least mean square technique to calculate expansion coefficients that depend on the SPND'S signals. The final power or neutron flux distribution is obtained by a combination of certains functions or expansion modes that are provided from diffusion calculation with the CITATION code. The MAP program is written in PASCAL language and will be used in IEA-R1 reactor for assisting its operation. (author) [pt

Neutron activation analysis in reconnaissance geochemical survey of Northwestern Mindoro

International Nuclear Information System (INIS)

Santos, G. Jr.; Fernandez, L.G.

1987-01-01

Instrumental neutron activation analysis (NAA) technique was used to analyze stream sediments collected in Northwestern Mindoro. The concentration levels of 18 elements were determined. It was noted that NAA is suitable for the determination of rare earth, gold, arsenic and cobalt among others because of favorable high neutron cross sections. Samples collected in regional reconnaissance geochemical surveys could be analyzed usng NAA technique to complement other non-nuclear techniques, such as atomic absorption and X-ray fluorescence analysis. (Author). 11 figs.; 2 tabs.; 12 refs

Self-SALP/self-assessment at Calvert Cliffs nuclear power plant

International Nuclear Information System (INIS)

Pieringer, P.; Slider, J.E.

1992-01-01

In recent years, utilities have undertaken various forms of self-assessment. These methods have as their goal early detection of weaknesses in plan performance to permit timely corrective action. Ideally, self-assessment methods identify problems at a lower threshold than do outside agencies, such as the US Nuclear Regulatory Commission (NRC) and the Institute of Nuclear Power Operations (INPO). One of the more popular methods of self-assessment has been the self-SALP (systematic assessment of licensee performance). This method emulates the procedures of the NRC's SALP to render a subjective, overall assessment of performance. The experience of Baltimore Gas and Electric Company (BG and E) in applying self-SALP methods at its Calvert Cliffs nuclear power plant (CCNPP) provides insights useful to other utilities considering use of self-SALP methods. The factors leading to BG and E's decision to use a different approach may help others evaluate their self-assessment choices. This paper summarizes BG and E's experience with self-SALP methods and describes the methods that will be used to evaluate performance in the future

Measurement of neutron activation cross-sections for elements Co, Ni, Y, Nb, Tm and Au between 12 and 20 MeV

International Nuclear Information System (INIS)

Iwasaki, S.; Matsuyama, S.; Ohkubo, T.; Fukuda, H.; Sakuma, M.; Kitamura, M.; Odano, N.

1995-01-01

Neutron activation cross-sections for cobalt, nickel, yttrium, niobium, thulium and gold have been measured in the neutron energies from 12 to 20 MeV with the reference cross section of NEA 93 Nb(n,2n) 92m Nb at Tohoku Dynamitron Facility. (author)

Polarized neutron reflectivity from monolayers of self-assembled magnetic nanoparticles.

Science.gov (United States)

Mishra, D; Petracic, O; Devishvili, A; Theis-Bröhl, K; Toperverg, B P; Zabel, H

2015-04-10

We prepared monolayers of iron oxide nanoparticles via self-assembly on a bare silicon wafer and on a vanadium film sputter deposited onto a plane sapphire substrate. The magnetic configuration of nanoparticles in such a dense assembly was investigated by polarized neutron reflectivity. A theoretical model fit shows that the magnetic moments of nanoparticles form quasi domain-like configurations at remanence. This is attributed to the dipolar coupling amongst the nanoparticles.

Formulation of detector response function to calculate the power density profiles using in-core neutron detectors

International Nuclear Information System (INIS)

Ahmed, S. A.; Peter, J. K.; Semmler, W.; Shultis, J. K.

2007-01-01

By measuring neutron fluxes at different locations throughout a core, it's possible to derive the power-density profile P k (W cm - 3), at an axial depth z of fuel rod k. Micro-pocket fission detectors (MPFD) have been fabricated to perform such in-core neutron flux measurements. The purpose of this study is to develop a mathematical model to obtain axial power density distributions in the fuel rods from the in-core responses of the MPFDs

Medical Radioisotope Production in a Power-Flattened ADS Fuelled with Uranium and Plutonium Dioxides

Directory of Open Access Journals (Sweden)

Gizem Bakır

2016-01-01

Full Text Available This study presents the medical radioisotope production performance of a conceptual accelerator driven system (ADS. Lead-bismuth eutectic (LBE is selected as target material. The subcritical fuel core is conceptually divided into ten equidistant subzones. The ceramic (natural U, PuO2 fuel mixture and the materials used for radioisotope production (copper, gold, cobalt, holmium, rhenium, thulium, mercury, palladium, thallium, molybdenum, and yttrium are separately prepared as cylindrical rods cladded with carbon/carbon composite (C/C and these rods are located in the subzones. In order to obtain the flattened power density, percentages of PuO2 in the mixture of UO2 and PuO2 in the subzones are adjusted in radial direction of the fuel zone. Time-dependent calculations are performed at 1000 MW thermal fission power (Pth for one hour using the BURN card. The neutronic results show that the investigated ADS has a high neutronic capability, in terms of medical radioisotope productions, spent fuel transmutation and energy multiplication. Moreover, a good quasiuniform power density is achieved in each material case. The peak-to-average fission power density ratio is in the range of 1.02–1.28.

High-power electron beam tests of a liquid-lithium target and characterization study of (7)Li(p,n) near-threshold neutrons for accelerator-based boron neutron capture therapy.

Science.gov (United States)

Halfon, S; Paul, M; Arenshtam, A; Berkovits, D; Cohen, D; Eliyahu, I; Kijel, D; Mardor, I; Silverman, I

2014-06-01

A compact Liquid-Lithium Target (LiLiT) was built and tested with a high-power electron gun at Soreq Nuclear Research Center (SNRC). The target is intended to demonstrate liquid-lithium target capabilities to constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals. The lithium target will produce neutrons through the (7)Li(p,n)(7)Be reaction and it will overcome the major problem of removing the thermal power >5kW generated by high-intensity proton beams, necessary for sufficient therapeutic neutron flux. In preliminary experiments liquid lithium was flown through the target loop and generated a stable jet on the concave supporting wall. Electron beam irradiation demonstrated that the liquid-lithium target can dissipate electron power densities of more than 4kW/cm(2) and volumetric power density around 2MW/cm(3) at a lithium flow of ~4m/s, while maintaining stable temperature and vacuum conditions. These power densities correspond to a narrow (σ=~2mm) 1.91MeV, 3mA proton beam. A high-intensity proton beam irradiation (1.91-2.5MeV, 2mA) is being commissioned at the SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator. In order to determine the conditions of LiLiT proton irradiation for BNCT and to tailor the neutron energy spectrum, a characterization of near threshold (~1.91MeV) (7)Li(p,n) neutrons is in progress based on Monte-Carlo (MCNP and Geant4) simulation and on low-intensity experiments with solid LiF targets. In-phantom dosimetry measurements are performed using special designed dosimeters based on CR-39 track detectors. © 2013 Elsevier Ltd. All rights reserved.

