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Journal Articles

Estimation of temporal variation of discharged inventory of radioactive strontium $$^{90}$$Sr ($$^{89}$$Sr) from port of Fukushima Daiichi Nuclear Power Plant; Analysis of the temporal variation from the accident to March 2022 and evaluation of its impact on Fukushima coast and offshore areas

Machida, Masahiko; Iwata, Ayako; Yamada, Susumu; Otosaka, Shigeyoshi*; Kobayashi, Takuya; Funasaka, Hideyuki*; Morita, Takami*

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(4), p.119 - 139, 2023/11

We estimate monthly discharged inventory of $$^{90}$$Sr from port of Fukushima Daiichi Nuclear Power Plant (1F) from Jun. 2013 to Mar. 2022 by using the Voronoi tessellation method inside the port, following the monitoring of $$^{90}$$Sr sea water radioactivity concentration inside the port. The results suggest that the closure of sea side impermeable wall is the most effective for the reduction of discharged one. In addition, the results roughly reveal the monthly discharged inventory required to observe visible enhancement of the sea radioactivity concentration from the background level in each area. Such outcome is significant for considering environmental impacts on the planned future releasing of the treated water accumulated in 1F site.

Journal Articles

Survey of air dose rate distribution inside and outside of wooden houses in Fukushima Prefecture; Actual condition of dose reduction factor

Kim, M.; Malins, A.*; Machida, Masahiko; Yoshimura, Kazuya; Saito, Kimiaki; Yoshida, Hiroko*

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(4), p.156 - 169, 2023/11

Dose reduction factor of a Japanese house is important information in the external exposure estimation of returning residents. In 2019, a total of 19 wooden houses were surveyed in Iitate Village and Namie Town using a gamma plotter that can continuously measure the air dose rate. In addition, the characteristics of the reduction factor were investigated from the measured air dose rate. In the vicinity of houses, uncontaminated areas exist underneath houses and, the ratio of paved surfaces such as asphalt roads is relatively high; furthermore, the pavement has a tendency for the radiation source to decay quickly. Therefore, the air dose rate near the house showed a relatively low value in common at all sites. Air dose rates above unpaved surfaces showed higher values and larger variations than those above paved surfaces within a radius of 50 m form the center of a house. The reduction factor was widely distributed even for one house, if the ratio of every air dose rate observed inside and outside the house is considered. It is suggested that a realistic reduction factor may not be obtained when the reduction factor is obtained based on the measured values at a small number of points that do not have the representativeness of the radiation field to be measured.

Journal Articles

Effect of decay heat on pyrochemical reprocessing of minor actinide transmutation nitride fuels

Hayashi, Hirokazu; Tsubata, Yasuhiro; Sato, Takumi

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(3), p.97 - 107, 2023/08

The Japan Atomic Energy Agency has chosen nitride fuel as the first candidate for the transmutation of long-lived minor actinides (MA) using accelerator-driven systems (ADS). The pyrochemical method has been considered for reprocessing spent MA nitride fuels, because their decay heat should be very large for aqueous reprocessing. This study was conducted to investigate the effect of decay heat on the pyrochemical reprocessing of MA nitride fuels. On the basis of the estimated decay heats and the temperature limits of the materials that are to be handled in pyrochemical reprocessing, quantities adequate for handling in argon gas atmosphere were evaluated. From these considerations, we proposed that an electrorefiner with a diameter of 26 cm comprising 12 cadmium (Cd) cathodes with a diameter of 4 cm is suitable. On the basis of the size of the electrorefiner, the number necessary to reprocess spent MA fuels from 1 ADS in 200 days was evaluated to be 25. Furthermore, the amount of Cd-actinides (An) alloy to produce An nitrides by the nitridation-distillation combined reaction process was proposed to be about one-quarter that of Cd-An cathode material. The evaluated sizes and required numbers of equipment support the feasibility of pyrochemical reprocessing for MA nitride fuels.

