Showing papers in "Journal of Nuclear Science and Technology in 1981"
TL;DR: The exposure rate due to natural radiation composed of terrestrial radiation, cosmic ray as well as a slight contribution from nuclear tests was estimated at 1,115 sites by an ionization chamber and sodium iodide scintillation surveymeters.
Abstract: The exposure rate due to natural radiation composed of terrestrial radiation, cosmic ray as well as a slight contribution from nuclear tests was estimated at 1,115 sites by an ionization chamber and sodium iodide scintillation surveymeters. The distribution of the exposure rates along with the population-weighted annual collective dose in each area were obtained. The mean exposure rate of all Japan was calculated to be 9.0 μR/h. The population-weighted mean dose for an individual person was 68.7 mrad/yr. Uncertainties involved in the field measurements and calibrations were estimated to be satisfactorily low. However, the result cannot be interpreted as to its very small variation because the sampling error could have been significant. Nevertheless the gross feature is meaningful. There is a general trend that South-West Japan has higher level than North- East Japan. It was also revealed that the population size is a more important factor than the radiation level in determining the distribution of the pop...
67 citations
TL;DR: In this article, the authors used the permeation time-lag method to determine the permeabilities, diffusion coefficients and solubilities of hydrogen and deuterium in type 304 stainless steel of three kinds of surface treated specimens; with oxide film, reduced by H2 gas and Pd coated.
Abstract: The permeation time-lag method have been used to determine the permeabilities, diffusion coefficients and solubilities of hydrogen and deuterium in type 304 stainless steel of three kinds of surface treated specimens; with oxide film, reduced by H2 gas and Pd coated. For the specimen with a thin oxide film the permeability increased abruptly at the temperature higher than 1,050 K owing to reduction of the oxide film by H2 gas introduced. The permeability and diffusion coefficient for hydrogen-reduced specimens agreed with those for Pd-coated specimens showing the data of bulk diffusion limited. The isotope effects for permeability ΦH/ΦD, diffusion coefficient D H/D D and solubility were about 1.4, 1.2 and 1.1, respectively. From these values θ(=hν/k)=1,530±50K and θ′(=hν/k)=2,740±20K were obtained by the quantum effect of a harmonic approximation.
50 citations
TL;DR: In this paper, the inner surface of Zircaloy-4 cladding is exposed to the environmental atmosphere provided by steam flowing at a constant rate in the range of 0.13 − 1.6 g/cm2-min.
Abstract: Embrittlement of Zircaloy-4 cladding by oxidation of the inner surface occurring in an LWR loss-of-coolant accident was studied using simulated fuel containing of A12O3 pellets sheathed in Zircaloy-4 specimen cladding, filled with Ar gas, and sealed. This simulated fuel rod was heated from outside until the isothermal oxidation temperature between 880 and 1,167°C was obtained after the cladding burst. This exposed the inner surface of the cladding to the environmental atmosphere, provided by steam flowing at a constant rate in the range of 0.13–1.6 g/cm2-min. The embrittlement of the specimen due to inner surface oxidation is influenced primarily by the amount of hydrogen absorbed by the Zircaloy-4. Ring compression tests conducted at 100°C on test pieces constituted of sliced sections of oxidized specimen showed that Zircaloy containing more than 200–300 wt.ppm of absorbed hydrogen became brittle when oxidized at temperatures above 1,000°C. In the range of oxidation temperature 932 to 972°C, brittleness ...
49 citations
TL;DR: In this paper, an experimental study was conducted on pool-boiling heat transfer from an electrically heated horizontal wire to saturated liquid nitrogen at atmospheric pressure Experimental results of heat transfer characteristics in both nucleate and filmboiling regimes, critical heat flux and minimum heat flux were analyzed and compared with various correlations in addition, photography was used to obtain information concerning the vapor bubble and vapor-film behaviors around the heating wire.