Evaluation of in-core measurements with self-powered detectors

International Nuclear Information System (INIS)

Spoden, E.

1977-01-01

A calculational method is presented for transformation of n-β currents into relative thermal neutron fluxes taking into account an epithermal correction. The difference between epithermal corrections with cross section ratios is demonstrated based on calculations with NR approximation and multiplied by the ratio of mean epithermal and thermal first collision probabilities as well as on calculations using the model of infinite dilution. The effects of this difference on axial power density distribution, relative power of the fuel assemblies, and the inhomogeneity coefficients are discussed in the light of a characteristic measurement with rhodium emitter in a WWER-type reactor. The effect of errors in the nuclear data on the mean error of relative thermal neutron flux is evaluated. (author)

Nuclear power and physics

International Nuclear Information System (INIS)

Xu Mi

2006-01-01

During the 30s and 40s of the last century atomic physicists discovered the fission of uranium nuclei bombarded by neutrons and realized the first self-sustaining controlled fission chain reaction, which ushered in the atomic age. After 50 years of electricity production, in 2003 nuclear power plants were generating 16% of the total electricity in the world. Of these, thermal neutron reactors make up over 99%. For the large scale production of nuclear power, say up to hundreds of GWe, it is very important to speed up the development and deployment of fast breeder reactors to avoid the future lack of uranium resources. (authors)

Neutron irradiation control in the neutron transmutation doping process in HANARO using SPND

Energy Technology Data Exchange (ETDEWEB)

Kang, Gi-Doo; Kim, Myong-Seop [Korea Atomic Energy Research Institute, Yuseong, Daejeon, 305-353, (Korea, Republic of)

2015-07-01

The neutron irradiation control method by using self-powered neutron detector (SPND) is developed for the neutron transmutation doping (NTD) application in HANARO. An SPND is installed at a fixed position of the upper part of the sleeve in HANARO NTD hole for real-time monitoring of the neutron irradiation. It is confirmed that the SPND is significantly affected by the in-core condition and surroundings of the facility. Furthermore, the SPND signal changes about 15% throughout a whole cycle according to the change of the control rod position. But, it is also confirmed that the variation of the neutron flux on the silicon ingots inside the irradiation can is not so big while moving of the control rod. Accordingly, the relationship between the ratio of the neutron flux to the SPND signal output and the control rod position is established. In this procedure, the neutron flux measurement by using zirconium foil is utilized. The real NTD irradiation experiments are performed using the established relationship. The irradiated neutron fluence can be controlled within ±1.3% of the target one. The mean value of the irradiation/target ratio of the fluence is 0.9992, and the standard deviation is 0.0071. Thus, it is confirmed that the extremely accurate irradiation would be accomplished. This procedure can be useful for the SPND application installed at the fixed position to the field requiring the extremely high accuracy. (authors)

Cobalt-Doped Nickel Phosphite for High Performance of Electrochemical Energy Storage.

Science.gov (United States)

Li, Bing; Shi, Yuxin; Huang, Kesheng; Zhao, Mingming; Qiu, Jiaqing; Xue, Huaiguo; Pang, Huan

2018-03-01

Compared to single metallic Ni or Co phosphides, bimetallic Ni-Co phosphides own ameliorative properties, such as high electrical conductivity, remarkable rate capability, upper specific capacity, and excellent cycle performance. Here, a simple one-step solvothermal process is proposed for the synthesis of bouquet-like cobalt-doped nickel phosphite (Ni 11 (HPO 3 ) 8 (OH) 6 ), and the effect of the structure on the pseudocapacitive performance is investigated via a series of electrochemical measurements. It is found that when the cobalt content is low, the glycol/deionized water ratio is 1, and the reaction is under 200 °C for 20 h, the morphology of the sample is uniform and has the highest specific surface area. The cobalt-doped Ni 11 (HPO 3 ) 8 (OH) 6 electrode presents a maximum specific capacitance of 714.8 F g -1 . More significantly, aqueous and solid-state flexible electrochemical energy storage devices are successfully assembled. The aqueous device shows a high energy density of 15.48 mWh cm -2 at the power density of 0.6 KW cm -2 . The solid-state device shows a high energy density of 14.72 mWh cm -2 at the power density of 0.6 KW cm -2 . These excellent performances confirm that the cobalt-doped Ni 11 (HPO 3 ) 8 (OH) 6 are promising materials for applications in electrochemical energy storage devices. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Staged Z-pinch Experiments at the 1MA Zebra pulsed-power generator: Neutron measurements

Science.gov (United States)

Ruskov, Emil; Darling, T.; Glebov, V.; Wessel, F. J.; Anderson, A.; Beg, F.; Conti, F.; Covington, A.; Dutra, E.; Narkis, J.; Rahman, H.; Ross, M.; Valenzuela, J.

2017-10-01

We report on neutron measurements from the latest Staged Z-pinch experiments at the 1MA Zebra pulsed-power generator. In these experiments a hollow shell of argon or krypton gas liner, injected between the 1 cm anode-cathode gap, compresses a deuterium plasma target of varying density. Axial magnetic field Bz neutron Time of Flight (nTOF) detectors are augmented with a large area ( 1400 cm2) liquid scintillator detector to which fast gatedPhotek photomultipliers are attached. Sample data from these neutron diagnostics systems is presented. Consistently high neutron yields YDD >109 are measured, with highest yield of 2.6 ×109 . A pair of horizontally and vertically placed plastic scintillator nTOFs suggest isotropic i.e. thermonuclear origin of the neutrons produced. nTOF data from the liquid scintillator detector was cross-calibrated with the silver activation detector, and can be used for accurate calculation of the neutron yield. Funded by the Advanced Research Projects Agency - Energy, under Grant Number DE-AR0000569.