Journal Articles

Promotive effects of siltfence on deposition of suspended particles; Evaluations of diffusion suppression of radionuclides in rectangular open channel via numerical simulations

Yamada, Susumu; Machida, Masahiko; Arikawa, Taro*

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(2), p.73 - 86, 2023/06

no abstracts in English

Journal Articles

Machine learning sintering density prediction model for MOX fuel pellet

Kato, Masato; Nakamichi, Shinya; Hirooka, Shun; Watanabe, Masashi; Murakami, Tatsutoshi; Ishii, Katsunori

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(2), p.51 - 58, 2023/04

Uranium and Plutonium mixed oxide (MOX) pellets used as fast reactor fuels have been produced from several raw materials by mechanical blending method through processes of ball milling, additive blending, granulation, pressing, sintering and so on. It is essential to control the pellet density which is one of the important fuel specifications, but it is difficult to understand relationships among many parameters in the production. Database for MOX production was prepared from production results in Japan, and input data of eighteen types were chosen from production process and made a data set. Machine learning model to predict sintered density of MOX pellet was derived by gradient boosting regressor, and represented the measured sintered density with coefficient of determination of R$$^{2}$$=0.996

Journal Articles

Evaluation of external dose exposure of workers during house demolition and dose reduction work in a difficult-to-return zone

Sanada, Yukihisa; Tokiyoshi, Masanori*; Nishiyama, Kyohei*; Sato, Rina; Yoshimura, Kazuya; Funaki, Hironori; Abe, Tomohisa; Ishida, Mutsushi*; Nagamine, Haruo*; Fujisaka, Motoyuki*

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(2), p.87 - 96, 2023/04

Since the accident at the Fukushima Daiichi Nuclear Power Plant, many decontamination works have been carried out, but it is difficult to say that much data on workers' exposure has necessarily been analyzed in detail. In this paper, based on the GPS location information carried by the workers together with their personal dosimeters, the air dose rate in the work area and the characteristics of each type of work were analyzed. The results showed that more than 50% of the measured actual doses were more than twice the median planned dose calculated from the air dose rate and actual working hours. Furthermore, as a result of the analysis by work type, it was found that the exposure doses of demolition workers tended to be high, and that this was due to the fact that most of the work was carried out before the work was carried out to reduce the dose at the work site. In addition, when the conversion from air dose to effective dose was taken into account, there were many cases of underestimation where the planned values were lower than the measured values, and it is considered important for management to set appropriate working factor.

Journal Articles

Tritium inventory and its temporal variation in Fukushima Front Sea Area; Comparison between coastal and offshore tritium inventories and 1f treated water and operational target values for discharge per year

Machida, Masahiko; Iwata, Ayako; Yamada, Susumu; Otosaka, Shigeyoshi*; Kobayashi, Takuya; Funasaka, Hideyuki*; Morita, Takami*

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(1), p.12 - 24, 2023/01

We estimate inventory of tritium in two sea areas corresponding to coastal and offshore ones around Fukushima Daiichi Nuclear Power Plant (1F) based on the measurement results of sea-water tritium concentration monitored constantly from 2013 to Jan. 2021 by using Voronoi tessellation scheme. The obtained results show that the offshore area inventory and its temporal variation amount correspond to approximately 1/5 and 1/40 of that of the treated-water accumulated inside 1F, respectively. These results just suggest that the presence of tritium already included in sea-water as the background is non negligible in evaluating the environmental impact by releasing the accumulated treated-water into the sea region. We also estimate the offshore area inventory before 1F accident and find that it had exceeded over 1F stored inventory over about 30 years from 1960s to 1980s with approximately 4 times larger in the peak decade, 1960s. This fact means that we had already experienced more contaminated situation over 30 years in the past compared to the conservative case appeared by just releasing whole the present 1F inventory. Here, it should be also emphasized that the past contaminated situation was shared by the entire world. We further extend the estimation region into a wider region including an offshore area from Miyagi to Chiba prefectures and find that the area average inventory is now comparable to a half of the present 1F one. Finally, we estimate internal dose per year via ingesting fishes caught inside the area when 1F inventory is just added inside the area and kept for a year. The result indicates that it approximately corresponds to 1.0$$^-6$$ of the dose from natural radiation sources. From these estimation results, it is found that all the tritium inventory stored inside 1F never contribute to significant dose increment even when it is instantly released into the area.

