Estimates of Maternal Mortality Ratio and the associated medical causes in Orissa and Rajasthan States - A cross sectional study

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Abha Rani Aggarwal

2015-03-01

Full Text Available Background: Maternal Mortality Ratio (MMR is an important indicator of reproductive health and its reduction remains a challenge in India. Aims &Objective: This study was conducted with the aim of estimating MMR in two states Orissa and Rajasthan having high MMR as well as to identify the associated medical causes of maternal mortality. Material Methods: This survey was conducted from October 2010-June 2012 on a sample of 13 Primary Health Centres (PHCs in Orissa and 15 PHCs in Rajasthan. These numbers have been derived after estimating the total number of live births using MMR and birth rate from Sample Registration System. 1997-2003.An adapted snowball technique was adopted wherein maternal deaths were captured by snowball technique and the numbers of live births were taken from the available records from the various health facilities in the study.  Results: The overall birth rate in Orissa was found to be 19 per 1000 population while in Rajasthan it was 24 per 1000 population. The study revealed that 17% additional maternal deaths could be captured by snowball technique as against the official record. The overall weighted estimate of MMR was 252 per one lakh live births (95% CI: 246-259 per 1,00,000 live births in Orissa and 209 per one lakh live births (95% CI: 207-211 per one lakh live births in Rajasthan. The main causes of maternal deaths were post-partum haemorrhage, anaemia and septicaemia. More than 25% maternal deaths could be attributed to indirect causes including suicide, accident and infectious diseases. Conclusion: There appears to be a positive trend towards reduction of maternal mortality in Orissa and Rajasthan. Greater care is essential to reduce medical as well as incidental causes of death during pregnancy.

Estimates of Maternal Mortality Ratio and the associated medical causes in Orissa and Rajasthan States - A cross sectional study

Directory of Open Access Journals (Sweden)

Abha Rani Aggarwal

2015-03-01

Full Text Available Background: Maternal Mortality Ratio (MMR is an important indicator of reproductive health and its reduction remains a challenge in India. Aims &Objective: This study was conducted with the aim of estimating MMR in two states Orissa and Rajasthan having high MMR as well as to identify the associated medical causes of maternal mortality. Material Methods: This survey was conducted from October 2010-June 2012 on a sample of 13 Primary Health Centres (PHCs in Orissa and 15 PHCs in Rajasthan. These numbers have been derived after estimating the total number of live births using MMR and birth rate from Sample Registration System. 1997-2003.An adapted snowball technique was adopted wherein maternal deaths were captured by snowball technique and the numbers of live births were taken from the available records from the various health facilities in the study.  Results: The overall birth rate in Orissa was found to be 19 per 1000 population while in Rajasthan it was 24 per 1000 population. The study revealed that 17% additional maternal deaths could be captured by snowball technique as against the official record. The overall weighted estimate of MMR was 252 per one lakh live births (95% CI: 246-259 per 1,00,000 live births in Orissa and 209 per one lakh live births (95% CI: 207-211 per one lakh live births in Rajasthan. The main causes of maternal deaths were post-partum haemorrhage, anaemia and septicaemia. More than 25% maternal deaths could be attributed to indirect causes including suicide, accident and infectious diseases. Conclusion: There appears to be a positive trend towards reduction of maternal mortality in Orissa and Rajasthan. Greater care is essential to reduce medical as well as incidental causes of death during pregnancy.

Awareness of workplace hazards and preventive measures among sandstone mineworkers in Rajasthan, India: A cross-sectional study

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Absar Ahmad

2017-03-01

Full Text Available Objective: The aim of this study was to assess awareness of workplace hazards and personal protective equipments (PPEs among mineworkers employed in a sandstone quarry in Rajasthan, India. Materials and Methods: A cross-sectional study of 218 miners was conducted in Karauli, Rajasthan, India. We used a standardized semi-structured questionnaire that was administered to each subject by face-to-face interviews. Descriptive statistics and Pearson chi-square test were used to show frequency distributions and associations between variables. Results: Almost all respondents were aware of at least one hazard in mining occupation (93.6%, but no of them was trained by a recent (within 1 year health and safety training course. However, mineworkers recognized only the risk of injury (74.3% and exposure to crystalline silica dust (40.4%. A high percentage of mineworkers were aware of PPEs (87.6%, but an only 16.5% of them used PPEs during their employment. The only PPEs mentioned by mineworkers was the dust protective mask. Occurrence of at least one occupational injury during work-life was associated with use of dust masks, while work-related diseases were associated with a low level of education, underweight (BMI < 18.5 kg/m2, and current smoking. Awareness of workplace hazards was associated with age less than 60, young age of starting work in mines (< 30 years, hours work per day (< 8 hr, and no availability of drinking water facility. Failure to use PPEs at work was statistically significant associated with belonging to scheduled castes or scheduled tribes, lower distance from home to workplace (1‒3 Km, hours work per day (< 8 hr, and no availability of safe drinking water. Discussion and Conclusion: In Rajasthan, India, there is a certain level of awareness about workplace hazards but usage of PPEs by sandstone mineworkers is very low. Policy makers should implement health and safety training programmes to promote use of PPEs among mine workers.

Experience with valves for PHWR reactors

International Nuclear Information System (INIS)

Narayan, K.; Mhetre, S.G.

1977-01-01

Material specifications and inspection and testing requirements of the valves meant for use in nuclear reactors are mentioned. In the heavy water systems (both primary and moderator) of a PHWR type reactor, the valves used are gate valves, globe valves, diaphragm valves, butterfly valves, check valves and relief valves. Their locations and functions they perform in the Rajasthan Atomic Power Station Unit-1 are described. Experience with them is given. The major problems encountered with them have been : (1) leakage from the stem seals and body bonnet joint, (2) leakage due to failure of diaphragm and/or washout of the packing and (3) malfunctioning. Measures taken to solve these are discussed. Finally a mention has been made of improved versions of valves, namely, metal diaphragm valve and inverted relief valve. (M.G.B.)

Methods of selection and training of personnel for the Rajasthan atomic power station

International Nuclear Information System (INIS)

Sarma, M.S.R.; Wagadarikar, V.K.

1975-01-01

Personnel selected to work in a nuclear electric generating station rarely have the necessary knowledge and experience in all the related fields. A station can be operated and maintained and at the same time radiation doses absorbed by station personnel can be kept to a minimum only if the operating personnel are familiar with, and can be used for, all phases of station operation and the maintainers have more than one skill or trade. More technical knowledge and more diversified skills, in addition to those required in other industries, are needed because of the nature of the nuclear reactor and the associated radiation environment and high automation. A training programme has been developed at the Nuclear Training Centre (NTC) near the Rajasthan Atomic Power Station (RAPS), Kota, India, to cater to the needs of the operation and maintenance personnel for nuclear power stations including the Madras Atomic Power Station. This programme has been in operation for the last five years. The paper describes the method of recruitment/selection of various categories of personnel and the method of training them to meet the job requirements. (author)

Geobotanical studies on uranium deposits of Udaipur, Rajasthan, India

International Nuclear Information System (INIS)

Aery, N.C.; Jain, G.S.

1995-01-01

Geobotanical studies were carried out on known uranium deposits of Udaisagar region in the district of Udaipur, Rajasthan. Releve method of Braun Blanquet was employed for community analysis. Though no species with an exclusive occurrence on uranium deposits was found, certain plant species registered higher constancy and fidelity on uranium rich soils in comparison to background soils. Obviously, these characteristic plant species have evolved tolerance to high uranium contents of the soils and might be neo-endemics. (author). 23 refs., 1 fig., 4 tabs

Assessment of tritium dose around Rajasthan Atomic Power Station (1989-1997)

International Nuclear Information System (INIS)

Verma, P.C.; Vijaykumar, B.; George, Thomas; Sankhla, Rajesh; Roy, Alpana; Vyas, P.V.; Gurg, R.P.

1999-01-01

Tritium monitoring in the atmospheric and aquatic environment forms an integral part of the environmental radiological measurements conducted around PHWR type Nuclear Power Plant sites as tritium is one of the predominantly produced radionuclides in such systems. This paper presents the tritium concentration levels in the atmospheric and aquatic environs of Rajasthan Atomic Power Station (RAPS) during 1989-1997. The committed effective doses due to tritium are computed on annual basis for different radial zones around RAPS. It is observed that the maximum dose due to tritium at 1.6 km post fence has been less than 1% of the dose limit (1 mSv) set by regulatory body for the members of the public and beyond 5 km distance it is only 0.4% of the limit. (author)

Neglected health literacy undermining fluorosis control efforts: A pilot study among schoolchildren in an endemic village of rural Rajasthan, India

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Neeti Rustagi

2017-01-01

Full Text Available Background: Ingestion of excess fluoride can cause fluorosis which adversely affects teeth and bones. Fluorosis is a major public health problem in the state of Rajasthan with all its 32 districts having variable fluoride contamination, and many initiatives are currently implemented in this region to mitigate the fluorosis burden. Objective: The objective of this study was to assess fluorosis, its risk factors, and the awareness among school students and teachers in endemic villages of Jodhpur district, Rajasthan. Materials and Methods: A representative sample of 300 students of age 12–15 years were enrolled for cross-sectional study in selected villages. Data were collected regarding awareness, behavioral and preventive practices about fluorosis and clinical assessment and fluoride estimation in water and urine samples was done. Results: Dental fluorosis through Dean's index was observed in 24.5% of students. The awareness and practices for fluorosis prevention and its risk factors were poor among both students and teachers. Intake of piped water supply was reported by majority of students (95.8%. High fluoride concentration was found in 35 (81.3% out of 43 urine samples. Conclusion: Improvement in drinking water supply in the endemic village of Rajasthan has decreased the burden of fluorosis, but low level of awareness and prevailing dietary and behavioral practices still pose them at risk of high fluoride intake. This signifies the need to address nonconventional sources of fluoride intake (diet and toothpaste and early screening of disease by involving teachers and family physicians in fluoride mitigation efforts.

Dynamic behaviour of CANDU reactor

International Nuclear Information System (INIS)

Subramanian, M.G.; Srikantiah, G.; Pai, M.A.

1976-01-01

Understanding of the dynamic behaviour of a reactor system in a power station is essential for evolving control stragies as well as design modifications. The dynamic behaviour of Rajasthan Atomic Power Station is studied. Mathematical models for the reactor, the steam generator and the steam drum with the natural circulation loop are developed from physical principles like conservation of mass, momentum and energy. Each of these models is then simulated on a digital computer to obtain the characteristics during transients. The models are then combined to yield a dynamic mathematical model of the system comprising the reactor, the steam generator and the steam drum and this results in a nonlinear model. Using this model, responses of the system for various disturbances like step change in the area of the steam valve, step change in the temperature of feed water are obtained and are discussed. These models could be used to devise new control laws using optimal control theory or to evaluate the performance of existing control schemes. (author)

How Do Age and Tooth Loss Affect Oral Health Impacts and Quality of Life?A Study Comparing Two State Samples of Gujarat and Rajasthan

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A. Mathur

2012-01-01

Full Text Available Objective: Age and tooth loss are expected to have a complex relationship with oral health-related quality of life. So the purpose of this study was to explain the impact of age and tooth loss on oral health-related quality of life using the short form 14-item oral health impact profile (OHIP-14 among two population samples of Gujarat and Rajasthan.Materials and Methods: A cross-sectional questionnaire-based survey was conducted among 1441 subjects collected from two major cities of Gujarat and Rajasthan. Both questionnaire approaches using OHIP-14 scale and clinical examination were conducted in accordance with WHO criteria using type III procedure on the same day. Chi square test, ANOVA and stepwise multiple regression analysis were applied using SPSS software version 15.0.Results: With the increase of age, OHIP mean score in both states increased, but that among Rajasthan state was higher, depicting poor oral health. Whereas, in the remaining 23-27 number of teeth both states showed higher OHIP mean, however again the score was much higher among Rajasthan subjects showing worse oral hygiene. Hence, overall all mean OHIP score for Gujarat was lower indicating good oral health; whereas, that among Rajasthan was higher indicating poor oral health-related quality of life.Conclusion: Both age and tooth loss are associated with each other, but they have an independent effect on the oral health-related quality of life. Thus, all studied populations with complete natural dentition showed good oral health-related quality of life.

Fission track dating and uranium estimation in pegmatitic minerals of Rajasthan state (India)

Energy Technology Data Exchange (ETDEWEB)

Singh, S; Virk, H S [Punjabi Univ., Patiala (India). Dept. of Physics

1978-12-01

Fission track geochronology of muscovite samples collected from some pegmatitic mines of Bhilwara and Ajmer districts of Rajasthan state (India) has been discussed. The ages obtained suggest the occurrence of Delhi Orogenic Cycle as the last major metamorphic activity in the region. The atomic fraction of uranium in muscovite samples is less than 1 p.p.b.

Reactivity requirements and safety systems for heavy water reactors

International Nuclear Information System (INIS)

Kati, S.L.; Rustagi, R.S.

1977-01-01

The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

Land use and second-generation biofuel feedstocks: The unconsidered impacts of Jatropha biodiesel in Rajasthan, India

International Nuclear Information System (INIS)

Findlater, K.M.; Kandlikar, M.

2011-01-01

Governments around the world see biofuels as a common solution to the multiple policy challenges posed by energy insecurity, climate change and falling farmer incomes. The Indian government has enthusiastically adopted a second-generation feedstock - the oilseed-bearing shrub, Jatropha curcas - for an ambitious national biodiesel program. Studies estimating the production capacity and potential land use implications of this program have typically assumed that the 'waste land' slated for Jatropha production has no economic value and that no activities of note will be displaced by plantation development. Here we examine the specific local impacts of rapid Jatropha plantation development on rural livelihoods and land use in Rajasthan, India. We find that in Jhadol Tehsil, Jatropha is planted on both government and private land, and has typically displaced grazing and forage collection. For those at the socioeconomic margins, these unconsidered impacts counteract the very benefits that the biofuel programs aim to create. The Rajasthan case demonstrates that local land-use impacts need to be integrated into decision-making for national targets and global biofuel promotion efforts. - Highlights: → Hardy biofuel crops like Jatropha replace edible feedstocks that use arable land. → In Rajasthan, Jatropha displaces grazing and forage on both public and private land. → As Jatropha plantations mature, the loss of grass becomes more pronounced. → Unconsidered impacts negate the benefits that the biodiesel program aims to create. → Local land-use impacts need to be integrated into decision-making.

Land use and second-generation biofuel feedstocks: The unconsidered impacts of Jatropha biodiesel in Rajasthan, India

Energy Technology Data Exchange (ETDEWEB)

Findlater, K.M. [Institute for Resources Environment and Sustainability, University of British Columbia, 429-2202 Main Mall, Vancouver, BC, V6T1Z4 (Canada); Kandlikar, M., E-mail: milind.k@ubc.ca [Liu Institute for Global Studies, University of British Columbia, 6476 NW Marine Drive, Vancouver, BC, V6T1Z2 (Canada)

2011-06-15

Governments around the world see biofuels as a common solution to the multiple policy challenges posed by energy insecurity, climate change and falling farmer incomes. The Indian government has enthusiastically adopted a second-generation feedstock - the oilseed-bearing shrub, Jatropha curcas - for an ambitious national biodiesel program. Studies estimating the production capacity and potential land use implications of this program have typically assumed that the 'waste land' slated for Jatropha production has no economic value and that no activities of note will be displaced by plantation development. Here we examine the specific local impacts of rapid Jatropha plantation development on rural livelihoods and land use in Rajasthan, India. We find that in Jhadol Tehsil, Jatropha is planted on both government and private land, and has typically displaced grazing and forage collection. For those at the socioeconomic margins, these unconsidered impacts counteract the very benefits that the biofuel programs aim to create. The Rajasthan case demonstrates that local land-use impacts need to be integrated into decision-making for national targets and global biofuel promotion efforts. - Highlights: > Hardy biofuel crops like Jatropha replace edible feedstocks that use arable land. > In Rajasthan, Jatropha displaces grazing and forage on both public and private land. > As Jatropha plantations mature, the loss of grass becomes more pronounced. > Unconsidered impacts negate the benefits that the biodiesel program aims to create. > Local land-use impacts need to be integrated into decision-making.

IAEA Leads Operational Safety Mission to Rajasthan Atomic Power Station 3 and 4

International Nuclear Information System (INIS)

2012-01-01

Full text: An international team of nuclear safety experts led by the International Atomic Energy Agency (IAEA) today completed a review of safety practices at Units 3 and 4 of the Rajasthan Atomic Power Station in Rawatbhata. The team noted a series of good practices and made recommendations and suggestions to reinforce safety practices. The IAEA assembled the Operational Safety Review Team (OSART) at the request of the Government of India. Led by the IAEA's Division of Nuclear Installation Safety, the team performed an in-depth operational safety review from 29 October to 14 November 2012. The team was comprised of experts from Canada, Belgium, Finland, Germany, Romania, Slovakia, Slovenia, Sweden and the IAEA. The team conducted an in-depth review of the aspects essential to the safe operation of the Power Plant. The conclusions of the review are based on the IAEA's Safety Standards and good international practices. The review covered the areas of Management, Organization and Administration; Training; Operations; Maintenance; Technical Support; Operating Experience; Radiation Protection; Chemistry; Emergency Planning and Preparedness; and Severe Accident Management. The OSART team identified a number of good practices of the plant. These will be shared in due course by the IAEA with the global nuclear industry for consideration. Examples include the following: - The Power Plant's safety culture cultivates a constructive work environment and a sense of accountability among the Power Plant personnel, and gives its staff the opportunity to expand skills and training; - The Power Plant's Public Awareness Programme provides educational opportunities to the local community about nuclear and radiation safety; - The Power Plant has a Management of Training and Authorization system for effective management of training activities; and - The Power Plant uses testing facilities and mockups to improve the quality of maintenance work and to reduce radiation doses. The OSART

Pattern of invasion by Adhatoda vasica in savannas of Sariska Tiger Reserve, Rajasthan, Western India

NARCIS (Netherlands)

Bhatt, Priyanka; Rawat, G.S.; Sankar, K.; Tomlinson, K.W.; Langevelde, van F.

2015-01-01

As part of global experiments on Savanna vegetation, we examined the ecological characteristics of an important
semiarid savanna in the Indian sub-continent i.e Sariska Tiger Reserve, Rajasthan, Western India during April 2009 to
May 2011. 149 plots across five line transects were sampled

Traditional knowledge on zootherapeutic uses by the Saharia tribe of Rajasthan, India

OpenAIRE

Jaroli DP; Mahawar Madan

2007-01-01

Abstract The present zootherapeutic study describes the traditional knowledge related to the use of different animals and animal-derived products as medicines by the Saharia tribe reside in the Shahabad and Kishanganj Panchayat Samiti's of Baran district of Rajasthan, India. A field survey was conducted from April to June 2006 by performing interview through structured questionnaire with 21 selected respondents, who provided information regarding use of animals and their products in folk medi...

Baseline radiological monitoring at proposed uranium prospecting site at Rohil Sikar, Rajasthan

International Nuclear Information System (INIS)

Kumar, Rajesh; Jha, V.N.; Sahoo, N.K.; Jha, S.K.; Tripathi, R.M.

2018-01-01

Once economically viable grades of uranium deposits are proposed for mining and processing by the industry radiological baseline studies are required for future comparison during operational phases. The information collected during such studies serve as connecting feature between regulatory compliance and technical information. Present paper summarizes the results of baseline monitoring of atmospheric 222 Rn and gamma level in the area at prospecting mining, milling and waste disposal sites of Rohil Rajasthan

Sequence characterization of cotton leaf curl virus from Rajasthan: phylogenetic relationship with other members of geminiviruses and detection of recombination.

Science.gov (United States)

Kumar, A; Kumar, J; Khan, J A

2010-04-01

Diseased cotton plants showing typical leaf curl symptoms were collected from experimental plot of Agriculture Research Station-Sriganganagar, Rajasthan. Complete DNA-A component from samples taken from two areas were amplified through rolling circle amplification (RCA) using templiphi kit (GE Healthcare) and characterized. DNA-A of one isolate consists of 2751 nucleotides and second isolate of 2759 nucleotide. Both sequences comprised six ORF's. Genome organization of DNA-A of one isolate shows high sequence similarity with other characterized local begomovirus isolates of Rajasthan, while other isolate shows high sequence similarity with CLCuV reported from Pakistan. The maximum similarity of first isolate, CLCuV-SG01, shows highest sequence identity with Cotton leaf curl Abohar (Rajasthan) virus, and second isolate, CLCuV-SG02, shows highest sequence identity with cotton leaf curl virus from Pakistan. Both isolates showed 85% similarities with each other. The sequence data revealed probable infiltration of some strains of Cotton leaf curl virus from Pakistan to India, or co-existence of different isolates under similar geographical conditions. While CLCuV-SG01 shows highest nt sequence similarity with CLCuV Rajasthan (Abohar), nt identity of V1 ORF (encoding coat protein) of SG01 shows the highest nt identity (100%) with CLCuV Multan (Bhatinda) and Abohar virus while AC1 region also showed difference. Complete nucleotide sequence of SG01 shows only 86% similarity with CLCuV Multan virus. Similarity search revealed significant difference in AV1 and AC1 regions with respect to DNA-A suggesting an evolutionary history of recombination. Computer based analysis, recombination detection Program (RDP) supports the recombination hypothesis, indicated that recombination with other begomoviruses had taken place within V1 ORF and AC1 ORF of CLCuV-SG01 and AC1 ORF of CLCuV-SG02 and also in noncoding intergenic region (IR).

Uranium and REE potential of the albitite-pyroxenite-microclinite belt of Rajasthan, India

International Nuclear Information System (INIS)

Singh, Govind; Sharma, D.K.; Yadav, O.P.; Jain, Rajan B.; Singh, Rajendra

1998-01-01

A number of radioactive albitite, pyroxenite and microclinite occurrences have been identified in north and central Rajasthan, along or in close proximity to major lineaments, from Dancholi - Mewara in the NE to Tal in the SW. With these new findings the total extent of Albitite belt of Rajasthan now stands at over 320 km. These occurrences have been evaluated on the basis of their U, Th and REE content to identify the potential areas for the second phase of uranium exploration programme. Further, based on the various characteristic features of radioactive host rocks, the Albitite Belt has been divided into five sectors. The U 3 O 8 content of albitites varies from 0.008 to 0.44% and of pyroxenites from 0.022 to 2.0% whereas ThO 2 varies from < 0.005 to 0.83% in albitites and <0.005 to 0.033% in pyroxenities. These albitites, microclinites and pyroxenites are also characterised by anomalous concentration of REEs. Uranium and REE bearing phases are represented by uraninite, brannerite, davidite, fergusonite, monazite, anatase, rutile, zircon, allanite and britholite. The data accrued so far suggest that U and REE potential of the Mewara-Maonda and Hurra Ki Dhani-Rohil sectors are very high and hence needs further detailed integrated exploration. (author)

Scrub typhus in rural Rajasthan and a review of other Indian studies.

Science.gov (United States)

Masand, Rupesh; Yadav, Ritesh; Purohit, Alok; Tomar, Balvir Singh

2016-05-01

Scrub typhus is an acute febrile illness which has been reported from various parts of India with Rajasthan recently joining the list of affected states. To report a series of paediatric scrub typhus cases from rural Rajasthan. Retrospective review of children with scrub typhus admitted to the wards and paediatric intensive care unit (PICU) of a tertiary-care hospital. The study was undertaken over an 8-month period from May to December 2013. All patients with a clinical presentation and/or serological confirmation of scrub typhus who tested negative for malaria, enteric fever, dengue, leptospirosis and urinary tract infection (UTI) were included. A range of investigations were undertaken including IgM-ELISA for scrub typhus, followed by appropriate medical management. Thirty patients satisfied the inclusion criteria. The mean (SD, range) age of the patients was 8·56 (3·43, 3-16) years. The most common clinical features were fever (n = 30, 100%), headache (n = 20, 66%), myalgia (n = 15, 50%), hepatosplenomegaly (n = 18, 60%) and pallor (n = 5, 16%). Typical features such as eschar and rash were observed in only one (3·3%) and three (10%) patients, respectively; none had generalised lymphadenopathy or conjunctival congestion. IgM-ELISA for scrub typhus was positive in 28 patients (93·3%) and 27 responded to doxycycline within 24-72 hours. One of the three patients who required PICU support responded to intravenous chloramphenicol and, of the other two (6·6%), one died of acute respiratory distress syndrome and the other owing to acute renal failure. A high index of suspicion is essential for early diagnosis and prevention of complications in scrub typhus together with prompt referral from rural areas to a higher centre. Awareness of the disease manifestations may further help to prevent excessive investigations in patients presenting with non-specific febrile illness and reduce the economic burden to the family and society in resource

Reproductive health and empowerment -- a Rajasthan perspective.

Science.gov (United States)

Pal, P; Joshi, V

1996-01-01

Reproductive health is one of the major issues of current feminist debates. The issue was brought to light because of population control policies which are being enforced through women's bodies and the spread of HIV/AIDS. In this context, women's organizations and activists are trying to focus upon the issue of reproductive health as part of the larger issue of the position of women in families, societies, and states. Policy makers and donor agencies are trying to address the problem as lack of awareness and knowledge of how to use contraceptives. The authors argue in this situation that it is important to study reproductive health relative to the status of women in society. This paper looks at the existing social construct of patriarchy and population control policies in relation to reproductive health. Women and self, the reproductive role of women, preference for male children, family planning decision making, family planning programs and reproductive health, and the Vikalp program in two districts of Rajasthan are discussed.

Forest structure, diversity and soil properties in a dry tropical forest in Rajasthan, Western India

Directory of Open Access Journals (Sweden)

J. I. Nirmal Kumar

2011-06-01

Full Text Available Structure, species composition, and soil properties of a dry tropical forest in Rajasthan Western India, were examined by establishment of 25 plots. The forest was characterized by a relatively low canopy and a large number of small-diameter trees. Mean canopy height for this forest was 10 m and stands contained an average of 995 stems ha-1 (= 3.0 cm DBH; 52% of those stems were smaller than 10 cm DBH. The total basal area was 46.35 m2ha-1, of which Tectona grandis L. contributed 48%. The forest showed high species diversity of trees. 50 tree species (= 3.0 cm DBH from 29 families were identified in the 25 sampling plots. T. grandis (20.81% and Butea monosperma (9% were the dominant and subdominant species in terms of importance value. The mean tree species diversity indices for the plots were 1.08 for Shannon diversity index (H´, 0.71 for equitability index (J´ and 5.57 for species richness index (S´, all of which strongly declined with the increase of importance value of the dominant, T. grandis. Measures of soil nutrients indicated low fertility, extreme heterogeneity. Regression analysis showed that stem density and the dominant tree height were significantly correlated with soil pH. There was a significant positive relationship between species diversity index and soil available P, exchangeable K+, Ca2+ (all p values < 0.001 and a negative relationship with N, C, C:N and C:P ratio. The results suggest that soil properties are major factors influencing forest composition and structure within the dry tropical forest in Rajasthan.

Medicinal plant diversity of Sitamata wildlife sanctuary, Rajasthan, India.

Science.gov (United States)

Jain, Anita; Katewa, S S; Galav, P K; Sharma, Pallavi

2005-11-14

The present study has been carried out in Sitamata wildlife sanctuary of Chittorgarh and Udaipur district located in south-west region of Rajasthan. A field survey of the study area was carried out during 2002-2004 to document the medicinal utility of herbs occurring in this area. Two hundred fourty-three genera belonging to 76 families have been reported which are used by the tribals of about 50 villages around the sanctuary as means of primary health care to cure various ailments. The study revealed the new ethnobotanical uses of 24 plant species belonging to 20 genera. A list of plant species along with their local name, plant part/s used and mode of administration for effective control in different ailments of ethnomedicinal plants are given.

Assisting community management of groundwater: Irrigator attitudes in two watersheds in Rajasthan and Gujarat, India

Science.gov (United States)

Varua, M. E.; Ward, J.; Maheshwari, B.; Oza, S.; Purohit, R.; Hakimuddin; Chinnasamy, P.

2016-06-01

The absence of either state regulations or markets to coordinate the operation of individual wells has focussed attention on community level institutions as the primary loci for sustainable groundwater management in Rajasthan and Gujarat, India. The reported research relied on theoretical propositions that livelihood strategies, groundwater management and the propensity to cooperate are associated with the attitudinal orientations of well owners in the Meghraj and Dharta watersheds, located in Gujarat and Rajasthan respectively. The research tested the hypothesis that attitudes to groundwater management and farming practices, household income and trust levels of assisting agencies were not consistent across the watersheds, implying that a targeted approach, in contrast to default uniform programs, would assist communities craft rules to manage groundwater across multiple hydro-geological settings. Hierarchical cluster analysis of attitudes held by survey respondents revealed four statistically significant discrete clusters, supporting acceptance of the hypothesis. Further analyses revealed significant differences in farming practices, household wealth and willingness to adapt across the four groundwater management clusters. In conclusion, the need to account for attitudinal diversity is highlighted and a framework to guide the specific design of processes to assist communities craft coordinating instruments to sustainably manage local aquifers described.

Economics of Garlic Production in Baran District of Rajasthan; Break Even Analysis

OpenAIRE

Meena, Lokesh Kumar; Sen, Chandra; Bairwa, Shoji lal; Arunjhajharia; Raghuwanshi, N. K.

2013-01-01

The study focuses on economic analysis of garlic production in the Baran District of Rajasthan. The study is carried out to determine break even analysis and constraints of garlic production in the study area. Break even analysis is carried out to arrive at that minimum level at which optimum conditions of cost and returns is equated that is no profit no loss point. In this study selected small, medium and large farmers will not be at loss even if their actual yield of garlic is decline by 56...

Uranium exploration in albitised rocks of North Delhi Fold Belt in Rajasthan and Haryana, India

International Nuclear Information System (INIS)

Pandey, P.; Khandelwal, M.; Bhairam, C.; Parihar, P.

2014-01-01

Uranium deposits in Na-metasomatised granites and metasediments are reported from several places in the world. In India, uranium mineralization associated with soda metasomatic activity has been recognized at a number of places in North Delhi Fold Belt (NDFB) in Rajasthan and adjoining Haryana. Exploration activities for uranium in Khetri Sub Basin (KSB) of North Delhi Fold Belt (NDFB) in last six decades have resulted in locating number of uranium occurrences in the albitites and albitised metasediments at Sior, Siswali, Maonda, Hurra ki Dhani, Diara, Saladipura, Khandela, Rohil, Ghateshwar, Bichun, Sakhun, Ladera and Chota Udaipur in parts of Rajasthan and Dhancholi, Raghunathpura, Rambas and Gorir, in parts of Haryana. Incidentally, the occurrences fall along a NNE-SSW trending “Albitite line”, which comprises a 170 km long, structurally weak zone/lineament and axial trace of major folds in the KSB extending from Raghunathpura in Mahendragarh district of southern Haryana to Ladera-Sakun-Bichun in Rajasthan. Lithounits of KSB comprise lower Alwar Group consisting quartzite, amphibole quartzite, subordinate phyllite and schist and upper Ajabgarh Group consisting schist, phyllite, marble, quartzite and carbon phyllite. The post-Delhi magmatic activity in NDFB is represented by alkali granites, pegmatites, aplites and albitites. The rocks of Delhi supergroup have undergone low to medium grade metamorphism (amphibolite facies) and polyphase deformation. First two deformations with N-S to NNE-SSW axial plane are coaxial while the third phase have E-W axial plane. Prominent shear zones are developed along the N-S to NNE-SSW axial planes, characterized by intense silicification, brecciation and ferruginisation. The NE-SW trending disposition of albitised granites indicate that the metasomatic fluids originated during reactivation of the NE-SW trending Khetri lineament, caused pervasive albitisation of the preexisting rocks, the deformed lithounits providing conduits

Operating Experience with the BR-5 Reactor; Experience acquise aupres du reacteur BR-5; Opyt ehkspluatatsii reaktora BR-5; Experiencia practica con el reactor BR-5

Energy Technology Data Exchange (ETDEWEB)

Lejpunskij, A. I.; Kazachkovskij, O. D.; Pinkhasik, M. S.; Aristarkhov, N. N.; Karpov, A. V.; Larin, E. P.; Efimov, I. A.

1963-10-15

The paper discusses the carrying-out of repair and maintenance work on the radioactive liquid-metal circuit of the BR-5 fast neutron reactor. Attention is also given to problems of reactor operation after achievement of the planned 2% fuel burn-up with some disturbance of leak-tightness in individual fuel elements. An account is given of experience in discharging the active section, examining the condition and leak-tightness of the fuel elements, and decontaminating the equipment and piping of the first radioactive circuit after reaching 5% fuel burn-up. (author) [French] Dans ce memoire les auteurs decrivent l'execution des reparations et des travaux d'entretien dans le circuit radioactif liquide-metal du reacteur a neutrons rapides BR-5. Ils etudient egalement les problemes lies au fonctionnement du reacteur au taux de combustion de 2% prevu avec quelques defauts d'etancheite dans des elements combustibles particuliers. Ils decrivent le dechargementen zone active et examinent les conditions d'etancheite des elements combustibles. Ainsi que la decontamination de l'appareillage et des tuyauteries du premier circuit radioactif apres avoir atteint un taux de combustion de 5%. (author) [Spanish] En la memoria se examinan los problemas planteados por el mantenimiento del circuito radiactivo de metal liquido del reactor de neutrones rapidos BR-5. Se tratan cuestiones relacionadas con la explotacion del reactor una vez alcanzado el grado de combustion de 2%, previsto en el proyecto y luego de producirse ciertas alteraciones de la densidad de determinados elementos combustibles. Se describen la experiencia adquirida durante la descarga del cuerpo del reactor, las investigaciones del estado general y de la hermeticidad de los elementos combustibles y las operaciones de descontaminacion de la instalacion y de las tuberias del circuito radiactivo primario despues de alcanzado un grado de combustion de 5%. (author) [Russian] V doklade rassmatrivayutsya voprosy proizvodstva

Irradiation effects on Zr-2.5Nb in power reactors

Energy Technology Data Exchange (ETDEWEB)

Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

A status survey of common water-borne diseases in desert city Bikaner (NW Rajasthan, India).

Science.gov (United States)

Saxena, M M; Chhabra, Chetna

2004-03-01

Water is scarce and, in general, a low quality resource in desert areas and the Indian desert is no exception. With this in view, the present study was taken up to survey the status of common water-borne diseases epidemiological trends in the desert city Bikaner (NW Rajasthan). In the city, 15.5 per cent population and 44.5 per cent families were found to suffer from one or more common water-borne diseases including amoebiasis, diarrhoea, dysentery, jaundice and typhoid. No case of fluorosis was recorded. The highest incidence was that of diarrhoea (5.4 per cent population). The worst affected and safe zones in the city were identified and the trends of different diseases in different zones of the city are discussed. The highest incidence of diseases was noted during summer (58.8 per cent) followed by winter (34.1 per cent) and monsoon (7.0 per cent). Relationship of diseases with population attributes like age, education, economy and family size are also discussed. Attributes for contamination of drinking water have been tried to identify and safety measures suggested.

A Comparative Study of Job Satisfaction among Management Teachers of MBA Colleges in South Rajasthan

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Anil K. Bhatt

2015-03-01

Full Text Available Management teaching is a noble profession, a continuous process and indicator of economic and social development. The management teachers are the pre-requisite of the success of management programmes, their role in society has been increasing with the rapid growth of trade and finance in the whole world. Society is continuously observing Management teachers while there are various concerns in their role and their satisfaction. Only satisfied and well-adjusted management teacher can think of the wellbeing of the future managers. The major source of satisfaction for the management teachers comes from their own institution, also known as hygiene factors. In the light of this background, the aim of this study is to analyze the job satisfaction level among the MBA teachers in selected management colleges of south Rajasthan. For this purpose, the Job related hygiene factors were identified and then the satisfactions of 220 management teachers were sought on various dimensions. The data were analysed by using multiple regression and ANOVA analysis with SPSS-19 software to identify the factors responsible for satisfaction. The analysis revealed that three factors Physical Teaching condition, Flexible working hours and Environment providing hint of Job Security revealed the job satisfaction in Management teachers of South Rajasthan.

Impact of micronutrients sprinkle on weight and height of children aged 6-36 months in Tonk district of Rajasthan state

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Vijay Jyoti

2014-12-01

Full Text Available Introduction: In India, multiple micronutrient deficiencies continue to be a major public health problem, especially for children below three years of age. It is a crucial period for the growth and development of children. There is lack of data from Rajasthan state on the effect of micronutrients supplementation on growth of children below three years of age. Aims & Objectives: To assess the impact of ICDS supplementary food with or without micronutrients sprinkles on weight and height of children aged 6-36 months in the Tonk district, Rajasthan state. Materials & Methods: The trial was conducted in the 15 Angan wadi centers, each from Tonk (rural and Malpura blocks of Tonk District in Rajasthan state. Children from both blocks were considered as experimental and control groups. Experimental (N=790 and Control groups (N=540 received ICDS supplementary food for six months with or without micronutrients sprinkles. Anthropometric measurements were taken using standard techniques. Results: At baseline, children with severe underweight, severe stunting and severe wasting in experimental group stood at 19.2%, 19.3%, 7.3%, respectively, which declined to 14.9%, 15.3% and 6.3%, after intervention. Significant difference was observed in the mean weights of post intervention children between experimental and control groups, whereas, there was no significant difference in mean heights. In experimental group, statistical significant difference was also noted in the mean weights and heights of children between pre and post intervention periods. Conclusion: Micronutrients sprinkles can be effective in reducing malnutrition amongst vulnerable population.