Abstract: An experimental study was conducted on pool-boiling heat transfer from an electrically heated horizontal wire to saturated liquid nitrogen at atmospheric pressure Experimental results of heat transfer characteristics in both nucleate- and film-boiling regimes, critical heat flux and minimum heat flux were analyzed and compared with various correlations In addition, photography was used to obtain information concerning the vapor-bubble and vapor-film behaviors around the heating wire These data of microcharacteristics were utilized for evaluating theoretical models of nucleate-boiling mechanism proposed by other investigators Transient conduction to the relatively cold (saturated) liquid, which came into the space vacated by the departing bubble and contacted with the heating surface, was found to be important for high heat transfer rates associated with nucleate boiling
39 citations
TL;DR: In this article, it was shown that a full adoption of recent publications of decay schemes to derive average energies of β- and γ-rays, Ēβ, and Ēγ, leads to a large underestimation of the γ -ray component of the decay heat and to an overestimation of α-ray component.
Abstract: Fission product decay heat of 235U was calculated for short cooling-time on the basis of a preliminary version of a new decay data library recently completed by the Japanese Nuclear Data Committee. It was shown that a full adoption of recent publications of decay schemes to derive average energies of β- and γ-rays, Ēβ, and Ēγ, leads to a large underestimation of the γ-ray component of the decay heat and to an overestimation of the β-ray component. In order to remedy this, theoretical values of Ēβ and Ēγ were introduced for high Q-value decays, which were obtained with a gross theory of β-decay. It improved remarkably the agreement between calculation and experiment not only for the 235U decay heat but for 239Pu and 241Pu as well. It was concluded that a large part of decay schemes recently published for high Q-value nuclides are inappropriate to use in calculations of Ēβ and Ēγ, because they fail to reproduce the effect of β- strengths at high excitations, which makes Ēβ small and Ēγ large. The use of the...
24 citations
TL;DR: The STP analysis in the present study intends to clarify through fluctuation analysis the cause-effect relationship between measured signals in both qualitative and quantitative manner and defines a new function called signal power contribution ratio which represents relative contributions to the output from the input.
Abstract: This paper is concerned with a method to evaluate signal transmission paths (STP) in dynamic systems. The STP analysis in the present study intends to clarify through fluctuation analysis the cause-effect relationship between measured signals in both qualitative and quantitative manner. In order to facilitate the analysis, we define a new function called signal power contribution ratio which represents relative contributions to the output from the input. The analysis is executed through comparative evaluation of the coherence, noise power contribution and signal power contribution functions calculated based upon autoregres-sive modeling. It is shown that comparison of these three functions provides information useful to evaluate the signal transmission paths. The present method has been applied to noise signals collected at Japan Power Demonstration Reactor (JPDR)-II. The result of the analysis suggested its effectiveness for investigation of at-power reactor noise, especially in identifying possible nois...
23 citations
Journal Article•
21 citations
TL;DR: In this article, a correlation of a heat transfer coefficient was derived for the transition flow region between the quench front and the dispersed flow region during the reflood phase, which showed fairly good agreement with the experimental data.
Abstract: A correlation of a heat transfer coefficient was derived for the transition flow region between the quench front and the dispersed flow region during the reflood phase. The correlation was compared with data from a small scale reflood experiment with a 4×4 heater-rod bundle and FLECHT experiments and showed fairly good agreement with the experimental data. The predicted errors were nearly within ±30% for all the data of the referred small scale reflood experiment, ±20% for all data except for the data of high flooding rate cases (higher than 7.5×l0−2m/s) and the data measured at the location just below the spacer grid, and ±20% for two selected FLECHT tests.
19 citations
TL;DR: In this article, the inner surface oxidation of Zircaloy claddings was investigated and the degradation of fracture load and the deflection at 373 K were largely influenced by the hydrogen absorption during the inner-surface oxidation.
Abstract: Under a LOCA transient in LWRs, ruptured Zircaloy clad tubes would be oxidized at the inner surface as well as outer surface by steam flowing in reactor core. In order to determine whether the inner-surface oxidation influences mechanical properties of clad tubes during the reflood stage of a LOCA, the ring compression tests were conducted on oxidized tube specimens and segments sectioned from ruptured claddings obtained from rod-burst/oxidation tests ranging from 1,200 to 1,500 K. The degradation of the fracture load and the deflection at 373 K are largely influenced by the hydrogen absorption during the inner-surface oxidation. The oxidation temperature and the time dependence of the ductility loss of Zircaloy clad tubes can be represented by iso-deflection lines. The iso-deflection lines due to the hydrogen absorption limit the permissible oxidation condition to a lower temperature and shorter period of exposure time, comparing with sole oxygen uptake.