Measurement and study of the self-diffusion of transition elements in 13X and 5A zeolites; Dosage et etude de l’auto-diffusion d’elements de transition dans les zeolithes 13X et 5A

Energy Technology Data Exchange (ETDEWEB)

Mulkay, P. [Université Lovanium, Kinshasa (Congo, The Democratic Republic of the); Roels, J. [Centre nucléaire TRICO, Kinshasa (Congo, The Democratic Republic of the)

1970-01-15

The authors measured cobalt, manganese and nickel in 13X zeolite and studied self-diffusion of cobalt in 5A zeolite. Ion exchange revealed the following order of selectivity: Co≥Mn>Ni. The metals were labelled in situ. For the self-diffusion study, cobalt activated by neutron bombardment was adsorbed in suitable quantities on the zeolite, which had been previously charged with natural cobalt by ion exchange. The measurements were carried out by gamma spectrometry. The activation energy for self-diffusion was determined from the linear plot of log{sub 10} D versus l/T °K, using the Arrhenius equation. A value of 13.4 kcal/deg/Lon g was found. The values obtained are compared with those for other divalent cations. (author) [French] Les auteurs ont réalisé le dosage du cobalt, du manganese et du nickel dans le zéolithe 13X et ont suivi l'auto-diffusion du cobalt dans le zéolithe 5A. L’échange ionique présentait la sélectivité suivante: Co≥Mn>Ni. Les métaux étaient marqués in situ. Pour I'auto-diffusion, le cobalt activé par bombardement neutronique a été fixé en quantité convenable sur le zéolithe, qui a été préalablement chargé de cobalt naturel par échange. Les mesures ont été réalisées par spectrométrie gamma- L’énergie d’activation pour le phénomène d'auto-diffusion est obtenue à partir du tracé linéaire de log{sub 10} D en fonction de l/T °K en se basant sur l’équation d’Arrhénius. Les auteurs ont trouvé une valeur de 13,4 kcal/degré/ion g. Les valeurs obtenues sont comparées à celles obtenues pour d'autres cations divalents. (author)

The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

International Nuclear Information System (INIS)

Zheng Youqi; Zu Tiejun; Wu Hongchun; Cao Liangzhi; Yang Chao

2012-01-01

In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.

Determination of chromium, cobalt and nickel in tissue samples by radiochemical activation analysis

International Nuclear Information System (INIS)

Reisell, A.; Lakomaa, E.L.

1983-03-01

A radiochemical neutron activation analysis method for the determination of chromium, cobalt and nickel in tissue samples. A radiochemical neutron activation analysis method for the determination of chromium, cobalt and nickel has been developed. The destruction device used consisted of a combined wet-ashing-distillation and ion-exchange system. Six samples could be treated at the same time. The samples were wet-ashed with H*L2SO*L4-H*L2O*L2 mixture. Volatile elements were distilled as bromide compounds with HBr*H-. The distillation residue in 8M HCl was passed through hydrated antimony pentoxide (HAP) in order to remove disturbing *H2*H4Na-activity and through a Dowex 2 x 8 column so as to retain *H6*H0Co (formed from *H5*H8Ni). Chromium was elutriated from the column and precipitated as Cr(OH)*L3 for the removal of disturbing *H3*H2P-activity. The standards and samples were treated in a similar manner each so that the yield determination is not necessarily needed. The yields by tracer experiments were (43 +- 5) % for Cr, (93 +- 4) % for Co and (88 +- 14) % for Ni. The precision and accuracy of the method were studied by using reference materials of the National Bureau of Standards (NBS) and the International Atomic Energy Agency (IAEA)

Neutron fluence measurement in the cavity of Balakovo nuclear power plant, unit 3

International Nuclear Information System (INIS)

Voorbraak, W.P.; Baard, J.H.; Paardekooper, A.; Nolthenius, H.J.

1996-12-01

An international benchmark exercise has been organized by the Russian GOSATOMNADZOR. The aim was to reduce the uncertainty of fluence measurements in Nuclear Power Plants in particular VVER-1000 reactors. The benchmark was set up in the cavity of the Balakovo NPP 3. Eight institutes were involved. This report presents the results obtained by ECN. From this report, it can be concluded that the results of the relative large monitor set (13 different reaction rates with overlapping response regions) point to possible imperfections in the calculated neutron spectra. However the experimental information is not powerful enough to reduce the uncertainty of the neutron fluence rate especially in the energy region between 0.1 and 0.5 MeV below 50 percent. (orig.)

Effects of long-time elevated temperature exposures on hot-isostatically-pressed power-metallurgy Udimet 700 alloys with reduced cobalt contents

Science.gov (United States)

Hart, F. H.

1984-01-01

Because almost the entire U.S. consumption of cobalt depends on imports, this metal has been designated "strategic'. The role and effectiveness of cobalt is being evaluated in commercial nickel-base superalloys. Udiment 700 type alloys in which the cobalt content was reduced from the normal 17% down to 12.7%, 8.5%, 4.3%, and 0% were prepared by standard powder metallurgy techniques and hot isostatically pressed into billets. Mechanical testing and microstructural investigations were performed. The mechanical properties of alloys with reduced cobalt contents which were heat-treated identically were equal or better than those of the standard alloy, except that creep rates tended to increase as cobalt was reduced. The effects of long time exposures at 760 C on mechanical properties and at 760 C and 845 C on microstructures were determined. Decreased tensile properties and shorter rupture lives with increased creep rates were observed in alloy modifications. The exposures caused gamma prime particle coarsening and formation of sigma phase in the alloys with higher cobalt contents. Exposure at 845 C also reduced the amount of MC carbides.

Self-powered radiation detector

International Nuclear Information System (INIS)

Goldstein, N.P.; Todt, W.H.