Groundwater Quality in the Shallow Aquifers of the Hadauti Plateau of the District of Baran, Rajasthan, India

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Kumar Lokesh

2014-07-01

Full Text Available With the rapid pace of agricultural development, industrialization and urbanization, the commonly observed geogenic contaminants in groundwater are fluoride and nitrate, whereas nitrate is the dominant anthropogenic contaminant in the south-eastern plains of Rajasthan, India. Samples obtained using a tube well and hand pump in November, 2012, demonstrate that Na-Cl is the dominant salt in the groundwater, and the total salinity of the water is between 211-1056 mg L-1. Moreover, the observed sodium adsorption ratio (SAR and residual sodium carbonate (RSC values ranged between 0.87 to 26.22 meq L-1 and -12.5 to 30.5 meq L-1 respectively. The study further shows that 6% of the total samples contain high amounts of nitrate, and 49% contain fluoride. A water quality index (WQI rating was carried out using nine parameters to quantify the overall groundwater quality status of the area.

ESTIMATION OF EXTRACELLULAR LIPOLYTIC ENZYME ACTIVITY BY THERMOPHILIC BACILLUS SP. ISOLATED FROM ARID AND SEMI-ARID REGION OF RAJASTHAN, INDIA

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Deeksha Gaur

2012-10-01

Full Text Available Thermophilic organisms can be defined as, micro-organisms which are adapted to survive at high temperatures. The enzymes secreted by thermophilic bacteria are capable of catalyzing biochemical reactions at high temperatures. Thermophilic bacteria are able to produce thermostable lipolytic enzymes (capable of degradation of lipid at temperatures higher than mesophilic bacteria. Therefore, the isolation of thermophilic bacteria from natural sources and their identification are quite beneficial in terms of discovering thermostable lipase enzymes. Due to great temperature fluctuation in hot arid and semi-arid region of Rajasthan, this area could serve as a good source for new thermophilic lipase producing bacteria with novel industrially important properties. The main objective of this research is the isolation and estimation of industrially important thermophilic lipase enzyme produced by thermophilic bacteria, isolated from arid and semi-arid region of Rajasthan. For this research purpose soil samples were collected from Churu, Sikar and Jhunjunu regions of Rajasthan. Total 16 bacterial strains were isolated and among only 2 thermostable lipolytic enzyme producing bacteria were charcterized. The thermostable lipolytic enzyme was estimated by qualitative and quantitative experiments. The isolates were identified as Bacillus sp. by microscopic, biochemical and molecular characterization. The optimum enzyme activity was observed at pH 8, temperature 60°C and 6% salt concentrations at 24 hrs time duration. Lipolytic enzyme find useful in a variety of biotechnological fields such as food and dairy (cheese ripening, flavour development, detergent, pharmaceutical (naproxen, ibuprofen, agrochemical (insecticide, pesticide and oleochemical (fat and oil hydrolysis, biosurfactant synthesis industries. Lipolytic enzyme can be further used in many newer areas where they can serve as potential biocatalysts.

Production and quality control of concrete for the Rajasthan Atomic Power Station - [Part 1

International Nuclear Information System (INIS)

Singha Roy, P.K.; Sukhtankar, K.D.; Prasad, K.

1975-01-01

The production and quality control of concrete and concrete materials for the construction of the twin-reactor Rajasthan Atomic Power Station with its 400 MW net capacity posed many challenges since many of the requirements for the properties of concrete were new and were being laid down for the first time in India. Some of the conditions for the concrete included leak-tightness against gas pressure, total absence of shrinkage in the containment even when the ambient temperature during concreting was as high as 45degC, placing concrete at a temperature as low as 8degC, the use of non-shrink and high strength grout, absolute impermeability against water, high density for radiation shielding, controlled modulus of elasticity for large machine foundations, high strength with high slump for the prestressed concrete dome, etc. Though the total quantity of concrete was not very much compared with a large river valley or steel plant project, (e.g., about 1.2 X 10 6 m 3 for a 2-million tonne steel plant) it was quite significant, being about 70,000 m 3 of normal density and 2,100 m 3 of high density concrete. The production of these quantities entailed intensive material study and investigation, development of new mixes with additives not tried out before in the country, and design and quality control techniques which were unique in many respects. The paper deals with the production and quality control of concrete, including grouts used in the projects, but the actual concreting and construction operations are not discussed. (author)

Thermo Gravimetric and Differential Thermal Analysis of Clay of Western Rajasthan (india)

Science.gov (United States)

Shekhawat, M. S.

The paper presents the study of thermo gravimetric and differential thermal analysis of blended clay. Western part of Rajasthan (India) is rich resource of Ball clays and it is mainly used by porcelain, sanitary ware, and tile industry. The quality and grade of clay available in the region vary from one deposit to other. To upgrade the fired colour and strength properties, different variety of clays may be blended together. The paper compares the results of thermal analysis one of blended clay B2 with reference clay of Ukraine which is imported by industries owners. The result revealed that the blended clay is having mineral kaolinite while the Ukrainian clay is Halloysite.

Factors Influencing Institutional-Based Pediatric Rehabilitation Services among Caregivers of Children with Developmental Delay in Southwestern Rajasthan

OpenAIRE

Mishra, Kriti; Siddharth, V.

2018-01-01

Context: A limited number of caregivers of children with developmental delay access rehabilitation facilities in India. The study explored utilization of rehabilitation services at a tertiary care setup in southwestern Rajasthan and various factors influencing it. Aims: The aim of this study is to explore rehabilitation service utilization among children with developmental delay at a tertiary care setup and to ascertain factors that influence this pattern. Settings: This study was conducted a...

Forest structure, diversity and soil properties in a dry tropical forest in Rajasthan, Western India

OpenAIRE

J. I. Nirmal Kumar,; Kanti Patel,; Rohit Bhoi Kumar

2011-01-01

Structure, species composition, and soil properties of a dry tropical forest in Rajasthan Western India, were examined by establishment of 25 plots. The forest was characterized by a relatively low canopy and a large number of small-diameter trees. Mean canopy height for this forest was 10 m and stands contained an average of 995 stems ha-1 (≥ 3.0 cm DBH); 52% of those stems were smaller than 10 cm DBH. The total basal area was 46.35 m2ha-1, of which Tectona grandis L. contributed 48%. The fo...

Inequities in coverage of preventive child health interventions: the rural drinking water supply program and the universal immunization program in Rajasthan, India.

Science.gov (United States)

Mohan, Pavitra

2005-02-01

I assessed whether the Rural Drinking Water Supply Program (RDWSP) and the Universal Immunization Program (UIP) have achieved equitable coverage in Rajasthan, India, and explored program characteristics that affect equitable coverage of preventive health interventions. A total of 2460 children presenting at 12 primary health facilities in one district of Rajasthan were enrolled and classified into economic quartiles based on possession of assets. Immunization coverage and prime source of drinking water were compared across quartiles. A higher access to piped water by wealthier families (P< .001) was compensated by higher access to hand pumps by poorer families (P<.001), resulting in equal access to a safe source (P=.9). Immunization coverage was inequitable, favoring the wealthier children (P<.001). The RDWSP has achieved equitable coverage, while UIP coverage remains highly inequitable. Programs can make coverage more equitable by formulating explicit objectives to ensure physical access to all, promoting the intervention's demand by the poor, and enhancing the support and monitoring of frontline workers who deliver these interventions.

Dental caries experience and treatment needs of green marble mine laborers in Udaipur district, Rajasthan, India

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Duraiswamy Prabu

2008-01-01

Full Text Available Background and Objectives: The study was undertaken at Kesariyaji, located in Udaipur district of Rajasthan. There are about 3 million workers who marble mine at Rajasthan. Living conditions of these workers are substandard and most of them are immigrant workers living in tiny shacks. Majority of them belong to lower socioeconomic status with poor educational background. The present study was carried out to estimate dental caries prevalence and treatment needs of laborers working in the green marble mines of Udaipur district. Basic Research Design: The data was collected using the methods and standards recommended by the WHO. Dentition status and treatment needs along with decayed, missing, and filled teeth (DMFT index, and decayed, missing, and filled surfaces score were recorded. Standard error of mean was calculated for all the mean values of treatment needs. There were three examiners, who were trained before the survey for inter-examiner variability, and the reliability was tested by means of weighted kappa statistics, which was 90%. Participants: The study population comprised 513 men in four age groups of 18-25, 26-34, 35-44, and 45-54 years, respectively. Results: The mean DMFT for all age groups was 3.13 with highest mean of 4.0 for the age group of 45-54 years. Mean decayed teeth were 2.60, 3.33, 1.46, and 1.5 for the age groups 15-24, 25-34, 35-44, and 45-54 years, respectively. Filled component was nil for all age groups. Most of the subjects required one surface filling with a very less proportion needing pulp care. Conclusions: The missing component constituted the major part of DMFT index in the 45-54 years age group and the absence of filled component in the whole study population implies that the treatment needs of the study population are unmet. Thus, intervention in the form of oral health promotion and curative services are the need of the hour.

PREVALENCE OF ANEMIA AND ITS SOCIO - DEMOGRAPHIC DETERMINANTS IN PREGNANT WOMEN AT A TERTIARY CARE HOSPITAL IN JAIPUR, RAJASTHAN

OpenAIRE

Prabhjot Singh; Swati; Rajat; Urvahi; Karnika; Isha; Navsangeet; Arihant

2015-01-01

AIM: Prevalence of anemia and its socio - demographic determinants in pregnant women at a tertiary care hospital in Jaipur, Rajasthan . MATERIALS AND METHODS: All the pregnant women aged 25 to 35 years , registered at antenatal clinic at Department of Obstetrics an d Gynaecology , Mahatma Gandhi Medical College , Jaipur were included. A predesigned and pre tested questionnaire was used to elicit the information. Various possible causes of anaemi...

Antenatal care among currently married women in Rajasthan, India

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Alok Chauhan

2012-10-01

Full Text Available Objective: To assess the utilization of antenatal care (ANC services among currently married women in Rajasthan. Methods: The data have been extracted from District Level Household and Facility Survey (DLHS-3 which was conducted during 2007-2008, all over India. A total of 12 458 currently married women in the age group of 15-49 were taken as the sample for the study. Cross tabulation and binary logistic regression method were applied to determine the factors influencing ANC. Results: Out of 12 458 respondents, 43.4 percent women not received even a single ANC during their pregnancy period. 45.1 percent of the women not received tetanus toxoid injection and 13.0 percent of the women not received Iron folic acid tablets during their pregnancy period. Only 6.6 percent of women fulfilled the minimum recommendation with regard ANC services. Conclusions: The study points to the avenues through which policy makers can formulate and implement policies on a realistic basis by identifying critical variables and target groups for effective utilisation of ANC.

Comparative assessment of Oral Hygiene and Periodontal status among children who have Poliomyelitis at Udaipur city, Rajasthan, India

OpenAIRE

Tak, Mridula; Nagarajappa, Ramesh; Sharda, Archana; Asawa, Kailash; Tak, Aniruddh; Jalihal, Sagar

2012-01-01

Objective: To assess and compare the oral hygiene and periodontal status among children with Poliomyelitis having upper limb disability, lower limb disability and both upper and lower disability at Udaipur city, Rajasthan, India. Study design: Total sample comprised of 344 Poliomyelitis children (upper limb disability: 33.4%; lower limb disability: 33.7%; both upper and lower limb disability: 32.9%) in the age group of 12-15 years. Clinical examination included recording Simplified Oral Hygie...

Start-up analysis of INET-5 MW district heating prototype reactor

International Nuclear Information System (INIS)

Li Tianshu

1991-09-01

The main features and thermohydraulic design parameters of the INET-5 MW reactor (INET: Institute of Nuclear Technology of Tsinghua University, Beijing) are presented. The start-up process and the effect of thermohydraulic instability on start-up process have been analyzed. The main obstacle of start-up process of INET-5 MW reactor is to pass the instability region from 1 atm to normal operation condition. For avoiding instability, the start-up process should be divided into two steps. The results of three different start-up proposals calculated by DACOL code are given and compared. The possibility of instabilities for each proposal has been checked. The checked results show that there is no instability during start-up of the three proposals. So, it is supposed that the INET-5 MW reactor can safely and stably reach the operation conditions. Finally, some conclusions about the effect of instability on start-up in boiling mode of INET-5MW reactor are given

Unusual nocturnal feeding by Brown Rock-chat Cercomela fusca (Passeriformes: Muscicapidae in Bikaner, Rajasthan, India

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P. Singh

2009-04-01

Full Text Available Brown Rock-chat is a diurnal insectivorous bird commonly associated with human habitations. I report here nocturnal foraging of the species in and around Bikaner in Rajasthan. The birds showed a marked bimodal activity during their nocturnal foraging which peaked in early morning and late evening hours. Bright sodium-vapour lights that attracted a horde of insects during monsoons offered ideal foraging opportunities for the birds. This behavior is explained here as an adaptation to maximize their food intake during the period when the birds breed and their nutritional requirements are naturally high.

Seroprevalence of Hepatitis B surface antigen, antibodies to the Hepatitis C virus, and human immunodeficiency virus in a hospital-based population in Jaipur, Rajasthan

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Sood Smita

2010-01-01

Full Text Available Background: Hepatitis B, hepatitis C, and HIV infections are a serious global and public health problem. To assess the magnitude and dynamics of disease transmission and for its prevention and control, the study of its seroprevalence is important. A private hospital catering to the needs of a large population represents an important center for serological surveys. Available data, at Rajasthan state level, on the seroprevalence of these bloodborne pathogens is also very limited. Objective: A study was undertaken to estimate the seroprevalence of hepatitis B surface antigen (HBsAg and antibodies to hepatitis C (anti-HCV Ab and human immunodeficiency virus (anti-HIV Ab in both the sexes and different age groups in a hospital-based population in Jaipur, Rajasthan. Materials and Methods: Serum samples collected over a period of 14 months from patients attending OPDs and admitted to various IPDs of Fortis Escorts Hospital, Jaipur, were subjected within the hospital-based lab for the detection of HBsAg and anti-HCV Ab and anti-HIV Ab using rapid card tests. This was followed by further confirmation of all reactive samples by a microparticle enzyme immunoassay (Abbott AxSYM at Super Religare Laboratories (formerly SRL Ranbaxy Reference Lab, Mumbai. Results: The seroprevalence of HBsAg was found to be 0.87%, of anti-HCV Ab as 0.28%, and of anti-HIV Ab as 0.35%. Conclusion: The study throws light on the magnitude of viral transmission in the community in the state of Rajasthan and provides a reference for future studies.

Fluoride, boron and nitrate toxicity in ground water of northwest Rajasthan, India.

Science.gov (United States)

Chaudhary, Veena; Kumar, Mukesh; Sharma, Mukesh; Yadav, B S

2010-02-01

The study was carried out to access the fluoride, boron, and nitrate concentrations in ground water samples of different villages in Indira Gandhi, Bhakra, and Gang canal catchment area of northwest Rajasthan, India. Rural population, in the study site, is using groundwater for drinking and irrigation purposes, without any quality test of water. All water samples (including canal water) were contaminated with fluoride. Fluoride, boron, and nitrate were observed in the ranges of 0.50-8.50, 0.0-7.73, and 0.0-278.68 mg/l, respectively. Most of the water samples were in the categories of fluoride 1.50 mg/l, of boron 2.0-4.0 mg/l, and of nitrate availability of these compounds in groundwater was due to natural reasons and by the use of chemical fertilizers.

Uranium in ancient slag from Rajasthan

International Nuclear Information System (INIS)

Pradeepkumar, T.B.; Fahmi, Sohail; Sharma, S.K.

2008-01-01

Anomalous radioactivity was recorded in two ancient slag dumps spread on the surface near Bansda (24 deg 35'N lat., 70 deg 09'E long.) and Dhavadiya (24 deg 30'N lat., 70 deg 05'E long.) villages, Udaipur District, Rajasthan. The slag, with a range of high to low radioactivity levels, is the remnant of ancient smelting in the area, probably for copper. Six samples showing low radioactivity in Bansda contain an average of 0.030% U 3 O 8 , while five moderately radioactive samples analysed contain 0.225% U 3 O 8 and four highly radioactive samples analysed contain 1.15% U 3 O 8 . The 15 samples contain on an average 0.627% copper, 719 ppm zinc, 329 ppm cobalt and 133 ppm vanadium. Fifteen samples from Dhavadiya slag assayed on an average contain 0.040% U 3 O 8 , 0.297% Cu, 292 ppm Zn and 250 ppm Co. The extent of crystallization seen in the slag is intriguing because an over-cooled melt generally forms glass. The high rate of crystal formation may be attributed to high amounts of volatiles, particularly CO 2 and SO 4 , released during the breakdown of limestone (added as flux during smelting) and sulphide minerals in the ore. The high order of radioactivity recorded in the slags of Bansda and Dhavadiya points to the presence of ore-grade uranium concentration associated with sulphide mineralization in the vicinity of the basement Banded Gneissic Complex, intrusive granites and the cover sequence of the Bhinder basin. (author)

Socio-economic effect on socially-deprived communities of developing drinking water quality problems in arid and semi-arid area of central Rajasthan

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I. Husain

2014-09-01

Full Text Available Rajasthan is well known for its Great Thar desert. Central Rajasthan has an arid to semi-arid environment. The area faces either scarcity of water or poor quality of drinking water. In some areas water is transported 2 km or more, which uses time, energy and money. Rich people have their own sources, which is restricted for use by others. Such conditions are affecting socially-deprived communities, both socially and economically. Groundwater is a major source of drinking water due to the unavailability of surface water. There is a lack of groundwater quality knowledge in the community and the data available is hard to understand by consumers. The CCME Water Quality Index is a tool to simplify the water quality report by rating the water on quality standards. It provides meaningful summaries of overall water quality and trends, which is accessible to non-technical lay people. In the present study the objective is to examine the groundwater quality of six districts (Ajmer, Bhilwara, Pali, Rajasamand, Nagaur and Jodhpur, centrally located in Rajasthan, with arid and semi-arid conditions. CCME WQI is also evaluated to produce quality data in a form to be understood by the community. A total of 4369 groundwater sources in 1680 villages from six districts (76 546 km2 were collected and examined. Results are outlined in the Bureau of Indian Standards (BIS: 10500, 2012 and 2952 sources are unsafe for drinking. According to CCME WQI groundwater of 93 villages is poor, 343 villages are marginal, and 369 villages are fair in quality. Toxicological studies of unsafe drinking water and their remedial measures are also discussed. A tentative correlation between prevailing water-borne diseases and quality parameter has also been shown

Awareness and Attitude of Physicians in Academia towards Human Stem Cell Research (HSCR) and Related Policies in Rajasthan, India

OpenAIRE

Nitin K Joshi; Latika Nath; Vibha Joshi; Anil Purohit

2015-01-01

Introduction: In India, several science agencies are promoting Stem Cell Research (SCR). There is paucity of studies which document the perception of doctors about SCR, especially physicians in academia. This study was carried out to assess perception of physicians in academia towards Human Stem Cell Research (HSCR) and related policies in India. Methods: We interviewed 200 doctors from three different government medical colleges of Rajasthan. A semi-structured questionnaire was used to disce...

Applicability of RELAP5 for safety analysis of AP600 and PIUS reactors

International Nuclear Information System (INIS)

Motloch, C.G.; Modro, S.M.

1990-01-01

An assessment of the applicability of using RELAP5 for performing safety analyses of the AP600 and PIUS advanced reactor concepts is being performed. This ongoing work is part of a larger safety assessment of advanced reactors sponsored by the United States Nuclear Regulatory Commission. RELAP5 models and correlations are being reviewed from the perspective of the new AP600 and PIUS phenomena and features that could be important to reactor safety. The purpose is to identify those areas in which new mathematical models of physical phenomena would be required to be added to RELAP5. In most cases, the AP600 and PIUS designs and systems and the planned and off-normal operations are similar enough to current Pressurized Water Reactors (PWR) that RELAP5 safety analysis applicability is unchanged. However, for AP600 the single most important systemic and phenomenological difference between it and current PWRs is in the close coupling between the reactor system and the containment during postulated Loss of Coolant Accident (LOCA) events. This close coupling may require the addition of some thermal-hydraulic models to RELAP5. And for PIUS, the most important new feature is the thermal density locks. These and other important safety-related features are discussed. This document presents general descriptions of RELAP5, AP600, and PIUS, describes the new features and phenomena of the reactors, and discusses the code/reactors safety-related issues. 32 refs., 4 figs., 2 tabs

Dental caries and treatment needs of children (6-10 years in rural Udaipur, Rajasthan

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Dhar Vineet

2009-01-01

Full Text Available Objective : This study was done to examine caries prevalence and corresponding treatment needs in school children of rural areas of Udaipur, Rajasthan. Materials and Methods : A total of 750 children of rural areas in the age group of 6-10 years were examined, using WHO index, to record the prevalence of dental caries and treatment needs. The results were subjected to statistical analysis using chi square test. Results : Dental caries was found in 63.20% children, and 85.07% children needed dental treatment. The highest need was of one surface filling (85.73% followed by sealant (51.20%. Conclusion : Dental caries showed to be a significant health issue in the rural population requiring immediate attention.

Prevalence, Virulence Potential, and Antibiotic Susceptibility Profile of Listeria monocytogenes Isolated From Bovine Raw Milk Samples Obtained From Rajasthan, India.

Science.gov (United States)

Sharma, Sanjita; Sharma, Vishnu; Dahiya, Dinesh Kumar; Khan, Aarif; Mathur, Manisha; Sharma, Amit

2017-03-01

Listeriosis is a serious foodborne disease of a global concern, and can effectively be controlled by a continuous surveillance of the virulent and multidrug-resistant strains of Listeria monocytogenes. This study was planned to investigate prevalence of L. monocytogenes in bovine raw milk samples. A total of 457 raw milk samples collected from 15 major cities in Rajasthan, India, were analyzed for the presence of L. monocytogenes by using standard microbiological and molecular methods. Five of the 457 samples screen tested positive for L. monocytogenes. Multiplex serotyping showed that 3/5 strains belonged to serotype 4b followed by one strain each to 1/2a and to 1/2c. Further virulence potential assessment indicated that all strains possessed inlA and inlC internalins, and, in addition, two strains also possessed the gene for inlB. All strains were positive for Listeriolysin O (LLO) and showed phosphatidylinositol-specific phospholipase C (PI-PLC) activity on an in vitro agar medium with variations in production levels among the strains. A good correlation between the in vitro pathogenicity test and the chick embryo test was observed, as the strains showing higher LLO and PI-PLC activity were found to be lethal to fertilized chick embryos. All strains were resistant to the majority of antibiotics and were designated as multidrug-resistant strains. However, these strains were susceptible to 9 of the 22 tested antibiotics. The maximum zone of inhibition (mm) and acceptable minimum inhibitory concentration were observed with azithromycin, and thus it could be the first choice of a treatment. Overall, the presence of multidrug-resistant L. monocytogenes strains in the raw milk of Rajasthan region is an indicator of public health hazard and highlighting the need of consumer awareness in place and implementation of stricter food safety regulations at all levels of milk production.

Gastrointestinal parasitic infection in diverse species of domestic ruminants inhabiting tribal rural areas of southern Rajasthan, India.

Science.gov (United States)

Choubisa, S L; Jaroli, V J

2013-10-01

A total of 415 adult domesticated ruminants, 130 cattle (Bos taurus), 108 buffaloes (Bubalus bubalis), 94 goats (Capra hircus) and 83 sheep (Ovis aries) inhabiting tribal rural areas of southern Rajasthan, India were investigated for evidence of gastrointestinal protozoan and helminthic infections. In southern Rajasthan humid ecosystem is predominant and has number of perennial freshwater bodies. Fresh faecal samples of these animals were examined microscopically by direct wet smear with saline and 1 % Lugol's iodine and formalin ether concentration. Of these 296 (71.32 %) were found to be infected with different species of gastrointestinal parasites. The highest (93.84 %) prevalence of these parasitic infections was found in cattle followed by goats (82.97 %), sheep (55.42 %) and buffaloes (46.29 %). Except cattle no other ruminants revealed protozoan infection. A total 8 species of gastrointestinal parasites were encountered. Among these parasites Fasciola hepatica was the commonest (15.18 %) followed by Haemonchus contortus (11.32 %), Ancylostoma duodenale (10.36 %), Trichuris trichiura (9.15 %), Amphistome species (7.95 %), Moniezia expansa (6.98 %), Strongyloides stercoralis (4.57 %) and Balantidium coli (3.37 %). The prevalence rate of these parasitic infections also varied seasonally. The highest prevalence rate was found in rainy season (84.21 %) followed by winter (73.9 %) and summer (52.8 %). The possible causes for variation in prevalence of parasitic infections are also discussed.

Oral health knowledge, attitude and practices of children and adolescents of orphanages in jodhpur city rajasthan, India.

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Hans, Rinki; Thomas, Susan; Dagli, Rushabh; Bhateja, Geetika Arora; Sharma, Akanksha; Singh, Amarpreet

2014-10-01

This study had twin objectives of assessing the oral health knowledge, attitude and practices and to assess the dental caries status and treatment needs among the orphan children of orphanages of Jodhpur city, Rajasthan, India. This cross- sectional study was carried out on 100 children to assess the oral health knowledge, attitude and practices of children and adolescents of orphanages in Jodhpur city, Rajasthan, India. The data was collected on a pre-tested questionnaire which included 20 closed ended multiple-choice questions on perceived oral health status, knowledge of oral health and attitude, oral health practices, dietary habits and behaviour towards dental treatment. On completion of the questionnaire, each child underwent an oral examination and Dentition status and treatment needs index (WHO Oral Health Surveys- 1997) was recorded for each subject. Almost 93% of the children felt the necessity of maintaining oral hygiene. There were 69% of the children who believed that it was necessary to brush teeth after every meal, 51% children believed that regular tooth-brushing prevents all tooth problems and 93% children knew that tobacco is carcinogenic in nature. Also, it was found that 77% of the children believed that regular dental visits help in maintaining oral hygiene. Many of them had acquired knowledge on oral health. More than half of the study subjects were aware of the importance of keeping good oral hygiene, regular dental visits and harmful effects of tobacco.

Nuclear research reactor 0.5 to 3 MW

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-05-15

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor

Nuclear research reactor 0.5 to 3 MW

International Nuclear Information System (INIS)

1992-05-01

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation

Techno-economic felling cycles for selected energy plantation species in the arid areas of western Rajasthan

Energy Technology Data Exchange (ETDEWEB)

Kalla, J.C.; Gyan, C.; Vyas, D.L.; Gehlot, N.S.

1978-01-01

Analysis of data collected from research plots at Pali, Rajasthan, predicted that the maximum fuel yields per tree would be reached at rotations of 50 years for (a) Azadirachta indica (51 kg), 25 years for (b) Acacia tortilis (39 kg), 14 years for (c) Albizzia (Abizia) lebbek (15 kg) and 13 years for (d) Acacia nilotica (6 kg). The best economic rotations would be 23 years for (a) yielding 24 kg/tree, 8 years for (b) yielding 20 kg/tree and 11 years for (c) yielding 6 kg/tree. Harvesting (d) for fuel would not be economic.

The tribal girl child in Rajasthan.

Science.gov (United States)

Bhanti, R

1995-01-01

This article describes the status of the girl child among tribes in India. Tribes have son preference but do not discriminate against girls by female infanticide or sex determination tests. Girls do not inherit land, but they are not abused, hated, or subjected to rigid social norms. Girls are not veiled and are free to participate in dancing and other recreational programs. There is no dowry on marriage. The father of the bridegroom pays a brideprice to the father of the girl. Widowed or divorced women are free to marry again. Daughters care for young children, perform housework, and work in the field with their brothers. In the tribal village of Choti Underi girls were not discriminated against in health and nutrition, but there was a gender gap in education. Both girls and boys were equally exposed to infection and undernourishment. Tribals experience high rates of infant and child mortality due to poverty and its related malnutrition. Child labor among tribals is a way of life for meeting the basic needs of the total household. A recent report on tribals in Rajasthan reveals that 15-20% of child labor involved work in mines that were dangerous to children's health. Girl children had no security provisions or minimum wages. Tribal children were exploited by human service agencies. Child laborers were raped. Government programs in tribal areas should focus on improving living conditions for children in general. Special programs for girls are needed for providing security in the workplace and increasing female educational levels. More information is needed on the work burden of tribal girls that may include wage employment as well as housework.

The heating operational summarization in three winters of a 5 MW test heating reactor

International Nuclear Information System (INIS)

Wang Dazhong; Dong Duo; Su Qingshan; Zhang Yajun

1992-09-01

The 5 MW THR (5 MW test heating reactor) is a new type reactor with inherent safety developed by INET (Institute of Nuclear Energy Technology). It is the first 'pressure vessel type' heating reactor in operation in the world. It was put into operation in November, 1989. Since then it has operated for three winter seasons. The total operation time has reached to 8174 hours and its availability of heating has reached to 99%. The advanced technology of this reactor has been proved in the past three years operation. The characteristics of power regulating, load following, reactivity disturbance and the variation of parameters under the condition of ATWS (anticipated transients without scram) were studied with experiments in 5 MW THR. The 5 MW THR is an ideal heating reactor and has outstanding performances

Rb-Sr and Ar-Ar systematics of Malani volcanic rocks of southwest Rajasthan: evidence for a younger post-crystallization thermal event

International Nuclear Information System (INIS)

Rathore, S.S.; Srivastava, R.K.

1996-01-01

A new Rb-Sr age of 779 ± 10 Ma has been obtained for a suite of andesite-dacite-rhyolite from the Malani igneous province of southwestern Rajasthan, dated earlier at 745 ± 10 Ma by Crawford and Compston (1970). The associated basalts may be slightly younger than the felsic volcanics and have a mantle source. The felsic volcanics on the other hand were most probably derived by fractional crystallization of a crustal magma. 40 Ar- 39 Ar systematics of three samples viz., a basalt, a dacite and a rhyolite show disturbed age spectra indicating a thermal event around 500-550 Ma ago. This secondary thermal event is quite wide-spread and possibly related to the Pan-African thermo-tectonic episode observed in the Himalayas and south India. (author). 38 refs., 5 figs., 2 tabs

Validation study of the reactor physics lattice transport code WIMSD-5B by TRX and BAPL critical experiments of light water reactors

International Nuclear Information System (INIS)

Khan, M.J.H.; Alam, A.B.M.K.; Ahsan, M.H.; Mamun, K.A.A.; Islam, S.M.A.

2015-01-01

Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to

Optimal operating parameters of the reactor oscillator in the channel of 6.5/10 MW reactor; Odredjivanje optimalnih radnih tacaka za reaktorski oscilator u kanalu na reaktoru 6,5/10 MW

Energy Technology Data Exchange (ETDEWEB)

Lolic, B; Zecevic, V [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1961-07-01

Operating conditions for the reactor oscillator in the central vertical experimental channel (VK5) in the RA reactor were studied during 1960. Channel VK5 was chosen because the sensitivity of the reactor is highest in this case. The central vertical experimental channel is placed in the center of the core and its bottom is placed 200 mm from the bottom of the reactor core. Diameter of the channel is is 110 mm and its length 5706 mm. During operation of the reactor oscillator with total modulation of the reactor power, it is very important to determine the oscillator operating point and the oscillation amplitude in such a way to avoid any change in reactor power level. Positive reactivity changes originating from oscillations of the samples should be compensated by the negative reactivity changes so that the effect should be nil. Operating points of the reactor oscillator are in the middle of the straight part of the figure showing the reactivity change dependent on the position of the absorber.

Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model

International Nuclear Information System (INIS)

Griggs, D.P.

1993-03-01

The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A six-loop RELAP5 K Reactor model is used to analyze the reactor system behavior dozing the Flow Instability (FI) phase of the LOCA, which comprises only the first 5 seconds following the DEGB. The RELAP5 K Reactor model includes tank and plenum nodalizations having five radial rings and six azimuthal sectors. The reactor system analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code, which models a single fuel assembly in detail. RELAP5 also performs the system analysis for the latter phase of the LOCA, denoted the Emergency Cooling System (ECS) phase. Results from the RELAP analysis are used to provide boundary conditions to the FLOWTRAN-TF code, which is an advanced two-phase version of FLOWTRAN. The RELAP5 K Reactor model has been tested for LOCA-FI and Loss-of-Pumping Accident analyses and the results compared with equivalent analyses performed with the TRAC-PF1/MOD1 code (TRAC). An equivalent RELAP5 six-loop, five-ring, six-sector L Reactor model has been benchmarked against qualified single-phase system data from the 1989 L-Area In-Reactor Test Program. The RELAP5 K and L Reactor models have also been subjected to an independent Quality Assurance verification

Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

International Nuclear Information System (INIS)

Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

2010-01-01

In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

Case study of fly ash brick manufacturing units at Kota in Rajasthan

Science.gov (United States)

Sharma, Y.; Saxena, B. K.; Rao, K. V. S.

2018-03-01

Kota Super Thermal Power Station of 1240 MW is located at Kota in Rajasthan, India. The quantity of fly ash generated by it is about 1.64 to 2.03 million tonnes per year. This fly ash is being utilized for making bricks, tiles, portland pozzolana cement, construction of highways, and other purposes. 1.79 million tonnes of fly ash was utilized for different applications in one year duration from April 01st, 2015 to March 31st, 2016. Out of this total utilization, 0.6439 million tonnes (36.06 %) of fly ash was used for making bricks, blocks, and tiles. In this paper, a case study of two fly ash brick manufacturing units using fly ash produced from Kota Super Thermal Power Station is described. These units produce about 15,000 and 20,000 bricks respectively by employing 10 and 16 workers each and are making a profit of about Rs. 6,000 and Rs. 8,000 per day in one shift.

Forecasting incidence of dengue in Rajasthan, using time series analyses.

Science.gov (United States)

Bhatnagar, Sunil; Lal, Vivek; Gupta, Shiv D; Gupta, Om P

2012-01-01

To develop a prediction model for dengue fever/dengue haemorrhagic fever (DF/DHF) using time series data over the past decade in Rajasthan and to forecast monthly DF/DHF incidence for 2011. Seasonal autoregressive integrated moving average (SARIMA) model was used for statistical modeling. During January 2001 to December 2010, the reported DF/DHF cases showed a cyclical pattern with seasonal variation. SARIMA (0,0,1) (0,1,1) 12 model had the lowest normalized Bayesian information criteria (BIC) of 9.426 and mean absolute percentage error (MAPE) of 263.361 and appeared to be the best model. The proportion of variance explained by the model was 54.3%. Adequacy of the model was established through Ljung-Box test (Q statistic 4.910 and P-value 0.996), which showed no significant correlation between residuals at different lag times. The forecast for the year 2011 showed a seasonal peak in the month of October with an estimated 546 cases. Application of SARIMA model may be useful for forecast of cases and impending outbreaks of DF/DHF and other infectious diseases, which exhibit seasonal pattern.

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

International Nuclear Information System (INIS)

Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

Robotic dismantlement systems at the CP-5 reactor D and D project

International Nuclear Information System (INIS)

Seifert, L. S.

1998-01-01

The Chicago Pile 5 (CP-5) Research Reactor Facility is currently undergoing decontamination and decommissioning (D and D) at the Argonne National Laboratory (ANL) Illinois site. CP-5 was the principle nuclear reactor used to produce neutrons for scientific research at Argonne from 1954 to 1979. The CP-5 reactor was a heavy-water cooled and moderated, enriched uranium-fueled reactor with a graphite reflector. The CP-5 D and D project includes the disassembly, segmentation and removal of all the radioactive components, equipment and structures associated with the CP-5 facility. The Department of Energy's Robotics Technology Development Program and the Federal Energy Technology Center, Morgantown Office provided teleoperated, remote systems for use in the dismantlement of the CP-5 reactor assembly for tasks requiring remote dismantlement as part of the EM-50 Large-Scale Demonstration Program (LSDP). The teleoperated systems provided were the Dual Arm Work Platform (DAWP), the Rosie Mobile Teleoperated Robot Work System (ROSIE), and a remotely-operated crane control system with installed swing-reduction control system. Another remotely operated apparatus, a Brokk BM250, was loaned to ANL by the Princeton Plasma Physics Laboratory (PPPL). This machine is not teleoperated and was not part of the LSDP, but deserves some mention in this discussion. The DAWP is a robotic dismantlement system that includes a pair of Schilling Robotic Systems Titan III hydraulic manipulator arms mounted to a specially designed support platform: a hydraulic power unit (HPU) and a remote operator console. The DAWP is designed to be crane-suspended for remote positioning. ROSIE, developed by RedZone Robotics, Inc. is a mobile, electro-hydraulic, omnidirectional platform with a heavy-duty telescoping boom mounted to the platform's deck. The work system includes the mobile platform (locomotor), a power distribution unit (PDU) and a remote operator console. ROSIE moves about the reactor building

Three-dimensional numerical investigation of a Molten Salt reactor concept with the code CFX-5.5

International Nuclear Information System (INIS)

Yamaji, B.; Csom, G.; Aszodi, A.