17 citations
TL;DR: In this paper, the most suitable condition has been established for fabricating uranium-plutonium mixed carbide pellets by the mechanical-blend carbothermic reduction and sintering method.
Abstract: Uranium-plutonium mixed monocarbide pellets, which had densities over 90% T.D. and near-stoichiometric compositions, were fabricated in the gloveboxes of the atmosphere of purified Ar. The carbide powder was first synthesized by reducing mechanically blended oxides with graphite at 1,510°C in a vacuum of 0.2 Pa. This condition was selected so as to minimize the loss of Pu during the reduction in vacuum. The synthesized carbide was milled and compacted, followed by sintering in flowing Ar at 1,540 to 1,750°C. The experimental conditions covered wide ranges of milling time, compacting pressure, sintering temperature and sintering time. The most suitable condition has been established for fabricating uranium-plutonium mixed carbide pellets by the mechanical-blend carbothermic reduction and sintering method. For the case of Ni addition as a sintering aid, the efficacy of the addition for fabricating dense pellets was confirmed, but it was noticed that the composition tended to hyperstoichiometry and the sesqu...
17 citations
TL;DR: In this paper, post-irradiation examinations were performed on three types of uranium carbide UC, UC2 and U2C3, that were irradiated with a relatively high thermal neutron flux (7×1013 n/cm2·s or 9×l012 fission/cm3·s for UC) and in a dose range between 2.7×l015 and 3.3×1018 fission /cm3 (i.e. 2.1×l016 and 2.4× 1019 nvt) in the JMTR (Japan
Abstract: Post-irradiation examinations were performed on three types of uranium carbide UC, UC2 and U2C3, that were irradiated with a relatively high thermal neutron flux (7×1013 n/cm2·s or 9×l012 fission/cm3·s for UC) and in a dose range between 2.7×l015 and 3.3×1018 fission/cm3 (i.e. 2.1×l016 and 2.4× 1019 nvt) in the JMTR (Japan Material Testing Reactor). On UC and UC2, trends similar to previous works were obtained in the irradiation effects. New preliminary results, however, were obtained on U2C3 for changes in the electrical resistivity and the lattice parameter which showed reduced values after attaining a maximum at 1017 fission/cm3. Successive annealing effects on the resistivity and the lattice parameter following the reactor irradiations were examined by pulse heating. Two major steps were observed, at around 400 and 600°C, in the recovery processes. A low temperature step, which was revealed previously at about 200°C in UC and UC2, was missing in this study, because of higher irradiation temperatures (...
TL;DR: In this article, a computer code is developed for effective and exact simulation of cryogenic distillation columns, which incorporates the differences of latent heat of vaporization among the six molecular species, pressure drop, decay heat of tritium, heat transfer through the column wall, Raoult's law deviation of the solutions of hydrogen isotopes, multiple feeds and multiple sidestreams.
Abstract: A computer code is developed for effective and exact simulation of cryogenic distillation columns. The simulation model incorporates the differences of latent heat of vaporization among the six molecular species, pressure drop, decay heat of tritium, heat transfer through the column wall, Raoult's law deviation of the solutions of hydrogen isotopes, multiple feeds and multiple sidestreams. The calculational procedure developed by Tomich for general equilibrium stage processes is further modified to match this complicated simulation model. A computer simulation study is performed for the typical cryogenic distillation column cascade and separation characteristics of each column are clarified. The effects of pressure drop are found not significant. If the tritium concentration in the column is considerably high, the lower performance caused by decay heat effects cannot be neglected any longer. Increase of the reflux ratio or refrigeration of the stripping section, is required for improvement. The present st...
TL;DR: In this article, the SAND-II code was used to evaluate the thermal-to-intermediate transition energy region for spectrum unfolding, with the cross section data for all the detector foils taken from ENDF/B-IV self-shielding corrections are successfully applied on the activation cross-section data of the foils.