Journal Articles

Overview of event progression of evaporation to dryness caused by boiling of high-level liquid waste in Reprocessing Facilities

Yamaguchi, Akinori*; Yokotsuka, Muneyuki*; Furuta, Masayo*; Kubota, Kazuo*; Fujine, Sachio*; Mori, Kenji*; Yoshida, Naoki; Amano, Yuki; Abe, Hitoshi

Nihon Genshiryoku Gakkai Wabun Rombunshi, 21(4), p.173 - 182, 2022/09

Risk information obtained from probabilistic risk assessment (PRA) can be used to evaluate the effectiveness of measures against severe accidents in nuclear facilities. The PRA methods used for reprocessing facilities are considered immature compared to those for nuclear power plants, and to make the methods mature, reducing the uncertainty of accident scenarios becomes crucial. In this paper, we summarized the results of literature survey on the event progression of evaporation to dryness caused by boiling of high-level liquid waste (HLLW) which is a severe accident in reprocessing facilities and migration behavior of associated radioactive materials. Since one of the important characteristics of Ru is its tendency to form volatile compounds over the course of the event progression, the migration behavior of Ru is categorized into four stages based on temperature. Although no Ru has been released in the waste in the high temperature region, other volatile elements such as Cs could be released. Sufficient experimental data, however, have not been obtained yet. It is, therefore, necessary to further clarify the migration behavior of radioactive materials that predominantly depends on temperature in this region.

Journal Articles

Re-evaluation of electricity generation cost of HTGR

Fukaya, Yuji; Ohashi, Hirofumi; Sato, Hiroyuki; Goto, Minoru; Kunitomi, Kazuhiko

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 21(2), p.116 - 126, 2022/06

An improvement electricity generation cost evaluation method for High Temperature Gas-cooled Reactors (HTGRs) has been performed. Japan Atomic Energy Agency (JAEA) had completed the commercial HTGR concept named Gas Turbine High Temperature Reactor (GTHTR300) and the electricity generation cost evaluation method approximately a decade ago. The cost evaluation was developed based on the method of Federation of Electric Power Companies (FEPC). The FEPC method was drastically revised after the Fukushima Daiichi nuclear disaster. Moreover, the escalation of material and labor cost for the decade should be consider to evaluate the latest cost. Therefore, we revised the cost evaluation method for GTHTR300 and the cost was compared with that of Light Water Reactor (LWR). As a result, it was found that the electricity generation cost of HTGR of 7.9 yen/kWh is cheaper than that of LWR of 11.7 yen/kWh by approximately 30% at the capacity factor of 70%.

Journal Articles

Safety improvements in demolition and removal activities with air-fed suit for Plutonium Fuel Facility Decommissioning

Kikuchi, Haruka; Hirano, Hiroshi*; Kitamura, Akihiro

Nihon Genshiryoku Gakkai Wabun Rombunshi, 21(1), p.50 - 63, 2022/03

The air fed suit is a kind of personal protective equipment that provides purified air through a hose and that protects a worker from radiation hazards. In the Nuclear Fuel Cycle Engineering Laboratories of the JAEA, the suit is used for size reduction and dismantlement of radioactively contaminated, in particular with plutonium, gloveboxes and equipment. Although the suit has been widely adopted in the similar activities, there still exist potential hazards due to the limiting features of the suit itself and its supplemental system. In fact, we had faced with several unexpected problems regarding such restricted aspects during the dismantling activities. To address these failure potentials, we have implemented various countermeasures and improvements to enhance the workers safety. We describe the disadvantages of the air fed suit system and positive feedbacks we have implemented.