2002-01-01

Partitioning and transmutation of actinides and long-lived fission products is a promising option to extend the possibilities and enhance the environmentally acceptable capabilities of nuclear energy. Also the possible implementation of the thorium cycle is considered as a way to reduce the problem of energy resources in the future. For both objectives different molten salt reactor concepts were proposed mainly based on the Molten Salt Reactor Experiment of the Oak Ridge National Laboratory. Not only critical reactors but also accelerator-driven subcritical systems (ADSs) have advantages worth considering for those aims, especially those ones with liquid fuel, such as molten salts. By using liquid fuel which is the coolant medium, too, a basically different thermalhydraulic behavior is expected than in the case of solid fuel and water coolant. In this work our purpose is to present the possible use of Computational Fluid Dynamics (CFD) technology in molten salt thermal hydraulics. The simulations were performed with the three-dimensional code CFX-5.5.(author)

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

International Nuclear Information System (INIS)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

International Nuclear Information System (INIS)

Bowman, S.M.; Suto, T.

1996-10-01

ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k eff of 1. 0040±0.0005

Project and characteristics of a 5MW experimental fast reactor

International Nuclear Information System (INIS)

Ishiguro, Y.; Nascimento, J.A. do.

1986-05-01

Characteristics of a 5 MW experimental fast reactor are reported. The reactor is designed with emphasis on fuel and materials irradiation and uses fuel assemblies of a standard structure. The reference core consist of 37 fuel assemblies, each of which contains 19 pins of metallic Pu/Zr fuel. With a core height of 17.6 cm the core volume is 11.4 liter and the central fast (E >=100 KeV) flux is 0.9 x 10 15 n/cm 2 sec. In addition to twelve control rod assemblies with a total reactivity worth of 5.5% Δk, 42 assemblies for reactivity compensation are placed in the two rings outside the core. Replacing these assemblies with driver, blanket, or refletor-shield assemblies, large reactivities can be added to make the central assembly position available for test irradiations and to assure high levels of burnup of driver assemblies. (Author) [pt

Traditional knowledge on zootherapeutic uses by the Saharia tribe of Rajasthan, India.

Science.gov (United States)

Mahawar, Madan Mohan; Jaroli, D P

2007-06-05

The present zootherapeutic study describes the traditional knowledge related to the use of different animals and animal-derived products as medicines by the Saharia tribe reside in the Shahabad and Kishanganj Panchayat Samiti's of Baran district of Rajasthan, India. A field survey was conducted from April to June 2006 by performing interview through structured questionnaire with 21 selected respondents, who provided information regarding use of animals and their products in folk medicine. A total of 15 animal species were recorded and they are used for different ethnomedical purposes, including cough, asthma, tuberculosis, paralysis, earache, herpes, weakness, muscular pain etc. The zootherapeutic knowledge was mostly based on domestic animals, but some protected species like the peacock (Pavo cristatus,), hard shelled turtle (Kachuga tentoria), sambhar (Cervus unicolor) were also mentioned as medicinal resources. We would suggest that this kind of neglected traditional knowledge should be included into the strategies of conservation and management of faunistic resources. Further studies are required for experimental validation to confirm the presence of bioactive compounds in these traditional remedies and also to emphasize more sustainable use of these resources.

Optimization of 10 kW solar photovoltaic – diesel generator hybrid energy system for different load factors at Jaisalmer location of Rajasthan, India

Science.gov (United States)

Saraswat, S. K.; Rao, K. V. S.

2018-03-01

Jaisalmer town in Rajasthan, India is having annual average solar insolation of 5.80 kWh/m2/day and 270 – 300 clear sky days in a year. A 10 kW off-grid hybrid energy system (HES) consisting of solar photovoltaic panels – diesel generator – bidirectional converter and batteries with zero percentage loss of load for Jaisalmer is designed using HOMER (version 3.4.3) software. Different system load factors of 0.33, 0.50, 0.67, 0.83 and 1 corresponding to fraction of running hours per day of the system are considered. The system is analyzed for all three aspects, namely, electrical, economic and emission point of view. Least levelized cost of electricity (LCOE) of Rs. 8.43/kWh is obtained at a load factor value of 0.5. If diesel generator alone (without Solar PV) is used to fulfil the demand for a load factor of 0.5the value of LCOE is obtained Rs.19.23/kWh. Comparison of results obtained for HES and diesel generator are made for load factor of 0.5 and 1.

Assessment of age-dependent radiation dose due to intake of uranium and thorium in drinking water from Sikar District, Rajasthan, India

International Nuclear Information System (INIS)

Duggal, Vikas; Rani, Asha; Balaram, V.

2016-01-01

The concentrations of 238 U and 232 Th have been determined in drinking water samples collected from the Sikar district of Rajasthan State, India. The samples have been analysed by using high-resolution inductively coupled plasma mass spectrometry. 238 U content in water samples ranged from 8.20 to 202.63 μg l -1 and 232 Th content ranged from 0.57 to 1.46 μg l -1 . The measured 238 U content in 25 % of the analysed samples exceeded the World Health Organization (WHO) and United States Environmental Protection Agency drinking water guidelines of 30 μg l -1 and 12.5 % of the samples exceeded the 60 μg l -1 Indian maximum acceptable concentration recommended by the Atomic Energy Regulatory Board, India. The annual effective doses (μSv y -1 ) due to ingestion of 238 U and 232 Th for different age groups were also calculated. The results compared with the recommended value reported by the WHO. (authors)

Determination of palladium content in palladium-alumina/palladium-silica/palladium-tin oxide catalyst for nuclear reactor applications

International Nuclear Information System (INIS)

Sharma, P.K.; Bassan, M.K.T.; Avhad, D.K.; Singhal, R.K.

2012-01-01

Alumina and silica act as support for finely divided palladium metal powder in synthesis of catalyst. These catalyst (Pd-Al 2 O 3 , Pd-SiO 2 and Pd-SnO 2 ) used in nuclear power reactor (moderator cover gas system) for the conversion of hydrogen. In Indian nuclear power programme these catalyst are regularly used in Kaiga 1 and 2 and Rajasthan atomic power plant 3 and 4. The performance of the catalyst, solely depends on the concentration of palladium, which is the active component in this catalyst composition. Therefore it is highly desirable to have rouged analytical methodology for the accurate estimation of palladium. Leaching of Pd from the bulk matrix is tedious due to the less reactive nature of Pd therefore complete solubilization of the matrix is carried out by fusion method

Effluents and releases of tritium from Novo-Voronezh-5 reactor

International Nuclear Information System (INIS)

Babenko, A.G.; Mekhedov, B.N.; Podporinova, L.E.; Popov, S.V.; Shalin, A.N.

1990-01-01

Results of systematic measurements of tritium concentration within technological systems of reactor of Novo-Voronezh NPP conducted to evaluate tritium effluents and releases and radiation doses to population from these effluents and releases are given. It is shown that 68% concerning tritium total amount were disposed into sewerage while 17% - through vent tube and 15% - with water and steam from secondary circuit systems. Standartized tritium effluents from WWER-1000 reactor for 5 year run constitute 15±1.9 GBq/MWxyear and it corresponds to mean value of effluents for foreign NPPs. Tritium concentration in the atmosphere constituted according to calculations (4.1-20)x10 -5 Bq/l. Conclusion is made about insignificant dose to population from tritium gaseous effluents. Detail study is necessary for dose connected with tritium contained in water effluents

Source and Assessment of Metal Pollution at Khetri Copper Mine Tailings and Neighboring Soils, Rajasthan, India.

Science.gov (United States)

Punia, Anita; Siddaiah, N Siva; Singh, Saurabh K

2017-11-01

We present here the results of the study on metal pollution by identifying source, abundance and distribution in soil and tailings of Khetri copper complex (KCC) mines, Rajasthan India. The region is highly contaminated by copper (Cu) with higher values in the soil near overburden material (1224 mg/kg) and tailings (111 mg/kg). The average Cu (231 mg/kg) concentration of soil is ~9, 5 and 32 times higher than upper crust, world average shale (WAS) and local background soil (LS), respectively. However this reaches to ~82, 46 and 280 times higher in case of tailing when compared. The correlation and principal component analysis for soil reveals that the source of Cu, Zn, Co, Ni, Mn and Fe is mining and Pb and Cd could be result of weathering of parent rocks and other anthropogenic activities. The source for Cr in soil is both mining activities and weathering of parent rocks. The values of index of geo-accumulation (I geo ) and pollution load index for soil using LS as background are higher compared to values calculated using WAS. The metal rich sulphide bearing overburden material as well as tailings present in the open environment at KCC mines region warrants a proper management to minimize their impact on the environment.

Assessment of Fluoride Concentration of Soil and Vegetables in Vicinity of Zinc Smelter, Debari, Udaipur, Rajasthan.

Science.gov (United States)

Bhat, Nagesh; Jain, Sandeep; Asawa, Kailash; Tak, Mridula; Shinde, Kushal; Singh, Anukriti; Gandhi, Neha; Gupta, Vivek Vardhan

2015-10-01

As of late, natural contamination has stimulated as a reaction of mechanical and other human exercises. In India, with the expanding industrialization, numerous unsafe substances are utilized or are discharged amid generation as cleans, exhaust, vapours and gasses. These substances at last are blended in the earth and causes health hazards. To determine concentration of fluoride in soils and vegetables grown in the vicinity of Zinc Smelter, Debari, Udaipur, Rajasthan. Samples of vegetables and soil were collected from areas situated at 0, 1, 2, 5, and 10 km distance from the zinc smelter, Debari. Three samples of vegetables (i.e. Cabbage, Onion and Tomato) and 3 samples of soil {one sample from the upper layer of soil (i.e. 0 to 20 cm) and one from the deep layer (i.e. 20 - 40 cm)} at each distance were collected. The soil and vegetable samples were sealed in clean polythene bags and transported to the laboratory for analysis. One sample each of water and fertilizer from each distance were also collected. The mean fluoride concentration in the vegetables grown varied between 0.36 ± 0.69 to 0.71 ± 0.90 ppm. The fluoride concentration in fertilizer and water sample from various distances was found to be in the range of 1.4 - 1.5 ppm and 1.8 - 1.9 ppm respectively. The fluoride content of soil and vegetables was found to be higher in places near to the zinc smelter.

Hydrogen production using Rhodopseudomonas palustris WP 3-5 with hydrogen fermentation reactor effluent

International Nuclear Information System (INIS)

Chi-Mei Lee; Kuo-Tsang Hung

2006-01-01

The possibility of utilizing the dark hydrogen fermentation stage effluents for photo hydrogen production using purple non-sulfur bacteria should be elucidated. In the previous experiments, Rhodopseudomonas palustris WP3-5 was proven to efficiently produce hydrogen from the effluent of hydrogen fermentation reactors. The highest hydrogen production rate was obtained at a HRT value of 48 h when feeding a 5 fold effluent dilution from anaerobic hydrogen fermentation. Besides, hydrogen production occurred only when the NH 4 + concentration was below 17 mg-NH 4 + /l. Therefore, for successful fermentation effluent utilization, the most important things were to decrease the optimal HRT, increase the optimal substrate concentration and increase the tolerable ammonia concentration. In this study, a lab-scale serial photo-bioreactor was constructed. The reactor overall hydrogen production efficiency with synthetic wastewater exhibiting an organic acid profile identical to that of anaerobic hydrogen fermentation reactor effluent and with effluent from two anaerobic hydrogen fermentation reactors was evaluated. (authors)

Relap5 simulation for severe accident analysis of RSG-GAS Reactor

International Nuclear Information System (INIS)

Andi Sofrany Ekariansyah; Endiah P-Hastuti; Sudarmono

2018-01-01

The research reactor in the world is to be known safer than power reactor due to its simpler design related to the core and operational characteristics. Nevertheless, potential hazards of research reactor to the public and the environment can not be ignored due to several special features. Therefore the level of safety must be clearly demonstrated in the safety analysis report (SAR) using safety analysis, which is performed with various approaches and methods supported by computational tools. The purpose of this research is to simulate several accidents in the Indonesia RSG-GAS reactor, which may lead to the fuel damage, to complement the severe accident analysis results that already described in the SAR. The simulation were performed using the thermal hydraulic code of RELAP5/SCDAP/Mod3.4 which has the capability to model the plate-type of RSG-GAS fuel elements. Three events were simulated, which are loss of primary and secondary flow without reactor trip, blockage of core subchannels without reactor trip during full power, and loss of primary and secondary flow followed by reactor trip and blockage of core subchannel. The first event will harm the fuel plate cladding as showed by its melting temperature of 590 °C. The blockage of one or more subchannels in the one fuel element results in different consequences to the fuel plates, in which at least two blocked subchannels will damage one fuel plate, even more the blockage of one fuel element. The combination of loss of primary and secondary flow followed by reactor trip and blockage of one fuel element has provided an increase of fuel plate temperature below its melting point meaning that the established natural circulation and the relative low reactor power is sufficient to cool the fuel element. (author)

Crystallochemical studies on davidite from Bichun, Jaipur District, Rajasthan, India

Science.gov (United States)

Singh, Yamuna; Saxena, Anubhooti; Bhatt, A. K.; Viswanathan, R.; Shaji, T. S.; Nanda, L. K.

2018-02-01

Crystallochemical data on metamict davidite from albitites and albitised rocks from the Bichun area (Jaipur district, Rajasthan, India) of Banded Gneissic Complex (BGC) are provided. Davidite occurs as euhedral, subhedral to anhedral crystals in the form of disseminated grains and also as fracture filled veins. The crystals of davidite are up to 8 cm in length and 6 cm in width. The powder X-ray diffraction (XRD) pattern of the heat-treated davidite (at 900{°}C) reveals well-defined reflections of crystallographic planes. The calculated unit-cell parameters of the heat treated davidite are: a0 = b0 = 10.3556 Å and c0 = 20.9067 Å, with unit-cell volume (V) = 1941.6385 Å3; and α=β= 90° and γ= 120°, which are in agreement with the values of davidite standard. Geochemical data reveals that the investigated davidite contains 51.5-52.6% TiO2, 14.8-15.1% Fe2 O3, 9.8-10.2% FeO, 6.97-7.12% U3 O8, 6.72-6.92% RE2 O3, 3.85-3.61% K2O, 0.9-1.4% Al2 O3, and 0.8-1.2% SiO2. The calculated structural formulae of the two davidite crystals are: D-1: K_{0.0044/0.004} Ba_{0.0044/0.005} Ca_{0.20/0.20} Na_{0.012/0.012} Mn_{0.053/0.053} Mg_{0.14/0.14} Pb_{0.0076/0.008} Fe_{2.675/2.675} Fe_{1.59/1.59} Y_{0.1175/0.118} P_{0.053/0.053} Nb_{0.008/0.008} Sn_{0.001/0.001} Zr_{0.033/0.033} U_{0.468/0.468} Th_{0.009/0.009} REE_{0.6829/0.683})_{6.05/6.05} (Ti_{12.15/12.15} Fe_{1.9022/1.903} Si_{0.372/0.372} Al_{0.517/0.517} Cr_{0.018/0.018} Co_{0.009/0.009} Ni_{0.027/0.027})_{15/15} O_{36/36} (OH_{0.319/0.319[]1.681/1.681})_{2/2} and D-2: (K_{0.004/0.004} Ba_{0.005/0.005} Ca_{0.20/0.20} Na_{0.012/0.012} Mn_{0.05/0.05} Mg_{0.094/0.094} Pb_{0.007/0.007} Fe_{2.58/2.58} Fe_{1.71/1.71} Y_{0.112/0.112} P_{0.106/0.106} Nb_{0.006/0.006} Sn_{0.001/0.001} Zr_{0.03/0.03} U_{0.48/0.48} Th_{0.009/0.009} REE_{0.665/0.665})_{6.088/6.088} (Ti_{12.48/12.48} Fe_{1.87/1.87} Si_{0.249/0.249} Al_{0.334/0.334} Cr_{0.019/0.019} Co_{0.008/0.008} Ni_{0.04/0.04})_{15/15} O_{36/36} (OH_{0

Dental anomalies of the deciduous dentition among Indian children: a survey from Jodhpur, Rajasthan, India.

Science.gov (United States)

Deolia, Shravani Govind; Chhabra, Chaya; Chhabra, Kumar Gaurav; Kalghatgi, Shrivardhan; Khandelwal, Naresh

2015-01-01

Anomalies and enamel hypoplasia of deciduous dentition are routinely encountered by dental professionals and early detection and careful management of such conditions facilitates may help in customary occlusal development. The aim of this study was to determine the prevalence of hypodontia, microdontia, double teeth, and hyperdontia of deciduous teeth among Indian children. The study group comprised 1,398 children (735 boys, 633 girls). The children were examined in department of Pedodontics and Preventive Dentistry in Jodhpur Dental College General Hospital, Jodhpur, Rajasthan, India. Clinical data were collected by single dentist according to Kreiborg criteria, which includes double teeth, hypodontia, microdontia, and supernumerary teeth. Statistical analysis of the data was performed using the descriptive analysis and chi-square test. Dental anomalies were found in 4% of children. The distribution of dental anomalies were significantly more frequent (P = 0.001) in girls (5.8%, n = 38) than in boys (2.7%, n = 18). In relation to anomaly frequencies at different ages, significant difference was found between 2 and 3 years (P = 0.001). Double teeth were the most frequently (2.3%) observed anomaly. The other anomalies followed as 0.3% supernumerary teeth, 0.6% microdontia, 0.6% hypodontia. Identification of dental anomalies at an early age is of great importance as it prevents malocclusions, functional and certain psychological problems.

Dental anomalies of the deciduous dentition among Indian children: A survey from Jodhpur, Rajasthan, India

Directory of Open Access Journals (Sweden)

Shravani Govind Deolia

2015-01-01

Full Text Available Background: Anomalies and enamel hypoplasia of deciduous dentition are routinely encountered by dental professionals and early detection and careful management of such conditions facilitates may help in customary occlusal development. Objective: The aim of this study was to determine the prevalence of hypodontia, microdontia, double teeth, and hyperdontia of deciduous teeth among Indian children. Materials and Methods: The study group comprised 1,398 children (735 boys, 633 girls. The children were examined in department of Pedodontics and Preventive Dentistry in Jodhpur Dental College General Hospital, Jodhpur, Rajasthan, India. Clinical data were collected by single dentist according to Kreiborg criteria, which includes double teeth, hypodontia, microdontia, and supernumerary teeth. Statistical analysis of the data was performed using the descriptive analysis and chi-square test. Results: Dental anomalies were found in 4% of children. The distribution of dental anomalies were significantly more frequent (P = 0.001 in girls (5.8%, n = 38 than in boys (2.7%, n = 18. In relation to anomaly frequencies at different ages, significant difference was found between 2 and 3 years (P = 0.001. Conclusion: Double teeth were the most frequently (2.3% observed anomaly. The other anomalies followed as 0.3% supernumerary teeth, 0.6% microdontia, 0.6% hypodontia. Identification of dental anomalies at an early age is of great importance as it prevents malocclusions, functional and certain psychological problems.

Anomalous uranium concentration in Archaean basement Shear at Dhani Basri and its significance on Southern Margin of Alwar sub-basin, Rajasthan

International Nuclear Information System (INIS)

Panigrahi, B.; Shaji, T.S.; Sharma, G.S.; Yadav, O.P.; Nanda, L.K.

2008-01-01

Prominent shear zones cutting through the basement and cover rocks of Delhi Supergroup have been recognized in Dhani Basri - Ramewala sector of Dausa district, Rajasthan. One such shear zone traversing the granite gneiss (Archaean basement) has been observed at Dhani Basri. The sheared rock is exposed in the form of a small hump and gives appearance of quartzite due to intense silicification. Grab samples collected from the shear zone rock analysed upto 93 ppm U 3 O 8 and <10 ppm ThO 2 , which is anomalous in comparison to unsheared rock which analysed 51 ppm eU 3 O 8 , upto 5 ppm U 3 O 8 and 80 ppm ThO 2 . Gamma-ray logging of boreholes drilled by GSI across this shear zone indicated uranium mineralization of the order of 0.030% eU 3 O 8 x 5.40 m and the primary radioactive mineral has been identified as uraninite. The extension of Dhani Basri shear zone inside the cover rocks of Meso-Proterozoic Delhi Supergroup of rocks of Alwar sub-basin is of paramount importance in locating unconformity related as well as hydrothermal vein type uranium mineralization. (author)

Experimental temperature analysis of simple & hybrid earth air tunnel heat exchanger in series connection at Bikaner Rajasthan India

Science.gov (United States)

Jakhar, O. P.; Sharma, Chandra Shekhar; Kukana, Rajendra

2018-05-01

The Earth Air Tunnel Heat Exchanger System is a passive air-conditioning system which has no side effect on earth climate and produces better cooling effect and heating effect comfortable to human body. It produces heating effect in winter and cooling effect in summer with the minimum power consumption of energy as compare to other air-conditioning devices. In this research paper Temperature Analysis was done on the two systems of Earth Air Tunnel Heat Exchanger experimentally for summer cooling purpose. Both the system was installed at Mechanical Engineering Department Government Engineering College Bikaner Rajasthan India. Experimental results concludes that the Average Air Temperature Difference was found as 11.00° C and 16.27° C for the Simple and Hybrid Earth Air Tunnel Heat Exchanger in Series Connection System respectively. The Maximum Air Temperature Difference was found as 18.10° C and 23.70° C for the Simple and Hybrid Earth Air Tunnel Heat Exchanger in Series Connection System respectively. The Minimum Air Temperature Difference was found as 5.20° C and 11.70° C for the Simple and Hybrid Earth Air Tunnel Heat Exchanger in Series Connection System respectively.

Loss of coolant acident analyses on Osiris research reactor using the RELAP5 code

International Nuclear Information System (INIS)

Soares, Humberto Vitor; Costa, Antonella Lombardi; Lima, Claubia Pereira Bezerra; Veloso, Maria Auxiliadora Fortini

2011-01-01

RELAP5/MOD 3.3 code is widely used for thermal hydraulic studies of commercial nuclear power plants. However, several current investigations have shown that RELAP5 code can also be applied for thermal hydraulic analysis of nuclear research systems with good predictions. In this paper, a nodalization of the core and the most important components of the primary cooling system of the OSIRIS reactor developed for RELAP5 thermal hydraulic code are presented as well as results of steady state and transient simulations. OSIRIS has thermal power of 70 MW and it is an open pool type research reactor moderated and cooled by water. The OSIRIS reactor characteristics have been used as a base for the development of a model for the Multipurpose Brazilian Reactor (RMB). The aim of the present work is to investigate the behavior of the core during a loss of coolant accident and the possible damage of the fuel elements due an inadequate heat removal. Although the core coolant reached the saturation point due the large break, the fuel element conditions were out of the damage zone. (author)

The AMPS 1.5 MW low-pressure compact reactor

International Nuclear Information System (INIS)

Hewitt, J.S.

1987-01-01

The 1.5-MWt reactor of the Autonomous Marine Power Source (AMPS) is designed to meet the unusual requirements of its first application. To provide for 100 kWe (net) on board self-sustaining manned submersible vehicles, the AMPS reactor must deliver safely, reliably and without direct operator surveillance, its thermal output to freon Rankine-cycle engines at thermodynamically useful temperatures. It must also conform to space and weight limits on the order of less than 50 cubic metres and 70 tonnes. The safety requirements are met by (i) limiting lifetime excess reactivity requirements by incorporation of burnable poison in the U-Zr-H fuel, (ii) maintaining nominal pressures in the light-water primary system at about 1 atmosphere, and (iii) maintaining a large volume of primary reserve coolant at temperature depressed relative to that of the circulating coolant. The latter averages 90 degrees celsius as it is pumped around loops that include the reactor core and the freon evaporators during normal operation. In the event of loss of pumped flow, the system defaults by intrinsic means to core cooling through natural convective exchange with the reserve coolant. In the post-shutdown situation, this passive cooling mode continues to operate regardless of vessel orientation and decay heat is safely dissipated to the sea. The design of the AMPS system, including the reactor, the freon engines, the control and monitoring system, the safety shut-down system and the power source container, are in advanced stages of design. (author)

The 5th surveillance testing for Kori unit 2 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-03-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

Traditional knowledge on zootherapeutic uses by the Saharia tribe of Rajasthan, India

Directory of Open Access Journals (Sweden)

Jaroli DP

2007-06-01

Full Text Available Abstract The present zootherapeutic study describes the traditional knowledge related to the use of different animals and animal-derived products as medicines by the Saharia tribe reside in the Shahabad and Kishanganj Panchayat Samiti's of Baran district of Rajasthan, India. A field survey was conducted from April to June 2006 by performing interview through structured questionnaire with 21 selected respondents, who provided information regarding use of animals and their products in folk medicine. A total of 15 animal species were recorded and they are used for different ethnomedical purposes, including cough, asthma, tuberculosis, paralysis, earache, herpes, weakness, muscular pain etc. The zootherapeutic knowledge was mostly based on domestic animals, but some protected species like the peacock (Pavo cristatus,, hard shelled turtle (Kachuga tentoria, sambhar (Cervus unicolor were also mentioned as medicinal resources. We would suggest that this kind of neglected traditional knowledge should be included into the strategies of conservation and management of faunistic resources. Further studies are required for experimental validation to confirm the presence of bioactive compounds in these traditional remedies and also to emphasize more sustainable use of these resources.

Age of initial cohort of dengue patients could explain the origin of disease outbreak in a setting: a case control study in Rajasthan, India.

Science.gov (United States)

Angel, Annette; Angel, Bennet; Yadav, Karuna; Sharma, Neha; Joshi, Vinod; Thanvi, Indu; Thanvi, Sharad

2017-06-01

Dengue fever (DF) and dengue hemorrhagic fever (DHF) is a public health problem with 390 million cases reported in world annually. In Rajasthan, DF with DHF is being reported for about two decades. For undertaking interventions into disease transmission, locating origin of transmission is very important. Present paper reports retrospective analysis of the hospital reported cases of dengue during the year 2013-2014 undertaken in Barmer, Rajasthan. To address task of investigating outbreak, detailed analysis of the data on serological test results (Mac-ELISA assay of NS1, IgG and IgM) performed by local hospital, Balotra was made. The domestic breeding containers were examined for the presence of larvae and adult forms of Aedes aegypti by visiting individual households as well as common places of human aggregation like schools and hospitals. The analysis showed that first dengue cases started from the lot of school going children and then followed by adults and finally during peak period of infection only children around 1-2 years got infected. The subsequent entomological investigations during the outbreak showed school as principal source of mosquito breeding. Present investigations highlight that schools (March to April) play the role of primary sites of disease transmission and should be preferred for undertaking vector control operations to prevent dengue transmission from getting aggravated.

Annual report of the Department of Atomic Energy, 1976-77

International Nuclear Information System (INIS)

1977-01-01

Research and development work in various research units, and activities and achievements of various public undertakings of the Department of Atomic Energy, India, during 1976-77 are reported. Construction of the 100 MW-thermal research reactor at Trombay and the Fast Breeder Test Reactor at Kalpakkam is in progress. Work on desalination, MHD and in seismology in continued. Report on performance of the Tarapur and Rajasthan Atomic Power Stations and progress of construction of the nuclear power stations at Kalpakkam and Narora is given. Fuelling machine carriage and shielding and plug assemblies for the second unit of the Rajasthan Atomic Power Station have been indigenously fabricated. A novel technique for prospecting nuclear minerals, termed as BARC-TEFUREX has been evolved and is being used successfully. The country-wide radiological protection programme covers 42,000 radiation workers in 2,280 institutions. (M.G.B.)

Application of RELAP5-3D code for thermal analysis of the ADS reactor core

International Nuclear Information System (INIS)

Fernandes, Gustavo Henrique Nazareno

2018-01-01

Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

Science.gov (United States)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

Directory of Open Access Journals (Sweden)

Khedr Ahmed

2005-01-01

Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

International Nuclear Information System (INIS)

Hussein, M.S; Lewis, B.J.; Bonin, H.W.

2013-01-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k eff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k eff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k eff calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S, E-mail: mohamed.hussein@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k{sub eff} calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k{sub eff} calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k{sub eff} calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

The 5th surveillance testing for Kori unit 1 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-08-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

2016-08-15

Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

International Nuclear Information System (INIS)

Shi, Chengbin; Cheng, Maosong; Liu, Guimin

2016-01-01

Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

Application of isotopes and chemistry in unsaturated zone in arid areas of Rajasthan, India

International Nuclear Information System (INIS)

Navada, S.V.; Nair, A.R.; Sinha, U.K.; Kulkarni, U.P.; Joseph, T.B.

2001-01-01

Isotopes and chemistry of unsaturated zone soil profiles in Saraf ki Dani, Sihania and Lakrasar site in Barmer district of western Rajasthan, and precipitation samples collected at Chohtan, the nearest meteorological station, have been used to study groundwater recharge and to characterise the behaviour of contaminants in the subsurface. Soil water was extracted by vacuum distillation and analysed for 2 H and 18 O, whereas elutriation methods were used for chemistry. Based on the observed chloride profiles, groundwater recharge rates are estimated to be ∼10 to 18 mm per annum for the above sites. δ 18 O profiles at Sihania and Lakrasar (1999), reveal depleted δ 18 O values observed at a depth of ∼2 m, possibly due to an episode of heavy rains during 1998. Depthwise distribution of of F - , SO 4 -2 and NO 3 - at Sihania site II (1999) and δ 18 O signatures confirm that evaporative concentration occurs during dry periods. Enrichment of NO 3 - in the profiles may be related to fixation of nitrogen by leguminous plants. (author)

PREVALENCE, CLINICAL PRESENTATION, DIAGNOSIS AND TREATMENT OF ACUTE PULMONARY OEDEMA IN SEVERE PREGNANCY-INDUCED HYPERTENSION AND ECLAMPSIA CASES IN TRIBAL POPULATION OF SOUTH RAJASTHAN

Directory of Open Access Journals (Sweden)

(Brig. Pradeep Kuma

2016-05-01

Full Text Available BACKGROUND Pulmonary oedema in severe pregnancy-induced hypertension is a life threatening complication with high maternal mortality, particularly in tribal population of South Rajasthan. METHODS Thirteen cases which occurred in the duration of two and half years were analysed through medical records and findings were recorded. RESULTS Maximum cases 10(76.92% were in less than 20 years of age. 12 (92.30% cases were nulliparous. Out of 13 cases of PIH, pulmonary oedema developed in 5 (38.46% cases of eclampsia and 8 (61.54% cases of severe pregnancy-induced hypertension. 10 (76.92%cases were 28 to 30 weeks of gestation and 3 (23.08% were 31 to 34 weeks of gestation. 8 (61.54% cases were severely anaemic. 12 (92.30% were unbooked cases. CONCLUSION Regular antenatal checkups, early diagnosis, prompt treatment of hypertension and pulmonary oedema and termination of pregnancy is required to prevent maternal death.

Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

Energy Technology Data Exchange (ETDEWEB)

Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

1997-07-01

Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

Factors associated with second trimester abortion in rural Maharashtra and Rajasthan, India.

Science.gov (United States)

Zavier, A J Francis; Jejeebhoy, Shireen; Kalyanwala, Shveta

2012-01-01

Many married women in India experience abortion in their second trimester of pregnancy. While there is an impression that second trimester abortions are now overwhelmingly used for sex selection, little is known about the extent to which second trimester abortions are indeed associated with son preference and sex selection motives, relative to other factors. Using data from a community-based study in rural Maharashtra and Rajasthan, research highlights the role of limited access in explaining second trimester abortion. While women with a single child who was a daughter were indeed more likely than other women to have terminated a pregnancy carrying a female foetus in the second trimester, more strikingly, exclusion from abortion-related decision-making, unsuccessful prior attempts to terminate the pregnancy, and distance from the facility in which their abortion was performed, were significantly associated with second trimester abortion, even after controlling for confounding factors. The study calls for greater efficiency in implementing the PCPNDT Act and addressing deep-rooted son preference. At the same time, findings that poverty and limited access to facilities are as, if not more, important drivers of second trimester abortion, highlight the need to meet commitments to ensure accessible abortion facilities for poor rural women.

Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

International Nuclear Information System (INIS)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

2013-01-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca, E-mail: lewis-b@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

HECTR [Hydrogen Event Containment Transient Response] Version 1.5N: A modification of HECTR Version 1.5 for application to N Reactor

International Nuclear Information System (INIS)

Camp, A.L.; Dingman, S.E.

1987-05-01

This report describes HECTR Version 1.5N, which is a special version of HECTR developed specifically for application to the N Reactor. HECTR is a fast-running, lumped-parameter containment analysis computer program that is most useful for performing parametric studies. The main purpose of HECTR is to analyze nuclear reactor accidents involving the transport and combustion of hydrogen, but HECTR can also function as an experiment analysis tool and can solve a limited set of other types of containment problems. Version 1.5N is a modification of Version 1.5 and includes changes to the spray actuation logic, and models for steam vents, vacuum breakers, and building cross-vents. Thus, all of the key features of the N Reactor confinement can be modeled. HECTR is designed for flexibility and provides for user control of many important parameters, if built-in correlations and default values are not desired

Safety of 5 MW district heating reactor (DHR) and hydraulic dynamic pressure drive control rods

International Nuclear Information System (INIS)

Wu Yuanqiang; Wang Dazhong

1991-11-01

The principles and movement characteristic of the hydraulic dynamic pressure drive for control rods in 5 MW district heating reactor are described with stress on analysis of its effects on reactor safety features. The drive is different from electric-magnetic drive for PWR or hydraulic drive for BWR. The drive cylinder is driven by dynamic pressure. In the new drive system, the reactor coolant (water) used as actuating medium is pressed by pump, then injected into a step cylinder which is set in the reactor core. The cylinder will move step by step by controlling flow, then the cylinder drives the neutron absorber and controls nuclear reaction. The drive is characterized by simplicity in structure, high reliability, inherent safety, reduction in reactor height, economy, etc

Transporting spent fuel and reactor waste in Sweden experience from 5 years of operation

International Nuclear Information System (INIS)

Dybeck, P.; Gustafsson, B.

1990-01-01

This paper reports that since the Final Repository for Reactor Waste, SFR, was taken into operation in 1988, the SKB sea transportation system is operating at full capacity by transporting spent fuel and now also reactor waste from the 12 Swedish reactors to CLAB and SFR. Transports from the National Research Center, Studsvik to the repository has recently also been integrated in the system. CLAB, the central intermediate storage for spent fuel, has been in operation since 1985. The SKB Sea Transportation System consists today of the purpose built ship M/s Sigyn, 10 transport casks for spent fuel, 2 casks for spent core components, 27 IP-2 shielded steel containers for reactor waste and 5 terminal vehicles. During an average year about 250 tonnes of spent fuel and 3 -- 4000 m 3 of reactor waste are transported to CLAB and SFR respectively, corresponding to around 30 sea voyages

RELAP5 kinetics model development for the Advanced Test Reactor

International Nuclear Information System (INIS)

Judd, J.L.; Terry, W.K.

1990-01-01

A point-kinetics model of the Advanced Test Reactor has been developed for the RELAP5 code. Reactivity feedback parameters were calculated by a three-dimensional analysis with the PDQ neutron diffusion code. Analyses of several hypothetical reactivity insertion events by the new model and two earlier models are discussed. 3 refs., 10 figs., 6 tabs

Evolution of the design of fuel handling control system in 220 MWe Indian PHWRs

International Nuclear Information System (INIS)

Dhruvanarayana, L.; Gupta, H.; Bharathkumar, M.

1996-01-01

Following two CANDU type reactors at Rajasthan (RAPS-1 and 2), three nuclear power stations, each of two units of 220 MWe has been in operation at Rajasthan (RAPS-1 and 2). Madras (MAPS-1 and 2). Narora (NAPS-1 and 2) and Kakrapar (KAPS-1 and 2). Two more stations, also of 220 MWe capacity, are under construction at Rajasthan (RAPP-3 and 4) and Kaiga (Kaiga-1 and 2). These are natural uranium fuelled pressurized heavy water cooled and heavy water moderated reactors (PHWRs). The two units at Rajasthan viz RAPS-1 and 2, were built with the technical collaboration with Canada, and the rest of the units have been designed and built indigenously, incorporating a number of modifications, particularly in the on-power refuelling system. The evolution of the design of the Fuel Handling Control systems of these reactors, taking into consideration operational needs, safety aspects and maintainability are highlighted in this paper. A combination of hydraulic and electronic control has been provided to enable the operations. In RAPS-1 and 2, hardwired electronic controls were provided, while in MAPS-1 and 2, the hardwired system was improved. From NAPS onwards, a computerized control system with hardwired interlock logic has been provided. New devices like coarse-fine potentiometers, special oil filled potentiometer assembly, rectilinear potentiometers etc., were specified from NAPS onwards. Positioning logic is computerized providing flexibility and expendability. Digital panel meters and indicating lamps have been provided for manual mode operations, while CRT (cathode-ray tube) monitors help in computer mode operations. Hydraulic controls which comprise D 2 0 hydraulics, H 2 0 hydraulics and oil hydraulics have been improved from NAPS onwards. Hydraulic panels have been relocated in accessible areas to reduce radiation doses and for better maintainability. All electric drives including X and Y drives were modified as hydraulic drives for better control. New types of valves

Evolution of the design of fuel handling control system in 220 MWe Indian PHWRs

Energy Technology Data Exchange (ETDEWEB)

Dhruvanarayana, L; Gupta, H; Bharathkumar, M [Nuclear Power Corporation of India Ltd., Mumbai (India)

1997-12-31

Following two CANDU type reactors at Rajasthan (RAPS-1 and 2), three nuclear power stations, each of two units of 220 MWe has been in operation at Rajasthan (RAPS-1 and 2). Madras (MAPS-1 and 2). Narora (NAPS-1 and 2) and Kakrapar (KAPS-1 and 2). Two more stations, also of 220 MWe capacity, are under construction at Rajasthan (RAPP-3 and 4) and Kaiga (Kaiga-1 and 2). These are natural uranium fuelled pressurized heavy water cooled and heavy water moderated reactors (PHWRs). The two units at Rajasthan viz RAPS-1 and 2, were built with the technical collaboration with Canada, and the rest of the units have been designed and built indigenously, incorporating a number of modifications, particularly in the on-power refuelling system. The evolution of the design of the Fuel Handling Control systems of these reactors, taking into consideration operational needs, safety aspects and maintainability are highlighted in this paper. A combination of hydraulic and electronic control has been provided to enable the operations. In RAPS-1 and 2, hardwired electronic controls were provided, while in MAPS-1 and 2, the hardwired system was improved. From NAPS onwards, a computerized control system with hardwired interlock logic has been provided. New devices like coarse-fine potentiometers, special oil filled potentiometer assembly, rectilinear potentiometers etc., were specified from NAPS onwards. Positioning logic is computerized providing flexibility and expendability. Digital panel meters and indicating lamps have been provided for manual mode operations, while CRT (cathode-ray tube) monitors help in computer mode operations. Hydraulic controls which comprise D{sub 2}0 hydraulics, H{sub 2}0 hydraulics and oil hydraulics have been improved from NAPS onwards. Hydraulic panels have been relocated in accessible areas to reduce radiation doses and for better maintainability. All electric drives including X and Y drives were modified as hydraulic drives for better control. New types of

RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

2003-04-01

The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

Operating reactors licensing actions summary. Volume 5, No. 2

International Nuclear Information System (INIS)

1985-04-01

The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the Operating Reactors Licensing Actions Program

Decentralizing Governance of Natural Resources in India: Lessons from the Case Study of Thanagazi Block, Alwar, Rajasthan, India - Comment

Directory of Open Access Journals (Sweden)

Maria Costanza Torri

2010-09-01

Full Text Available Numerous countries have undergone decentralisation reforms in the management of natural resources. However, the policies implemented are often not applied in ways compatible with the democratic potential with which decentralisation is conceived. The paper analyses the issue of decentralisation in resource management, in Thanagazi block, Alwar District, Rajasthan. In this paper I present a case of community initiated decentralisation carried out through village organisations. The aim is to contrast it with the state initiated decentralisation system carried out through the local administrative unit, the Gram Panchayat. Some conclusive remarks will be made on the importance of promoting more inclusive and democratic institutions which take into account the local needs and priorities regarding the management of natural resources and development interventions.