Abstract: To evaluate the neutron energy spectra in the irradiation field of a materials testing reactor, the corresponding spectra are determined on a critical facility of identical core configulation by multi-foil activation measurements The validity of this method is examined In the present work, emphasis is laid on the thermal-to-intermediate transition energy region For spectrum unfolding, the SAND-II code is utilized, with the cross section data for all the detector foils taken from ENDF/B-IV Self-shielding corrections are successfully applied on the activation cross section data of the foils, being based on a mono-energetic model An examination is made of the error incurred by this expedient The propriety of the correction depends on the manner in which the scattering effect is treated The spectrum obtained can be considered unique from the fact that the thermal-to-intermediate facing coeficient varied only by 2%, the thermal neutron temperature by 15°C, and the unit lethargy flux level of intermediat
TL;DR: In this paper, the effective formation cross sections of 90mZr and 207mPb for natural Zr and Pb have also been estimated as 475±40mb and 900±80mb, respectively.
Abstract: The cross sections for the reactions 90Zr (n, n′) 90mZr, 91Zr (n, n′)2n) 90mZr, 207Pb (n, n′) 207mPb and 208Pb (n, n′)2n) 207mPb for neutron energy of 14.8 MeV have been measured by means of the activation method. The cross sections were determined with reference to the known 27A1 (n, α) 24Na reaction cross section. Isotropically enriched isotopes 90Zr, 91Zr, 207Pb and 208Pb were used as target materials. The irradiations and the activity measurements were repeated 100 times using a fast pneumatic transport system. The following cross sections were obtained: 410±40 mb for 90Zr (n, n′) 90mZr, 775±65mb for 91Zr (n, 2n) 90mZr 160±23 mb for 207Pb(n, n′) 207mPb, 1,220±100 mb for 208Pb(n, 2n) 207mPb. From the results of the present work, the effective formation cross sections of 90mZr and 207mPb for natural Zr and Pb have also been estimated as 475±40mb and 900±80mb, respectively
TL;DR: In this paper, the removal of solubilized solvent extractants from aqueous waste stream was discussed, and the results showed that the extractants can be used in nuclear power plants.
Abstract: (1981). Removal of Solubilized Solvent Extractants from Aqueous Waste Streams. Journal of Nuclear Science and Technology: Vol. 18, No. 7, pp. 561-562.
TL;DR: In this article, the authors measured the capture reaction rate of 232Th in the energy range from 1 to 408 keV, and with 55 and 146 keV neutrons by thick silicon filtered neutron beams, using a linac time-of-flight technique.
Abstract: Measurement of the neutron capture reaction rate of 232Th has been made (1) in the energy range from 1 to 408 keV, and (2) with 55 and 146 keV neutrons by thick silicon filtered neutron beams, using a linac time-of-flight technique. The capture reaction rate of 232Th has been measured relative to that for the 10B(n, αγ) reaction, whose cross section values have been taken from ENDF/B-V File. The results have been normalized to the 24 keV value previously measured by the present authors. Capture γ-rays were detected by a pair of C6D6 scintillators located at 12m on the neutron flight path. Thorium samples were chemically purified in order to reduce the γ-ray background from the daughter nuclides in the thorium decay chain. The point data measured with the 55 and 146 keV neutrons have been in good agreement with the energy dependent cross section for the 232Th(n, γ) reaction. It has been found that the results support recent measurements and the ENDF/B-V values of the capture cross section, while the discre...
TL;DR: In this paper, a computer code is developed for analysis of dynamic behavior of a single cryogenic distillation column in a typical column cascade which consists of four columns and two catalytic equilibrators.
Abstract: A computer code is developed for analysis of dynamic behavior of a single cryogenic distillation column. A dynamic simulation study is performed for the lead column in a typical column cascade which consists of four columns and two catalytic equilibrators. The atom fraction of protium in the bottom product and the reflux ratio are regarded as the controlled variable and the manipulated variable, respectively. Fluctuation or change of the feed composition is considered as the disturbance and the PI control scheme is introduced for a constant value control. The criteria are derived, which can be used in setting the choice of the two PI controller parameters. Investigated are the effects of the controller parameters on the response of the controlled variable to a step change in the feed composition, and the validity of the criteria is verified. The parameter setting method proposed in the present study has two advantages: it accounts for the strong nonlinearity of the column and it predicts the unstable regi...