Journal Articles

Estimation of temporal variation of discharged tritium from port of Fukushima Dai-ichi Nuclear Power Plant; Analysis of the temporal variation and comparison with released tritium from Japan and major nuclear facilities worldwide

Machida, Masahiko; Iwata, Ayako; Yamada, Susumu; Otosaka, Shigeyoshi*; Kobayashi, Takuya; Funasaka, Hideyuki*; Morita, Takami*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 21(1), p.33 - 49, 2022/03

We estimate monthly discharge inventory of tritium from the port of Fukushima Daiichi Nuclear Power Plant (1F) from Jun 2013 to Mar 2020 by using the Voronoi tessellation scheme, following that the tritium monitoring inside the port has started since Jun 2013. As for the missing period from the initial month, Apr 2011 to May 2013, we calculate it by utilizing the concentration ratio of tritium to that of $$^{137}$$Cs in stagnant contaminant water during the initial direct discharged period to Jun 2011 and the discharge inventory correlation between tritium and $$^{137}$$Cs for the next-unknown continuously-discharged period up to May 2013. From the all- estimated results over 9 years, we find that the monthly discharge inventory sharply dropped just after closing the sea-side impermeable sea-wall in Oct. 2015 and subsequently coincided well with the sum of those of drainage and subdrain. By comparing the estimated results with those in the normal operation period before the accident, we point out that the discharge inventory from 1F port is not so large compared to those during the normal operation. Even the estimated one in year 2011 is found to be comparable to the maximum of operating pressurized water reactors discharging relatively large inventory in the order. In the nation level, the whole Japan domestic discharge inventory significantly decreased after the accident due to operation shutdown of most plants. Furthermore, 1F and even Japanese total discharge inventory are found to be entirely minor when comparing those of nuclear reprocessing plants and heavy-water reactors in world-wide level. From the above, we suggest that various scenarios can be openly discussed on the management in tritium stored inside 1F with help of the present estimated data and its comparison with the past discharge inventory.

Journal Articles

Applicability of equivalent linear analysis to reinforced concrete shear walls; 3D FEM simulation of experiment results of seismic wall ultimate behavior

Ichihara, Yoshitaka*; Nakamura, Naohiro*; Moritani, Hiroshi*; Horiguchi, Tomohiro*; Choi, B.

Nihon Genshiryoku Gakkai Wabun Rombunshi, 21(1), p.1 - 14, 2022/03

In this study, we aim to approximately evaluate the effect of nonlinearity of reinforced concrete structures through seismic response analysis using the equivalent linear analysis method. A simulation analysis was performed for the ultimate response test of the shear wall of the reactor building used in an international competition by OECD/NEA in 1996. The equivalent stiffness and damping of the shear wall were obtained from the trilinear skeleton curves proposed by the Japan Electric Association and the hysteresis curves proposed by Cheng et al. The dominant frequency, maximum acceleration response, maximum displacement response, inertia force-displacement relationship, and acceleration response spectra of the top slab could be simulated well up to a shear strain of approximately $$gamma$$=2.0$$times$$10$$^{-3}$$. The equivalent linear analysis used herein underestimates the maximum displacement response at the time of ultimate fracture of approximately $$gamma$$=4.0$$times$$10$$^{-3}$$. Moreover, the maximum shear strain of the shear wall could not capture the locally occurring shear strain compared with that of the nonlinear analysis. Therefore, when employing this method to evaluate the maximum shear strain and test results, including those during the sudden increase in displacement immediately before the fracture, sufficient attention must be paid to its applicability.