Operating reactors licensing actions summary. Volume 5, No. 6

International Nuclear Information System (INIS)

1985-08-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published for internal NRC use in managing the Operating Reactors Licensing Actions Program. Its content will change based on NRC management informational requirements

Decolourization of remazol black-5 textile dyes using moving bed bio-film reactor

Science.gov (United States)

Pratiwi, R.; Notodarmojo, S.; Helmy, Q.

2018-01-01

The desizing and dyeing processes in the textile industries produces wastewaster containing high concentration of organic matter and colour, so it needs treatment before released to environment. In this research, removal of azo dye (Remazol Black 5/RB 5) and organic as COD was performed using Moving Bed Biofilm Reactor (MBBR). MBBR is biological treatment process with attached growth media system that can increase removal of organic matter in textile wastewater. The effectiveness of ozonation as pre-treatment process to increase the removal efficiency in MBBR was studied. The results showed that in MBBR batch system with detention time of 1 hour, pre-treatment with ozonation prior to MBBR process able to increase the colour removal efficiency of up to 86.74%. While on the reactor without ozone pre-treatment, the colour removal efficiency of up to 68.6% was achieved. From the continuous reactor experiments found that both colour and COD removal efficiency depends on time detention of RB-5 dyes in the system. The higher of detention time, the higher of colour and COD removal efficiency. It was found that optimum removal of colour and COD was achieved in 24 hour detention time with its efficiency of 96.9% and 89.13%, respectively.

RELAP5 - a new tool for pressurized water reactor safety analysis

International Nuclear Information System (INIS)

Perneczky, L.

1988-11-01

The RELAP type pressurized water reactor safety system codes are used world wide for the loss of coolant accident analyses. In this paper the RELAP5, the advanced generation of the code family is presented. The relationship to RELAP4/mod6 version is discussed. The capability of the RELAP5/mod1-EUR version for small, medium and large break LOCA is investigated based on international user experience. (author) 30 refs

Adherence to evidence based care practices for childbirth before and after a quality improvement intervention in health facilities of Rajasthan, India.

Science.gov (United States)

Iyengar, Kirti; Jain, Motilal; Thomas, Sunil; Dashora, Kalpana; Liu, William; Saini, Paramsukh; Dattatreya, Rajesh; Parker, Indrani; Iyengar, Sharad

2014-08-13

After the launch of Janani Suraksha Yojana, a conditional cash transfer scheme in India, the proportion of women giving birth in institutions has rapidly increased. However, there are important gaps in quality of childbirth services during institutional deliveries. The aim of this intervention was to improve the quality of childbirth services in selected high caseload public health facilities of 10 districts of Rajasthan. This intervention titled "Parijaat" was designed by Action Research & Training for Health, in partnership with the state government and United Nations Population Fund. The intervention was carried out in 44 public health facilities in 10 districts of Rajasthan, India. These included district hospitals (9), community health centres (32) and primary health centres (3). The main intervention was orientation training of doctors and program managers and regular visits to facilities involving assessment, feedback, training and action. The adherence to evidence based practices before, during and after this intervention were measured using structured checklists and scoring sheets. Main outcome measures included changes in practices during labour, delivery or immediate postpartum period. Use of several unnecessary or harmful practices reduced significantly. Most importantly, proportion of facilities using routine augmentation of labour reduced (p = 0), episiotomy for primigravidas (p = 0.0003), fundal pressure (p = 0.0003), and routine suction of newborns (0 = 0.0005). Among the beneficial practices, use of oxytocin after delivery increased (p = 0.0001) and the practice of listening foetal heart sounds during labour (p = 0.0001). Some practices did not show any improvements, such as dorsal position for delivery, use of partograph, and hand-washing. An intervention based on repeated facility visits combined with actions at the level of decision makers can lead to substantial improvements in quality of childbirth practices at health facilities.

RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Debrecin, N.; Foretic, D.

2003-01-01

In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

Operating reactors licensing actions summary. Volume 5, No. 7

International Nuclear Information System (INIS)

1985-09-01

The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

International Nuclear Information System (INIS)

Rimpault, G.

2004-01-01

Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO 2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

RELAP5 analyses of overcooling transients in a pressurized water reactor

International Nuclear Information System (INIS)

Bolander, M.A.; Fletcher, C.D.; Ogden, D.M.; Stitt, B.D.; Waterman, M.E.

1983-01-01

In support of the Pressurized Thermal Shock Integration Study sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.5 computer code. These analyses were performed for Oconee Plants 1 and 3, which are pressurized water reactors of Babcock and Wilcox lowered-loop design. Results of the RELAP5 analyses are presented, including a comparison with plant data. The capabilities and limitations of the RELAP5/MOD1.5 computer code in analyzing integral plant transients are examined. These analyses require detailed thermal-hydraulic and control system computer models

An approach to estimate the reactivity worth of R-5 poison tube system and experimental verification in ZERLINA reactor

International Nuclear Information System (INIS)

Khosla, S.K.; Paul, O.P.K.; Sengupta, S.N.

1976-01-01

It is proposed to employ a liquid poison injection system as an emergency shut down device for R-5 reactor. The liquid poison consists of gadolinium nitrate solution, which is injected into twenty poison tubes made of zircaloy that are located in between the regular lattice positions in R-5 reactor. The calculational model adopted to estimate the reactivity worth of the poison tubes so as to hold the reactor subcritical by 50 mk at full tank, is described. Similar reactivity estimates have also been carried out for R-5 poison tubes installed in Zerlina reactor in order to assess the adequacy of the calculational mode. The results of the calculations are compared with experimental values for single poison tubes. (author)

Epidemiological, bacteriological and molecular studies on caseous lymphadenitis in Sirohi goats of Rajasthan, India.

Science.gov (United States)

Kumar, Jyoti; Singh, Fateh; Tripathi, Bhupendra Nath; Kumar, Rajiv; Dixit, Shivendra Kumar; Sonawane, Ganesh Gangaram

2012-10-01

Corynebacterium pseudotuberculosis is the causative agent of caseous lymphadenitis (CL), a chronic debilitating disease of goats. In the present study, a total of 575 goats of Sirohi breed on an organized farm situated in the semi-arid tropical region of Rajasthan, India were clinically examined. Pus samples from superficial lymph nodes of 27 (4.7%) adult goats presenting clinical lesions suggestive of CL were collected for bacteriological and molecular analyses. Of these goats, 51.9% yielded C. pseudotuberculosis on the basis of morphological, cultural and biochemical characteristics. A polymerase chain reaction (PCR) assay targeting proline iminopeptidase gene specific to C. pseudotuberculosis was developed that confirmed all 14 bacterial isolates. The specificity of the PCR product was confirmed by sequencing of the 551-bp amplicon in both senses, showing 98-100% homology with published sequences. Thus, overall prevalence rate based on clinical, bacterial culture and PCR assay were found to be 4.7%, 2.4% and 2.4%, respectively. The PCR assay developed in this study was found to be specific and rapid, and could be used for confirmation of CL in goats as an alternative method to generally cumbersome, time-consuming and less reliable conventional methods.

Emergency planning and emergency drill for a 5 MW district heating reactor

International Nuclear Information System (INIS)

Shi Zhongqi; Wu Zhongwang; Hu Jingzhong; Feng Yuying; Li Zhongsan; Dong Shiyuan

1991-01-01

The authors describes the main contents of the emergency planning for a 5 MW nuclear district heating reactor and some considerations for the planning's making, and presents the situation on implementing emergency preparedness and an emergency drill that has been carried out

Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

Directory of Open Access Journals (Sweden)

Skrzypek Maciej

2015-09-01

Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

Science.gov (United States)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo codes for transient reactor analysis

International Nuclear Information System (INIS)

Hoogenboom, J.E.

2013-01-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branch-less collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires the coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3*3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3*3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail. (authors)

WIMSD5, Deterministic Multigroup Reactor Lattice Calculations

International Nuclear Information System (INIS)

2004-01-01

1 - Description of program or function: The Winfrith improved multigroup scheme (WIMS) is a general code for reactor lattice cell calculation on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered the choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are included in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a successor version of WIMS-D/4. 2 - Method of solution: The treatment of resonances is based on the use of equivalence theorems with a library of accurately evaluated resonance integrals for equivalent homogeneous systems at a variety of temperatures. The collision theory procedure gives accurate spectrum computations in the 69 groups of the library for the principal regions of the lattice using a simplified geometric representation of complicated lattice cells. The computed spectra are then used for the condensation of cross-sections to the number of groups selected for solution of the transport equation in detailed geometry. Solution of the transport equation is provided either by use of the Carlson DSN method or by collision probability methods. Leakage calculations including an allowance for streaming asymmetries may be made using either diffusion theory or the more elaborate B1-method. The output of the code provides Eigenvalues for the cases where a simple buckling mode is applicable or cell-averaged parameters for use in overall reactor calculations. Various reaction rate edits are provided for direct comparison with experimental measurements. 3 - Restrictions on the complexity of

Operating reactors licensing actions summary. Volume 5, Number 1

International Nuclear Information System (INIS)

1985-03-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

International Nuclear Information System (INIS)

Scari, Maria Elizabeth

2017-01-01

Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

International Nuclear Information System (INIS)

Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

1978-12-01

This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

Maintenance of Structures, Systems and Components of the RSG-GAS Reactor as an Implementation of BAPETEN Regulation No. 5 Year 2011

International Nuclear Information System (INIS)

Aep Saepudin Catur; Dede Solehudin Fauzi; Djunaidi

2012-01-01

Maintenance activities for structures, systems and components of the reactor, consider prerequisites for the operation of the non power reactor. This activity is intended to ensure that the structures, systems and components function properly. Implementation of reactor maintenance carried out starting from programs establishment, scheduling, maintenance accomplishment and maintenance report. This paper will describe the implementation of reactor maintenance non power as required by BAPETEN regulation no.5, year 2011. By understanding correctly of this regulation, it is expected that maintenance activity of structures, systems and components of the reactor can be successfully performed. The RSG-GAS reactor has implemented various types of reactor maintenance based on BAPETEN regulation no.5 year 2011 properly. As a result failure of the structures, systems and components of the reactor can be minimized then they can be kept reliable. (author)

Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

International Nuclear Information System (INIS)

González Mantecón, Javier

2015-01-01

The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

High Flux Isotope Reactor system RELAP5 input model

International Nuclear Information System (INIS)

Morris, D.G.; Wendel, M.W.

1993-01-01

A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model

Process and impact evaluation of a community gender equality intervention with young men in Rajasthan, India.

Science.gov (United States)

Freudberg, Halima; Contractor, Sana; Das, Abhijit; Kemp, Christopher G; Nevin, Paul E; Phadiyal, Ashima; Lal, Jagdish; Rao, Deepa

2018-02-01

This paper reports on the results of a process and impact evaluation to assess the effects of a project aiming to engage men in changing gender stereotypes and improving health outcomes for women in villages in Rajasthan, India. We conducted seven focus group discussions with participants in the programme and six in-depth interviews with intervention group leaders. We also conducted 137 pre- and 70 post-intervention surveys to assess participant and community knowledge, attitudes and behaviours surrounding gender, violence and sexuality. We used thematic analysis to identify process and impact themes, and hierarchical mixed linear regression for the primary outcome analysis of survey responses. Post-intervention, significant changes in knowledge and attitudes regarding gender, sexuality and violence were made on the individual level by participants, as well as in the community. Moderate behavioural changes were seen in individuals and in the community. Study findings offer a strong model for prevention programmes working with young men to create a community effect in encouraging gender equality in social norms.

On the anisotropy of in-reactor creep of Zr-2.5Nb tubes

International Nuclear Information System (INIS)

Causey, A.R.; Holt, R.A.

1993-06-01

Creep specimens made from cold-worked Zr-2.5Nb tubes, fabricated with two different microstructures and crystallographic textures, were irradiated in the Osiris reactor in France in a fast-neutron flux of about 1.8 x 10 18 n.m -2 .s -1 , E > MeV, at 553 and 585 K. The hoop stresses from internal Fluences, up to 4 x 10 25 n.m -2 , more than double those achieved an any other creep test on cold-worked Zr-2.5Nb in which both axial and transverse strain were measured. Creep rates were obtained from strain versus fluence plots, and creep compliances were obtained from plots of the strain rates against hoop stress for each material at each temperature. The ratio of creep rates at 583 K to those at 553 K was ∼ 1.36, a little higher than that extrapolated from stress relaxation results at temperatures between 523 and 568 K. The ratio of the biaxial creep compliance in the axial direction to that in the transverse directions is different for the two test materials: 0.0 to -0.1 for the fuel sheathing texture and 0.5 to 0.6 for the pressure tube texture. The results were analysed using a self-consistent model developed to account for the contributions to the creep anisotropy of the three microstructure parameters involved and to account for the grain interaction effects. The model, which was normalized to test reactor and power reactor creep data for cold-worked Zr-2.5Nb tubes, predicted the ratio of the creep compliancies to be -0.26 and 0.63, respectively. Thus the creep anisotropy of Zr-2.5Nb tubes with pressure-tube-like crystallographic texture can be adequately predicted. (author). 18 refs., 4 tabs., 13 figs

Household Coverage of Fortified Staple Food Commodities in Rajasthan, India.

Directory of Open Access Journals (Sweden)

Grant J Aaron

Full Text Available A spatially representative statewide survey was conducted in Rajasthan, India to assess household coverage of atta wheat flour, edible oil, and salt. An even distribution of primary sampling units were selected based on their proximity to centroids on a hexagonal grid laid over the survey area. A sample of n = 18 households from each of m = 252 primary sampling units PSUs was taken. Demographic data on all members of these households were collected, and a broader dataset was collected about a single caregiver and a child in the first 2 years of life. Data were collected on demographic and socioeconomic status; education; housing conditions; recent infant and child mortality; water, sanitation, and hygiene practices; food security; child health; infant and young child feeding practices; maternal dietary diversity; coverage of fortified staples; and maternal and child anthropometry. Data were collected from 4,627 households and the same number of caregiver/child pairs. Atta wheat flour was widely consumed across the state (83%; however, only about 7% of the atta wheat flour was classified as fortifiable, and only about 6% was actually fortified (mostly inadequately. For oil, almost 90% of edible oil consumed by households in the survey was classified as fortifiable, but only about 24% was fortified. For salt, coverage was high, with almost 85% of households using fortified salt and 66% of households using adequately fortified salt. Iodized salt coverage was also high; however, rural and poor population groups were less likely to be reached by the intervention. Voluntary fortification of atta wheat flour and edible oil lacked sufficient industry consolidation to cover significant portions of the population. It is crucial that appropriate delivery channels are utilized to effectively deliver essential micronutrients to at-risk population groups. Government distribution systems are likely the best means to accomplish this goal.

Status of water pollution in relation to industrialization in Rajasthan.

Science.gov (United States)

Rajput, Ritu Singh; Pandey, Sonali; Bhadauria, Seema

2017-09-26

India is a large and densely populated country; its economy is largely agricultural. Making the best use of the country's manpower has always posed a challenge. Industrialization could become a dominant component of the economy and displace agriculture. Traditional livelihoods of occupational groups are threatened by the practice of disposing untreated industrial waste into rivers and bodies of water. These uncontrolled disposals impact local natural resources with negative long-term effects. Industrialization is the development of intellectual and financial trade that changes a predominantly rustic culture into a modern one. Many industrial units discharge wastewater locally without treatment. Many industries directly discharged their waste into lakes, rivers and ocean. Water contamination impacts the environment. Pesticides, chemical, waste oil and heavy metals are regularly transported into their waters. Humans and other living organisms can accumulate heavy metals from industrial discharges in their tissues. Industrial waste may be reactive, corrosive, flammable, or toxic. When untreated sewage is emptied into rivers, it causes diseases like typhoid, dysentery and cholera. Natural elements and plant supplements like nitrate and phosphates stimulate growth of algae on the water surface. The algae reduce the oxygen in the water and cause eutrophication. It is harmful to the water ecosystem. In Rajasthan proper, there are a number of sites bordering rivers and lakes where the pace of industrialization has proceeded far beyond the ability of regulators to establish and enforce meaningful limits on the amount of point source pollution permitted to the various industrial complexes, which include cement, chemical, fertilizer, textile, mining, quarrying, dyeing and printing facilities. The scale of the problem is obvious to the casual observer, but actual documentation of the total impact remains to be done.

Nyay karo ya jail bharo] Sathins of Rajasthan demand justice.

Science.gov (United States)

Abha; Geetanjali; Radha; Ratna

1992-01-01

The experiences of Bhanwari, a 4-year old worker for the Women's Development Program in the Kumhar community in Rajasthan, India, is recounted. On September 22, 1992, 2 men from the Gujar community gang raped her and brutally assaulted her husband. The actions were viewed as a reprisal against her for trying to end child marriages. Legal and medical efforts were ineffective in protecting her human rights, and the actions of the men, who were known and identified within the community, were effectively concealed and condoned. The woman and her husband made a concerted effort to obtain a medical examination to prove that rape had occurred. The police accused her of making false allegations. A male doctor refused to examine her. The medical jurist refused to conduct an examination for rape unless orders were issued by the magistrate, who refused. The next day a doctor performed the examination and confirmed that Bhanwari (a mother of four) was not a virgin. Her clothing was taken as evidence, and she was forced to wear the blood-stained clothes of her husband. Since the rape, continued questioning by the police has taken the form of harassment. The National Commission for Women condemned the police behavior and concluded that evidence was tampered with and that the police were protecting the rapists. No arrests have been made. In protest over the mistreatment, a rally was held in Jaipur. During the public meeting, many accounts of sexual abuse by men were related. The issue of gender equality was raised as well as sexual harassment of female workers, who have not legal recourse. A comprehensive law is needed, and women need to speak out against the powerful alliance of politicians and state agencies.

Graphite moderator annealing of the experimental reactor for irradiation (0.5 MW)

International Nuclear Information System (INIS)

Oliveira Avila, Carlos Alberto de; Pires, Luis Fernando Goncalves

1995-01-01

This work describes an operational procedure for the annealing of the graphite moderator in the 0,5 MW Experimental Reactor for Irradiation. A theoretical methodology has been developed for calculating the temperature field during the annealing process. The equations for mass, momentum, and energy conservation for the coolant as well as for the energy conservation in the moderator are solved numerically. The energy stored in the graphite and released in the annealing is accounted for by the use of a modified source term in the energy conservation equation for the moderator. A good agreement has been found for comparisons of the calculations with annealing data from the BEPO reactor. The major parameters affecting annealing have also been determined. (author). 8 refs, 11 figs

Studies on the recovery of REEY from a hard-rock deposit of Rajasthan

International Nuclear Information System (INIS)

Karan, Ram; Giri, Nitai; Anand Rao, K.; Sreenivas, T.

2017-01-01

Dantala area in Banner district of Rajasthan is endowed with poly metallic mineralization hosted in tuffaceous rocks consisting of REE, Zr, Nb, U Th, Zn and Ag. The region has potential to become a hardrock rare earth resource in India. The chemical and XRF analysis of a representative borehole sample revealed that the ore has about 68% SiO_2, 7.7% Al_2O_3, 7.4% Fe_2O_3, 4.75% CaO along with 305 ppm Nb, 109 ppm Th, 3738 ppm TREE of which, 2580 ppm is LREE and 1158 ppm is HREE including 777 ppm Y. The deportment profile of REEY in the sample determined by sequential chemical extraction technique. Major distribution of REEY is dissolvable only under highly acidic conditions as the REEY are bound to iron oxides and are associated with refractory phases (zircon?). Heavy media separation of coarsely crushed material showed presence of about 90% of REEY values in specific gravity d'' 2.8. Therefore, the REEY values could not be preconcentrated by conventional physical beneficiation techniques. Direct 'whole ore' leaching using HCL and H_2SO_4 in 'pug cure leaching' mode with H_2SO_4 yielded about 50% dissolution of REEY values. The paper discusses the process complexities in the recovery of REEY from this hard rock ore sample

Estimation of radon concentration in soil and groundwater samples of Northern Rajasthan, India

Directory of Open Access Journals (Sweden)

Sudhir Mittal

2016-04-01

Full Text Available In the present investigation, analysis of radon concentration in 20 water and soil samples collected from different locations of Bikaner and Jhunjhunu districts of Rajasthan, India has been carried out by using RAD7 an electronic Radon detector. The measured radon concentration in water samples lies in the range from 0.50 to 22 Bq l−1 with the mean value of 4.42 Bq l−1, which lies within the safe limit from 4 to 40 Bq l−1 recommended by United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR, 2008. The total annual effective dose estimated due to radon concentration in water ranges from 1.37 to 60.06 μSV y−1 with the mean value of 12.08 μSV y−1, which is lower than the safe limit 0.1 mSv y−1 as set by World Health Organization (WHO, 2004 and European Council (EU, 1998. Radon measurement in soil samples varies from 941 to 10,050 Bq m−3 with the mean value of 4561 Bq m−3, which lies within the range reported by other investigators. It was observed that the soil and water of Bikaner and Jhunjhunu districts are suitable for drinking and construction purpose without posing any health hazard.

Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.

2003-01-01

The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)

Some suggestions based on the instrumentation installation experience at RAPP

International Nuclear Information System (INIS)

Raghunath, M.R.; Singh, S.; Jain, V.K.

1977-01-01

Suggestions regarding installation of reactor instrumentation have been made based on the instrumentation installation experience at the Rajasthan Atomic Power Plant. It has been mentioned that the instrumentation installation work has to proceed simultaneously with that of the heavy equipment and piping errection work, to meet the commissioning target dates. (S.K.K.)

Knowledge regarding antibiotic drug action and prescription practices among dentist in Jaipur city, Rajasthan

Directory of Open Access Journals (Sweden)

Dushyant Pal Singh

2015-01-01

Full Text Available Introduction: Dentists prescribe antibiotics routinely to manage oral and dental infections. Unscrupulous antibiotic prescriptions can be associated with unfavorable side effects and the development of resistance. Thus, the aim of this study was to assess the level of knowledge regarding antibiotic prescription use among dentists in Jaipur City, Rajasthan. Materials and Methods: A questionnaire survey was conducted among 300 dentists in Jaipur city. A validated, self-designed, 21-item, closed-ended questionnaire was used to collect data on knowledge regarding antibiotic prescription. Descriptive statistics were calculated. Results: A total of 300 dental practitioners were included in the study. The majority of the respondents seem to prescribe antibiotics that are broad spectrum or the ones that are commonly used. A considerable percentage of the respondents were not aware of the pregnancy drug risk categories by Food and Drug Administration. The most of the respondents said that they prescribe antibiotics on the basis of the diagnosis, whereas more than two-thirds of the respondents said that they never advise culture sensitivity test before prescribing the antibiotics. Conclusion: Our findings suggest the knowledge of dentists regarding antibiotic prescription is inadequate and more focus should be given to the ongoing training regarding the pharmacological aspects, pertinent medical conditions, and prophylactic use of antibiotics in dentistry.

Development and validation of a model TRIGA Mark III reactor with code MCNP5; Desarrollo y validacion de un modelo del reactor Triga Mark III con el codigo MCNP5

Energy Technology Data Exchange (ETDEWEB)

Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K{sub eff} was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

Operating reactors licensing actions summary. Volume 5, No. 9

International Nuclear Information System (INIS)

1985-11-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Operating reactors licensing actions summary. Volume 5, No. 8

International Nuclear Information System (INIS)

1985-10-01

This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

The effect of an affordable daycare program on health and economic well-being in Rajasthan, India: protocol for a cluster-randomized impact evaluation study

Directory of Open Access Journals (Sweden)

Arijit Nandi

2016-06-01

Full Text Available Abstract Background The provision of affordable and reliable daycare services is a potentially important policy lever for empowering Indian women. Access to daycare might reduce barriers to labor force entry and generate economic opportunities for women, improve education for girls caring for younger siblings, and promote nutrition and learning among children. However, empirical evidence concerning the effects of daycare programs in low-and-middle-income countries is scarce. This cluster-randomized trial will estimate the effect of a community-based daycare program on health and economic well-being over the life-course among women and children living in rural Rajasthan, India. Methods This three-year study takes place in rural communities from five blocks in the Udaipur District of rural Rajasthan. The intervention is the introduction of a full-time, affordable, community-based daycare program. At baseline, 3177 mothers with age eligible children living in 160 village hamlets were surveyed. After the baseline, these hamlets were randomized to the intervention or control groups and respondents will be interviewed on two more occasions. Primary social and economic outcomes include women’s economic status and economic opportunity, women’s empowerment, and children’s educational attainment. Primary health outcomes include women’s mental health, as well as children’s nutritional status. Discussion This interdisciplinary research initiative will provide rigorous evidence concerning the effects of daycare in lower-income settings. In doing so it will address an important research gap and has the potential to inform policies for improving the daycare system in India in ways that promote health and economic well-being. Trial registration (1 The ISRCTN clinical trial registry (ISRCTN45369145, http://www.isrctn.com/ISRCTN45369145 , registered on May 16, 2016 and (2 The American Economic Association’s registry for randomized controlled trials

Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

Energy Technology Data Exchange (ETDEWEB)

C. B. Davis

2006-07-01

The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

Research activities on high-temperature gas-cooled reactors (HTRs) in the 5. EURATOM RTD Framework programme

International Nuclear Information System (INIS)

Martin-Bermejo, J.; Hugon, M.; Van Goethem, G.

2002-01-01

One of the areas of research of the 'nuclear fission' key action of the 5. EURATOM RTD Framework Programme (FP5) is the safety and efficiency of future systems. The main objective of this area is to investigate and evaluate new or revisited concepts (both reactors and alternative fuels) for nuclear energy that offer potential longer term benefits in terms of cost, safety, waste management, use of fissile material, less risk of diversion and sustainability. Several projects related to high-temperature gas-cooled reactors (HTRs) were retained by the European Commission (EC) services. They address important issues such as HTR fuel technology, HTR fuel cycle, HTR materials, power conversion systems and licensing. Most of these projects have already started and are progressing according to the schedule. They are the initial core of activities of a European Network on 'High-temperature Reactor Technology' (HTR-TN) recently set up by 18 EU organisations. (authors)

Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor

International Nuclear Information System (INIS)

Bhattacharyya, S. K.; Boing, L. E.

2000-01-01

The aging of research reactors worldwide has resulted in a heightened awareness in the international technical decommissioning community of the timeliness to review and address the needs of these research institutes in planning for and eventually performing the decommissioning of these facilities. By using the reactors already undergoing decommissioning as test beds for evaluating enhanced or new/innovative technologies for decommissioning, it is possible that new techniques could be made available for those future research reactor decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the research institutes in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research reactor decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to use in the research reactors. The decommissioning of the CP-5 Research Reactor is currently in the final phase of dismantlement. In this paper the authors present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors

Small propulsion reactor design based on particle bed reactor concept

International Nuclear Information System (INIS)

Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

1989-01-01

In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

Research reactor 'A' 6.5/10 MW; Istrazivacki reaktor 'A' 6,5/10 MW

Energy Technology Data Exchange (ETDEWEB)

Petrovic, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1961-07-01

This booklet includes a short description of the RA research reactor with basic properties of its components: control and safety system, heavy water system, technical water cooling system, heavy water distillation system, cover gas system, dosimetry control system, power supply system. It is used for fundamental research in reactor and nuclear physics, isotope production, materials testing.

Neutronics modeling of TRIGA reactor at the University of Utah using agent, KENO6 and MCNP5 codes

International Nuclear Information System (INIS)

Yang, X.; Xiao, S.; Choe, D.; Jevremovic, T.

2010-01-01

The TRIGA reactor at the University of Utah is modelled in 2D using the AGENT state-of-the-art methodology based on the Method of Characteristics (MOC) and R-function theory supporting detailed reactor analysis of reactor geometries of any type. The TRIGA reactor is also modelled using KENO6 and MCNP5 for comparison. The spatial flux and reaction rates distribution are visualized by AGENT graphics support. All methodologies are in use in to study the effect of different fuel configurations in developing practical educational exercises for students studying reactor physics. At the University of Utah we train graduate and undergraduate students in obtaining the Nuclear Regulatory Commission license in operating the TRIGA reactor. The computational models as developed are in support of these extensive training classes and in helping students visualize the reactor core characteristics in regard to neutron transport under various operational conditions. Additionally, the TRIGA reactor is under the consideration for power uprate; this fleet of computational tools once benchmarked against real measurements will provide us with validated 3D simulation models for simulating operating conditions of TRIGA. (author)

Development of a coupled reactor with a catalytic combustor and steam reformer for a 5 kW solid oxide fuel cell system

International Nuclear Information System (INIS)

Kang, Sanggyu; Lee, Kanghun; Yu, Sangseok; Lee, Sang Min; Ahn, Kook-Young

2014-01-01

Highlights: • Proposes the scale-up strategy to develop a large-scale coupled reactor. • Investigation of performance of steam reformer coupled with catalytic combustor. • Experimental parameters are inlet temp., air excess ratio, SCR, fuel utilization. • Evaluation of the heat transfer distribution along the gas flow direction. • The mean value of methane conversion rate is approximately 93.4%. - Abstract: The methane (CH 4 ) conversion rate of a steam reformer can be increased by thermal integration with a catalytic combustor, called a coupled reactor. In the present study, a 5 kW coupled reactor has been developed based on a 1 kW coupled reactor in previous work. The geometric parameters of the space velocity, diameter and length of the coupled reactor selected from the 1 kW coupled reactor are tuned and applied to the design of the 5 kW coupled reactor. To confirm the scale-up strategy, the performance of 5 kW coupled reactor is experimentally investigated with variations of operating parameters such as the fuel utilization in the solid oxide fuel cell (SOFC) stack, the inlet temperature of the catalytic combustor, the excess air ratio of the catalytic combustor, and the steam to carbon ratio (SCR) in the steam reformer. The temperature distributions of coupled reactors are measured along the gas flow direction. The gas composition at the steam reformer outlet is measured to find the CH 4 conversion rate of the coupled reactor. The maximum value of the CH 4 conversion rate is approximately 93.4%, which means the proposed scale-up strategy can be utilized to develop a large-scale coupled reactor

Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

International Nuclear Information System (INIS)

Tuunanen, J.

1993-01-01

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

In-reactor creep of zirconium-2.5 wt% niobium at 570 K

International Nuclear Information System (INIS)

Coleman, C.E.; Causey, A.R.; Fidleris, V.

1976-01-01

The effect of fast neutron flux at 570 K on the creep rate of specimens of zirconium-2.5 wt% niobium alloy taken from tubes in various metallurgical conditions has been measured using both constant load tensile creep machines and bent-beam stress relaxation. Creep rates calculated from stress relaxation fit on the trend line for the constant load creep data. Between 114 MPa and 450 MPa the creep rate is proportional to neutron flux. The creep rate of specimens from the longitudinal direction is about twice that of specimens from the circumferential direction of a tube. This anisotropy in creep strength is attributed partly to crystallographic texture and partly to deformation substructure. Cold-work is detrimental to in-reactor creep strength; as-extruded material has higher creep strength. In cold-worked material at stresses below 100 MPa the stress exponent, n, is about 1; n gradually increases with stress being about 10 at 525 MPa and about 100 at 660 MPa. In laboratory tests, rupture ductility correlates inversely with n; the lower n the higher the ductility. In-reactor tests support this correlation thus pressure tubes in CANDU reactors, operating at 117 MPa where n approximately 1, should have good ductility. (Auth.)

Sensitivity analysis to a RELAP5 nodalization developed for a typical TRIGA research reactor

International Nuclear Information System (INIS)

Reis, Patrícia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M.; Veloso, Maria Auxiliadora F.

2012-01-01

Highlights: ► We investigated how much the code results are affected by the code user. ► Two essential modifications were made on a previously validated nodalization. ► We used the RELAP5 code to predict the results. ► Results highlight the necessity of sensitivity analysis to have the ideal modeling. - Abstract: The main aim of this work is to identify how much the code results are affected by the code user in the choice of, for example, the number of thermal hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previously validated nodalization for analysis of steady-state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of modeling. The results highlight the necessity of sensitivity analysis to obtain the ideal modeling to simulate a specific system.

Experimental and MCNP5 based evaluation of neutron and gamma flux in the irradiation ports of the University of Utah research reactor

Directory of Open Access Journals (Sweden)

Noble Brooklyn

2012-01-01

Full Text Available Neutron and gamma flux environment of various irradiation ports in the University of Utah training, research, isotope production, general atomics reactor were experimentally assessed and fully modeled using the MCNP5 code. The experimental measurements were based on the cadmium ratio in the irradiation ports of the reactor, flux profiling using nickel wire, and gamma dose measurements using thermo luminescence dosimeter. Full 3-D MCNP5 reactor model was developed to obtain the neutron flux distributions of the entire reactor core and to compare it with the measured flux focusing at the irradiation ports. Integration of all these analysis provided the updated comprehensive neutron-gamma flux maps of the existing irradiation facilities of the University of Utah TRIGA reactor.

Relap5 Analysis of Processes in Reactor Cooling Circuit and Reactor Cavity in Case of Station Blackout in RBMK-1500

International Nuclear Information System (INIS)

Kaliatka, A.

2007-01-01

Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA) to a severe accident are discussed

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

Thermal hydraulic and neutron kinetic simulation of the Angra 2 reactor using a RELAP5/PARCS coupled model

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

2015-07-01

The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)

Degradation of Reactive Black 5 dye using anaerobic/aerobic membrane bioreactor (MBR) and photochemical membrane reactor

International Nuclear Information System (INIS)

You, Sheng-Jie; Damodar, Rahul A.; Hou, Sheng-Chon

2010-01-01

Three different types of advance treatment methods were evaluated for the degradation of Reactive Black 5 (RB5). The performance of two stage anaerobic SBR-aerobic MBR, anaerobic MBR with immobilized and suspended biocells and an integrated membrane photocatalytic reactor (MPR) using slurry UV/TiO 2 system were investigated. The results suggest that, nearly 99.9% color removal and 80-95% organic COD and TOC removal can be achieved using different reactor systems. Considering the Taiwan EPA effluent standard discharge criteria for COD/TOC, the degree of treatment achieved by combining the anaerobic-aerobic system was found to be acceptable. Anew, Bacilluscereus, high color removal bacterium was isolated from Anaerobic SBR. Furthermore, when this immobilized into PVA-calcium alginate pellets, and suspended in the anaerobic MBR was able to achieve high removal efficiencies, similar to the suspended biocells system. However, the immobilized cell Anaerobic MBR was found to be more advantageous, due to lower fouling rates in the membrane unit. Results from slurry type MPR system showed that this system was capable of mineralizing RB5 dyes with faster degradation rate as compared to other systems. The reactor was also able to separate the catalyst effectively and perform efficiently without much loss of catalyst activity.

Description of the two-loop RELAP5 model of the L-Reactor at the Savannah River Site

International Nuclear Information System (INIS)

Cozzuol, J.M.; Davis, C.B.