TL;DR: In this article, the background levels were classified into seventeen groups from which a contour map was made by simple interpolations, and the general pattern obtained showed that high levels are prominent in Hida Mountains-Wakasa Bay area and Joetsu-northern Kanto provinces, and low levels were prominent in Hokkaido and northern Tohoku districts along with Kanto Plain.
Abstract: Natural background radiation in Japan was analyzed based on surveyed data which were rearranged from macroscopic points of view. The background levels were classified into seventeen groups from which a contour map was made by simple interpolations. The general pattern obtained showed that high levels are prominent in Hida Mountains-Wakasa Bay area and Joetsu-northern Kanto provinces, and low levels are prominent in Hokkaido and northern Tohoku districts along with Kanto Plain. The whole area was further divided into two classes according to whether the level is higher than the country-wide mean level. A boundary between the two classes seen in Chubu District was found almost exactly coincides with Itoigawa-Shizuoka Tectonic Line. Kanto Mountainland was found to have higher levels than its neighborhood, which supports the current idea that old geology of granitic rocks gives high level radiation. A good correspondence could be seen between the radiation contours and every tectonic line proposed as the eastern border of Fossa Magna which is important in the tectonics of Japan Islands. It was then anticipated that informations of the natural radiation will give valuable side evidences to infer the geologic history of Japan Islands.
TL;DR: A geological interpretation on the survey data of natural radiation in Japan was done in this paper, where the mean exposure rate on each geologic belt, rock type and soil type were calculated.
Abstract: A geological interpretation on the survey data of natural radiation in Japan was done. As terrestrial radiation is the major source of natural radiation, a systematic correlation was expected between the distribution of geology and the radiation. With a hope to find the correlation, Japan was divided into 376 pieces of 30 km meshes and a mean exposure rate of each mesh was obtained by averaging all values in it. The geographical distribution was then made clearer, where the differences between North-East Japan and Southwest Japan can be recognized. The radiation level of the latter is higher. This feature seems to be controlled mainly by granitic rocks and volcanism. The condition of weathering is implicit but important factor. Based on those meshes the mean exposure rate on each geologic belt, rock type and soil type were calculated. Every analysis has brought about a consistency in that the higher radiation level area corresponds to the old weathered region of grantic rocks. On the other hand, the lower...
TL;DR: In this paper, ampoule irradiations using 60Co γ-rays were undertaken on CC12F2, C2C14F2 and CC2C13F3 to gain information concerning the radiation damage of fluorocarbon, used as solvent in the recovery of 85Kr from the off-gas of fuel reprocessing plants operating on the solvent absorption process.
Abstract: Ampoule irradiations using 60Co γ-rays were undertaken on CC12F2, C2C14F2 and C2C13F3 to gain information concerning the radiation damage of fluorocarbon, used as solvent in the recovery of 85Kr from the off-gas of fuel reprocessing plants operating on the solvent absorption process. The dose rate was varied in the range of 2.6×104~8.0×105 rad/h and the irradiation temperature in the range of −100~70°C. Identified among the numerous degradation products were six compounds deriving from CC12F2, eleven from C2C14F2, and eleven also from C2C13F3. The decomposition yield of each of these fluorocarbons increased with irradiation temperature, and decreased with dose rate. It rose roughly in proportion to the absorbed dose. Comparison of decomposition behavior between the four fluorocarbons (including CC13F covered in the preceding report) was made in reference to the irradiation temperatures corresponding to those at which these fluorocarbons would respectively be applied in practice as solvent for 85Kr absorpt...
Journal Article•
TL;DR: In this article, the authors investigated the mechanism of the limit cycle oscillation observed in 1 MW SG without the thermal insulated downcomer of Power Reactor and Nuclear Fuel Development Corp., so that the effect of boiling in downcomers on the flow instability is studied experimentally.
Abstract: The steam generator for LMFBR is composed of multiple tubes with the different flow conditions from each other. Most of the computer codes developed to evaluate the stability analyze a representative tube on the basis of the assumption of the constant pressure drop between the feedwater header and the steam one as single channel. Although the method selecting the representative channel has never been investigated, it is one of the purposes of this paper to examine the approximate method to evaluate the stable boundary exactly. It is another purpose of this paper to research the mechanism of the limit cycle oscillation observed in 1 MW SG without the thermal insulated downcomer of Power Reactor and Nuclear Fuel Development Corp., so that the effect of boiling in downcomer on the flow instability is studied experimentally. The slug excursion instability, which does not belong to Boure's classification of two-phase flow instability, is observed in our experimental apparatus. Then, this mechanism is studied u...