Journal Articles

${it In situ}$ spectrometry of terrestrial gamma rays using portable germanium detectors in area of 80 km radius around the Fukushima Daiichi Nuclear Power Plant

Mikami, Satoshi; Tanaka, Hiroyuki*; Okuda, Naotoshi*; Sakamoto, Ryuichi*; Ochi, Kotaro; Uno, Kiichiro*; Matsuda, Norihiro; Saito, Kimiaki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 20(4), p.159 - 178, 2021/12

In order to know the background radiation level where the area affected by the Fukushima Daiichi Nuclear Power Plant accident in 2011, terrestrial gamma rays had been measured by using portable germanium detectors repeatedly from 2013 through 2019, at 370 locations within 80 km radius area centered on the Fukushima Daiichi Nuclear Power Plant. Radioactive concentrations of Uranium 238, Thorium 232, Potassium 40 and kerma rates in air due to terrestrial gamma rays were obtained at those locations based on the method of ICRU report 53. Averaged concentrations of $$^{238}$$U, $$^{232}$$Th and $$^{40}$$K were 18.8, 22.7, 428 Bq/kg, respectively, and kerma rate in air over the area was found to be 0.0402 $$mu$$Gy/h. The obtained kerma rates in air were compared to those reported in literatures. It was confirmed that the data were correlated with each other, and were agreed within the range of their uncertainty. This is because the kerma rate in air due to terrestrial gamma rays is depend on geology. The similar trend to previous findings was observed that the kerma rate in air at locations geologically classified as Mesozoic era, Granite and Rhyolite were statistically significantly higher than the others.

Journal Articles

Study on solubility of cesium iodide and cesium molybdate in water at around room temperature

Imoto, Jumpei; Nakajima, Kunihisa; Osaka, Masahiko

Nihon Genshiryoku Gakkai Wabun Rombunshi, 20(4), p.179 - 187, 2021/12

Some of the Cs inside the Fukushima Daiichi Nuclear Power Station would be deposited in chemical forms such as CsI and Cs$$_{2}$$MoO$$_{4}$$. Since Cs compounds are generally water-soluble, it is predicted that the migration of Cs through the aqueous phase occurs in the long term. Knowledge of the solubility in water is required as basic data for such migration behavior evaluation. Therefore, this study was conducted to investigate the dissolution properties of CsI and Cs$$_{2}$$MoO$$_{4}$$ in water at 20$$^{circ}$$C and 25$$^{circ}$$C. The solubilities of CsI at 25$$^{circ}$$C calculated using thermodynamic data and the Pitzer ion interaction model were in good agreement with the literature value. It was found that the literature value of CsI at around room temperature is highly reliable. The experimental value of CsI at 20$$^{circ}$$C obtained by the OECD test guideline 105 flask method (test guideline) was also in good agreement with the literature value. The measured solubility of Cs$$_{2}$$MoO$$_{4}$$ was 256.8 $$pm$$ 6.2 (g/100 g H$$_{2}$$O) at 20$$^{circ}$$C using the test guideline. This measured solubility of Cs$$_{2}$$MoO$$_{4}$$ was found to be comparable to those of other alkaline molybdates and considered to be more reliable than the literature value.

Journal Articles

Radiation monitoring and evaluation of exposure doses to lift the evacuation orders for the zones designated for reconstruction and recovery

Sanada, Yukihisa; Kurikami, Hiroshi; Funaki, Hironori; Yoshimura, Kazuya; Abe, Tomohisa; Ishida, Mutsushi*; Tanimori, Soichiro*; Sato, Rina

Nihon Genshiryoku Gakkai Wabun Rombunshi, 20(2), p.62 - 73, 2021/06

Japanese government starts to consider radiation protection in the "specific reconstruction reproduction base area" of which evacuation order will be lifted by 2023. It is essential to grab the present situations of radiation contamination and evaluate exposure dose in the area to realize the plan. Many surveys have evaluated the distributions of air dose rate and exposure dose has been estimated based on the results since the Fukushima Daiichi Nuclear Power Plant accident. Nevertheless, more detailed information on exposure is needed for the areas because its radiation level is relatively high. That is also to help make prudent evaluation plan. This study aimed to evaluate the detailed contamination situation there and estimate exposure dose with considering areal circumstances. Investigations were carried out for (1) airborne survey of air dose rate using an unmanned helicopter (2) evaluation of airborne radiocesium and (3) estimation of external/internal effective doses for typical activity patterns assumed.