1989-12-01

A two-loop RELAP5 input model of the L-Reactor at the Savannah River Site (SRS) was developed to support thermal-hydraulic analysis of SRS reactors. The model was developed to economically evaluate potential design changes. The primary simplifications in the model were in the number of loops and the detail in the moderator tank. The six loops in the reactor were modeled with two loops, one representing a single loop and the other representing five combined loops. The model has undergone a quality assurance review. This report describes the two-loop model, its limitations, and quality assurance. 29 refs., 18 figs., 10 tabs

Prevalence of spheroidal degeneration of cornea and its association with other eye diseases in tribes of Western Rajasthan

Directory of Open Access Journals (Sweden)

Amit Mohan

2017-01-01

Full Text Available Purpose: To determine the prevalence of spheroidal degeneration of cornea (SDC and its association with other eye diseases in the tribes of South-West Rajasthan. Methods: A total of 5012 patients were examined on slit lamp for the diagnosis of SDC. Diagnosis of SDC was made based on presence of amber granules in the superficial stroma of peripheral interpalpebral cornea with increasing opacification, coalescence and central spread or nodular and hazy surrounding stroma and divided in three stages. Results: The prevalence of SDC was 10.7%. Around 55% of the total of 535 cases examined were found to have Stage I followed by Stage II (32% and Stage III (13%. The prevalence is greatest in both men and women over 70 years of age. The severity of SDC is greater in men. SDC was significantly associated with pterygium and pseudocapsular exfoliation. Conclusion: Extreme temperature, low humidity, dust, high wind, and microtrauma caused by sand particles are the probable etiologies for higher prevalence of this kind of degeneration in this region.

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

International Nuclear Information System (INIS)

Szczurek, J.

1995-01-01

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

Energy Technology Data Exchange (ETDEWEB)

Szczurek, J. [Inst. of Atomic Energy, Swierk (Poland)

1995-12-31

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

Energy Technology Data Exchange (ETDEWEB)

Szczurek, J [Inst. of Atomic Energy, Swierk (Poland)

1996-12-31

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

RELAP5 thermal-hydraulic analyses of overcooling sequences in a pressurized water reactor

International Nuclear Information System (INIS)

Bolander, M.A.; Fletcher, C.D.; Davis, C.B.; Kullberg, C.M.; Stitt, B.D.; Waterman, M.E.; Burtt, J.D.

1984-01-01

In support of the Pressurized Thermal Shock Integration Study, sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.6 and MOD2.0 computer codes. These analyses were performed for the H.B. Robinson Unit 2 pressurized water reactor, which is a Westinghouse 3-loop design plant. Results of the RELAP5 analyses are presented. The capabilities of the RELAP5 computer code as a tool for analyzing integral plant transients requiring a detailed plant model, including complex trip logic and major control systems, are examined

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

Integrated analysis for a small break LOCA in the IRIS reactor using MELCOR and RELAP5 codes

International Nuclear Information System (INIS)

Del Nevo, A.; Manfredini, A.; Oriolo, F.; Paci, S.; Oriani, L.

2004-01-01

The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. This paper's focus is on the use of well known computer codes for integrated (reactor vessel and containment) calculations of the IRIS response to a small break loss of coolant accident (LOCA). In IRIS, large break LOCA events are eliminated by the use of a layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than designing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. The novel IRIS approach requires development of evaluation models that are different from those used for the current generation of pressurized water reactors. An analysis of small break LOCAs for IRIS is documented in two companion papers, and has been developed using a preliminary evaluation model based on the explicit coupling of the RELAP5 and GOTHIC codes. The objective of this paper is to compare the results obtained via the coupled RELAP/GOTHIC code with different computational tools. A reference case from the preliminary IRIS safety assessment was selected, and the same small break LOCA sequence is analyzed using

Analysis of a total flow blockage of a Fuel Assembly in a typical MTR Research Reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Adorni, M.; Salah, A.B.; Di Maro, B.; Pierro, F.; D'Auria, F.; Hamidouche, T.

2004-01-01

The lack of full understanding of complex mechanisms connected with the interaction between thermal-hydraulics and neutronics still challenge the design and the operation of nuclear reactors by the adoption of conservative safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience imposes the revisiting of those areas and eventually the identification of design/safety requirements that can be relaxed [1]. Currently, the enlarged commercial exploitation of nuclear Research Reactors (RR) has increased the consideration to their corresponding safety issues. Almost all of the safety analyses have so far been performed using conservative computational tools [2]. Nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity in order to increase their commercial productivity. In this framework, an attempt is made to apply the BE technique to perform a safety evaluation under research reactors operational conditions. In fact, this technique has been largely verified and validated for power reactors using coupled system thermal-hydraulic and three-dimensional neutron kinetics [1]. For this purpose, as typical representative of research reactors, the IAEA 10 MW MTR Research Reactors problem [3] is considered. The system thermal-hydraulic RELAP5 [4] code was developed to simulate transient scenarios in Power reactors such PWR, BWR, VVER, etc. However, only limited work was performed to access the applicability of the code to Research Reactors operating conditions (low pressure, mass flow rates, power, etc) [5]. Previous works performed in this field are reported in [5], [6] and [7]. In this framework, total and partial blockage of a single Fuel Assembly cooling channel are investigated. As a first attempt the calculations are performed by applying the BE thermal-hydraulic system code RELAP5 alone using its point kinetic model to derive the instantaneous core

RA research nuclear reactor operation in forced regime, Annex 5; Prilog 5 - Rad istrazivackog nuklearnog reaktora RA u forsiranom rezimu, Prvo saopstenje

Energy Technology Data Exchange (ETDEWEB)

Mitrovic, S [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1964-12-15

In order to increase the flux and experimental potentials of the RA reactor, properties of the reactor operating in the forced regime have been inspected four times, twice with the spent fuel (core C), and twice with partly fresh fuel (cores E and F). Forced regime means power higher than nominal 6.5 MW. From the analyses of technology parameters during operation in the forced regime at power levels of 7 and 11 MW, state and behaviour of the components, efficiency of the biological protection and other parameters it was concluded that the reactor operation is possible at power levels of 10 MW during a whole year, and under certain conditions even longer operation at power levels up to 15 MW. This paper examines the reactor operation conditions in the forced regime and gives conclusions related to real increase of the experimental possibilities, isotope production and economic factors. Radi povecanja fluksa i eksperimentalnih mogucnosti reaktora RA, cetiri puta su proverene osobine reaktora u forsiranom rezimu, tj. na snagama vecim od nominalne (6,5 MW) i to dva puta sa istrosenim gorivom (jezgro C) i dva puta sa delimicno svezim (jezgra E i F). Iz analize tehnoloskih parametara u toku eksperimentalno rada reaktora u forsiranom rezimu na snagama od 7 do 11 MW, stanja i ponasanja opreme, efikasnosti bioloske zastite i ostalih parametara, zakljucuje se da je moguc permanentan rad reaktora tokom cele godine na snagama do 10 MW, a pod izvesnim uslovima i duzi rad na snagama do 15 MW. U radu su diskutovani uslovi rada reaktora u forsiranom rezimu i donose se zakljucci o realnom povecanju eksperimentalnih mogucnosti, proizvodnje radioizotopa i o ekonomskim faktorima (author)

CFX-10 and RELAP5-3D simulations of coolant mixing phenomena in RPV of VVER-1000 reactors

International Nuclear Information System (INIS)

Terzuoli, F.; Moretti, F.; Melideo, D.; D'Auria, F.; Shkarupa, O.

2006-01-01

The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed with the ANSYS CFX-10 CFD code and with the RELAP5-3D system code. In particular, the attention focused on the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. The results have been compared against experimental data from V1000CT-2 Benchmark. (author)

Thermal, thermo-hydraulic and thermo-mechanic analysis for fuel elements of IEA-R1 reactor at 5MW

International Nuclear Information System (INIS)

Teixeira e Silva, A.; Silva Macedo, L.V. da

1989-01-01

In connection with the on going conversion of IEA-R1 Research Reactor, operated by IPEN-CNEN/SP, from the use of highly enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel, steady-state thermal and thermo-hydraulic analysis of both existing HEU and proposed LEU cores under 2 MW operating conditions have been carried out. Keeping in mind the possibility of power upgrading, steady-state thermal, thermo-hydraulic and thermomechanical analysis of proposed LEU core under 5 MW operating conditions have also been carried out. The thermal and thermo-hydraulic analysis at 2 MW show that the conversion of the existing HEU core to be proposed LEU core will not change the reactor safety margins. Although the upgrading of the reactor power to 5 MW will result in safety margins lower than in case of 2MW, these will be still sufficient for optimum operation and safe behaviour. The thermomechanical analysis at 5 MW show that the thermal stresses induced in the fuel element will satisfy the design limits for mechanical strenght and elastic stability. (author) [pt

Development and validation of a model TRIGA Mark III reactor with code MCNP5

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Aguilar H, F.

2015-09-01

The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K eff was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

Analysis of loss of flow events on Brazilian multipurpose reactor by RELAP5 code

International Nuclear Information System (INIS)

Soares, Humberto V.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Aronne, Ivan D.; Rezende, Guilherme P.

2011-01-01

The Brazilian Multipurpose Reactor (BMR) is currently being projected and analyzed. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly, with planar plates. BMR will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of BMR using the RELAP5 model and also three transient cases of loss of flow accident (LOFA), in the primary cooling system. A LOFA may arise through failures associated with the primary cooling system pumps or through events resulting in a decrease in the primary coolant flow with the primary cooling system pumps functioning normally. The cases presented in this paper are: primary cooling system pump shaft seizure, failure of one primary cooling system pump motor and failure of both primary cooling system pump motors. In the shaft seizure case, the flow reduction is sudden, with the blocking of the flow coast down The motor failure cases, deal with the failure of one or two pump motor due to, for example, malfunction or interruption of power and differently of the shaft seizure it can be observed the flow coast down provided by the pump inertia. It is shown that after all initiating events the reactor reaches a safe new steady state keeping the integrity of the fuel elements. (author)

Analysis of loss of flow events on Brazilian multipurpose reactor by RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Soares, Humberto V.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria Auxiliadora F., E-mail: antonella@nuclear.ufmg.br, E-mail: laubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, UFMG, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores, CNPq (Brazil); Aronne, Ivan D.; Rezende, Guilherme P., E-mail: aroneid@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte (Brazil).

2011-07-01

The Brazilian Multipurpose Reactor (BMR) is currently being projected and analyzed. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly, with planar plates. BMR will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of BMR using the RELAP5 model and also three transient cases of loss of flow accident (LOFA), in the primary cooling system. A LOFA may arise through failures associated with the primary cooling system pumps or through events resulting in a decrease in the primary coolant flow with the primary cooling system pumps functioning normally. The cases presented in this paper are: primary cooling system pump shaft seizure, failure of one primary cooling system pump motor and failure of both primary cooling system pump motors. In the shaft seizure case, the flow reduction is sudden, with the blocking of the flow coast down The motor failure cases, deal with the failure of one or two pump motor due to, for example, malfunction or interruption of power and differently of the shaft seizure it can be observed the flow coast down provided by the pump inertia. It is shown that after all initiating events the reactor reaches a safe new steady state keeping the integrity of the fuel elements. (author)

The CEA research reactors

International Nuclear Information System (INIS)

Schwartz, J.P.

1993-01-01

Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

Volatility of V15Cr5Ti fusion reactor alloy

International Nuclear Information System (INIS)

Neilson, R.M. Jr.

1986-01-01

One potential hazard from the presence of activation products in fusion facilities is accidental oxidation-driven volatility of those activation products. Scoping experiments were conducted to investigate the oxidation and elemental volatility of candidate fusion reactor alloy V15Cr5Ti as a function of time, temperature, and test atmosphere. Experiments in air and in argon carrier gases containing 10 4 to 10 1 Pa (10 -1 to 10 -4 atm) oxygen were conducted to investigate the lower oxygen partial pressure limit for the formation of a low melting point (approximately 650 0 C), high volatility, oxide layer and its formation rate. Experiments to determine the elemental volatility of alloy constituents in air at temperatures of 700 0 C to greater than 1600 0 C. Some of these volatility experiments used V15Cr5Ti that was arc-remelted to incorporate small quantities (<0.1 wt. %) of Sc and Ca. Incorporation of Sc and Ca in test specimens permitted volatility measurement of radioactive constituents present only after activation of V15Cr5Ti

The status and distribution of major aquatic fauna in the National Chambal Gharial Sanctuary in Rajasthan with special reference to the Gangetic Dolphin Platanista gangetica gangetica (Cetartiodactyla: Platanistidae

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A.K. Nair

2009-03-01

Full Text Available This paper records observation on the status and distribution of Gangetic Dolphin, Gharial, Mugger and other aquatic animals, and birds in the National Chambal Gharial Sanctuary in Rajasthan during the Chambal river expedition conducted with the Indian Army in May 1998. A total of five Gangetic Dolphins, nine Gharials, 14 Indian Mugger crocodiles and 118 species of birds were sighted during the survey of 350km-long stretch of the river Chambal from Keshoraipatan-Bundi to Dhaulpur. The current status of the riverine habitat in view of disturbance and other anthropogenic factors is discussed and suggestions made to safeguard the sanctuary from various threats.

Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)

International Nuclear Information System (INIS)

Bottimore, R.R.

1980-12-01

The Standard Technical Specifications for General Electric Boiling Water Reactors (GE-STS) is a generic document prepared by the US NRC for use in the licensing process of current General Electric Boiling Water Reactors. The GE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

A preliminary neutronic evaluation and depletion study of VHTR and LS-VHTR reactors using the codes: WIMSD5 and MCNPX

International Nuclear Information System (INIS)

Silva, Fabiano C.; Pereira, Claubia; Costa, Antonella Lombardi; Veloso, Maria Auxiliadora Fortini

2009-01-01

It is expected that, in the future, besides electricity generation, reactors should also develop secondary activities, such as hydrogen generation and seawater desalinization. Generation IV reactors are expected to possess special characteristics, like high safety, minimization of radioactive rejects amount and ability to use reprocessed fuel with non-proliferating projects in their cycles. Among the projects of IV generation reactors available nowadays, the (High Temperature Reactors) HTR, are highlighted due to these desirable characteristics. Under such circumstances, such reactor may be able to have significant higher thermal power ratings to be used for hydrogen production, without loose of safety, even in an emergency. For this work, we have chosen two HTR concepts of a prismatic reactor: (Very High Temperature Reactor) VHTR and the (Liquid Salted -Very High Temperature Reactor) LS-VHTR. The principal difference between them is the coolant. The VHTR uses helium gas as a coolant and have a burnup of 101,661 MWd/THM while the LS-VHTR uses low-pressure liquid coolant molten fluoride salt with a boiling point near 1500 de C working at 155,946 MWd/THM. The ultimate power output is limited by the capacity of the passive decay system; this capacity is limited by the reactor vessel temperature. The goal was to evaluate the neutronic behavior and fuel composition during the burnup using the codes (Winfrith Improved Multi-Group Scheme) WIMSD5 and the MCNPX2.6. The first, deterministic and the second, stochastic. For both reactors, burned fuel type 'C' coming from Angra-I nuclear plant, in Brazil, was used with 3.1% of initial enrichment, burnup to 33,000 MWd/THM using the ORIGEN2.1 code, divided in three steps of 11,000 MWd/THM, with an average density power of 37.75 MWd/THM and 5 years of cooling in pool. Finally, the fuel was reprocessed by Purex technique extracting 99.9% of Pu, and the desired amount of fissile material (15%) to achieve the final mixed oxide was

Probability safety assessment of the Kozloduy-5 and Kozloduy-6 reactors

Energy Technology Data Exchange (ETDEWEB)

Boyadzhiev, A; Manchev, B [Risk Engineering Ltd., Sofia (Bulgaria)

1996-12-31

A probability safety assessment (PSA) of Level 1 (assessment of plant failures leading to the determination of core damage frequency) has been carried out for the NPP Kozloduy Units 5 and 6 (reactors WWER-1000). The scope of the study includes all significant accident initiators including seismic (earthquake) and fire initiators. Event trees for all initiators and fault trees for front line systems, support systems and major safety systems have been built. A distribution of the different initiators has been established as follows: internal initiators - 85%, seismic initiators - 5%, fire initiators- 10%. The loss of offsite power was identified as main contributor from the internal initiators with frequency 1,1.10{sup -4}/y. It is concluded that the safety functions of WWER-1000 are adequately covered by the safety systems. 4 refs., 2 tabs.

Palynology of Lower Palaeogene (Thanetian-Ypresian) coastal deposits from the Barmer Basin (Akli Formation, Western Rajasthan, India): palaeoenvironmental and palaeoclimatic implications

Energy Technology Data Exchange (ETDEWEB)

Tripathi, S.K.M.; Kumar, M.; Srivastava, D. [Birbal Sahni Instititue of Paleobotany, Lucknow (India)

2009-03-15

The 32-m thick sedimentary succession of the Paleocene-Eocene Akli Formation (Barmer basin, Rajasthan, India), which is exposed in an open-cast lignite mine, interbed several lignite seams that alternate with fossiliferous carbonaceous clays, green clays and widespread siderite bands and chert nodules. The palynofloral assemblages consist of spore, pollen and marine dinoflagellate cysts that indicate a Thanetian to Ypresian age. The assemblage is dominated by angiospermic pollen and specimens showing affinity with the mangrove Palm Nypa are also very abundant. The Nypa-like pollen specimens exhibit a wide range of morphological variation, some of the recorded morphotypes being restricted to this Indian basin. Preponderance of these pollen taxa indicates that the sediments were deposited in a coastal swamp surrounded by thick, Nypa-dominated mangrove vegetation. The dispersed organic matter separated from macerated residues indicates the dominance of anoxic conditions throughout the succession, although a gradual transition to oxic conditions is recorded in the upper part.

Application of non-thermal plasma reactor for degradation and detoxification of high concentrations of dye Reactive Black 5 in water

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Dojčinović Biljana P.

2016-01-01

Full Text Available Degradation and detoxification efficiency of high concentrations of commercially available reactive textile dye Reactive Black 5 solution (40, 80, 200, 500, 1000 mg L-1, were studied. Advanced oxidation processes in water falling film based dielectric barrier discharge as a non-thermal plasma reactor were used. For the first time, this reactor was used for the treatment of high concentrations of organic pollutants such as reactive textile dye Reactive Black 5 in water. Solution of the dye is treated by plasma as thin water solution film that is constantly regenerated. Basically, the reactor works as a continuous flow reactor and the electrical discharge itself takes place at the gas-liquid interphase. The dye solution was recirculated through the reactor with an applied energy density of 0-374 kJ L-1. Decolorization efficiency (% was monitored by UV-VIS spectrophotometric technique. Samples were taken after every recirculation (~ 22 kJ L-1 and decolorization percent was measured after 5 min and 24 h of plasma treatment. The efficiency of degradation (i.e. mineralization and possible degradation products were also tracked by determination of the chemical oxygen demand (COD and by ion chromatography (IC. Initial toxicity and toxicity of solutions after the treatment were studied with Artemia salina test organisms. Efficiency of decolorization decreased with the increase of the dye concentration. Complete decolorization, high mineralization and non-toxicity of the solution (<10 % were acomplished after plasma treatment using energy density of 242 kJ L-1, while the initial concentrations of Reactive Black 5 were 40 and 80 mg L-1. [Projekat Ministarstva nauke Republike Srbije, br. 172030 i br. 171034

Recurrent Early Cretaceous, Indo-Madagascar (89-86 Ma) and Deccan (66 Ma) alkaline magmatism in the Sarnu-Dandali complex, Rajasthan: 40Ar/39Ar age evidence and geodynamic significance

Science.gov (United States)

Sheth, Hetu; Pande, Kanchan; Vijayan, Anjali; Sharma, Kamal Kant; Cucciniello, Ciro

2017-07-01

The Sarnu-Dandali alkaline complex in Rajasthan, northwestern India, is considered to represent early, pre-flood basalt magmatism in the Deccan Traps province, based on a single 40Ar/39Ar age of 68.57 Ma. Rhyolites found in the complex are considered to be 750 Ma Malani basement. Our new 40Ar/39Ar ages of 88.9-86.8 Ma (for syenites, nephelinite, phonolite and rhyolite) and 66.3 ± 0.4 Ma (2σ, melanephelinite) provide clear evidence that whereas the complex has Deccan-age (66 Ma) components, it is dominantly an older (by 20 million years) alkaline complex, with rhyolites included. Basalt is also known to underlie the Early Cretaceous Sarnu Sandstone. Sarnu-Dandali is thus a periodically rejuvenated alkaline igneous centre, active twice in the Late Cretaceous and also earlier. Many such centres with recurrent continental alkaline magmatism (sometimes over hundreds of millions of years) are known worldwide. The 88.9-86.8 Ma 40Ar/39Ar ages for Sarnu-Dandali rocks fully overlap with those for the Indo-Madagascar flood basalt province formed during continental breakup between India (plus Seychelles) and Madagascar. Recent 40Ar/39Ar work on the Mundwara alkaline complex in Rajasthan, 120 km southeast of Sarnu-Dandali, has also shown polychronous emplacement (over ≥ 45 million years), and 84-80 Ma ages obtained from Mundwara also arguably represent post-breakup stages of the Indo-Madagascar flood basalt volcanism. Remnants of the Indo-Madagascar province are known from several localities in southern India but hitherto unknown from northwestern India 2000 km away. Additional equivalents buried under the vast Deccan Traps are highly likely.

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

5. symposium on nuclear reactor remote monitoring

International Nuclear Information System (INIS)

1987-01-01

17 papers deal with the data-technological concept and mode of operation of nuclear-reactor remote-monitoring (RM) systems from the perspectives of users in Baden-Wurttemberg, Sleswig-Holstein, Bavaria and Belgium, with the requirements on measuring devices and equipment in NRM systems, computer-controlled evaluation and processing of measured data, in particular the LASAT and OLDES systems. (DG) [de

Early postpartum maternal morbidity among rural women of Rajasthan, India: a community-based study.

Science.gov (United States)

Iyengar, Kirti

2012-06-01

The first postpartum week is a high-risk period for mothers and newborns. Very few community-based studies have been conducted on patterns of maternal morbidity in resource-poor countries in that first week. An intervention on postpartum care for women within the first week after delivery was initiated in a rural area of Rajasthan, India. The intervention included a rigorous system of receiving reports of all deliveries in a defined population and providing home-level postpartum care to all women, irrespective of the place of delivery. Trained nurse-midwives used a structured checklist for detecting and managing maternal and neonatal conditions during postpartum-care visits. A total of 4,975 women, representing 87.1% of all expected deliveries in a population of 58,000, were examined in their first postpartum week during January 2007-December 2010. Haemoglobin was tested for 77.1% of women (n=3,836) who had a postnatal visit. The most common morbidity was postpartum anaemia--7.4% of women suffered from severe anaemia and 46% from moderate anaemia. Other common morbidities were fever (4%), breast conditions (4.9%), and perineal conditions (4.5%). Life-threatening postpartum morbidities were detected in 7.6% of women--9.7% among those who had deliveries at home and 6.6% among those who had institutional deliveries. None had a fistula. Severe anaemia had a strong correlation with perinatal death [pcaste or tribe [p<0.000, AOR=2.47 (95% CI 1.83-3.33)], and parity of three or more [p<0.000, AOR=1.52 (95% CI 1.18-1.97)]. The correlation with antenatal care was not significant. Perineal conditions were more frequent among women who had institutional deliveries while breast conditions were more common among those who had a perinatal death. This study adds valuable knowledge on postpartum morbidity affecting women in the first few days after delivery in a low-resource setting. Health programmes should invest to ensure that all women receive early postpartum visits after

Radiographic inspection and densitometric evaluation of CP-5 reactor fuel

International Nuclear Information System (INIS)

Staroba, J.F.; Knoerzer, T.W.

1978-02-01

This report covers the radiographic and densitometric techniques used as part of a quality verification program for CP-5 reactor fuel by the Nondestructive Assay Section of the Special Materials Division. Other nondestructive tests used were ultrasonic and gamma-ray spectrometry. The main objectives were to perform a one-hundred percent radiographic inspection of the fuel tubes and to derive a quantitative relationship between fuel thickness and film density with the use of fabricated fuel step wedges. By the use of tangential x-ray techniques, measurements were made of fuel peaks or ''hot spots'' that protruded above the main fuel line. Other general problems in radiographic inspection and solutions for the upgrading of the total radiographic inspection program are also discussed

Comparison of Technological Options for Distributed Generation-Combined Heat and Power in Rajasthan State of India

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Ram Kumar Agrawal

2013-01-01

Full Text Available Distributed generation (DG of electricity is expected to become more important in the future electricity generation system. This paper reviews the different technological options available for DG. DG offers a number of potential benefits. The ability to use the waste heat from fuel-operated DG, known as combined heat and power (CHP, offers both reduced costs and significant reductions of CO2 emissions. The overall efficiency of DG-CHP system can approach 90 percent, a significant improvement over the 30 to 35 percent electric grid efficiency and 50 to 90 percent industrial boiler efficiency when separate production is used. The costs of generation of electricity from six key DG-CHP technologies; gas engines, diesel engines, biodiesel CI engines, microturbines, gas turbines, and fuel cells, are calculated. The cost of generation is dependent on the load factor and the discount rate. It is found that annualized life cycle cost (ALCC of the DG-CHP technologies is approximately half that of the DG technologies without CHP. Considering the ALCC of different DG-CHP technologies, the gas I.C. engine CHP is the most effective for most of the cases but biodiesel CI engine CHP seems to be a promising DG-CHP technology in near future for Rajasthan state due to renewable nature of the fuel.

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

Abortion and sex determination: conflicting messages in information materials in a District of Rajasthan, India.

Science.gov (United States)

Nidadavolu, Vijaya; Bracken, Hillary

2006-05-01

Public information campaigns are an integral component of reproductive health programmes, including on abortion. In India, where sex selective abortion is increasing, public information is being disseminated on the illegality of sex determination. This paper presents findings from a study undertaken in 2003 in one district in Rajasthan to analyse the content of information materials on abortion and sex determination and people's perceptions of them. Most of the informational material about abortion was produced by one abortion service provider, but none by the public or private sector. The public sector had produced materials on the illegality of sex determination, some of which failed to distinguish between sex selection and other reasons for abortion. In the absence of knowledge of the legal status of abortion, the negative messages and strong language of these materials may have contributed to the perception that abortion is illegal in India. Future materials should address abortion and sex determination, including the legal status of abortion, availability of providers and social norms that shape decision-making. Married and unmarried women should be addressed and the participation of family members acknowledged, while supporting independent decisions by women. Sex determination should also be addressed, and the conditions under which a woman can and cannot seek an abortion clarified, using media and materials accessible to low-literate audiences. Based on what we learned in this research, a pictorial booklet and educator's manual were produced, covering both abortion and sex determination, and are being distributed in India.

Actualización del sistema SCADA y de control para los reactores MQ5 y MQ6 de la planta de Pinturas Condor, Sherwin Williams Ecuador

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Jonathan Reinoso

2013-12-01

Full Text Available El presente documento describe la actualización del sistema SCADA para los reactores MQ5 y MQ6 de la planta de Pinturas Condor mediante el software Intouch y la actualización del sistema de control del reactor MQ5 implementado en un controlador lógico programable (PLC de marca SCHNEIDER, además de la arquitectura de control realizada en el proyecto. El sistema SCADA y de control de los reactores permiten la visualización y control de los datos y variables más relevantes durante las diferentes fases de producción de resinas en los reactores MQ5 y MQ6.

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

International Nuclear Information System (INIS)

Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

2014-01-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

2014-07-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

Engineering grouts - materials and applications with specific examples from Ra asthan Atomic Power Pro ect

International Nuclear Information System (INIS)

Singha Roy, P.K.; Sachchidanand; Sukhthankar, K.D.

1978-01-01

Grouting, though not very significant costwise, is an important aspect6 of construction in most of the major projects. According to modern construction technology, grouts have very stringent and diverse uses. The materials and practices generally found in India for grouting, mainly for structural grouts upto the middle of this decade are outlined along with details of specific grouts used in the construction of the twin reactor (440 MWe) of Rajasthan Atomic Power Station, one reactor of which is already operational. Some guidance and tables for selection of grout for a specific use have also been given. (auth.)

Nuclear cycle of thorium and molt salts reactors. PE 5.8

International Nuclear Information System (INIS)

Doubre, H.

2004-01-01

In the framework of the nuclear industry development, many scenario are studied from the standard reactors using enriched uranium to the IV generation reactors. The study of new systems for the future of the nuclear needs to develop new simulation tools. The research programs of the IPN of Orsay are presented. (A.L.B.)

Annual report of the Department of Atomic Energy 1977-78

International Nuclear Information System (INIS)

1978-01-01

The activities during the financial year 1977-78 of the research organizations and laboratories, various projects underway and public sector undertakings of the Department of Atomic Energy (India) have been reported. The R and D Work of the Bhabha Atomic Research Centre, Bombay and Reactor Research Centre, Kalpakkam in the fields of nuclear physics, radio- and radiation chemistry and other physical sciences; biological sciences including nuclear medicine, food irradiation and plant breeding by radiation mutation; reactor engineering and application of radiation and radioisotopes has been surveyed. The progress of heavy water projects, MHD project, nuclear power plant projects and 100 Mw thermal research reactor R-5 project has been described. Performance of Tarapur and Rajasthan Atomic Power Stations, Nuclear Fuel Complex and Electronics Corporation of India Ltd., both at Hyderabad, Uranium Corporation of India Ltd. at Jaduguda and Indian Rare Earths Ltd. has been reported. Major achievements during the period of report are : (1) completion of construction work of the Power Reactor Fuel Reprocessing Plant at Tarapur and (2) Commissioning of the Variable Energy Cyclotron, Calcutta for the internal circulating beam of alpha particles. (M.G.B.)

Shore and off-shore monitoring of Rana Pratap Sagar lake

International Nuclear Information System (INIS)

Verma, P.C.; Roy, Alpana; Hegde, A.G.

2005-01-01

The Rana Pratap Sagar (RPS) is a man made fresh water reservoir on the river Chambal and is located about 65 km away from Kota city in the state of Rajasthan. It is a balancing reservoir between Gandhi Sagar, upstream and Jawahar Sagar, downstream. On the eastern bank of RPS there exist Rajasthan Atomic Power Station (RAPS) Site, comprising of four operating pressurized Heavy Water Reactors (PHWRs) of 220 MWe each. The low level radioactive effluents from RAPS, are released in controlled manner to RPS. On release the effluents undergo dispersion in the aquatic environment of RPS and may reach man through, various pathways. Being fresh water reservoir, considerable emphasis is laid on the aquatic monitoring aspects relating to RPS and downstream reservoirs. This paper presents the monitored levels of radioactivity prevailing in the aquatic samples collected from near shore and off shore locations of RPS. (author)

TREATMENT OF METHANOLIC WASTEWATER BY ANAEROBIC DOWN-FLOW HANGING SPONGE (ANDHS) REACTOR AND UASB REACTOR

Science.gov (United States)

Sumino, Haruhiko; Wada, Keiji; Syutsubo, Kazuaki; Yamaguchi, Takashi; Harada, Hideki; Ohashi, Akiyoshi

Anaerobic down-flow hanging sponge (AnDHS) reactor and UASB reactor were operated at 30℃ for over 400 days in order to investigate the process performance and the sludge characteristics of treating methanolic wastewater (2 gCOD/L). The settings OLR of AnDHS reactor and of UASB reactor were 5.0 -10.0 kgCOD/m3/d and 5.0 kgCOD/m3/d. The average of the COD removal demonstrated by both reactors were over 90% throughout the experiment. From the results of methane producing activities and the PCR-DGGE method, most methanol was directly converted to methane in both reactors. The conversion was carried out by different methanogens: one closely related to Methanomethylovorans hollandica in the AnDHS retainted sludge and the other closely related to Methanosarcinaceae and Metanosarciales in the UASB retainted sludge.

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

Reactor lattice codes

International Nuclear Information System (INIS)

Kulikowska, T.

2001-01-01

The description of reactor lattice codes is carried out on the example of the WIMSD-5B code. The WIMS code in its various version is the most recognised lattice code. It is used in all parts of the world for calculations of research and power reactors. The version WIMSD-5B is distributed free of charge by NEA Data Bank. The description of its main features given in the present lecture follows the aspects defined previously for lattice calculations in the lecture on Reactor Lattice Transport Calculations. The spatial models are described, and the approach to the energy treatment is given. Finally the specific algorithm applied in fuel depletion calculations is outlined. (author)

Computerized reactor power regulation with logarithmic controller

International Nuclear Information System (INIS)

Gossanyi, A.; Vegh, E.

1982-11-01

A computerized reactor control system has been operating at a 5 MW WWR-SM research reactor in the Central Research Institute for Physics, Budapest, for some years. This paper describes the power controller used in the SPC operating mode of the system, which operates in a 5-decade wide power range with +-0.5% accuracy. The structure of the controller easily limits the minimal reactor period and produces a reactor transient with constant period if the power demand changes. (author)

The Simulator Development for RDE Reactor

Science.gov (United States)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

Statistical analysis and modelling of in-reactor diametral creep of Zr-2.5Nb pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Jyrkama, Mikko I., E-mail: mjyrkama@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada); Bickel, Grant A., E-mail: grant.bickel@cnl.ca [Canadian Nuclear Laboratories, Chalk River Laboratories, Chalk River, ON, Canada K0J 1J0 (Canada); Pandey, Mahesh D., E-mail: mdpandey@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada)

2016-04-15

Highlights: • New and simple statistical model of pressure tube diametral creep. • Based on surveillance data of 328 pressure tubes from eight different CANDU reactors. • Uses weighted least squares (WLS) to regress out operating conditions. • The shape of the diametral creep profiles are predicted very well. • Provides insight and relative ranking of strain behaviour of in-service tubes. - Abstract: This paper presents the development of a simplified regression approach for modelling the diametral creep over time in Zr-2.5 wt% Nb pressure tubes used in CANDU reactors. The model is based on a large dataset of in-service inspection data of 328 different pressure tubes from eight different CANDU reactor units. The proposed weighted least squares (WLS) regression model is linear in time as a function of flux and temperature, with a temperature-dependent variance function. The model predicts the shape of the observed diametral creep profiles very well, and is useful not merely for prediction, but also for assessing tube-to-tube variability and manufacturing properties among the inspected tubes.

Recommendations of the Reactor Safety Commission (RSK) 1981-1983. Vol. 5

International Nuclear Information System (INIS)

1984-02-01

Since the reorganization of the RSK in 1971 the recommendations are published in the ''Bundesanzeiger'' (BAnz). This makes the work of the RSK transparent for the public. After publication of the recommendations in the Bundesanzeiger, the Office of the RSK publishes them once more as a summary report by order of the Federal Minister of Interior. The reports are divided generally into two parts: Part I contains the recommendations which are given by the RSK; part II contains the official notices concerning the RSK. There is also a subject index. In addition volume 5 contains as part III all papers which were held on the occasion of the 25th anniversary of the Reactor Safety Commission. (orig./HP) [de

Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5; Calculo de activacion neutronica de barras de control de un reactor nuclear, utilizando MCNP5

Energy Technology Data Exchange (ETDEWEB)

Pena V, J.D.

2016-07-01

The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)

Comparative assessment of Oral Hygiene and Periodontal status among children who have Poliomyelitis at Udaipur city, Rajasthan, India

Science.gov (United States)

Nagarajappa, Ramesh; Sharda, Archana; Asawa, Kailash; Tak, Aniruddh; Jalihal, Sagar

2012-01-01

Objective: To assess and compare the oral hygiene and periodontal status among children with Poliomyelitis having upper limb disability, lower limb disability and both upper and lower disability at Udaipur city, Rajasthan, India. Study design: Total sample comprised of 344 Poliomyelitis children (upper limb disability: 33.4%; lower limb disability: 33.7%; both upper and lower limb disability: 32.9%) in the age group of 12-15 years. Clinical examination included recording Simplified Oral Hygiene Index and Community Periodontal Index. Analysis of variance (ANOVA), multiple logistic and stepwise linear regression were used for statistical analysis. Results: The mean OHI-S (2.52±1.05) score was found to be highest among children who had both upper and lower limb disability (poral hygiene and periodontal status was limb involved in the disability. Conclusion: The results of the study depicted an overall poor oral hygiene and periodontal status of the group. It was recognized that limbs involved in the disability had an impact on the oral hygiene and periodontal condition. The situation in this specialized population draws immediate attention for an integrated approach in improving the oral health and focus towards extensive research. Key words:Poliomyelitis, upper limb disability, lower limb disability, oral hygiene, periodontal status. PMID:22549671

Variation of annual effective dose due to radon level in indoor air in Marwar region of Rajasthan, India

Energy Technology Data Exchange (ETDEWEB)

Rani, Asha, E-mail: ashasachdeva78@gmail.com [Department of Applied Science, Ferozepur College of Engineering and Technology, Farozshah, Ferozepur-142052, Punjab (India); Mittal, Sudhir, E-mail: sudhirmittal03@gmail.com [Department of Applied Sciences, Punjab Technical University, Jalandhar-144601, Punjab (India); Mehra, Rohit [Department of Physics, Dr. B.R.Ambedkar National Institute of Technology, Jalandhar-144011 (India)

2015-08-28

In the present work, indoor radon and thoron measurements have been carried out from different locations of Jodhpur and Nagaur districts of Northern Rajasthan, India using RAD7, a solid state alpha detector. The radon and thoron concentration in indoor air varies from 8.75 to 61.25 Bq m{sup −3} and 32.7 to 147.2 Bq m{sup −3} with the mean value of 32 and 73 Bq m{sup −3} respectively. The observed indoor radon concentration values are well below the action level recommended by International Commission on Radiological Protection (200-300 Bq m{sup −3}) and Environmental Protection Agency (148 Bq m{sup −3}). The survey reveals that the thoron concentration values in the indoor air are well within the International Commission on Radiological Protection (2005). The calculated total annual effective dose due to radon level in indoor air varies from 0.22 to 1.54 mSv y{sup −1} with the mean value of 0.81 mSv y{sup −1} which is less than even the lower limit of action level 3-10 mSv y{sup −1} recommended by International Commission on Radiological Protection (2005)

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Dosimetry and radiation shielding at the RA reactor, Annual report 1975, Annex 5

International Nuclear Information System (INIS)

Ninkovic, M.