TL;DR: In this article, the authors investigated the mechanism of the limit cycle oscillation observed in 1 MW SG without the thermal insulated downcomer of Power Reactor and Nuclear Fuel Development Corp., so that the effect of boiling in downcomers on the flow instability is studied experimentally.
Abstract: The steam generator for LMFBR is composed of multiple tubes with the different flow conditions from each other. Most of the computer codes developed to evaluate the stability analyze a representative tube on the basis of the assumption of the constant pressure drop between the feedwater header and the steam one as single channel. Although the method selecting the representative channel has never been investigated, it is one of the purposes of this paper to examine the approximate method to evaluate the stable boundary exactly. It is another purpose of this paper to research the mechanism of the limit cycle oscillation observed in 1 MW SG without the thermal insulated downcomer of Power Reactor and Nuclear Fuel Development Corp., so that the effect of boiling in downcomer on the flow instability is studied experimentally. The slug excursion instability, which does not belong to Boure's classification of two-phase flow instability, is observed in our experimental apparatus. Then, this mechanism is studied u...
TL;DR: In this paper, an optical apparatus for direct observation of transient fuel behavior in a water environment in actual in-reactor experiments was developed, which consists of a test capsule and connected upper containments.
Abstract: A great effort has been made to develop an optical apparatus for direct observation of transient fuel behavior in a water environment in actual in-reactor experiments. There are many difficulties to be overcome, such as high radiation fluence, limited space, shock pressure generation, fission products release. Through the extensive irradiation tests of various kinds of glasses, optical fibers and film and ex-reactor simulation tests, an in-reactor fuel behavior observation system was designed and fabricated for NSRR (Nuclear Safety Research Reactor) experiments. The system consists of a test capsule and connected upper containments. As an image guide is installed inside the containment a periscope with non-browning lenses whose lower part is covered by a stainless steel pipe with anti-shock window made of quartz in the test capsule, and the top of the periscope is connected with a high speed camera. A high intense lamp is immersed directly into water in the capsule as a light source. The motion pictures t...
TL;DR: It was shown from the numerical results that the proposed approach could provide some useful informations to make an actual management plan, and was introduced to make a decision through an interactive manner.
Abstract: A mathematical procedure is proposed to make a radioactive waste management plan comprehensively. Since such planning is relevant to some different goals in management, decision making has to be formulated as a multiobjective optimization problem. A mathematical programming method was introduced to make a decision through an interactive manner which enables us to assess the preference of decision maker step by step among the conflicting objectives. The reference system taken as an example is the radioactive waste management system at the Research Reactor Institute of Kyoto University (KUR). Its linear model was built based on the experience in the actual management at KUR. The best-compromise model was then formulated as a multiobjective linear programming by the aid of the computational analysis through a conventional optimization. It was shown from the numerical results that the proposed approach could provide some useful informations to make an actual management plan.
TL;DR: In this article, the effects of coolant flow on fuel behavior under reactivity initiated accident (RIA) conditions performed in the Nuclear Safety Research Reactor (NSRR) were investigated.
Abstract: This paper describes the in-pile experimental results to study the influences of coolant flow on fuel behaviors under reactivity initiated accident (RIA) conditions performed in the Nuclear Safety Research Reactor (NSRR). A single PWR type test fuel rod was irradiated by a large neutron pulse in the NSRR to simulate a prompt power excursion of RIA's. The effects of coolant flow were studied at a coolant flow velocity of 0.3~1.8m/s and a coolant temperature of 20~90°C under the atmospheric pressure. It was found that the cooling conditions had considerable influences on fuel thermal behaviors under prompt heat-up. The increase of coolant flow velocity and subcooling enhanced heat transfer coefficient at cladding surface during film boiling, which resulted in large decrease of maximum cladding temperature and film boiling duration, and consequently in the increase of fuel failure threshold energy. The data tendencies are summarized and the influences of coolant flow are discussed with some computer analyses.