1979-01-01

Self-powered gamma radiation detector composed of a conducting emitter surrounded by an insulating medium and a conducting tubular collector, the emitter being a hollow tube containing an electrical insulator [fr

The low power miniature neutron source reactors: Design, safety and applications

International Nuclear Information System (INIS)

Ahmed, Y.A.; Ewa, I.O.B.; Umar, M.; Bezboruah, T.; Johri, M.; Akaho, E.H.K.

2006-04-01

The Chinese Miniature Neutron Source Reactor (MNSR) is a low power research reactor with maximum thermal neutron flux of 1 x 10 12 n.cm -2 .s -1 in one of its inner irradiation channels and thermal power of approximately 30kW. The MNSR is designed based on the Canadian SLOWPOKE reactor and is one of the smallest commercial research reactors presently available in the world. Its commercial versions currently in operation in China, Ghana, Iran, Nigeria, Pakistan and Syria, is considered as an excellent tool for Neutron Activation Analysis (NAA), training of Scientist, and Engineers in nuclear science and technology and small scale radioisotope production. The paper highlights the basic design and theory of the commercial MNSR, its safety features, applications and advantages over the Chinese Prototype. The experimental flux characteristics determined in this work and in similar studies by other authors reveal that the commercial MNSR has more flux stability, longer life span, higher negative temperature coefficient of reactivity and low under-moderation compared to its prototype in China. The result shows that the facility is safe for reactor physics experiments, teaching and training of students and also ideal for application of NAA for the determination of elemental composition of biological and environmental samples. It can also be a useful tool for geochemical and soil fertility mapping. (author)

Equipment for thermal neutron flux measurements in reactor R2

Energy Technology Data Exchange (ETDEWEB)

Johansson, E; Nilsson, T; Claeson, S

1960-04-15

For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

Cobalt metabolism and toxicology—A brief update

International Nuclear Information System (INIS)

Simonsen, Lars Ole; Harbak, Henrik; Bennekou, Poul

2012-01-01

Cobalt metabolism and toxicology are summarized. The biological functions of cobalt are updated in the light of recent understanding of cobalt interference with the sensing in almost all animal cells of oxygen deficiency (hypoxia). Cobalt (Co 2+ ) stabilizes the transcriptional activator hypoxia-inducible factor (HIF) and thus mimics hypoxia and stimulates erythropoietin (Epo) production, but probably also by the same mechanism induces a coordinated up-regulation of a number of adaptive responses to hypoxia, many with potential carcinogenic effects. This means on the other hand that cobalt (Co 2+ ) also may have beneficial effects under conditions of tissue hypoxia, and possibly can represent an alternative to hypoxic preconditioning. Cobalt is acutely toxic in larger doses, and in mammalian in vitro test systems cobalt ions and cobalt metal are cytotoxic and induce apoptosis and at higher concentrations necrosis with inflammatory response. Cobalt metal and salts are also genotoxic, mainly caused by oxidative DNA damage by reactive oxygen species, perhaps combined with inhibition of DNA repair. Of note, the evidence for carcinogenicity of cobalt metal and cobalt sulfate is considered sufficient in experimental animals, but is as yet considered inadequate in humans. Interestingly, some of the toxic effects of cobalt (Co 2+ ) have recently been proposed to be due to putative inhibition of Ca 2+ entry and Ca 2+ -signaling and competition with Ca 2+ for intracellular Ca 2+ -binding proteins. The tissue partitioning of cobalt (Co 2+ ) and its time-dependence after administration of a single dose have been studied in man, but mainly in laboratory animals. Cobalt is accumulated primarily in liver, kidney, pancreas, and heart, with the relative content in skeleton and skeletal muscle increasing with time after cobalt administration. In man the renal excretion is initially rapid but decreasing over the first days, followed by a second, slow phase lasting several weeks, and

Self-regulating characteristics of cold neutron source with annular cylindrical moderator cell

International Nuclear Information System (INIS)

Kawai, Takeshi; Lee Chien-Hsiung; Chan Yea-Kuang; Guung Tai-Cheng; Yoshino, Hirofumi; Kawabata, Yuji; Hino, Masahiro

2001-01-01

The conditions, in which the ORPHEE type cold neutron source with an annular cylindrical moderator cell could have self-regulating characteristics, were obtained through thermodynamic considerations. From a viewpoint of engineering, it is not easy to establish these conditions because three parameters are involved even in an idealized system without the effect of the mass transfer resistance in the moderator transfer tube between the condenser and the moderator cell. The inner shell of the ORPHEE moderator cell is open in the bottom, but it is expected that only hydrogen vapor is contained in the inner shell and liquid hydrogen in the outer shell. The thermodynamic considerations show that such a state is maintained only when a liquefaction capacity of the condenser is large compared to heat lead and three parameters are optimized with a good balance. We proposed another type of a moderator cell, which has an inner cylindrical cavity with no hole in the bottom but a vapor inlet opening at the uppermost part of the cavity. In this structure, a self-regulating characteristic is established easily and the liquid level in the outer shell is maintained almost constant against thermal disturbances. Therefore it is enough to control one parameter, that is, the reservoir tank pressure corresponding to the liquefaction capacity of the condenser given by the refrigeration power of the helium refrigerator. (author)

When dispositional and role power fit: implications for self-expression and self-other congruence.

Science.gov (United States)

Chen, Serena; Langner, Carrie A; Mendoza-Denton, Rodolfo

2009-03-01

Integrating and extending the literatures on social power and person-environment fit, 4 studies tested the hypothesis that when people's dispositional beliefs about their capacity to influence others fit their assigned role power, they are more likely to engage in self-expression-that is, behave in line with their states and traits-thereby increasing their likelihood of being perceived by others in a manner congruent with their own self-judgments (i.e., self-other congruence). In Studies 1-3, dispositionally high- and low-power participants were randomly assigned to play a high- or low-power role in an interaction with a confederate. When participants' dispositional and role power fit (vs. conflicted), they reported greater self-expression (Study 1). Furthermore, under dispositional-role power fit conditions, the confederate's ratings of participants' emotional experiences (Study 2) and personality traits (Study 3) were more congruent with participants' self-reported emotions and traits. Study 4's results replicated Study 3's results using an implicit manipulation of power and outside observers' (rather than a confederate's) ratings of participants. Implications for research on power and person perception are discussed.