1976-01-01

In the working environment at the RA reactor, the level of gamma radiation is measured continuously by the built-in stationary system. According to the needs, measurement are done in the reactor hall every day. The level of gamma radiation is measured separately in typical points when the reactor is operated at nominal power and during intervals between two operating campaigns. The level of neutron radiation is measured according to the needs by means of a mobile spherical neutron detector. These measurements are done in the reactor hall around the horizontal experimental channels. Measured values of neutron radiation are three times lower than the relevant levels of gamma radiation [sr

Impact of proposed research reactor standards on reactor operation

Energy Technology Data Exchange (ETDEWEB)

Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

1974-07-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Impact of proposed research reactor standards on reactor operation

International Nuclear Information System (INIS)

Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

1974-01-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Isothermal calorimeter for reactor radiation dosimetry

Energy Technology Data Exchange (ETDEWEB)

Radak, B; Markovic, V [Institute of Nuclear Sciences Boris Kidric, Odeljenje za radijacionu hemiju, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

An isothermal calorimeter with thermistors for measuring absorbed dose rates from 10{sup 4}-5-6.10{sup 5} rad/h in reactor experimental holes has been designed. A kinetics method for determining the equilibrium temperature difference has been developed, and its application in isothermal calorimetry proved. The expected accuracy in measurements within {+-} 2-5% has been proved by measurements carried out in the reactor. Some data obtained by measurements in the reactor RA are presented (author)

The Role of Transdisciplinary Approach and Community Participation in Village Scale Groundwater Management: Insights from Gujarat and Rajasthan, India

Directory of Open Access Journals (Sweden)

Basant Maheshwari

2014-11-01

Full Text Available Sustainable use of groundwater is becoming critical in India and requires effective participation from local communities along with technical, social, economic, policy and political inputs. Access to groundwater for farming communities is also an emotional and complex issue as their livelihood and survival depends on it. In this article, we report on transdisciplinary approaches to understanding the issues, challenges and options for improving sustainability of groundwater use in States of Gujarat and Rajasthan, India. In this project, called Managed Aquifer Recharge through Village level Intervention (MARVI, the research is focused on developing a suitable participatory approach and methodology with associated tools that will assist in improving supply and demand management of groundwater. The study was conducted in the Meghraj watershed in Aravalli district, Gujarat, and the Dharta watershed in Udaipur district, Rajasthan, India. The study involved the collection of hydrologic, agronomic and socio-economic data and engagement of local village and school communities through their role in groundwater monitoring, field trials, photovoice activities and education campaigns. The study revealed that availability of relevant and reliable data related to the various aspects of groundwater and developing trust and support between local communities, NGOs and government agencies are the key to moving towards a dialogue to decide on what to do to achieve sustainable use of groundwater. The analysis of long-term water table data indicated considerable fluctuation in groundwater levels from year to year or a net lowering of the water table, but the levels tend to recover during wet years. This provides hope that by improving management of recharge structures and groundwater pumping, we can assist in stabilizing the local water table. Our interventions through Bhujal Jankaars (BJs, (a Hindi word meaning “groundwater informed” volunteers, schools

Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

International Nuclear Information System (INIS)

Lai, W.; Collins, E.K.

1977-01-01

This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

Trends and homogeneity of monthly, seasonal, and annual rainfall over arid region of Rajasthan, India

Science.gov (United States)

Meena, Hari Mohan; Machiwal, Deepesh; Santra, Priyabrata; Moharana, Pratap Chandra; Singh, D. V.

2018-05-01

Knowledge of rainfall variability is important for regional-scale planning and management of water resources in agriculture. This study explores spatio-temporal variations, trends, and homogeneity in monthly, seasonal, and annual rainfall series of 62 stations located in arid region of Rajasthan, India using 55 year (1957-2011) data. Box-whisker plots indicate presence of outliers and extremes in annual rainfall, which made the distribution of annual rainfall right-skewed. Mean and coefficient of variation (CV) of rainfall reveals a high inter-annual variability (CV > 200%) in the western portion where the mean annual rainfall is very low. A general gradient of the mean monthly, seasonal, and annual rainfall is visible from northwest to southeast direction, which is orthogonal to the gradient of CV. The Sen's innovative trend test is found over-sensitive in evaluating statistical significance of the rainfall trends, while the Mann-Kendall test identifies significantly increasing rainfall trends in June and September. Rainfall in July shows prominently decreasing trends although none of them are found statistically significant. Monsoon and annual rainfall show significantly increasing trends at only four stations. The magnitude of trends indicates that the rainfall is increasing at a mean rate of 1.11, 2.85, and 2.89 mm year-1 in August, monsoon season, and annual series. The rainfall is found homogeneous over most of the area except for few stations situated in the eastern and northwest portions where significantly increasing trends are observed. Findings of this study indicate that there are few increasing trends in rainfall of this Indian arid region.

IEA-R1 reactor core simulation with RELAP5 code

International Nuclear Information System (INIS)

Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

2005-01-01

This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

What occurred in the reactors

International Nuclear Information System (INIS)

Kudo, Kazuhiko

2013-01-01

Described is what occurred in the reactors of Fukushima Daiichi Nuclear Power Plant at the Tohoku earthquake and tsunami (Mar. 11, 2011) from the aspect of engineering science. The tsunami attacked the Plant 1 hr after the quake. The Plant had reactors in buildings no.1-4 at 10 m height from the normal sea level which was flooded by 1.5-5.5 m high wave. All reactors in no.1-6 in the Plant were the boiling water type, and their core nuclear reactions were stopped within 3 sec due to the first quake by control rods inserted automatically. Reactors in no.1-5 lost their external AC power sources by the breakdown and subsequent submergence (no.1-4) of various equipments and in no.1, 2 and 4, the secondary DC power was then lost by the battery death. Although the isolation condenser started to cool the reactor in no.1 after DC cut, its valve was then kept closed to heat up the reactor, leading to the reaction of heated Zr in the fuel tube and water to yield H 2 which was accumulated in the building: the cause of hydrogen explosion on 12th. The reactor in no.2 had the reactor core isolation cooling system (RCIC) which operated normally for few hrs, then probably stopped to heat up the reactor, resulting in meltdown of the core but no explosion occurred because of the opened door of the blowout panel on the wall by the blast of no.1 explosion. The reactor in no.3 had RCIC and high pressure coolant injection system, but their works stopped to result in the core damage and H 2 accumulation leading to the explosion on 14th. The reactor in no.4 had not been operated because of its periodical annual examination, but was explored on 15th, of which cause was thought to be due to backward flow of H 2 from no.3. Finally, the author discusses about this accident from the industrial aspect of the design of safety level (defense in depth) on international views, and problems and tasks given. (T.T.)

Drainage Characteristics of Tectonically Active Areas: An Example from Rajasthan, India

Directory of Open Access Journals (Sweden)

SWATI JAIN

2010-06-01

Full Text Available The morphotectonic studies help in deciphering the role of tectonics and neotectonics in morphological evolution of drainage basins. On the basis of remote sensing technique, the relationship between morphology and tectonics have been investigated in Bundi-Indergarh sector of southeast Rajasthan. The area selected for present study is drained by Mej river and its tributaries and occupies the southeastern part of the Aravalli Mountain Range (AMR. The course of Mej river is mostly controlled by the Great Boundary Thrust (GBT and associated tectonic elements. GBT separates the folded, faulted and metamorphosed older rocks of the AMR in the west and relatively undeformed Vindhyan rocks in the east. This study has been carried out using digital and hard copy product of IRS 1C/1D LISS III geocoded FCC data. The morphometric and morphotectonic aspects have been studied for identification of present day tectonic activities in the area. The remote sensing data interpretation indicates that the landforms of the area are structurally controlled and mainly covered by linear and parallel strike ridges and valleys. These valleys indicate sign of stream rejuvenation and occasional presence of dynamic ravines. General morphometric parameters, bifurcation ratio, stream length and shape parameters have been computed. Longitudinal river profiles can be quantified by normalizing the elevation and the distance along rivers. Several parameters such as profile shape (concavity, gradient fluctuations, river grade and valley incision have been derived from longitudinal river profile. These quantified parameters and their interrelations are useful in comparing different drainage basins and also help drawing inferences on neotectonism. The computed values suggest that the area is covered by resistant rock and drainage network, affected by tectonic distur-bance. The valley floor ratio is very low, indicating channel down cutting vis-a-vis ground uplift. The gradient index

Necessity of research reactors

International Nuclear Information System (INIS)

Ito, Tetsuo

2016-01-01

Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

Department of Atomic Energy [India]: Annual report 1979-1980

International Nuclear Information System (INIS)

1980-01-01

The work of the research establishments, projects undertaken and public sector undertakings of the Department of Atomic Energy during the financial year 1979-80 is surveyed. The research and development activities of the Bhabha Atomic Research Centre at Bombay, the Reactor Research Centre at Kalpakkam, the Tata Institute of Fundamental Research at Bombay, the Saha Institute of Nuclear Physics at Calcutta and the Tata Memorial Centre at Bombay are described. An account of the progress of heavy water production plant projects, the Madras and Narora Atomic Power Projects, the MHD project and the 100 MW thermal research reactor R-5 Project at Trombay is given. Performance of the Tarapur and Rajasthan Atomic Power Stations, Nuclear Fuel Complex at Hyderabad, Atomic Minerals Division, ISOMED (the radiation sterilisation plant for medical products) at Bombay, the Indian Rare Earths Ltd., the Uranium Corporation of India Ltd., and the Electronics Corporation of India Ltd., Hyderabad is reported. (M.G.B.)

Oxidation and deuterium uptake of Zr-2.5Nb pressure tubes in CANDU-PHW reactors

International Nuclear Information System (INIS)

Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

1989-01-01

Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm 2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

International Nuclear Information System (INIS)

Brand, P.; Wiechers, W.K.

1974-01-01

The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

Study of sources, dose contribution and control measures of Argon-41 at Kaiga Generating Station

International Nuclear Information System (INIS)

Venkata Ramana, K.; Shrikrishna, U.V.; Manojkumar, M.; Ramesh, R.; Madhan, V.; Varadhan, R.S.

2001-01-01

Air is used as a medium for cooling calandria vault and thermal shield systems in the earlier Pressurised Heavy Water Reactors (Rajasthan Atomic Power Station and Madras Atomic Power Station) in India. This leads to production of significant quantity of 41 Ar in calandria vault and thermal shield cooling systems due to neutron activation of 40 Ar present in air (∼1% v/v). The presence of 41 Ar in reactor building contributes significant external doses to plant personnel during reactor operation and the release of this radionuclide to the environment result in dose to the public in the vicinity of the plants. An attempt is made to eliminate Argon-41 production in Indian standard Pressurised Heavy Water Reactors (Narora Atomic Power Station, Kakrapar Atomic Power Station, Kaiga Generating Station -1 and 2 and Rajasthan Atomic Power Station-3 and 4), by filling the calandria vault with demineralized water and providing a separate Annulus Gas Monitoring System (AGMS) for detecting leaks from calandria tube or pressure tube using Carbon dioxide as a medium. However, 41 Ar is produced in the Annulus Gas Monitoring System, Primary Heat Transport cover gas system and moderator cover gas system due to ingress of air into the systems during operational transients or due to trace quantity of air present as an impurity in the gases used for the above systems. A study was conducted to identify and quantify the sources of 41 Ar in the work areas. This report brings out the sources of 41 Ar, reasons for 41 Ar production and the results of the measures incorporated to reduce the presence of 41 Ar in the above systems. (author)

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor, United States of America

Energy Technology Data Exchange (ETDEWEB)

Boing, L E; Bhattacharyya, S K [Technology Development Division, Decommissioning Program, Argonne National Laboratory, Argonne, IL (United States)

2002-02-01

The aging of research reactors worldwide has resulted in a heightened awareness in the international decommissioning community of the timeliness to review and address the needs of research reactor operators in planning for and eventually performing the decommissioning of these types of facilities. Many reactors already undergoing decommissioning can be used as test beds for evaluating enhanced or new/innovative technologies for decommissioning; it is possible that new techniques could be made available for future research reactor-decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the decommissioners in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to their use in actual research reactor decommissioning. The decommissioning of the CP-5 Research Reactor located at the ANL-East Site has been completed. In this paper we present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors. In addition, details are provided on other related U.S. D and D activities, which may be useful to the international research reactor D and D community. (author)

Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor, United States of America

International Nuclear Information System (INIS)

Boing, L.E.; Bhattacharyya, S.K.

2002-01-01

The aging of research reactors worldwide has resulted in a heightened awareness in the international decommissioning community of the timeliness to review and address the needs of research reactor operators in planning for and eventually performing the decommissioning of these types of facilities. Many reactors already undergoing decommissioning can be used as test beds for evaluating enhanced or new/innovative technologies for decommissioning; it is possible that new techniques could be made available for future research reactor-decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the decommissioners in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to their use in actual research reactor decommissioning. The decommissioning of the CP-5 Research Reactor located at the ANL-East Site has been completed. In this paper we present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors. In addition, details are provided on other related U.S. D and D activities, which may be useful to the international research reactor D and D community. (author)

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

Occupational radiation exposure at Commercial Nuclear Power reactors 1983. Volume 5. Annual report

International Nuclear Information System (INIS)

Brooks, B.G.

1985-03-01

This report presents an updated compilation of occupational radiation exposure at commercial nuclear power reactors for the years 1969 through 1983. The summary based on information received from the 75 light-water-cooled reactors (LWRs) and one high temperature gas-cooled reactor (HTGR). The total number of personnel monitored at LWRs in 1983 was 136,700. The number of workers that received measurable doses during 1983 and 85,600 which is about 1000 more than that found in 1982. The total collective dose at LWRs for 1983 is estimated to be 56,500 man-rems (man-cSv), which is about 4000 more man-rems (man-cSv) than that reported in 1982. This resulted in the average annual dose for each worker who received a measurable dose increasing slightly to 0.66 rems (cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv) to a value of 753 man-rems (man-cSv). The collective dose per megawatt of electricity generated by each reactor also increased slightly to an average value of 1.7 man-rems (man-cSv) per megawatt-year. Health implications of these annual occupational doses are discussed

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Directory of Nuclear Research Reactors 1994

International Nuclear Information System (INIS)

1995-08-01

The Directory of Nuclear Research Reactors is an output of the Agency's computerized Research Reactor Data Base (RRDB). It contains administrative, technical and utilization information on research reactors known to the Agency at the end of December 1994. The data base converted from mainframe to PC is written in Clipper 5.0 and the publication generation system uses Excel 4. The information was collected by the Agency through questionnaires sent to research reactor owners. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the RRDB. This system contains all the information and data previously published in the Agency's publication, Directory of Nuclear Research Reactor, as well as updated information

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

A multi-purpose reactor

International Nuclear Information System (INIS)

Changwen Ma

2000-01-01

An integrated natural circulation self pressurized reactor can be used for sea water desalination, electrogeneration, ship propulsion and district or process heating. The reactor can be used for ship propulsion because it has following advantages: it is a integrated reactor. Whole primary loop is included in a size limited pressure vessel. For a 200 MW reactor the diameter of the pressure vessel is about 5 m. It is convenient to arranged on a ship. Hydraulic driving facility of control rods is used on the reactor. It notably decreases the height of the reactor. For ship propulsion, smaller diameter and smaller height are important. Besides these, the operation reliability of the reactor is high enough, because there is no rotational machine (for example, circulating pump) in safety systems. Reactor systems are simple. There are no emergency water injection system and boron concentration regulating system. These features for ship propulsion reactor are valuable. Design of the reactor is based on existing demonstration district heating reactor design. The mechanic design principles are the same. But boiling is introduced in the reactor core. Several variants to use the reactor as a movable seawater desalination plant are presented in the paper. When the sea water desalination plant is working to produce fresh water, the reactor can supply electricity at the same time to the local electricity network. Some analyses for comprehensive application of the reactor have been done. Main features and parameters of the small (Thermopower 200 MW) reactor are given in the paper. (author)

Final report for the 5th surveillance test of the reactor pressure vessel material (capsule Y) of Yonggwang Nuclear Power Plant unit 2

Energy Technology Data Exchange (ETDEWEB)

Lee, Sam Lai; Kim, ByoungChul; Chang, Kee Ok (and others)

2006-02-15

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X, W and Y are 5.777E+18, 1.5371E+19, 3.7634E+19, 4.3045E+19, and 4.8662E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.953 for the 1st through 5th testing and the calculational uncertainty,7.2% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.659E+19n/cm{sup 2} based on the end of 13th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 48, 56 and 64EFPY would reach 3.625E+19, 5.293E+19, 6.127E+19 and 6.960E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the Pressurized Thermal Shock(PTS) during the operation until design life.

Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1989-01-01

The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs

Farewell to a reactor

International Nuclear Information System (INIS)

Skanborg, P.

1976-01-01

Denmark's second reactor, DR 2, whose first criticality took place the night of 18/19 December 1958 was shut down for the last time on 31 October 1975. It was a light-water moderrated and cooled reactor of swimming-pool type with a thermal power of 5 MW, using 90% enriched uranium. The operation is described. The reactor and auxiliary equipment are now being put 'in store' - all fuel elements sent for reprocessing, the reactor tank and cooling circuits emptied, and a lead shielding placed over the tank opening. The rest of the equipment will remain in place. (B.P.)

Fusion reactor problems

International Nuclear Information System (INIS)

Carruthers, R.

It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10 14 cm -3 sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm -2 , temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

Safety operation of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

2001-01-01

There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

Energy Technology Data Exchange (ETDEWEB)

El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

2017-11-15

In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

Brazilian multipurpose reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-07-01

The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U{sub 3}Si{sub 2} dispersion-type Al having a density of up to 3.5 gU/cm{sup 3} and enrichment of 19.75% by weight of {sup 23{sup 5}}U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive

Brazilian multipurpose reactor

International Nuclear Information System (INIS)

2014-01-01

The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U 3 Si 2 dispersion-type Al having a density of up to 3.5 gU/cm 3 and enrichment of 19.75% by weight of 23 5 U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive facilities are

Analyzing the loss of coolant accident in PWR nuclear reactors with elevation change in cold leg by RELAP5/MOD3.2 system code

International Nuclear Information System (INIS)

Kheshtpaz, H.; Alison, C.

2006-01-01

As, the Russian designed VVER-1000 reactor of the Bushehr Nuclear Power Plant by taking into account the change from German technology to that of Russian technology, and with the design of elevation change in the cold legs has been developed; therefore safety assessment of these systems for loss of coolant accident in elevation change in the cold legs and comparison results for non change elevation in the cold legs for a typical reactor (normal design of nuclear reactors) is the main important factor to be considered for the safe operation. In this article, the main objective is the simulation of the loss of coolant accident scenario by the RELAP5/MOD3.2 code in two different cases; first, the elevation change in the cold legs, and the second, non change in it. After comparing and analyzing these two code calculations the results have been generalized for a new design feature of Bushehr reactor. The design and simulation of the elevation change in the cold legs process with RELAP5/MOD3.2 code for PWR reactor is performed for the first time in the country, where it is introducing several important results in this respect

Spectral pathways for exploration of secondary uranium: An investigation in the desertic tracts of Rajasthan and Gujarat, India

Science.gov (United States)

Bharti, Rishikesh; Kalimuthu, R.; Ramakrishnan, D.

2015-10-01

This study aims at identifying potential zones of secondary uranium enrichment using hyperspectral remote sensing, γ-ray spectrometry, fluorimetry and geochemical techniques in the western Rajasthan and northern Gujarat, India. The investigated area has suitable source rocks, conducive past-, and present-climate that can facilitate such enrichment. This enrichment process involves extensive weathering of uranium bearing source rocks, leaching of uranyl compounds in groundwater, and their precipitation in chemical deltas along with duricrusts like calcretes and gypcretes. Spatial distribution of groundwater calcretes (that are rich in Mg-calcite) and gypcretes (that are rich in gypsum) along palaeochannels and chemical deltas were mapped using hyperspectral remote sensing data based on spectral absorptions in 1.70 μm, 2.16 μm, 2.21 μm, 2.33 μm, 2.44 μm wavelength regions. Subsequently based on field radiometric survey, zones of U anomalies were identified and samples of duricrusts and groundwater were collected for geochemical analyses. Anomalous concentration of U (2345.7 Bq/kg) and Th (142.3 Bq/kg) are observed in both duricrusts and groundwater (U-1791 μg/l, Th-34 μg/l) within the palaeo-delta and river confluence. The estimated carnotite Solubility Index also indicates the secondary enrichment of U and the likelihood of occurrence of an unconventional deposit.

Comparative assessment of oral hygiene and periodontal status among children who have Poliomyelitis at Udaipur city, Rajasthan, India.

Science.gov (United States)

Tak, Mridula; Nagarajappa, Ramesh; Sharda, Archana; Asawa, Kailash; Tak, Aniruddh; Jalihal, Sagar

2012-11-01

To assess and compare the oral hygiene and periodontal status among children with Poliomyelitis having upper limb disability, lower limb disability and both upper and lower disability at Udaipur city, Rajasthan, India. Total sample comprised of 344 Poliomyelitis children (upper limb disability: 33.4%; lower limb disability: 33.7%; both upper and lower limb disability: 32.9%) in the age group of 12-15 years. Clinical examination included recording Simplified Oral Hygiene Index and Community Periodontal Index. Analysis of variance (ANOVA), multiple logistic and stepwise linear regression were used for statistical analysis. The mean OHI-S (2.52 ± 1.05) score was found to be highest among children who had both upper and lower limb disability (phealthy sextants were found among those with only lower limb disability (4.53 ± 2.05) and among those with both upper and lower limb disability (0.77 ± 1.39), respectively (poral hygiene and periodontal status was limb involved in the disability. The results of the study depicted an overall poor oral hygiene and periodontal status of the group. It was recognized that limbs involved in the disability had an impact on the oral hygiene and periodontal condition. The situation in this specialized population draws immediate attention for an integrated approach in improving the oral health and focus towards extensive research.

Technical mechanics in constructional reactor safety

International Nuclear Information System (INIS)

Matthees, W.

1979-01-01

Reactor safety is based on close cooperation between a number of technical and scientific disciplines; most problems of reactor technology can be solved with the aid of technical mechanics. At the 5th International Conference on Structural Mechanics in Reactor Technology (5th SMIRT), one of the biggest conferences in the field of applied technical mechanics, about 800 papers were read giving the latest state of knowledge in the field of constructional reactor safety. The main subject of the conference was the analysis of material behaviour under high loads; the information and methods of these analysis go far beyond what is required in the conventional field. (orig./UA) [de

Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D; Estudo termofluidodinâmico de reatores nucleares avançados de alta temperatura utilizando o RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Scari, Maria Elizabeth

2017-07-01

Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF{sub 2}, the LiF-BeF{sub 2}, also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO

Task 24: Dynamic analysis of Kozloduy NPP unit 5 structures: Reactor building

International Nuclear Information System (INIS)

Zola, M.

1999-01-01

This report refers to the activities of a sub-contract to the Project RER/9/046, awarded to ISMES by the International Atomic Energy Agency (IAEA) of Vienna, to compare the results obtained from the experimental activities performed under previous contract by ISMES with those coming from analytical studies performed in the framework of the Coordinated Research Programme (CRP) on 'Benchmark Study for the Seismic Analysis and Testing of WWER-type Nuclear Power Plants' by other Institutions, relevant to Kozloduy Unit 5 reactor building. After a brief introduction to the problem in Chapter 1, the identification of the comparison positions and reference directions is given in Chapter 3. A very quick description of the performed experimental tests is given in Chapter 4, whereas the characteristics of both experimental and analytical data are presented in Chapter 5. The data processing procedures are reported in Chapter 6 and some simple remarks are given in Chapter 7. (author)

Estimation of doses received by operators in the 1958 RB reactor accident using the MCNP5 computer code simulation

Directory of Open Access Journals (Sweden)

Pešić Milan P.

2012-01-01

Full Text Available A numerical simulation of the radiological consequences of the RB reactor reactivity excursion accident, which occurred on October 15, 1958, and an estimation of the total doses received by the operators were run by the MCNP5 computer code. The simulation was carried out under the same assumptions as those used in the 1960 IAEA-organized experimental simulation of the accident: total fission energy of 80 MJ released in the accident and the frozen positions of the operators. The time interval of exposure to high doses received by the operators has been estimated. Data on the RB1/1958 reactor core relevant to the accident are given. A short summary of the accident scenario has been updated. A 3-D model of the reactor room and the RB reactor tank, with all the details of the core, created. For dose determination, 3-D simplified, homogenised, sexless and faceless phantoms, placed inside the reactor room, have been developed. The code was run for a number of neutron histories which have given a dose rate uncertainty of less than 2%. For the determination of radiation spectra escaping the reactor core and radiation interaction in the tissue of the phantoms, the MCNP5 code was run (in the KCODE option and “mode n p e”, with a 55-group neutron spectra, 35-group gamma ray spectra and a 10-group electron spectra. The doses were determined by using the conversion of flux density (obtained by the F4 tally in the phantoms to doses using factors taken from ICRP-74 and from the deposited energy of neutrons and gamma rays (obtained by the F6 tally in the phantoms’ tissue. A rough estimation of the time moment when the odour of ozone was sensed by the operators is estimated for the first time and given in Appendix A.1. Calculated total absorbed and equivalent doses are compared to the previously reported ones and an attempt to understand and explain the reasons for the obtained differences has been made. A Root Cause Analysis of the accident was done and

Estimation of radon concentration in soil and groundwater samples of Northern Rajasthan, India

International Nuclear Information System (INIS)

Mittal, Sudhir; Asha Rani; Mehra, Rohit

2015-01-01

In the present investigation, analysis of radon concentration in 20 water and soil samples collected from different locations of Bikaner and Jhunjhunu districts of Rajasthan, India has been carried out by using RAD7 an electronic Radon detector. The water samples are taken from hand pumps and tube wells having depths ranging from 50 to 600 feet. All the soil gas measurements have been carried out at 100 cm depth. The measured radon concentration in water samples lies in the range from 0.50 to 22 Bq l -1 with the mean value of 4.42 Bq l -1 . Only in one water sample radon concentration is found to be higher than the safe limit of 11 Bq l -1 recommended US Environmental Protection Agency (USEPA, 1991). The measured value of radon concentration in all ground water samples is within the safe limit from 4 to 40 Bq l -1 recommended by United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR, 2008). The total annual effective dose estimated due to radon concentration in water ranges from 1.37 to 60 μSV y -1 with the mean value of 12.08 μSV y -1 . The total annual effective dose from all locations of our studied area is found to be well within the safe limit 0.1 mSv y -1 recommended by World Health Organization (WHO, 2004) and European Council (ED, 1998). Radon measurement in soil samples varies from 941 to 10050 Bq m -3 with the mean value of 4561 Bq m -3 , The radon concentration observed from the soil samples from our study area lies within the range reported by other investigators. Moreover a positive correlation of radon concentration in water with soil samples has been observed. It was observed that the soil and water of Bikaner and Jhunjhunu districts are suitable for drinking and construction purpose without posing any health hazard. (author)

Modification to 200 MW(e) CANDU for improved dynamic behaviour

International Nuclear Information System (INIS)

Chamany, B.F.; Murthy, L.G.K.; Ray, R.N.

1976-01-01

Rajasthan Atomic Power Station is inherently suitable for base load operation. Its control philosophy is based on turbine following the reactor. However, due to load fluctuations and inherent limitation of the control system, there had been considerable number of outages of the station. This limitation is further enhanced by improper choice of the operating pressure range of the boilers. Besides, existing fuel design does not permit thermal cycling and hence there is no use in attempting to make the reactor follow the turbine. Design modifications have been suggested for incorporation in the further 200 MW(e) systems. The method adopted is complete decoupling of the reactor from the load. Dynamic behaviour of the station with the suggested modifications and its comparison with the existing situation has been brought out. (author)

Fusion reactor materials

International Nuclear Information System (INIS)

Anon.

1977-01-01

The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Martinc, R.; Cupac, S.; Stanic, A.

1990-01-01

RA reactor was not operated during the past five years due to the renewal and reconstruction of the reactor systems, which in underway. In the period from 1986-1990, reactor was operated only 144 MWh in 1986, for the need of testing the reactor systems and possibility of irradiating 125 I. Reactor will not be operated in 1991 because of the exchange of complete instrumentation which is planned to be finished by the end of 1991. It is expected to start operation in May 1992. That is why this annex includes the plan of reactor operation for period of nine months starting from from the moment of start-up. It is planned to operate the reactor at 0.02 MW power first three months, to increase the power gradually and reach 3.5 MW after 8 months of operation. It is foreseen to operate the reactor at 4.7 MW from the tenth month on [sr

Engineering reactors for catalytic reactions

Indian Academy of Sciences (India)

126, No. 2, March 2014, pp. 341–351. c Indian Academy of Sciences. ... enhancement was realized by catalyst design, appropriate choice of reactor, better injection and .... Gas–liquid and liquid–solid transport processes in catalytic reactors.5.

Future view of total energy system and reactor engineering and reactor physics

International Nuclear Information System (INIS)

Ozawa, T.

1974-01-01

This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

Production of Advanced Biofuels via Liquefaction - Hydrothermal Liquefaction Reactor Design: April 5, 2013

Energy Technology Data Exchange (ETDEWEB)

Knorr, D.; Lukas, J.; Schoen, P.

2013-11-01

This report provides detailed reactor designs and capital costs, and operating cost estimates for the hydrothermal liquefaction reactor system, used for biomass-to-biofuels conversion, under development at Pacific Northwest National Laboratory. Five cases were developed and the costs associated with all cases ranged from $22 MM/year - $47 MM/year.

Main safety lessons from 5-year operation of the renovated Dalat nuclear research reactor

International Nuclear Information System (INIS)

Anh, T.H.; Lam, P.V.; An, T.K.; Khang, N.P.; Tan, D.Q.

1989-01-01

The paper presents main safety related characteristics of the Dalat Nuclear Research Reactor (DNRR), which was reconstructed in 1982 at the site of the former TRIGA Mark II, while retaining some of its structures. Experience acquired from reactor operation is analysed. The programme of investigations aimed at better ensuring nuclear safety of the reactor, together with some of its results are presented. Finally some propositions to improve the present situation are suggested. (Authors). (2 Tables, 2 fig.)

Reactor Sharing Program

International Nuclear Information System (INIS)

Tehan, Terry

2002-01-01

Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

Evaluation of the Secretor Status of ABO Blood Group Antigens in Saliva among Southern Rajasthan Population Using Absorption Inhibition Method.

Science.gov (United States)

Metgud, Rashmi; Khajuria, Nidhi; Mamta; Ramesh, Gayathri

2016-02-01

The ABO blood group system was the significant element for forensic serological examination of blood and body fluids in the past before the wide adaptation of DNA typing. A significant proportion of individuals (80%) are secretors, meaning that antigens present in the blood are also found in other body fluids such as saliva. Absorption inhibition is one such method that works by reducing strength of an antiserum based on type and amount of antigen present in the stains. To check the efficacy of identifying the blood group antigens in saliva and to know the secretor status using absorption inhibition method among southern Rajasthan population. Blood and saliva samples were collected from 80 individuals comprising 20 individuals in each blood group. The absorption inhibition method was used to determine the blood group antigens in the saliva and then the results were correlated with the blood group of the collected blood sample. The compiled data was statistically analysed using chi-square test. Blood groups A & O revealed 100% secretor status for both males and females. While blood groups B and AB revealed 95% secretor status. Secretor status evaluation of the ABO blood group antigen in saliva using absorption inhibition method can be a useful tool in forensic examination.

Calculation of neutron spectra in the reactor cell of the RA experimental reactor in Vinca

International Nuclear Information System (INIS)

Bosevski, T.; Altiparmakov, D.; Marinkovic, N.

1974-01-01

In the frame of neutron properties of RA experimental reactor the study of energy neutron spectra in the reactor cell are planned. Complex reactor cell geometry, nine cylindrical regions causes high space-energy variations of neutron flux with a significant gradient both in energy and space variables. Treatment of such a complex problem needs adequate methodology which ensures reliable results and control of accuracy. This paper describes in detail the method for calculating group constants based on lattice cell calculation for the need of calculation of reactor core parameters. In 26 group approximation for the energy region from 0 - 10.5 MeV, values of neutron spectra are obtained in 18 space points chosen to describe, with high accuracy, integral reactor cell parameters of primary importance for the reactor core calculation. Obtained space-energy distribution of neutron flux in the reactor cell is up to now unique in the study of neutron properties of Ra reactor [sr

FBR type reactor

International Nuclear Information System (INIS)

Yamaoka, Mitsuaki

1988-01-01

Purpose: To enable to increase the burning period by enabling to decrease the reduction of burning reactivity and unifying the irradiation amount of fast neutrons. Constitution: A cylindrical reactor core made of fissile material-enriched fuel is constituted so as to form a plurality of layer-like enriched regions in which the enrichment degree of the fissile material is increased from the center to the radial and axial directions. Then, the ratio between the average enrichment degree for all of the enrichment regions other than the region at the reactor core center with the lowest enrichment degree and the enrichment degree of the enriched region formed at the center of the reactor core is made greater by 5 % or 20 % than the ratio at the initial burning stage where the power distribution of the reactor core is most flattened. (Kawakami, Y.)

Report on the specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors organized by the international atomic energy agency at Oarai Engineering Centre of power reactor and nuclear development corporation, Japan, November 5-7, 1991

International Nuclear Information System (INIS)

Paranjpe, S.R.

1992-01-01

As recommended by the International Working Group on Fast Reactors (IWGFR), the International Atomic Energy Agency organized a specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors. Specialists from all member countries of IWGFR-China, France, Germany, India, Italy, Japan, Russia, the United Kingdom, and the United States-participated in the meeting and made presentations as listed in Table 1. The Commission of European Communities also sent representatives to the meeting. Table 2 contains a list of participants. The meeting consisted of five sessions: (1) an overview, (2) safety characteristics of decay heal removal systems, (3) safely characteristics of reactor protection systems and reactor shutdown systems, (4) safely characteristics of reactor cores, and (5) general discussions antiformulation of recommendations

Predicting the in-reactor mechanical behavior of Zr-2.5Nb pressure tubes from postirradiation microstructural examination data

International Nuclear Information System (INIS)

Griffiths, M.; Davies, P.H.; Davies, W.G.; Sagat, S.

2002-01-01

Postirradiation microstructure examinations of Zr-2.5Nb pressure tubes removed from service in CANDU reactors have shown clear trends in the dislocation structure and the state of the β-phase, as a function of operating temperature, neutron flux, and time. These microstructural parameters correlate well with changes in the mechanical properties. For example, the rapid increase in dislocation loop density in the early stages of irradiation corresponds with a rapid increase in tensile strength and DHC velocity, and a corresponding decrease in fracture toughness. There is also a strong negative correlation between the degree of β-phase decomposition and DHC velocity. In addition to the effects of microstructure evolution on the mechanical properties, changes in the a-type and c-component dislocation loop densities also affect irradiation deformation (creep and growth). Statistical analyses of the irradiation microstructure data have been used to derive empirical relationships between dislocation densities and β-phase structure with temperature, flux, and time. The relationships thus derived are useful in predicting where the mechanical properties are most affected by the in-reactor operating conditions. The predictions are compared with mechanical test data for samples from various axial and circumferential locations of 42 pressure tubes removed from operating CANDU reactors. The results are discussed in terms of the mechanisms controlling tensile strength, fracture, delayed-hydride-cracking, and in-reactor deformation. (author)

Dosimetry and radiation shielding at the RA reactor, Annual report 1975, Annex 5; Prilog 5 - Dozimetrija i tehnicka zastita

Energy Technology Data Exchange (ETDEWEB)

Ninkovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1976-01-15

In the working environment at the RA reactor, the level of gamma radiation is measured continuously by the built-in stationary system. According to the needs, measurement are done in the reactor hall every day. The level of gamma radiation is measured separately in typical points when the reactor is operated at nominal power and during intervals between two operating campaigns. The level of neutron radiation is measured according to the needs by means of a mobile spherical neutron detector. These measurements are done in the reactor hall around the horizontal experimental channels. Measured values of neutron radiation are three times lower than the relevant levels of gamma radiation. [Serbo-Croat] Na reaktoru RA vrsi se stalna kontrola nivoa gama zracenja po tehnoloskim i radnim prostorijama, pomocu ugradjenog stacionarnog sistema. Svakodnevno se vrse merenja u reaktorskoj hali, prema ukazanim potrebama. Posebno se mere nivoi gamma zracenja u karakteristicnim tackama pri radu reaktora na nominalnoj snazi i u pauzama izmedju dva kampanje. Merenje nivoa neutronskog zracenja vrsi se diskontinualno pomocu mobilnog sfernog neutronskog dozimetra. Merenja se obavljaju u hali oko horizontalnih eksperimentalnih kanala. Izmerene vrednosti su u proseku tri puta manje od odgovarajucih nivoa gama zracenja.

Analysis of reactivity transient for the DIDO type research reactors using RELAP5

International Nuclear Information System (INIS)

Adorni, M.; Bousbia-Salah, A.; D'Auria, F.; Nabbi, R.