Journal Articles

Development of evaluation method for diffusion and filtration behavior of colloid in compacted bentonites using dendrimers

Endo, Takashi*; Tachi, Yukio; Ishidera, Takamitsu; Terashima, Motoki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 20(1), p.9 - 22, 2021/03

Evaluation method of colloid diffusion and filtration in compacted bentonites was developed using dendrimers. Diffusion and filtration behavior of PAMAM dendrimers with the size of 5.7$$sim$$7.2nm was investigated by the through-diffusion experiment in bentonite compacted to 0.8 Mg/m$$^{3}$$ and saturated with 0.005$$sim$$0.5mol/L NaCl. Effective diffusivities (De) and filtration ratios (Rf) of dendrimers were determined from the breakthrough curves and the depth profiles in compacted bentonite, respectively. The De values of negatively charged dendrimer increased when porewater salinity increased and dendrimer size decreased as influenced by anion exclusion effect in negatively charged clay surfaces. The Rf values increased when porewater salinity decreased and dendrimer size increased, demonstrating significant fractions of dendrimer were filtered by narrow pores in complex pore networks. These trends consistent with the previous studies emphasize the validity of the evaluation method using dendrimer.

Journal Articles

Determination of parameters for an equation to obtain natural background radiation using KURAMA-II loaded with C12137-01 type CsI(Tl) detector

Ando, Masaki; Matsuda, Norihiro; Saito, Kimiaki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 20(1), p.34 - 39, 2021/03

We measured count rates and air dose rates at 11 measurement points where the influence of the Fukushima Dai-ichi Nuclear Power Plant accident could be ignored to obtain parameters for a background equation applying to KURAMA-II loaded with the high sensitivity CsI(Tl) detector, C12137-01. It was found that the sensitivity of KURAMA-II (C12137-01) was about 10 times or more for background measurement, compared with KURAMA-II loaded with the standard type CsI(Tl) detector, C12137. A background equation for the energy range of 1400-2000 keV was determined as, y ($$mu$$Sv/h)=0.062 x (cps). We evaluated background air dose rates using KURAMA-II (C12137-01) for 71 municipalities and compared them with the previous study using KURAMA-II (C12137). Evaluated background air dose rates in this study were almost equal to those in the previous study. We confirmed that the background equation evaluated in this study was applicable for the KURAMA-II (C12137-01).

Journal Articles

Experiments of self-wastage phenomena elucidation in steam generator tube of sodium-cooled fast reactor

Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu

Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.234 - 244, 2020/12

Sodium-water reaction caused by failure of the steam generator tube of sodium-cooled fast reactor induce the wastage phenomenon, which has erosive and corrosive feature. In this report, the authors have performed the self-wastage experiments under high sodium temperature condition to evaluate the effect of wastage form/geometry by using two types of initial defect such as the micro fine pinhole and fatigue crack, and water leak rate on self-wastage rate. Based on the consideration of crack type influence, it was confirmed that self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide intervenes and inhibits the progress of self-wastage. The dependence of initial sodium temperature on self-wastage rate was clearly observed, and new self-wastage correlation was derived considering the initial sodium temperature.

Journal Articles

Status of investigation to ensure applicability of ECCS acceptance criteria to high-burnup fuel

Ozawa, Masaaki*; Amaya, Masaki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.185 - 200, 2020/12

no abstracts in English

Journal Articles

Pressure resistance thickness of disposal containers for spent fuel direct disposal

Sugita, Yutaka; Taniguchi, Naoki; Makino, Hitoshi; Kanamaru, Shinichiro*; Okumura, Taisei*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.121 - 135, 2020/09

A series of structural analysis of disposal containers for direct disposal of spent fuel was carried out to provide preliminary estimates of the required pressure resistance thickness of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body and then the lid of the disposal container. This work also provides additional analytical technical knowledge, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.

298 (Records 1-20 displayed on this page)