Neutron scattering. Lectures

Energy Technology Data Exchange (ETDEWEB)

Brueckel, Thomas; Heger, Gernot; Richter, Dieter; Roth, Georg; Zorn, Reiner (eds.)

2010-07-01

The following topics are dealt with: Neutron sources, neutron properties and elastic scattering, correlation functions measured by scattering experiments, symmetry of crystals, applications of neutron scattering, polarized-neutron scattering and polarization analysis, structural analysis, magnetic and lattice excitation studied by inelastic neutron scattering, macromolecules and self-assembly, dynamics of macromolecules, correlated electrons in complex transition-metal oxides, surfaces, interfaces, and thin films investigated by neutron reflectometry, nanomagnetism. (HSI)

Neutron scattering. Lectures

International Nuclear Information System (INIS)

Brueckel, Thomas; Heger, Gernot; Richter, Dieter; Roth, Georg; Zorn, Reiner

2010-01-01

The following topics are dealt with: Neutron sources, neutron properties and elastic scattering, correlation functions measured by scattering experiments, symmetry of crystals, applications of neutron scattering, polarized-neutron scattering and polarization analysis, structural analysis, magnetic and lattice excitation studied by inelastic neutron scattering, macromolecules and self-assembly, dynamics of macromolecules, correlated electrons in complex transition-metal oxides, surfaces, interfaces, and thin films investigated by neutron reflectometry, nanomagnetism. (HSI)

Valence states of cobalt and crystal structure peculiarities of solid solution YBa2Cu3-xCoxO6+σ

International Nuclear Information System (INIS)

Voronin, V.I.; Goshchinskij, B.N.; Mitberg, Eh.B.; Leonidov, I.A.; Kozhevnikov, V.L.

2000-01-01

Crystal structure of solid solution YBa 2 Cu 3-x Co x O 6+σ , where x = 0.2, 0.4, 0.6 and 0.8, is studied by the method of powder neutron diffraction. Charge states of the cation are calculated using the interatomic distances obtained. It is shown that cobalt in Cu1 position has valency 3 + and octahedral coordination at x = 0.2 and 0.4. Increase in doping degree involves both transition of a portion of cobalt ions in the positions mentioned to the state with valence 4 + and tetrahedral coordination and partial substitution of copper in Cu2 position [ru

Power generation using spinel manganese-cobalt oxide as a cathode catalyst for microbial fuel cell applications.

Science.gov (United States)

Mahmoud, Mohamed; Gad-Allah, Tarek A; El-Khatib, K M; El-Gohary, Fatma

2011-11-01

This study focused on the use of spinel manganese-cobalt (Mn-Co) oxide, prepared by a solid state reaction, as a cathode catalyst to replace platinum in microbial fuel cells (MFCs) applications. Spinel Mn-Co oxides, with an Mn/Co atomic ratios of 0.5, 1, and 2, were prepared and examined in an air cathode MFCs which was fed with a molasses-laden synthetic wastewater and operated in batch mode. Among the three Mn-Co oxide cathodes and after 300 h of operation, the Mn-Co oxide catalyst with Mn/Co atomic ratio of 2 (MnCo-2) exhibited the highest power generation 113 mW/m2 at cell potential of 279 mV, which were lower than those for the Pt catalyst (148 mW/m2 and 325 mV, respectively). This study indicated that using spinel Mn-Co oxide to replace platinum as a cathodic catalyst enhances power generation, increases contaminant removal, and substantially reduces the cost of MFCs. Copyright © 2011 Elsevier Ltd. All rights reserved.

Cobalt metabolism and toxicology-A brief update

Energy Technology Data Exchange (ETDEWEB)

Simonsen, Lars Ole, E-mail: LOSimonsen@dadlnet.dk; Harbak, Henrik; Bennekou, Poul

2012-08-15

Cobalt metabolism and toxicology are summarized. The biological functions of cobalt are updated in the light of recent understanding of cobalt interference with the sensing in almost all animal cells of oxygen deficiency (hypoxia). Cobalt (Co{sup 2+}) stabilizes the transcriptional activator hypoxia-inducible factor (HIF) and thus mimics hypoxia and stimulates erythropoietin (Epo) production, but probably also by the same mechanism induces a coordinated up-regulation of a number of adaptive responses to hypoxia, many with potential carcinogenic effects. This means on the other hand that cobalt (Co{sup 2+}) also may have beneficial effects under conditions of tissue hypoxia, and possibly can represent an alternative to hypoxic preconditioning. Cobalt is acutely toxic in larger doses, and in mammalian in vitro test systems cobalt ions and cobalt metal are cytotoxic and induce apoptosis and at higher concentrations necrosis with inflammatory response. Cobalt metal and salts are also genotoxic, mainly caused by oxidative DNA damage by reactive oxygen species, perhaps combined with inhibition of DNA repair. Of note, the evidence for carcinogenicity of cobalt metal and cobalt sulfate is considered sufficient in experimental animals, but is as yet considered inadequate in humans. Interestingly, some of the toxic effects of cobalt (Co{sup 2+}) have recently been proposed to be due to putative inhibition of Ca{sup 2+} entry and Ca{sup 2+}-signaling and competition with Ca{sup 2+} for intracellular Ca{sup 2+}-binding proteins. The tissue partitioning of cobalt (Co{sup 2+}) and its time-dependence after administration of a single dose have been studied in man, but mainly in laboratory animals. Cobalt is accumulated primarily in liver, kidney, pancreas, and heart, with the relative content in skeleton and skeletal muscle increasing with time after cobalt administration. In man the renal excretion is initially rapid but decreasing over the first days, followed by a second, slow

Design and fabrication of self-powered micro-harvesters rotating and vibrated micro-power systems

CERN Document Server

Pan, C T; Lin, Liwei; Chen, Ying-Chung

2013-01-01

Presents the latest methods for designing and fabricating self-powered micro-generators and energy harvester systems Design and Fabrication of Self-Powered Micro-Harvesters introduces the latest trends of self-powered generators and energy harvester systems, including the design, analysis and fabrication of micro power systems. Presented in four distinct parts, the authors explore the design and fabrication of: vibration-induced electromagnetic micro-generators; rotary electromagnetic micro-generators; flexible piezo-micro-generator with various widths; and PVDF electrospunpiezo-energy with