2005-01-01

Recent availability of high performance computers and computational methods together with the continuing increase in operational experience imposes revising some operational constrains and conservative safety margins. The application of Best-Estimate (BE) method constitutes a real necessity in the safety and design analysis and allows getting more realistic simulation of the processes taking place during the steady state operation and transients. In comparison to the conservative approaches, the application of Best-Estimate methods results in the mitigation of the constraining limits in design and operation. This paper presents the results of the application of the RELAP5/Mod3.3 system thermal-hydraulic code to the German FRJ-2 research reactor for a reactivity transient, which has been analyzed in the past using the verified system code CATHENA [1], [2], [3]. The work mainly aims checking the capability of RELAP5 [4] for research reactor transient analysis by the comparison of the results of the two codes and including modeling basis and analytical approaches. According to the existing references RELAP5 applications are concentrated on the transient analysis of nuclear power systems. The considered case consists of a simulation related to a hypothetical fast reactivity transient, which is assumed to be caused by the failure of one shutdown arm. The case has been chosen due to the importance of the models for the precise description of the complex phenomenon of subcooled boiling and two phase flow taking place during the transient. For this purpose, the fuel element assembly was modeled in detail according to design data. The primary circuit was included in the whole model in order to consider the interaction with individual fuel elements with core. In general the results of the two codes are in agreement and comparable during the initial phase of the transient. After reaching the flow regime with fully developed nucleate boiling and two phase flow RELAP5 exhibits

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

Designing a mini subcritical nuclear reactor; Diseno de un mini reactor nuclear subcritico

Energy Technology Data Exchange (ETDEWEB)

Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M., E-mail: rafelaescobedo@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Jardin Juarez 147, Col. Centro, 98000 Zacatecas, Zac. (Mexico)

2015-10-15

In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of {sup 239}PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of {sup 239}PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k{sub e}-f{sub f}, the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k{sub eff}, the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons {sup 239}PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k{sub eff} was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k{sub eff} of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five

Quantification of Net Erosion and Uplift Experienced by the Barmer Basin, Rajasthan Using Sonic Log

Science.gov (United States)

Mitra, K.; Schulz, S.; Sarkar, A.

2015-12-01

Barmer Basin of Rajasthan, Western India is a hydrocarbon rich sedimentary basin currently being explored by Cairn India Limited. The hydrocarbon bearing Fatehgarh Formation is being found at different depths in different oil fields (e.g. From south to north: Guda, Vijaya & Vandana, Air field High) of the basin. The net uplift and erosion in the Barmer Basin has been quantified using compaction methodology. The sonic log, which is strongly controlled by porosity, is an appropriate indicator of compaction, and hence used for quantification of net uplift and erosion from compaction. The compaction methodology has been applied to the shale rich Dharvi Dungar Formation of Barmer Basin of Late Paleocene age. The net uplift and erosion is also being checked with the help of AFTA-VR and seismic sections. The results show relatively no uplift in the southernmost part of the basin and a Guda field well is thus taken to be the reference well with respect to which the uplifts in different parts of the basin have been calculated. The northern part of the basin i.e. Air Field High wells experienced maximum uplift (~2150m). Interestingly, a few wells further south of the reference well show evidence for uplift. The study was able to point out errors in the report produced with the help of AFTA-VR which found out less uplift in Vijaya & Vandana oil fields as opposed to sonic log data. The process of finding out uplift using sonic log has a standard deviation of 200m as compared to about 500m error in AFTA-VR method. This study has major implications for hydrocarbon exploration. Maturation of source rock will be higher for any given geothermal history if net uplift and erosion is incorporated in maturation modeling. They can also be used for porosity predictions of reservoir units in undrilled targets.

Application of RELAP5-3D code for thermal analysis of the ADS reactor core; Aplicação do código RELAP5-3D para análise térmica do núcleo de um reator ADS

Energy Technology Data Exchange (ETDEWEB)

Fernandes, Gustavo Henrique Nazareno

2018-04-01

Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

Turning points in reactor design

International Nuclear Information System (INIS)

Beckjord, E.S.

1995-01-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems

Turning points in reactor design

Energy Technology Data Exchange (ETDEWEB)

Beckjord, E.S.

1995-09-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5

International Nuclear Information System (INIS)

Pena V, J.D.

2016-01-01

The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)

Thorium utilisation in thermal reactors

International Nuclear Information System (INIS)

Balakrishnan, K.

1997-01-01

It is now more or less accepted that the best way to use thorium is in thermal reactors. This is due to the fact that U233 is a good material in the thermal spectrum. Studies of different thorium cycles in various reactor concepts had been carried out in the early days of nuclear power. After three decades of neglect, the world is once again looking at thorium with some interest. We in India have been studying thorium cycles in most of the existing thermal reactor concepts, with greater emphasis on heavy water reactors. In this paper, we report some of the work done in India on different thorium cycles in the Indian pressurized heavy water reactor (PHWR), and also give a description of the design of the advanced heavy water reactor (AHWR). (author). 1 ref., 2 tabs., 5 figs

Integrated radioactive waste management from NPP, research reactor and back end of nuclear fuel cycle - an Indian experience

International Nuclear Information System (INIS)

Kumar, S.; Ali, S.S.; Chander, M.; Bansal, N.K.; Balu, K.

2001-01-01

India is one of the developing countries operating waste management facilities for entire nuclear fuel cycle for the last three decades. Over the years, the low and intermediate level (LIL) liquid waste streams arising from reactors and fuel reprocessing facilities have been well characterised and different processes for treatment, conditioning and disposal are being practised. LIL waste generated in nuclear facilities is treated by chemical treatment processes where majority of the activity is retained in the form of sludge. Decontamination factors ranging from 10 to 1000 are achieved depending upon the process employed and characteristics of the waste. At an inland PHWR site at Rajasthan, the LIL waste is concentrated by solar evaporation. To augment the treatment capability, a plant is being set up at Trombay to treat LIL waste based on reverse osmosis process. Alkaline waste of intermediate level activity is being treated by using indigenously developed resorcinol formaldehyde resin. Solid radioactive waste is volume reduced by compacting, baling and incineration depending on the nature of the waste. Cement matrix is employed for immobilisation of process concentrate such as chemical sludge, ash from incinerators etc. The solid waste, depending on the activity contents, is disposed in underground engineered trenches in near surface disposal facility. Bore well samples around the trench are drawn periodically to ascertain the effectiveness of the disposal system. The gaseous waste is treated at the source itself. High efficiency particulate air (HEPA) filter and impregnated activated carbon is employed to restrict the release of airborne activity to the environment. Radioactive waste discharges are kept well below the authorised limits prescribed by the regulatory authorities. This paper covers the waste management practices being adopted in India for treatment, conditioning, interim storage and disposal of low and intermediate level waste arising from the

DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

Directory of Open Access Journals (Sweden)

Ralind Re Marla

2015-03-01

Full Text Available Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN untuk jenis Pebble Bed Modular Reactor (PBMR dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL. Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5 dan dari hasil analisis ini diharapkan dapat memenuhi syarat dalam mendukung program percepatan pembangunan kelistrikan batubara 10.000 MWe khususnya untuk proses smelter, yang tersebar merata di wilayah Indonesia. Hasil penelitian menunjukkan bahwa, faktor perlipatan efektif (k-eff Reaktor jenis PBMR daya 70 MWe mengalami kondisi kritis pada pengkayaan 5,626 % dengan nilai faktor perlipatan efektif 1,00031±0,00087 dan nilai koefisien reaktivitas suhu pada -10,0006 pcm/K. Dari hasil analisis daat disimpulkan bahwa reaktor jenis PBMR daya 70 MWe adalah aman.   ABSTRACT The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR type with 70 MWe capacity power in Beginning of Life (BOL has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff with power 70 MWe critical condition at enrichment 5,626 %is 1,00031±0,00087, based on enrichment result, a value of the temperature coefficient reactivity is - 10,0006 pcm/K. Based on the results of these studies, it can beconcluded that the PBMR 70 MWe design is theoritically safe.

Iris reactor conceptual design

International Nuclear Information System (INIS)

Carelli, M.D.; Conway, L.E.; Petrovic, B.; Paramonov, D.V.; Galvin, M.; Todreas, N.E.; Lombardi, C.V.; Maldari, F.; Ricotti, M.E.; Cinotti, L.

2001-01-01

IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

Prospect of realizing nuclear fusion reactors

International Nuclear Information System (INIS)

1989-01-01

This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Reactor utilization; Eksploatacija reaktora

Energy Technology Data Exchange (ETDEWEB)

Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel.

Reactor for exothermic reactions

Science.gov (United States)

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Comparison of SERPENT and CASMO-5M for pressurized water reactors models

International Nuclear Information System (INIS)

Hursin, M.; Vasiliev, A.; Ferroukhi, H.; Pautz, A.

2013-01-01

The objective of this work is to perform a preliminary assessment of the capability of SERPENT to generate cross sections for a PWR Beginning-of-Life (BOL) isothermal mini-core by comparing a SERPENT/PARCS solution with the results obtained using a CASMO-5M/PARCS approach. The PARCS code is used instead of the usual SIMULATE-3 to analyze the Swiss Reactors, because interfaces with PARCS already exist to obtain neutronic data from SERPENT. For the PWR configurations, the differences between CASMO-5M and SERPENT solutions are within 200 pcm at the assembly level and thus rather small when considering the deterministic transport method (energy/angular/space discretization) in CASMO-5M versus the stochastic treatment of SERPENT, the statistical uncertainties in the Monte-Carlo approach as well as the eventual differences in nuclear data used by both codes. At the 2D mini-core level, no major difference is observed when comparing PARCS run with CASMO-5M versus SERPENT cross sections. For the generation of kinetic parameters, non trivial differences are observed due both to the methods and the data used. For the relatively limited number of configurations considered, it is hard to make any definitive statement on the benefits of using Monte Carlo codes in terms of nuclear data generation. (authors)

Cooperation of nuclear reactor controller ARM-5S and turbine TVER-02

International Nuclear Information System (INIS)

Wagner, K.; Lnenicka, B.; Pokorny, F.; Prochazka, F.

1985-01-01

Turbines of Czechoslovak make provided with controllers TVER-02 are installed in WWER-440 nuclear power plants under construction in Czechoslovakia. Reactor output is controlled using Soviet-made controllers ARM-5S which already comprise turbine controllers. The problems are analyzed of cooperation of both controllers, especially their parameters and transient processes in typical operating situations. The analysis uses the results of measurements performed during the power start-up of Unit 1 of the V-2 nuclear power plant at Jaslovske Bohunice. The results show that two types of control modes can be selected for the operation of the entire unit: the control to constant unit output, and control of unit output varying with turbine load selected on the TVER-02 controller or given by the demand of the power network. (Z.M.)

First TREAT [Transient Reactor Test Facility] transient overpower tests on U-Pu-Zr fuel: M5 and M6

International Nuclear Information System (INIS)

Robinson, W.R.; Bauer, T.H.; Wright, A.E.; Rhodes, E.A.; Stanford, G.S.; Klickman, A.E.

1987-01-01

Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) tests of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the Integral Fast Reactor concept. Similar tests on U-Fs fueled EBR-II driver pins were previously performed and reported [1,2]. Results from these earlier tests indicated a margin to failure of about 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuel types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety related fuel performance characteristics of U-Fs and U-Pu-Zr. This report described conditions, results, and conclusions of testing of these fuel types

Correlations between power and test reactor data bases

International Nuclear Information System (INIS)

Guthrie, G.L.; Simonen, E.P.

1989-02-01

Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2

International Nuclear Information System (INIS)

Ojeda S, J.; Morales S, J.; Chavez M, C.

2009-10-01

This work intends a model using the code Relap5/Mod 3.2, for the instability study in normal operation of a natural circulation reactor type ESBWR. A conceptual analysis is considered because all the information was obtained of the open literature, and some of reactor operation or dimension (not available) parameters were approached. As starting point was took the pattern developed for reactor type BWR, denominated Browns Ferry and changes were focused in elimination of bonds of forced recirculation, in modification of operation parameters, dimensions and own control parameters, according to internal code structure. Additionally the nodalization outline is described analyzing for separate the four fundamental areas employees in peculiar geometry of natural circulation reactor. Comparative analysis of results of stability behavior obtained with those reported in the open literature were made, by part of commercial reactor designer ESBWR. (Author)

Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

International Nuclear Information System (INIS)

Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K.; McCauley, E.W.

1977-01-01

An accurate Mark I 1 / 5 -scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

The Integral Fast Reactor

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.; Lineberry, M.J.

1990-01-01

Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs

Feasibility study for LEU conversion of the WWR-K reactor at the Institute of Nuclear Physics in Kazakhstan using a 5-tube fuel assembly

International Nuclear Information System (INIS)

Hanan, N.A.; Liaw, J.R.; Matos, J.E.

2005-01-01

A feasibility study by the RERTR program for possible LEU conversion of the 6 MW WWR-K reactor concludes that conversion is feasible using an LEU 5-tube Russian fuel assembly design. This 5-tube design is one of several LEU fuel assembly designs being studied (Ref. 1) for possible use in this reactor. The 5-tube assembly contains 200 g 235 U with an enrichment of 19.7% in four cylindrical inner tubes and an outer hexagonal tube with the same external dimensions as the current HEU (36%) 5-tube fuel assembly, which contains 112.5 g 235 U. The fuel meat material, LEU UO 2 -Al dispersion fuel with ∼ 2.5 g U/cm 3 , has been extensively irradiation tested in a number of reactors with uranium enrichments of 36% and 19.7%. Since the 235 U loading of the LEU assemblies is much larger than the HEU assemblies, a smaller LEU core with five rows of fuel assemblies is possible (instead of six rows of fuel assemblies in the HEU core). This smaller LEU core would consume about 60% as many fuel assemblies per year as the current HEU core and provide thermal neutron fluxes in the inner irradiation channels that are ∼ 17% larger than with the present HEU core. The current 21 day cycle length would be maintained and the average discharge burnup would be ∼ 42%. Neutron fluxes in the five outer irradiation channels would be smaller in the LEU core unless these channels can be moved closer to the LEU fuel assemblies. Results show that the smaller LEU core would meet the reactor's shutdown margin requirements and would have an adequate thermal-hydraulic safety margin to onset of nucleate boiling. (author)

Cryogenic system design for a compact tokamak reactor

International Nuclear Information System (INIS)

Slack, D.S.; Kerns, J.A.; Miller, J.R.

1988-01-01

The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

How Canada's and India's nuclear roles have been sadly misrepresented

International Nuclear Information System (INIS)

Bindon, G.; Mukerji, S.

1977-01-01

The authors are concerned to remove misunderstandings about the termination of nuclear cooperation between Canada and India following India's explosion of a nuclear device in 1974. The history of the cooperation between the two countries, leading to the construction of the Rajasthan-1 reactor, is recounted. It is suggested that India had long intended to explode a nuclear device, consistent with its refusal to sign the nonproliferation treaty; surprise was therefore unwarranted. The value of the Rajasthan contract to the Canadian nuclear industry is emphasized, on the grounds that it came at a crucial time, just after Douglas Point had reached full power in 1962. The authors seem to believe that Canadian industry will suffer in that it will be more difficult to achieve agreements with developing countries in the future. This is made the occasion for some general criticism of the Canadian nuclear programme, but not of the Indian, which is held to be more diversified, and therefore able to develop supposedly practical and better alternatives to nuclear energy. (N.D.H.)

Safety evaluation of the Dalat research reactor operation

International Nuclear Information System (INIS)

Long, V.H.; Lam, P.V.; An, T.K.

1989-01-01

After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

The contract of design of an atomic reactor system in Yonggwang - 5, 6 nuclear power plant

International Nuclear Information System (INIS)

1995-05-01

This is a contract of design of an atomic reactor system in Yonggwang 5, 6 nuclear power plant. It has the general contract condition. In the appendix, it indicates the detail regulations between two parties which are the coverage of division on the responsibility, schedule of the delivery, standard of the technology, guarantee, drawing and paper support of the Korea Electric Power Corporation, support of technology drill, test, regulations of code and standard and list of items and prices.

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Fusion reactor safety

International Nuclear Information System (INIS)

1987-12-01

Nuclear fusion could soon become a viable energy source. Work in plasma physics, fusion technology and fusion safety is progressing rapidly in a number of Member States and international collaboration continues on work aiming at the demonstration of fusion power generation. Safety of fusion reactors and technological and radiological aspects of waste management are important aspects in the development and design of fusion machines. In order to provide an international forum to review and discuss the status and the progress made since 1983 in programmes related to operational safety aspects of fusion reactors, their waste management and decommissioning concepts, the IAEA had organized the Technical Committee on ''Fusion Reactor Safety'' in Culham, 3-7 November 1986. All presentations of this meeting were divided into four sessions: 1. Statements on National-International Fusion Safety Programmes (5 papers); 2. Operation and System Safety (15 papers); 3. Waste Management and Decommissioning (5 papers); 4. Environmental Impacts (6 papers). A separate abstract was prepared for each of these 31 papers. Refs, figs, tabs

Preliminary scoping study of some neutronic aspects of new shim safety rods for a typical 5 MW research reactor by Monte Carlo simulation

Energy Technology Data Exchange (ETDEWEB)

Shoushtari, M.K.; Kakavand, T. [Faculty of Science, University of Zanjan, P.O. BOX 1415, Zanjan (Iran, Islamic Republic of); Ghaforian, H. [Faculty of Science and Technology of Marine, P.O. BOX 212 Tehran (Iran, Islamic Republic of); Kiai, S.M. Sadat [Nuclear Science and Technology Research Institute (NSTR), Nuclear Science Research, A.E.O.I. P.O. BOX 14155-1339, Tehran (Iran, Islamic Republic of)], E-mail: sadatkiai@yahoo.com

2009-02-15

A Monte Carlo simulation of a typical 5 MW research reactor (TRR) was carried out using MCNP4C code. The geometry of the reactor core was modeled including the details of all fuel elements, control rods, all irradiation channels, graphite reflectors, reactor pool and thermal column. The model predicted neutron flux distributions within the core, control rod (CR) worth, core reactivity ({rho}), shutdown margin, and some kinetic parameters when the control rod insert or withdraw. This study was carried out to reduce blockage probability of shim safety rod (SSR)s of the TRR. Two introduced more blackness SSRs were chosen and made thinner in a way adequate blackness, in comparison to the present rods, achieved.

Physical security at research reactors

International Nuclear Information System (INIS)

Clark, R.A.

1977-01-01

Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

Planning the Decommissioning of Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Podlaha, J., E-mail: pod@ujv.cz [Nuclear Research Institute Rez, 25068 Rez (Czech Republic)

2013-08-15

In the Czech Republic, three research nuclear reactors are in operation. According to the valid legislation, preliminary decommissioning plans have been prepared for all research reactors in the Czech Republic. The decommissioning plans shall be updated at least every 5 years. Decommissioning funds have been established and financial resources are regularly deposited. Current situation in planning of decommissioning of research reactors in the Czech Republic, especially planning of decommissioning of the LVR-15 research reactor is described in this paper. There appeared new circumstances having wide impact on the decommissioning planning of the LVR-15 research reactor: (1) Shipment of spent fuel to the Russian Federation for reprocessing and (2) preparation of processing of radioactive waste from reconstruction of the VVR-S research reactor (now LVR-15 research reactor). The experience from spent fuel shipment to the Russian Federation and from the process of radiological characterization and processing of radioactive waste from reconstruction of the VVR-S research reactor (now the LVR-15 research reactor) and the impact on the decommissioning planning is described in this paper. (author)

TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

International Nuclear Information System (INIS)

Solbrig, C.W.; Stevens, W.W.

1990-01-01

Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Operational safety and reactor life improvements of Kyoto University Reactor

International Nuclear Information System (INIS)

Utsuro, M.; Fujita, Y.; Nishihara, H.

1990-01-01

Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

Simulation development for TRIGA reactor

International Nuclear Information System (INIS)

Handoyo, D.

1997-01-01

A simulator of the dynamic of TRIGA reactor has been made. this simulator is meant to study the reactor kinetic behavior and for operator training to more assure the safety and the reliability of the real operation of TRIGA reactor. the simulator consists of PC (Personal Computer) for processing the calculation of reactivity, neutron flux, period, ect and control panel for regulating the input data such as the change of power range, control rod position as well as cooling flow rate. the result will be displayed on screen monitor of personal computer as given in the real control room of TRIGA reactor. the output of simulator will be verified by comparing with measurement result in the real TRIGA MARK II reactor of Musashi institute of technology. for the change of reactivity of 0.3, 0.5 and 0.7 the reactor power and fuel temperature between the simulator and measurements are comparable

International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

Digital integration of geological and aeroradiometric data for probability mapping of uranium occurrences in parts of south-eastern Rajasthan, India

International Nuclear Information System (INIS)

Chawla, A.S.; Katti, V.J.; Kak, S.N.; Das, S.K.

1993-01-01

Integration and evaluation of geological, radio geochemical, and magnetic information of Umra-Udaisagar and Sarara inlier area, Udaipur district, Rajasthan was attempted. Seventeen lithostructural variables from colour infrared (CIR) photo geological analogue maps were interpreted, radio geochemical and magnetic variables thematically evaluated from airborne gamma-ray spectrometric (AGRS) and aero magnetic (AM) digital data were co-registered using a sequential grid matrix of 500 m x 500 m. The variables were quantified using theme-specific equations and digitized in simple Boolean representation format, depending on the presence or absence of a variable, its positive or negative interest, and/or greater than or less than a theme-specific value in each cell. The database so generated was subjected to a software programme weighted modelling wherein, weights for each variable are computed based on conditional probability method, and favorability index maps are generated by discriminant objective analysis. Areas with high probability for uranium mineralisation were delineated by computing composite coincidence of cells with high favorability index value considering each variable as a control variable. Critical analysis of weights computed for each variable in different sets predicts the importance of control of the variable to uranium mineralisation. This attempt has resulted in delineating several new high-probability zones of uranium enrichment and indicated a regional structural control in the area. (author). 11 refs., 4 figs

Nuclear reactor multi-physics simulations with coupled MCNP5 and STAR-CCM+

International Nuclear Information System (INIS)

Cardoni, Jeffrey Neil; Rizwan-uddin

2011-01-01

The MCNP5 Monte Carlo particle transport code has been coupled to the computational fluid dynamics code, STAR-CCM+, to provide a high fidelity multi-physics simulation tool for pressurized water nuclear reactors. The codes are executed separately and coupled externally through a Perl script. The Perl script automates the exchange of temperature, density, and volumetric heating information between the codes using ASCII text data files. Fortran90 and Java utility programs assist job automation with data post-processing and file management. The MCNP5 utility code, MAKXSF, pre-generates temperature dependent cross section libraries for the thermal feedback calculations. The MCNP5–STAR-CCM+ coupled simulation tool, dubbed MULTINUKE, was applied to a steady state, PWR cell model to demonstrate its usage and capabilities. The demonstration calculation showed reasonable results that agree with PWR values typically reported in literature. Temperature and fission reaction rate distributions were realistic and intuitive. Reactivity coefficients were also deemed reasonable in comparison to historically reported data. The demonstration problem consisted of 9,984 CFD cells and 7,489 neutronic cells. MCNP5 tallied fission energy deposition over 3,328 UO_2 cells. The coupled solution converged within eight hours and in three MULTINUKE iterations. The simulation was carried out on a 64 bit, quad core, Intel 2.8 GHz microprocessor with 1 GB RAM. The simulations on a quad core machine indicated that a massively parallelized implementation of MULTINUKE can be used to assess larger multi-million cell models. (author)

RELAP5 analyses of two hypothetical flow reversal events for the advanced neutron source reactor

Energy Technology Data Exchange (ETDEWEB)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. [Oak Ridge National Lab., TN (United States)

1995-09-01

This paper presents RELAP5 results of two hypothetical, low flow transients analyzed as part of the Advanced Neutron Source Reactor safety program. The reactor design features four independent coolant loops (three active and one in standby), each containing a main curculation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and additionally, that the check valve in that loop remains stuck in the open position. This accident is considered extremely unlikely. Flow reverses in this loop, reducing the core flow because much of the coolant is diverted from the intact loops back through the failed loop. The second transient examines a 102-mm-diam instantaneous pipe break near the core inlet (the worst break location). A break is assumed to occur 90 s after a total loss-of-offsite power. Core flow reversal occurs because accumulator injection overpowers the diminishing pump flow. Safety margins are evaluated against four thermal limits: T{sub wall}=T{sub sat}, incipient boiling, onset of significant void, and critical heat flux. For the first transient, the results show that these limits are not exceeded (at a 95% non-exceedance probability level) if the pony motor battery lasts 30 minutes (the present design value). For the second transient, the results show that the closest approach of the fuel surface temperature to the local saturation temperature during core flow reversal is about 39{degrees}C. Therefore the fuel remains cool during this transient. Although this work is done specifically for the ANSR geometry and operating conditions, the general conclusions may be applicable to other highly subcooled reactor systems.

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Analyses for MARIA Research Reactor with RELAP/MOD3 code

International Nuclear Information System (INIS)

Szczurek, J.; Czerski, P.

2004-01-01

This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Blevins, J.D.; Stasko, R.R.

1989-09-01

An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

RA reactor exploitation, task 3.08/01

International Nuclear Information System (INIS)

Zecevic, V.

1963-01-01

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

Prospects for applications of ship-propulsion nuclear reactors

International Nuclear Information System (INIS)

Mitenkov, F.M.

1994-01-01

The use of ship-propulsion nuclear power reactors in remote areas of Russia is examined. Two ship reactors were analyzed: the KLT-40, a 170 MW-thermal reactor; and the KN-3, a 300 MW-thermal reactor. The applications considered were electricity generation, desalination, and drinking water production. Analyses showed that the applications are technically justified and could be economically advantageous. 5 refs., 9 figs., 1 tab

Maintenance and material aspects of DREAM reactor

International Nuclear Information System (INIS)

Ueda, S.; Nishio, S.; Yamada, R.; Seki, Y.; Kurihara, R.; Adachi, J.; Yamazaki, S.

2000-01-01

A concept of a commercial fusion power reactors (Fusion Power: 5.5 GW, electric output: 2.7 GW) having high environmental safety, high thermal efficiency and high availability has been studied in JAERI. The gross reactor configuration was designed to achieve good maintainability, high performance breeding blanket, high efficient power generation system and little radwastes. Design was based on the use of low activation structural material (SiC/SiC composites) and helium as a coolant. In this paper, maintenance and material aspects of DREAM reactor design is discussed. The concluding remarks are as follows. (1) The difficulty of development of maintenance tool is alleviated by sector replacement and the radiation dose environment less than 10 Gy/h in a reactor chamber. (2) Design requirement and present status of SiC/SiC composites was investigated. (3) The SiC/SiC composite development program is planned to satisfy the requirements of DREAM reactor

Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

Energy Technology Data Exchange (ETDEWEB)

Lai, W.; McCauley, E.W.

1978-01-04

Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

International Nuclear Information System (INIS)

Lai, W.; McCauley, E.W.

1978-01-01

Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90 0 torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this

Transmutation of actinide 237Np with a fusion reactor and a hybrid reactor

International Nuclear Information System (INIS)

Feng, K.M.; Huang, J.H.

1994-01-01

The use of fusion reactors to transmute fission reactor wastes to stable species is an attractive concept. In this paper, the feasibility of transmutation of the long-lived actinide radioactive waste Np-237 with a fusion reactor and a hybrid reactor has been investigated. A new waste management concept of burning HLW (High Level Waste), utilizing released energy and converting Np-237 into fissile fuel Pu-239 through transmutation has been adopted. The detailed neutronics and depletion calculation of waste inventories was carried out with a modified version of one-dimensional neutron transport and burnup calculation code system BISON1.5 in this study. The transmutation rate of Np with relationship to neutron wall loading, Pu and Np with relationship to neutron wall load, Pu and Np concentration in the transmutation zone have been explored as well as relevant results are also given

Thermionic reactors for space nuclear power

Science.gov (United States)

Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

1985-01-01

Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

Fusion reactors as a future energy source

International Nuclear Information System (INIS)

Seifritz, W.

A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

Energy Technology Data Exchange (ETDEWEB)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.

1996-02-01

In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

Vibration-proof FBR type reactor

International Nuclear Information System (INIS)

Kawamura, Yutaka.

1992-01-01

In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

Ecological observations on the Indian Spiny-tailed Lizard Saara hardwickii (Gray, 1827 (Reptilia: Squamata: Agamidae in Tal Chhapar Wildlife Sanctuary, Rajasthan, India

Directory of Open Access Journals (Sweden)

S.K. Das

2013-01-01

Full Text Available Observations on the Indian Spiny-tailed Lizard Saara hardwickii (Gray, 1827 were undertaken in Tal Chhapar Wildlife Sanctuary, Rajasthan, India during the monsoons (July following quadrat sampling that was time-constrained. The study revealed that the area is one of the preferable habitats for the species. A population analysis showed that the relative abundance of the subadults was higher, followed by juveniles and adults during the study period. The beginning of activity of the lizards was found to vary over the study period depending on prevailing weather conditions. The activity pattern was bimodal, except across rain events. The study revealed two important ecological findings about these lizards; complete sealing of burrow during rains which differed from partial sealing on normal days and complete diurnal cycle of body colour changes during the monsoon. Feeding was the predominant activity of this lizard followed by basking, resting and chasing each other. The adult lizards were found to be strictly herbivorous, in spite of an abundance of insects available in the area during the period. Subadults and juveniles were found to eat both plant parts, as well as insects. Microhabitat use such as inside grass clumps was found to be higher followed by barren ground, under shade and on stones.

Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

International Nuclear Information System (INIS)

Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

1981-01-01

Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

Overview of fusion reactor safety

International Nuclear Information System (INIS)

Cohen, S.; Crocker, J.G.

1981-01-01

Use of deuterium-tritium burning fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control, (2) neutron activation of structural materials, fluid streams and reactor hall environment, (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions, (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices, and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power

Research reactor spent fuel in Ukraine

International Nuclear Information System (INIS)

Trofimenko, A.P.

1996-01-01

This paper describes the research reactors in Ukraine, their spent fuel facilities and spent fuel management problems. Nuclear sciences, technology and industry are highly developed in Ukraine. There are 5 NPPs in the country with 14 operating reactors which have total power capacity of 12,800 MW

RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident

Energy Technology Data Exchange (ETDEWEB)

Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

1990-05-01

A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.

RELAP5/MOD2.5 analysis of the HFBR [High Flux Beam Reactor] for a loss of power and coolant accident

International Nuclear Information System (INIS)

Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

1990-05-01

A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs

Compact stellarators as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

2001-01-01

Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

Proceedings of the advisory committee on reactor safeguards workshop on future reactors

International Nuclear Information System (INIS)

2001-12-01

This report contains the information presented at the Advisory Committee on Reactor Safeguards Workshop on Future Reactors held at the Nuclear Regulatory Commission headquarters in Rockville, Maryland, on June 4-5, 2001. Included are the subject matter summaries, followed by the presentation material and selected participants discussions. The primary purpose of the workshop was to identify the regulatory challenges associated with future reactor designs. A list of such challenges was developed from the workshop notes, the various presentations, the panel discussions and the question and answer sessions. This list is included in the Introduction section of this document. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final workshop agenda

Proceedings of the advisory committee on reactor safeguards workshop on future reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-12-01

This report contains the information presented at the Advisory Committee on Reactor Safeguards Workshop on Future Reactors held at the Nuclear Regulatory Commission headquarters in Rockville, Maryland, on June 4-5, 2001. Included are the subject matter summaries, followed by the presentation material and selected participants discussions. The primary purpose of the workshop was to identify the regulatory challenges associated with future reactor designs. A list of such challenges was developed from the workshop notes, the various presentations, the panel discussions and the question and answer sessions. This list is included in the Introduction section of this document. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final workshop agenda.

Remote ultrasonic characterisation of an irradiated pressure tube from RAPS-II

Energy Technology Data Exchange (ETDEWEB)

Gangotra, S; Muralidhar, S; Raut, S D; Ouseph, P M; Ghosh, J K; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

1994-12-31

The Rajasthan Atomic Power Station Unit-2 (RAPS-2) has reached a stage of operation where the contacting pressure tubes are suspect to failure as a result of irradiation creep and displacement of the garter springs, the hot pressure tube coming in contact with the cold calandria tube. To study and assess the safety of these pressure tubes, two channels believed to be in contact with the calandria tubes, have been removed from the reactor for detailed full length post irradiation examination. Some of the test results are presented. 2 refs., 3 figs., 1 tab.

Status of nuclear power in India

International Nuclear Information System (INIS)

Srinivasan, M.R.

1976-01-01

The article traces the history of Indian atomic energy programme and deals with the various nuclear power reactors operating commercially and under construction. Development of atomic energy in India is firmly committed to the installation of PHWR's as the mainstay of the country's immediate nuclear power programme. Indian nuclear power programme upto 1982 and major efforts at indigenisation, as reflected by the diminishing foreign exchange component in capital cost of different projects are discussed. Operational highlights of Tarapur atomic power station and Rajasthan atomic power station are reported. (S.K.K.)

SCDAP/RELAP5 modeling of fluid heat transfer and flow losses through porous debris in a light water reactor

International Nuclear Information System (INIS)

Harvego, E. A.; Siefken, L. J.

2000-01-01

The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the U.S. Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems during severe accidents. This paper describes the modeling approach used in the SCDAP/RELAP5 code to calculate fluid heat transfer and flow losses through porous debris that has accumulated in the vessel lower head and core regions during the latter stages of a severe accident. The implementation of heat transfer and flow loss correlations into the code is discussed, and calculations performed to assess the validity of the modeling approach are described. The different modes of heat transfer in porous debris include: (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, (5) film boiling, and (6) transition from film boiling to convection to vapor. The correlations for flow losses in porous debris include frictional and form losses. The correlations for flow losses were integrated into the momentum equations in the RELAP5 part of the code. Since RELAP5 is a very general non-homogeneous non-equilibrium thermal-hydraulics code, the resulting modeling methodology is applicable to a wide range of debris thermal-hydraulic conditions. Assessment of the SCDAP/RELAP5 debris bed thermal-hydraulic models included comparisons with experimental measurements and other models available in the open literature. The assessment calculations, described in the paper, showed that SCDAP/RELAP5 is capable of calculating the heat transfer and flow losses occurring in porous debris regions that may develop in a light water reactor during a severe accident

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Backfitting of the FRG reactors

Energy Technology Data Exchange (ETDEWEB)

Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

1990-05-01

The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U{sub 3}Si{sub 2} fuel. Both cooling towers were repaired. Replacement of instrumentation is planned.

Backfitting of the FRG reactors

International Nuclear Information System (INIS)

Krull, W.

1990-01-01

The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

Generation IV reactors: international projects

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

2003-01-01

Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

Practical course on reactor instrumentation

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2004-06-01

This course is based on the description of the instrumentation of the TRIGA-reactor Vienna, which is used for training research and isotope production. It comprises the following chapters: 1. instrumentation, 2. calibration of the nuclear channels, 3. rod drop time of the control rods, 4. neutron flux density measurements using compensated ionization, 5. neutron flux density measurement with fission chambers (FC), 6. neutron flux density measurement with self-powered neutron detectors (SPND), 7. pressurized water reactor simulator, 8. verification of the radiation level during reactor operation. There is one appendix about neutron-sensitive thermocouples. (nevyjel)

Compact toroid refueling of reactors

International Nuclear Information System (INIS)

Gouge, M.J.; Hogan, J.T.; Milora, S.L.; Thomas, C.E.

1988-04-01

The feasibility of refueling fusion reactors and devices such as the International Thermonuclear Engineering Reactor (ITER) with high-velocity compact toroids is investigated. For reactors with reasonable limits on recirculating power, it is concluded that the concept is not economically feasible. For typical ITER designs, the compact toroid fueling requires about 15 MW of electrical power, with about 5 MW of thermal power deposited in the plasma. At these power levels, ideal ignition (Q = ∞) is not possible, even for short-pulse burns. The pulsed power requirements for this technology are substantial. 6 ref., 1 figs

Neutronic feasibility studies using U-Mo dispersion fuel (9 Wt % Mo, 5.0 gU/cm3) for LEU conversion of the MARIA (Poland), IR-8 (Russia), and WWR-SM (Uzbekistan) research reactors

International Nuclear Information System (INIS)

Bretscher, M.M.; Deen, J.R.; Hanan, N.A.; Matos, E.

2000-01-01

U-Mo alloys dispersed in an Al matrix offer the potential for high-density uranium fuels needed for the LEU conversion of many research reactors. On-going fuel qualification tests by the US RERTR Program show good irradiation properties of U-Mo alloy dispersion fuel containing 7-10 weight percent molybdenum. For the neutronic studies in this paper the alloy was assumed to contain 9 wt % Mo (U-9Mo) with a uranium density in the fuel meat of 5.00 gU/cm 3 which corresponds to 32.5 volume % U-9Mo. Fuels containing U-9Mo have been used in Russian reactors since the 1950's. For the three research reactors analyzed here, LEU fuel element thicknesses are the same as those for the Russian-fabricated HEU reference fuel elements. Relative to the reference fuels containing 80-90% enriched uranium, LEU U-9Mo Al-dispersion fuel with 5.00 gU/cm 3 doubles the cycle length of the MARIA reactor and increases the IR-8 cycle length by about 11%. For the WWR-SM reactor, the cycle length, and thus the number of fuel assemblies used per year, is nearly unchanged. To match the cycle length of the 36% enriched fuel currently used in the WWR-SM reactor will require a uranium density in the LEU U-9Mo Al-dispersion fuel of about 5.4 gU/cm 3 . The 5.00 gU/cm 3 LEU fuel causes thermal neutron fluxes in water holes near the edge of the core to decrease by (6-8)% for all three reactors. (author)

RA Research reactor, Annual report 1971

International Nuclear Information System (INIS)

Milosevic, D. et al.