Cobalt Fischer-Tropsch catalysts: influence of cobalt dispersion and titanium oxides promotion

Energy Technology Data Exchange (ETDEWEB)

Azib, H

1996-04-10

The aim of this work is to study the effect of Sol-Gel preparation parameters which occur in silica supported cobalt catalysts synthesis. These catalysts are particularly used for the waxes production in natural gas processing. The solids have been characterized by several techniques: transmission electron microscopy (TEM), X-ray absorption near edge spectroscopy (XANES), programmed temperature reduction (TPR), infrared spectroscopy (IR), ultraviolet spectroscopy (UV), Magnetism, thermodesorption of H{sub 2} (TPD). The results indicate that the control of the cobalt dispersion and oxide phases nature is possible by modifying Sol-Gel parameters. The catalytic tests in Fischer-Tropsch synthesis were conducted on a pilot unit under pressure (20 atm) and suggested that turnover rates were independent of Co crystallite size, Co phases in the solids (Co deg., cobalt silicate) and titanium oxide promotion. On the other methane, the C{sub 3}{sup +} hydrocarbon selectivity is increased with increasing crystallite size. Inversely, the methane production is favoured by very small crystallites, cobalt silicate increase and titanium addition. However, the latter, used as a cobalt promoter, has a benefic effect on the active phase stability during the synthesis. (author). 149 refs., 102 figs., 71 tabs.

Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

International Nuclear Information System (INIS)

Bernard, J.A.

1989-09-01

This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the 'MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power on spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs

Cobalt accumulation and circulation by blackgum trees

International Nuclear Information System (INIS)

Thomas, W.A.

1975-01-01

Blackgum (Nyssa sylvatica Marsh.) trees accumulate far greater concentrations of cobalt in mature foliage than do other species on the same site (363 ppM in ash of blackgum, compared with about 3 ppM by mockernut hickory and about 1 ppM by red maple, tulip tree, and white oak). Cobalt concentrations in dormant woody tissues of blackgum also significantly exceed those in the other four species. Inoculation of six blackgums with 60 Co revealed that cobalt remains mobile in the trees for at least 3 years. Foliar concentrations of stable cobalt increase uniformly until senescence. In late August, foliage accounts for only 9 percent of total tree weight but 57 percent of total tree cobalt. Losses of cobalt from trees occur almost entirely by leaf abscission, and the loss rates of weight and cobalt from decomposing litter are similar. Retention of cobalt in the biologically active soil layers perpetuates zones of cobalt concentration created by this species in woodlands

Electroplated zinc-cobalt alloy

International Nuclear Information System (INIS)

Carpenter, D.E.O.S.; Farr, J.P.G.

2005-01-01

Recent work on the deposition and use of ectrodeposited zinc-cobalt alloys is surveyed. Alloys containing lower of Nuclear quantities of cobalt are potentially more useful. The structures of the deposits is related to their chemical and mechanical properties. The inclusion of oxide and its role in the deposition mechanism may be significant. Chemical and engineering properties relate to the metallurgical structure of the alloys, which derives from the mechanism of deposition. The inclusion of oxides and hydroxides in the electroplate may provide evidence for this mechanism. Electrochemical impedance measurements have been made at significant deposition potentials, in alkaline electrolytes. These reveal a complex electrode behaviour which depends not only on the electrode potential but on the Co content of the electrolyte. For the relevant range of cathodic potential zinc-cobalt alloy electrodeposition occurs through a stratified interface. The formation of an absorbed layer ZnOH/sup +/ is the initial step, this inhibits the deposition of cobalt at low cathodic potentials, so explaining its 'anomalous deposition'. A porous layer of zinc forms on the adsorbed ZnOH/sup +/ at underpotential. As the potential becomes more cathodic, cobalt co- deposits from its electrolytic complex forming a metallic solid solution of Co in Zn. In electrolytes containing a high concentration of cobalt a mixed entity (ZnCo)/sub +/ is assumed to adsorb at the cathode from which a CoZn intermetallic deposits. (author)

Pulsed power supply system for neutron well logging

International Nuclear Information System (INIS)

Hopkinson, E.C.

1975-01-01

A variable D. C. power supply and a variable A. C. power supply are coupled onto the upper end of a single conductor cable at the earth's surface and from the lower end of said cable to a well logging instrument. The A. C. voltage is used to provide filament power for the ion source accelerator tube. The D. C. voltage is used to provide power to fixed D. C. loads maintained constant by regulators once a threshold is reached. The D. C. voltage is raised above the threshold to control the pulsed acceleration voltage and hence neutron output by first feeding into a unijunction relaxation oscillator in combination with an SCR output which output is transformer coupled into a voltage multiplier circuit. An antilatch feature is provided for the SCR by transformer coupling the pulses on the SCR anode back to the base of a transistor in series with the cathode of the SCR. Two outputs of the voltage multiplier circuit are connected to the cathode and anode, respectively, of an ion source accelerator tube, the cathode being connected through a resistor to retard the ripple pulsing of the cathode to allow ionization of the accelerator tube

Self-Powered Sun Sensor Microsystems

NARCIS (Netherlands)

Wu, H.; Emadi, A.; Graaf, G. de; Leijtens, J.A.P.; Wolffenbuttel, R.F.

2009-01-01

An analog sun sensor has been designed based on shade profile proportional to the angle of incidence of incoming light projected onto a 2×2 array of photodiodes. This concept enables an autonomous self-powered optical system with two the main functions (electrical power generation for the amplifier

Cobalt allergy in hard metal workers

Energy Technology Data Exchange (ETDEWEB)

Fischer, T; Rystedt, I

1983-03-01

Hard metal contains about 10% cobalt. 853 hard metal workers were examined and patch tested with substances from their environment. Initial patch tests with 1% cobalt chloride showed 62 positive reactions. By means of secondary serial dilution tests, allergic reactions to cobalt were reproduced in 9 men and 30 women. Weak reactions could not normally be reproduced. A history of hand eczema was found in 36 of the 39 individuals with reproducible positive test reactions to cobalt, while 21 of 23 with a positive initial patch test but negative serial dilution test had never had any skin problems. Hand etching and hand grinding, mainly female activities and traumatic to the hands, were found to involve the greatest risk of cobalt sensitization. 24 individuals had an isolated cobalt allergy. They had probably been sensitized by hard metal work, while the individuals, all women, who had simultaneous nickel allergy had probably been sensitized to nickel before their employment and then became sensitized to cobalt by hard metal work. A traumatic occupation, which causes irritant contact dermatitis and/or a previous contact allergy or atopy is probably a prerequisite for the development of cobalt allergy.

Pharmacokinetics of inorganic cobalt and a vitamin B12 supplement in the Thoroughbred horse: Differentiating cobalt abuse from supplementation.

Science.gov (United States)

Hillyer, L L; Ridd, Z; Fenwick, S; Hincks, P; Paine, S W

2018-05-01

While cobalt is an essential micronutrient for vitamin B 12 synthesis in the horse, at supraphysiological concentrations, it has been shown to enhance performance in human subjects and rats, and there is evidence that its administration in high doses to horses poses a welfare threat. Animal sport regulators currently control cobalt abuse via international race day thresholds, but this work was initiated to explore means of potentially adding to application of those thresholds since cobalt may be present in physiological concentrations. To devise a scientific basis for differentiation between presence of cobalt from bona fide supplementation and cobalt doping through the use of ratios. Six Thoroughbred horses were given 10 mL vitamin B 12 /cobalt supplement (Hemo-15 ® ; Vetoquinol, Buckingham, Buckinghamshire, UK., 1.5 mg B 12 , 7 mg cobalt gluconate = 983 μg total Co) as an i.v. bolus then an i.v. infusion (15 min) of 100 mg cobalt chloride (45.39 mg Co) 6 weeks later. Pre-and post-administration plasma and urine samples were analysed for cobalt and vitamin B 12 . Urine and plasma samples were analysed for vitamin B 12 using an immunoassay and cobalt concentrations were measured via ICP-MS. Baseline concentrations of cobalt in urine and plasma for each horse were subtracted from their cobalt concentrations post-administration for the PK analysis. Compartmental analysis was used for the determination of plasma PK parameters for cobalt using commercially available software. On administration of a vitamin B 12 /cobalt supplement, the ratio of cobalt to vitamin B 12 in plasma rapidly increased to approximately 3 and then rapidly declined below a ratio of 1 and then back to near baseline over the next week. On administration of 100 mg cobalt chloride, the ratio initially exceeded 10 in plasma and then declined with the lower 95% confidence interval remaining above a ratio of 1 for 7 days. For two horses with extended sampling, the plasma ratio remained above one for

Novel design of a self powered and self sensing magneto-rheological damper

International Nuclear Information System (INIS)

Ferdaus, Mohammad Meftahul; Rashid, M M; Bhuiyan, M M I; Muthalif, Asan Gani Bin Abdul; Hasan, M R

2013-01-01

Magneto-rheological (MR) dampers are semi-active control devices and use MR fluids. Magneto-rheological dampers have successful applications in mechatronics engineering, civil engineering and numerous areas of engineering. At present, traditional MR damper systems, require a isolated power supply and dynamic sensor. This paper presents the achievability and accuracy of a self- powered and self-sensing magneto-rheological damper using harvested energy from the vibration and shock environment in which it is deployed and another important part of this paper is the increased yield stress of the Magneto rheological Fluids. Magneto rheological fluids using replacement of glass beads for Magnetic Particles to surge yield stress is implemented here. Clearly this shows better result on yield stress, viscosity, and settling rate. Also permanent magnet generator (PMG) is designed and attached to a MR damper. For evaluating the self-powered MR damper's vibration mitigating capacity, an Engine Mount System using the MR damper is simulated. The ideal stiffness of the PMG for the Engine Mount System (EMS) is calculated by numerical study. The vibration mitigating performance of the EMS employing the self-powered and self sensing MR damper is theoretically calculated and evaluated in the frequency domain

Novel design of a self powered and self sensing magneto-rheological damper

Science.gov (United States)

Meftahul Ferdaus, Mohammad; Rashid, M. M.; Bhuiyan, M. M. I.; Muthalif, Asan Gani Bin Abdul; Hasan, M. R.

2013-12-01

Magneto-rheological (MR) dampers are semi-active control devices and use MR fluids. Magneto-rheological dampers have successful applications in mechatronics engineering, civil engineering and numerous areas of engineering. At present, traditional MR damper systems, require a isolated power supply and dynamic sensor. This paper presents the achievability and accuracy of a self- powered and self-sensing magneto-rheological damper using harvested energy from the vibration and shock environment in which it is deployed and another important part of this paper is the increased yield stress of the Magneto rheological Fluids. Magneto rheological fluids using replacement of glass beads for Magnetic Particles to surge yield stress is implemented here. Clearly this shows better result on yield stress, viscosity, and settling rate. Also permanent magnet generator (PMG) is designed and attached to a MR damper. For evaluating the self-powered MR damper's vibration mitigating capacity, an Engine Mount System using the MR damper is simulated. The ideal stiffness of the PMG for the Engine Mount System (EMS) is calculated by numerical study. The vibration mitigating performance of the EMS employing the self-powered & self sensing MR damper is theoretically calculated and evaluated in the frequency domain.

COBALT SALTS PRODUCTION BY USING SOLVENT EXTRACTION

Directory of Open Access Journals (Sweden)

Liudmila V. Dyakova

2010-06-01

Full Text Available The paper deals with the extracting cobalt salts by using mixtures on the basis of tertiary amine from multicomponent solutions from the process of hydrochloride leaching of cobalt concentrate. The optimal composition for the extraction mixture, the relationship between the cobalt distribution coefficients and modifier’s nature and concentration, and the saltingout agent type have been determined. A hydrochloride extraction technology of cobalt concentrate yielding a purified concentrated cobalt solution for the production of pure cobalt salts has been developed and introduced at Severonikel combine.

Calcium-assisted reduction of cobalt ferrite nanoparticles for nanostructured iron cobalt with enhanced magnetic performance

International Nuclear Information System (INIS)

Qi, B.; Andrew, J. S.; Arnold, D. P.

2017-01-01

This paper demonstrates the potential of a calcium-assisted reduction process for synthesizing fine-grain (~100 nm) metal alloys from metal oxide nanoparticles. To demonstrate the process, an iron cobalt alloy (Fe_6_6Co_3_4) is obtained by hydrogen annealing 7-nm cobalt ferrite (CoFe_2O_4) nanoparticles in the presence of calcium granules. The calcium serves as a strong reducing agent, promoting the phase transition from cobalt ferrite to a metallic iron cobalt alloy, while maintaining high crystallinity. Magnetic measurements demonstrate the annealing temperature is the dominant factor of tuning the grain size and magnetic properties. Annealing at 700 °C for 1 h maximizes the magnetic saturation, up to 2.4 T (235 emu/g), which matches that of bulk iron cobalt.

Calcium-assisted reduction of cobalt ferrite nanoparticles for nanostructured iron cobalt with enhanced magnetic performance

Energy Technology Data Exchange (ETDEWEB)

Qi, B. [University of Florida, Interdisciplinary Microsystems Group, Department of Electrical and Computer Engineering (United States); Andrew, J. S. [University of Florida, Department of Materials Science and Engineering (United States); Arnold, D. P., E-mail: darnold@ufl.edu [University of Florida, Interdisciplinary Microsystems Group, Department of Electrical and Computer Engineering (United States)

2017-03-15

This paper demonstrates the potential of a calcium-assisted reduction process for synthesizing fine-grain (~100 nm) metal alloys from metal oxide nanoparticles. To demonstrate the process, an iron cobalt alloy (Fe{sub 66}Co{sub 34}) is obtained by hydrogen annealing 7-nm cobalt ferrite (CoFe{sub 2}O{sub 4}) nanoparticles in the presence of calcium granules. The calcium serves as a strong reducing agent, promoting the phase transition from cobalt ferrite to a metallic iron cobalt alloy, while maintaining high crystallinity. Magnetic measurements demonstrate the annealing temperature is the dominant factor of tuning the grain size and magnetic properties. Annealing at 700 °C for 1 h maximizes the magnetic saturation, up to 2.4 T (235 emu/g), which matches that of bulk iron cobalt.

A Self-Powered Insole for Human Motion Recognition

Directory of Open Access Journals (Sweden)

Yingzhou Han

2016-09-01

Full Text Available Biomechanical energy harvesting is a feasible solution for powering wearable sensors by directly driving electronics or acting as wearable self-powered sensors. A wearable insole that not only can harvest energy from foot pressure during walking but also can serve as a self-powered human motion recognition sensor is reported. The insole is designed as a sandwich structure consisting of two wavy silica gel film separated by a flexible piezoelectric foil stave, which has higher performance compared with conventional piezoelectric harvesters with cantilever structure. The energy harvesting insole is capable of driving some common electronics by scavenging energy from human walking. Moreover, it can be used to recognize human motion as the waveforms it generates change when people are in different locomotion modes. It is demonstrated that different types of human motion such as walking and running are clearly classified by the insole without any external power source. This work not only expands the applications of piezoelectric energy harvesters for wearable power supplies and self-powered sensors, but also provides possible approaches for wearable self-powered human motion monitoring that is of great importance in many fields such as rehabilitation and sports science.

Self-oscillating resonant power converter

DEFF Research Database (Denmark)

2014-01-01

The present invention relates to resonant power converters and inverters comprising a self-oscillating feedback loop coupled from a switch output to a control input of a switching network comprising one or more semiconductor switches. The self-oscillating feedback loop sets a switching frequency...... of the power converter and comprises a first intrinsic switch capacitance coupled between a switch output and a control input of the switching network and a first inductor. The first inductor is coupled in-between a first bias voltage source and the control input of the switching network and has...... a substantially fixed inductance. The first bias voltage source is configured to generate an adjustable bias voltage applied to the first inductor. The output voltage of the power converter is controlled in a flexible and rapid manner by controlling the adjustable bias voltage....

Analysing power for neutron-proton scattering at 14.1 MeV

International Nuclear Information System (INIS)

Brock, J.E.; Chisholm, A.; Duder, J.C.; Garrett, R.; Poletti, J.L.

1981-01-01

The analysing power Asub(y)(theta) for neutron-proton scattering has been measured at 14.1 MeV for c.m. angles between 50 0 and 157 0 . A polarized neutron beam was produced by the reaction 3 H(d,n) 4 He at 110 keV, using polarized deuterons from an atomic beam polarized ion source. Liquid and plastic scintillators were used for proton targets and the scattered particles were detected in an array of platic scintillators. Use of the associated alpha technique, multi-parameter recording of events and off-line computer treatment led to very low backgrounds. The results differ significantly from the predictions of the phase-shift analyses of Yale IV, Livermore X and Arndt et al. We find, however, excellent agreement with the predictions of the Paris potential of Lacombe et al. Existing n-p analysing power results up to 30 MeV are surveyed and found to be consistent. An attempt was made to look for an isospin splitting of the triplet P-wave phase shifts. (orig.)

Spin density measurement of water-bridged Co-dimer using polarized neutrons

DEFF Research Database (Denmark)

Damgaard-Møller, Emil; Overgaard, Jacob; Chilton, Nick

present an experimentally determined spin density using polarized neutron diffraction in a simple water-bridged cobalt dimer [Co2(H2O)(piv)4(Hpiv)2(py)2] which is known to have a small ferromagnetic coupling between the spin centers. Visualizing the SDD could get us one step further in understanding...

Notes on neutron flux measurement; Notas sobre medida de flujos neutronicos

Energy Technology Data Exchange (ETDEWEB)

Alcala Ruiz, F

1984-07-01

The main purpose of this work is to get an useful guide to carry out topical neutron flux measurements. Although the foil activation technique is used in the majority of the cases, other techniques, such as those based on fission chambers and self-powered neutron detectors, are also shown. Special interest is given to the description and application of corrections on the measurement of relative and absolute induced activities by several types of detectors (scintillators, G-M and gas proportional counters). The thermal arid epithermal neutron fluxes, as determined in this work, are conventional or effective (West cots fluxes), which are extensively used by the reactor experimentalists; however, we also give some expressions where they are related to the integrated neutron fluxes, which are used in neutron calculations. (Author) 16 refs.