1971-12-01

During 1971, the RA Reactor was operated at nominal power of 6.5 MW for 190 days, and 50 days at lower power levels. Total production mounted to 31606 MWh which is 5.3% higher than planned, and the highest annual level since the reactor started operation. Reactor was used for irradiation and experiments according to the demand of 425 users, of which 370 from the Institute and 55 external users. This report contains detailed data about reactor power and experiments performed in 1971. Discrepancies from the action plan, meaning higher production was achieved in June and December due to special demand of the users. Total number of interruptions was lower than during all the previous years, and were caused mainly due to power cuts during reactor operation. There was no longer interruption caused by failures of the equipment. There was only on scram shutdown during this year caused by a false signal of the reactor control instrumentation. Shorter interruptions resulted from breaking of connectors in the technical water pipe system caused by soil sliding near the pumping station on the Danube. Total personnel exposure dose was lower than during previous years. There was no accident nor any event that could be called accidental. Decontamination od surfaces was less than during previous years. It was concluded that the successful operation in 1971 resulted from efficient work during past years. But, some of the activities were interrupted due to undefined policy concerned with operation of the RA reactor and financial issues. This involves study of the possibility to use highly enriched fuel that would increase the useful neutron flux and the reactor compatibility with similar reactors for the future ten years. Another project that has been interrupted is construction of the emergency core cooling system which is important for the reactor safety. Financial problems have influenced not only the reactor operation but the number of employees, which could cause negative

French experience in design, operation and revamping of nuclear research reactors, in support of advanced reactors development

International Nuclear Information System (INIS)

Barre, B.; Bergeonneau, P.; Merchie, F.; Minguet, J.L.; Rousselle, P.

1996-01-01

The French nuclear program is strongly based on the R and D work performed in the CEA nuclear research centers and particularly on the various experimental programs carried out in its research reactors in the frame of cooperative actions between the Commissariat a l'Energie Atomique (CEA), Framatome and Electricite de France (EDF). Several types of research reactors have been built by Technicatome and CEA to carry out successfully this considerable R and D work on fuels and materials, among them the socalled Materials Testing Reactors (MTR) SILOE (35 MW) and OSIRIS (70 MW) which are indeed very well suited for technological irradiations. Their simple and flexible design and the large irradiation space available around the core, the SILOE and OSIRIS reactors can be shared by several types of applications such as fuel and material testings for nuclear power plants, radioisotopes production, silicon doping and fundamental research. It is worthwhile recalling that Technicatome and CEA have also built research reactors fully dedicated to safety experimental studies, such as the CABRI, SCARABEE and PHEBUS reactors at Cadarache, and others dedicated to fundamental research such as ORPHEE (14 MW) and the Reacteur a Haut Flux -High Flux Reactor- (RHF 57 MW). This paper will present some of the most significant conceptual and design features of all these reactors as well as the main improvements brought to most of them in the last years. Based on this wide experience, CEA and Technicatome have specially designed for export a new multipurpose research reactor named SIRIUS, with two versions depending on the utilization spectrum and the power range (5 MW to 30 MW). At last, CEA has recently launched the preliminary project study of a new MTR, the Jules Horowitz Reactor, to meet the future needs of fuels and materials irradiations in the next 4 or 5 decades, in support of the French long term nuclear power program. (J.P.N.)

Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium

Directory of Open Access Journals (Sweden)

Baschwitz Anne

2016-01-01

Full Text Available International organizations regularly produce global energy demand scenarios. To account for the increasing population and GDP trends, as well as to encompass evolving energy uses while satisfying constraints on greenhouse gas emissions, long-term installed nuclear power capacity scenarios tend to be more ambitious, even after the Fukushima accident. Thus, the amounts of uranium or plutonium needed to deploy such capacities could be limiting factors. This study first considers light-water reactors (LWR, GEN III using enriched uranium, like most of the current reactor technologies. It then examines the contribution of future fast reactors (FR, GEN IV operating with an initial fissile load and then using depleted uranium and recycling their own plutonium. However, as plutonium is only available in limited quantity since it is only produced in nuclear reactors, the possibility of starting up these Generation IV reactors with a fissile load of enriched uranium is also explored. In one of our previous studies, the uranium consumption of a third-generation reactor like an EPR™ was compared with that of a fast reactor started up with enriched uranium (U5-FR. For a reactor lifespan of 60 years, the U5-FR consumes three times less uranium than the EPR and represents a 60% reduction in terms of separative work units (SWU, though its requirements are concentrated over the first few years of operation. The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand. The deployment of light-water reactors only is not sustainable in the long run. Generation IV reactors are therefore essential. Yet when started up with plutonium, the number of reactors that can be deployed is also limited. In a fleet deployment

Fast Reactor Physics. Vol. II. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

International Nuclear Information System (INIS)

1968-01-01

Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

The fuel of nuclear reactors

International Nuclear Information System (INIS)

1995-03-01

This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2002-01-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2002-04-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

International Nuclear Information System (INIS)

Hagrman, D.T.

1995-06-01

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident

Water treatment process for nuclear reactors

International Nuclear Information System (INIS)

Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

1992-01-01

Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigate water treatment process for nuclear reactor utilization. Analysis of output water chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerates to obtain the optimum quantity of pure water which reached to 15 cubic meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30 %. output water chemistry agree with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined.5 fig., 3 tab

Permeated defect detecting test method and device in reactor

International Nuclear Information System (INIS)

Sakurai, Yoshishige.

1996-01-01

The present invention provides a method of and a device capable of performing a test for entire inner surfaces of the reactor upon periodical inspection of a BWR type reactor while sufficiently taking countermeasures for radiation rays into consideration. Namely, the present invention comprises following steps. (1) A provisional step for taking a shroud head of a reactor core shroud and incore structural components above and below the shroud out of the reactor, discharging reactor water and water tightly closing openings such as reactor wall perforation holes, (2) a pretreatment step for washing exposed inner surfaces of the reactor and peeling deteriorated materials, (3) a first drying step for drying portions washed and peeled in the step (2), (4) a permeation step for applying a permeation liquid of a defect detecting medium on the exposed inner surfaces of the reactor, (5) a permeation liquid removing step for removing the an excess permeation liquid in the step (4), (6) a second drying step for drying corresponding portions after performing the step (5), and (7) a flaw detecting step for optically observing the corresponding portions after performing the step (6) and detecting flaws. (I.S.)

MLR reactor

International Nuclear Information System (INIS)

Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

1998-01-01

The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

1982-01-01

A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

Introduction to reactor internal materials for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

Introduction to reactor internal materials for pressurized water reactor

International Nuclear Information System (INIS)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application

International Nuclear Information System (INIS)

Freels, J.D.

1993-01-01

This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as ''the code''). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed

Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

International Nuclear Information System (INIS)

Rubini, L.A.

1978-02-01

An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

Perception of Rural and Urban Mothers about Consumption of Targeted Fortified Products in Jaipur, Rajasthan - India: A Cross-Sectional Study.

Science.gov (United States)

Nagaraj, Anup; Yousuf, Asif; Ganta, Shravani

2013-01-01

Food fortification is the addition of one or more essential nutrients to a food whether or not it is normally contained in the food, for the purpose of preventing or correcting a demonstrated deficiency of one or more nutrients in the population or specific population groups. The present cross-sectional study was conducted to obtain comprehensive information towards consumption of Targeted Fortified Products (TFP) among rural and urban mothers of children rural and urban mothers were selected from Primary Health Centre, Achrol Village and Uniara Hospital in Jaipur, Rajasthan. The data were collected using a selfadministered questionnaire. The current nutritional status of children was determined by anthropometric measurements. A total of 53.33% rural and 65.33% urban mothers had knowledge (P=0.046), amongst which 52.67% rural and 66.00% urban mothers (P= 0.026) consumed TFP directed towards mothers. In addition, 56% rural and 94.67% urban mothers had knowledge (P=0.000) about TFP directed towards children, amongst which 19.33% rural and 50.67% urban mothers (P=0.000) fed their children with TFP. There was significantly less awareness regarding consumption of TFP directed towards both pregnant women and children among rural when compared to urban population. Attitudes seemed to be governed by the traditional beliefs and family influences of region rather than the knowledge obtained. There is a need of intensive coordinated efforts to create awareness among mothers to enhance the maternal and child diets through TFP.

Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

International Nuclear Information System (INIS)

Tukiran S; Surian Pinem; Tagor MS; Lily S; Jati Susilo

2012-01-01

The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm 2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm 2 sand length of cycle is 57 days is the most optimal of IRR. (author)

Backfitting swimming pool reactors

International Nuclear Information System (INIS)

Roebert, G.A.

1978-01-01

Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2000-01-01

SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

Fusion Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2000-07-01

SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors

Energy Technology Data Exchange (ETDEWEB)

Khedr, A.; Abdel-Latif, Salwa H. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt); Abdel-Hadi, Eed A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; D' Auria, F. [Pisa Univ. (Italy)

2016-03-15

In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

International Nuclear Information System (INIS)

Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying; Lee, Hyung-Sool; Wang, Ai-Jie

2012-01-01

Highlights: ► A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. ► Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). ► PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. ► The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 ± 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m −3 d −1 ) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m −3 d −1 (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

Energy Technology Data Exchange (ETDEWEB)

Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China); Lee, Hyung-Sool [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West Waterloo, Ontario, Canada N2L 3G1 (Canada); Wang, Ai-Jie, E-mail: waj0578@hit.edu.cn [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China)

2012-11-15

Highlights: Black-Right-Pointing-Pointer A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. Black-Right-Pointing-Pointer Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). Black-Right-Pointing-Pointer PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. Black-Right-Pointing-Pointer The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 {+-} 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m{sup -3} d{sup -1}) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m{sup -3} d{sup -1} (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

Nuclear reactors to come

International Nuclear Information System (INIS)

Lung, M.

2002-01-01

The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

A friendly Maple module for one and two group reactor model

International Nuclear Information System (INIS)

Baptista, Camila O.; Pavan, Guilherme A.; Braga, Kelmo L.; Silva, Marcelo V.; Pereira, P.G.S.; Werner, Rodrigo; Antunes, Valdir; Vellozo, Sergio O.

2015-01-01

The well known two energy groups core reactor design model is revisited. A simple and friendly Maple module was built to cover the steps calculations of a plate reactor in five situations: 1. one group bare reactor, 2. two groups bare reactor, 3. one group reflected reactor, 4. 1-1/2 groups reflected reactor and 5. two groups reflected reactor. The results show the convergent path of critical size, as it should be. (author)

A friendly Maple module for one and two group reactor model

Energy Technology Data Exchange (ETDEWEB)

Baptista, Camila O.; Pavan, Guilherme A.; Braga, Kelmo L.; Silva, Marcelo V.; Pereira, P.G.S.; Werner, Rodrigo; Antunes, Valdir; Vellozo, Sergio O., E-mail: camila.oliv.baptista@gmail.com, E-mail: pavanguilherme@gmail.com, E-mail: kelmo.lins@gmail.com, E-mail: marcelovilelasilva@gmail.com, E-mail: rodrigowerner@hotmail.com, E-mail: neutron201566@yahoo.com, E-mail: vellozo@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear

2015-07-01

The well known two energy groups core reactor design model is revisited. A simple and friendly Maple module was built to cover the steps calculations of a plate reactor in five situations: 1. one group bare reactor, 2. two groups bare reactor, 3. one group reflected reactor, 4. 1-1/2 groups reflected reactor and 5. two groups reflected reactor. The results show the convergent path of critical size, as it should be. (author)

Ageing management of the BR2 research reactor

International Nuclear Information System (INIS)

Verpoortem, J. R.; Van Dyck, S.

2014-01-01

At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

TREAT Reactor Control and Protection System

International Nuclear Information System (INIS)

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

Assessment of torsatrons as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Painter, S.L.

1992-12-01

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R 0 = 6.6-8.8 m, on-axis magnetic field B 0 = 4.8-7.5 T, B max (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions

Consequence analysis for nuclear reactors, Yongbyon

International Nuclear Information System (INIS)

Kang, Taewook; Jae, Moosung

2017-01-01

Since the Fukushima nuclear power plant accidents in 2011, there have been an increased public anxiety about the safety of nuclear power plants in Korea. The lack of safeguards and facility aging issues at the Yongbyon nuclear facilities have increased doubts. In this study, the consequence analysis for the 5-MWe graphite-moderated reactor in North Korea was performed. Various accident scenarios including accidents at the interim spent fuel pool in the 5-MWe reactor have been developed and evaluated quantitatively. Since data on the design and safety system of nuclear facilities are currently insufficient, the release fractions were set by applying the alternative source terms made for utilization in the analysis of a severe accident by integrating the results of studies of severe accidents occurred before. The calculation results show the early fatality zero deaths and latent cancer fatality about only 13 deaths in Seoul. Thus, actual impacts of a radiological release will be psychological in terms of downwind perceptions and anxiety on the part of potentially exposed populations. Even considering the simultaneous accident occurrence in both 5-MWe graphite-moderated reactor and 100-MWt light water reactor, the consequence analysis using the MACCS2 code shows no significant damage to people in South Korea. (author)

Designing a mini subcritical nuclear reactor

International Nuclear Information System (INIS)

Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M.

2015-10-01

In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of 239 PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of 239 PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k e -f f , the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k eff , the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons 239 PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k eff was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k eff of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five cylinders not met. (Author)

Remote dismantlement tasks for the CP5 reactor: Implementation, operations, and lessons learned

International Nuclear Information System (INIS)

Noakes, M.W.

1998-01-01

This paper presents a developer's perspective on lessons learned from one example of the integration of new prototype technology into a traditional operations environment. The dual arm work module was developed by the Robotics Technology Development Program as a research and development activity to examine manipulator controller modes and deployment options. It was later reconfigured for the dismantlement of the Argonne National Laboratory Chicago Pile number-sign 5 reactor vessel as the crane-deployed dual arm work platform. Development staff worked along side operations staff during a significant part of the deployment to provide training, maintenance, and tooling support. Operations staff completed all actual remote dismantlement tasks. At the end of available development support funding, the Dual Arm Work Platform was turned over to the operations staff, who is still using it to complete their dismantlement tasks

Factors Influencing Institutional-Based Pediatric Rehabilitation Services among Caregivers of Children with Developmental Delay in Southwestern Rajasthan.

Science.gov (United States)

Mishra, Kriti; Siddharth, V

2018-01-01

A limited number of caregivers of children with developmental delay access rehabilitation facilities in India. The study explored utilization of rehabilitation services at a tertiary care setup in southwestern Rajasthan and various factors influencing it. The aim of this study is to explore rehabilitation service utilization among children with developmental delay at a tertiary care setup and to ascertain factors that influence this pattern. This study was conducted at the department of physical medicine and rehabilitation at tertiary care setup. This was an observational study. Children with developmental delay who were advised institutional-based rehabilitation were identified over span of 1 year. Those who failed to return for rehabilitation after the first visit were interviewed telephonically. The interview had semi-structured open-ended questions about their reasons for inability to avail services. SPSS statistics 22 was used for descriptive analysis and correlation of variables. Of 230 children with developmental delay visiting department in 1-year duration, 48 took regular rehabilitation. Parents of 129 children with complete records were asked regarding discontinuation. Factors cited by majority were long distance from institute and service at hospital. Other reasons for discontinuation were related to belief system, family issues, time issues, socioeconomic factors, etc. Socioeconomic status was significantly associated with parental education (C = 0.488, P = 0.000) and financial issues. Location of family had significant association with long distance (C = 0.315, P = 0.000), parental education (C = 0.251, P = 0.003), and belief system (C = 0.265, P = 0.002). Distance from institute and quality of hospital service determined rehabilitation service use at a tertiary institute. Other factors such as socioeconomic status, family support, and social belief system must also be addressed while delivering institutional rehabilitation to children.

Factors influencing institutional-based pediatric rehabilitation services among caregivers of children with developmental delay in Southwestern Rajasthan

Directory of Open Access Journals (Sweden)

Kriti Mishra

2018-01-01

Full Text Available Context: A limited number of caregivers of children with developmental delay access rehabilitation facilities in India. The study explored utilization of rehabilitation services at a tertiary care setup in southwestern Rajasthan and various factors influencing it. Aims: The aim of this study is to explore rehabilitation service utilization among children with developmental delay at a tertiary care setup and to ascertain factors that influence this pattern. Settings: This study was conducted at the department of physical medicine and rehabilitation at tertiary care setup. Design: This was an observational study. Subjects and Methods: Children with developmental delay who were advised institutional-based rehabilitation were identified over span of 1 year. Those who failed to return for rehabilitation after the first visit were interviewed telephonically. The interview had semi-structured open-ended questions about their reasons for inability to avail services. Statistical Analysis: SPSS statistics 22 was used for descriptive analysis and correlation of variables. Results: Of 230 children with developmental delay visiting department in 1-year duration, 48 took regular rehabilitation. Parents of 129 children with complete records were asked regarding discontinuation. Factors cited by majority were long distance from institute and service at hospital. Other reasons for discontinuation were related to belief system, family issues, time issues, socioeconomic factors, etc. Socioeconomic status was significantly associated with parental education (C = 0.488, P = 0.000 and financial issues. Location of family had significant association with long distance (C = 0.315, P = 0.000, parental education (C = 0.251, P = 0.003, and belief system (C = 0.265, P = 0.002. Conclusions: Distance from institute and quality of hospital service determined rehabilitation service use at a tertiary institute. Other factors such as socioeconomic status, family support, and

Frequency and correlation of lip prints, fingerprints and ABO blood groups in population of Sriganganagar District, Rajasthan.

Science.gov (United States)

Sandhu, Harpreet; Verma, Pradhuman; Padda, Sarfaraz; Raj, Seetharamaiha Sunder

2017-11-01

To investigate the frequency and uniqueness of different lip print patterns, fingerprint patterns in relation to gender and ABO Rh blood groups among a semi-urban population of Sriganganagar, Rajasthan. The study was conducted on 1200 healthy volunteers aged 18-30 years. The cheiloscopic and dermatographic data of each subject were obtained and were analysed according to the Suzuki and Tsuchihashi and Henry systems of classification, respectively. Two forensic experts analyzed the patterns independently. The ABO Rh blood group was also recorded for each subject. The Chi square statistical analysis was done and tests were considered significant when p value blood group was noted as most common in both genders while least common were A- among males and AB- in females. Type II lip pattern was most predominant while the least common was Type I' in males and Type I' and Type V in females. The UL fingerprint pattern was the most common, while RL was least noted in both genders. All the fingerprint patterns showed correlation with different lip print patterns. A correlation was found between different blood groups and lip print patterns except Type I (vertical) lip pattern. A positive correlation was observed between all the blood groups and fingerprint patterns, except for RL pattern. There is an association between lip print patterns, fingerprint patterns and ABO blood groups in both the genders. Thus, correlating the uniqueness of these physical evidences sometimes helps the forensic team members in accurate personal identification or it can at least narrow the search for an individual where there are no possible data referring to the identity of the subject. Copyright © 2017 by Academy of Sciences and Arts of Bosnia and Herzegovina.

Awareness and Attitude of Physicians in Academia towards Human Stem Cell Research (HSCR and Related Policies in Rajasthan, India

Directory of Open Access Journals (Sweden)

Nitin K Joshi

2015-12-01

Full Text Available Introduction: In India, several science agencies are promoting Stem Cell Research (SCR. There is paucity of studies which document the perception of doctors about SCR, especially physicians in academia. This study was carried out to assess perception of physicians in academia towards Human Stem Cell Research (HSCR and related policies in India. Methods: We interviewed 200 doctors from three different government medical colleges of Rajasthan. A semi-structured questionnaire was used to discern their awareness, attitudes towards utilization of SCR and their knowledge of related international and ethical policy issues. Results: Though mostly 177 (96.2% physicians acknowledged the public health benefits of promoting stem cell research in India, but 166 (66.2% were not aware of the stem cell research policy of the Government of India and 111 (60.3% were not aware of the ICMR guidelines for Human Stem Cell Research in India. There was a strong desire among academic physicians 152 (82.6% to incorporate a course on SCR to the students in the near future. Discussion: Physicians in academia have views that SCR should be encouraged to treat clinical diseases and this technology should be brought into India in a big way. They seem to believe that one of the ways to promote the benefits of SCR would be to raise awareness by publishing success stories in widely read Indian Medical Journals, giving updated information regarding its uses in clinical practices and its inclusion as a part of the curricula for health professionals.

Relocation and Livelihood Concerns of Sariska Tiger Project, Rajasthan: A Pride or Plight?

Directory of Open Access Journals (Sweden)

Muraree Lal Meena

2015-11-01

Full Text Available It is well known that Sariska Tiger Reserve is a home to the India’s national animal - the Bengal Tiger (Panthera tigris tigris. The crux of this research is to examine the role of local peoples in the conservation of Sariska Tiger project, which was declared a wildlife reserve way back in 1955 and then further raised to a status of a Tiger Reserve in 1978, and a National Park in 1982. According to the Government officials, the people around the reserve are not only responsible for degrading the reserve, it has also emerged as a safe haven for the poachers involved in illegal hunting of the tigers. One of the reasons identified to be the cause of tiger extinction is the human habitation in the core and in the peripheral areas surrounding the tiger reserves. Despite efforts being made to protect, the Sariska Tiger Reserve, located in Alwar (Rajasthan is seriously facing the problem of tiger extinction. It is estimated that there are around 12 thousand people residing inside the tiger reserve, with 11 villages in the Core Zone-1 area and about 170 villages situated along the peripheries of the reserve. Among the prime measures undertaken are diversions of traffic, relocation of the villages located inside the forest. The Tiger Task Force (2005 has recommended relocation of three key villages surrounding the central area of the Sariska Tiger Reserve. However, this relocation of villages is quite contrary to the life style of the people residing within these villages. Their social and cultural attributes, needs and aspirations have not been given due consideration and the villagers are strongly resenting this move by the government. These displaced villagers have been living in perfect harmony with their environment from time immemorial and it is hard to understand how these villagers can be held responsible for degrading their natural environment, which is their lifeline. The government has failed to take into consideration the role of the

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

Savannah River Site production reactor safety analysis report

International Nuclear Information System (INIS)

1996-01-01

The process water system (PWS) is designed to remove heat produced in the reactor from the fission process, gamma radiation absorption, and fission product decay. Heat removal is accomplished by circulating heavy water through the reactor. Cooling is provided for fuel assemblies, target assemblies, control rods, bulk moderator, deflector plate, reactor tank, and reactor structural components. Approximately 90% of the heat load is generated in the fuel and target assemblies, 5% in the moderator, and 5% in the shielding. In addition to serving as the-heat transfer medium, the process water moderates neutrons produced by fission in the fuel. D 2 O is used in this application because of its favorable moderating and neutron capture properties, which result in high neutron efficiency and reactor productivity. The PWS piping and components also provide a high-integrity leak barrier against loss of moderator and the radioactive fission and corrosion products. Components of the PWS are located in the reactor building between the -40-foot elevation and the 0-foot elevation. Specific locations include the process room, heat exchanger bay, motor rooms, and pump rooms. The system diagram is shown in Figure 5.1-2. PWS design data are presented in Table 5.1-1. The PWS consists of six parallel heat transfer loops. In each loop, approximately 25,000 gpm of D 2 O is circulated from one of six outlet nozzles in the bottom of the reactor tank through a motor-operated valve (MOV) to the suction side of the process water pump. Each pump is driven by an AC motor and a DC motor through a gear reducer unit. A 3-ton flywheel on the drive shaft of the AC motor provides gradual flow coastdown when power is lost. During reactor operation, the DC motors are operated continuously from the diesel generator sets as backup to the AC motors. Following shutdown, the DC motors are operated to provide adequate circulation and core cooling

WWER-440 type reactor core

International Nuclear Information System (INIS)

Mizov, J.; Svec, P.; Rajci, T.

1987-01-01

Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

International Nuclear Information System (INIS)

Obara, Toru; Sekimoto, Hiroshi

1997-01-01

The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

Chemical surveillance of commercial fast breeder reactors

International Nuclear Information System (INIS)

Stamm, H.H.; Stade, K.Ch.

1988-01-01

After BN-600 (USSR) and SUPERPHENIX (France) were started succesfully, the international development of LMFBRs is standing at the doorstep of commercial use. For commercial use of LMFBRs cost reductions for construction and operation are highly desirable and necessary. Several nations developing breeder reactors have joined in a common effort in order to reach this aim by standardization and harmonization. On the base of more than 20 years of operation experience of experimental reactors (EBR-II, FFTF, RAPSODIE, DFR, BR-5/BR-10, BOR-60, JOYO, KNK-II) and demonstration plants (PHENIX, PFR, BN-350), possibilities for standardization in chemical surveillance of commercial breeder reactors without any loss of availability, reliability and reactor safety will be discussed in the following chapters. Loop-type reactors will be considered as well as pool-type reactors, although all commercial plants under consideration so far (SUPERPHENIX II, BN-800, BN-1600, CFBR, SNR-2, EFR) include pool-type reactors only. Table 1 gives a comparison of the Na inventories of test reactors, prototype plants and commercial LMFBRs

Seventh regular meeting of the International Working Group on Reliability of Reactor Pressure Components, Vienna, 3-5 September 1985

International Nuclear Information System (INIS)

1986-07-01

The seventh regular meeting of the IAEA International Working Group on Reliability of Reactor Pressure Components was held at the Agency's Headquarters in Vienna from 3 to 5 September 1985. The representatives of Member States and of the Commission of the European Communities reported the status of the research programmes in this field (12 presentations). A separate abstract was prepared for each of the presentations

Safety analysis of RA reactor operation, I-II, Part I - RA reactor technical and operation characteristics; Analiza sigurnosti rada reaktora RA - I-III, I deo - Tehnicke i pogonske karakteristike reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

RA research reactor is a thermal, heavy water moderated system with graphite reflector having nominal power 6.5 MW. The 2% enriched metal uranium fuel in the reactor core produces mean thermal neutron flux of 2.9 10{sup 13} neutrons/cm{sup 2} s, and maximum neutron flux 5.5 10{sup 13} neutrons/cm{sup 2} s. main components of the reactor described in this report are: rector core, reflector, biological shield, heavy water cooling system, ordinary water cooling system, helium system, reactor control system, reactor safety system, dosimetry system, power supply system, and fuel transport system. Detailed reactor properties and engineering drawings of all the system are part of this volume.

Savannah River Site reactor hardware design modification study

International Nuclear Information System (INIS)

Fisher, J.E.

1990-01-01

A study was undertaken to assess the merits of proposed design modifications to the Savannah River Site (SRS) reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. System recovery potential was evaluated for break locations at the pump suction, the pump discharge, and the plenum inlet. The code version used was RELAP5/MOD2.5 version 3d3, a preliminary version of RELAP5/MOD3. The model was a three-dimensional representation of the K-Reactor water plenum and moderator tank. It included explicit representations of all six loops, which were based on the configuration of L-Reactor. A combination of features is recommended to ensure liquid inventory recovery for all break locations. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 7 refs., 10 figs., 2 tabs

Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 3: Hardware component failure data; Volume 5, Revision 4

International Nuclear Information System (INIS)

Reece, W.J.; Gilbert, B.G.; Richards, R.E.

1994-09-01

This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data

Reactor types for the future

International Nuclear Information System (INIS)

Hall, A.C.

1990-01-01

The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

Reactor types for the future

Energy Technology Data Exchange (ETDEWEB)

Hall, A C [PWR Power Projects Ltd., Knutsford, Cheshire (United Kingdom)

1990-06-01

The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

First preliminary design of an experimental fusion reactor

International Nuclear Information System (INIS)

1977-09-01

A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

Neutron behavior, reactor control, and reactor heat transfer. Volume four

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

Evaluation of environmental radioactivity monitoring data around the Kartini Reactor area

International Nuclear Information System (INIS)

Yazid, M; Sutrisno; Sukarman-Aminjoyo; Zaenal-Abidin

1996-01-01

Evaluation of environmental radioactivity monitoring data around the Kartini Reactor area has been done. The aim of this investigation is for tracing the possibility of radioactivity released in the environment during the operation of Kartini reactor. The data was evaluated were monthly monitored data taken from 1986 to 1994 period. The method of analysis was done by comparing the environmental radioactivity data before and after reactor commissioning, off side the reactor up to a radius of 5.000 meters and more than 5.000 meters from Kartini reactor and also compared to the maximum permissible radioactivity according to the current regulation. This evaluation showed that there was no indication of radioactivity release to the environment during this period of reactor operation

Perspectives on reactor safety

Energy Technology Data Exchange (ETDEWEB)

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

1994-03-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

Perspectives on reactor safety

International Nuclear Information System (INIS)

Haskin, F.E.

1994-03-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

Jules Horowitz reactor (RJH): its design

International Nuclear Information System (INIS)

Dupuy, J.P.

2002-01-01

This article presents the design of the new irradiation facility (Jules Horowitz reactor) that is planned to be built on the Cadarache site of Cea. 2 principles have been followed. The first one is based on a physical separation between the systems and activities related to the reactor and the experiments from one hand and the other systems and means dedicated to the treatment of the experimental devices before and after irradiation on the other hand. This first principle implies to build 2 buildings: the reactor building and the nuclear auxiliaries building. Inside the reactor building activities from the reactor itself are separated from those dedicated to experimentation. In order to maximize the efficiency of such a reactor, an important number of simultaneous experiments is expected, which will generate an endless flux of incoming and out-going experiments and as a consequence an important handling work between the different work posts. The second principle aims at easing any handling work without breaking the rules of confinement. The different storing pools, the water pits that lead to the 5 hot cells and the reactor tank will communicate through a water-filled canal that will link the 2 buildings. (A.C.)

Improving the proliferation resistance of research and test reactors

International Nuclear Information System (INIS)

Lewis, R.A.

1978-01-01

Elimination, or substantial reduction, of the trade in unirradiated highly-enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less-than-20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. To first order, this implies the requirement that the 235 U loading in the reduced-enrichment fuel elements be the same as the 235 U loading in the 90-93% enriched fuel elements. This can be accomplished by substitution of higher uranium density fuel technology for currently-used fuel technology in the fuel meat volume of the current fuel element design and/or by increasing the usable fuel meat volume. For research and test reactors of power greater than 5-10 megawatts, fuel technology does not currently exist that would permit enrichment reductions to below 20% utilizing this criterion. A program is now beginning in the U.S. to develop the necessary fuel technology. Currently-proven fuel technology is capable, however, of accommodating enrichment reductions to the 30-45% range (from 90-93%) for many reactors in the 5-50MW range. Accordingly the U.S. is proposing to convert existing reactors (and new designs) in the 5-50MW range from the use of highly-enriched fuel to the use of 30-45% enriched fuel, and reactors of less that about 5MW to less-than-20% enrichment, wherever this can be done without significant

Anaerobic digestion of cheese whey using up-flow anaerobic sludge blanket reactor

Energy Technology Data Exchange (ETDEWEB)

Yan, J.Q.; Lo, K.V.; Liao, P.H.

1989-01-01

Anaerobic treatment of cheese whey using a 17.5-litre up-flow anaerobic sludge blanket reactor was investigated in the laboratory. The reactor was studied over a range of influent concentration from 4.5 to 38.1 g chemical oxygen demand per litre at a constant hydraulic retention time of 5 days. The reactor start-up and the sludge acclimatization were discussed. The reactor performance in terms of methane production, volatile fatty acids conversion, sludge net growth and chemical oxygen demand reduction were also presented in this paper. Over 97% chemical oxygen demand reduction was achieved in this experiment. At the influent concentration of 38.1 g chemical oxygen demand per litre, an instability of the reactor was observed. The results indicated that the up-flow anaerobic sludge blanket reactor process could treat cheese whey effectively.

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

Advanced electrolytic cascade process for tritium recovery from irradiated heavy water moderator (Preprint No. PD-15)

International Nuclear Information System (INIS)

Ragunathan, P.; Mitra, S.K.; Jain, D.K.; Nayar, M.G.; Ramani, M.P.S.

1989-04-01

The paper briefly describes a design study of an electrolytic cascade process plant for enrichment and recovery of tritium from irradiated heavy water moderators from Rajasthan Atomic Power Station Reactors. In direct multistage electrolysis process, tritiated heavy water from the reactor units is fed to the electrolytic cell modules arranged in the form of a cascade where it is enriched and decomposed into O 2 gas stream and D 2 /DT gas stream. The direct electrolysis of tritiated heavy water allows tritium to be concentrated in the aqueous phase. Several stages are used to achieve the necessary enrichment. The cascade plant incorporates the advanced electrolyser technology developed in Bhabha Atomic Research Centre (Bombay) using porous nickel electrodes, capable o f high current density operation at reduced energy consumption for electrolysis. (author). 3 tabs

Safety features of TR-2 reactor

International Nuclear Information System (INIS)

Tuerker, T.

2001-01-01

TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay

International Nuclear Information System (INIS)

Dhami, P.S; Yadav, J.S; Agarwal, K.

2017-01-01

Exploitation of the abundant thorium resources to meet sustained energy demand forms the basis of the Indian nuclear energy programme. To gain reprocessing experience in thorium fuel cycle, thoria was irradiated in research reactor CIRUS in early sixties. Later in eighties, thoria bundles were used for initial flux flattening in some of the pressurized heavy water reactors (PHWRs). The research reactor irradiated thoria contained small content (∼ 2-3ppm) of "2"3"2U in "2"3"3U product, which did not pose any significant radiological problems during processing in Uranium Thorium Separation Facility (UTSF), Trombay. Thoria irradiated in PHWRs on discharge contained (∼ 0.5-1.5% "2"3"3U with significant "2"3"2U content (100-500 ppm) requiring special radiological attention. Based on the experience from UTSF, a new facility viz. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay was built which was hot commissioned in the year 2015

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Kinetic analysis of sub-prompt-critical reactor assemblies

International Nuclear Information System (INIS)

Das, S.

1992-01-01

Neutronic analysis of safety-related kinetics problems in experimental neutron multiplying assemblies has been carried out using a sub-prompt-critical reactor model. The model is based on the concept of a sub-prompt-critical nuclear reactor and the concept of instantaneous neutron multiplication in a reactor system. Computations of reactor power, period and reactivity using the model show excellent agreement with results obtained from exact kinetics method. Analytic expressions for the energy released in a controlled nuclear power excursion are derived. Application of the model to a Pulsed Fast Reactor gives its sensitivity between 4 and 5. (author). 6 refs., 4 figs., 1 tab

Nuclear reactor building

International Nuclear Information System (INIS)

Oshima, Nobuaki.

1991-01-01

The secondary container in a nuclear reactor building is made of a transparent structure having a shielding performance such as lead glass, by which the inside of the secondary container can be seen without undergoing radiation exposure. In addition, an operator transportation facility capable of carrying about 5 to 10 operators at one time is disposed, and the side of the facility on the secondary container is constituted with a transparent material such as glass, to provide a structure capable of observing the inside of the secondary container. The ventilation and air conditioning in the operator's transportation facility is in communication with the atmosphere of a not-controlled area. Accordingly, operators at the outside of the reactor building can reach the operator's transportation facility without taking and procedures for entering the controlled area and without undergoing radiation exposure. The inside of the secondary container in the reactor building can be seen from various directions through the transparent structure having the shielding performance. (N.H.)

Advances in Reactor Physics, Mathematics and Computation. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

Mirror reactor blankets

International Nuclear Information System (INIS)

Lee, J.D.; Barmore, W.L.; Bender, D.J.; Doggett, J.N.; Galloway, T.R.

1976-01-01

The general requirements of a breeding blanket for a mirror reactor are described. The following areas are discussed: (1) facility layout and blanket maintenance, (2) heat transfer and thermal conversion system, (3) materials, (4) tritium containment and removal, and (5) nuclear performance

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

Clinical experience with insulin detemir, biphasic insulin aspart and insulin aspart in people with type 2 diabetes: Results from the Rajasthan cohort of the A 1 chieve study

Directory of Open Access Journals (Sweden)

Akhil Joshi

2013-01-01

Full Text Available Background: The A 1 chieve, a multicentric (28 countries, 24-week, non-interventional study evaluated the safety and effectiveness of insulin detemir, biphasic insulin aspart and insulin aspart in people with T2DM (n = 66,726 in routine clinical care across four continents. Materials and Methods: Data was collected at baseline, at 12 weeks and at 24 weeks. This short communication presents the results for patients enrolled from Rajasthan, India. Results: A total of 477 patients were enrolled in the study. Four different insulin analogue regimens were used in the study. Patients had started on or were switched to biphasic insulin aspart (n = 340, insulin detemir (n = 90, insulin aspart (n = 37, basal insulin plus insulin aspart (n = 7 and other insulin combinations (n = 2. At baseline glycaemic control was poor for both insulin naïve (mean HbA 1 c: 8.3% and insulin user (mean HbA 1 c: 8.4% groups. After 24 weeks of treatment, both the groups showed improvement in HbA 1 c (insulin naïve: −0.9%, insulin users: −1.2%. Major hypoglycaemic events decreased from 0.5 events/patient-year to 0.0 events/patient-year in insulin naïve group while no change from baseline (1.3 events/patients-year was observed for insulin users. SADRs were not reported in any of the study patients. Conclusion: Starting or switching to insulin analogues was associated with improvement in glycaemic control with a low rate of hypoglycaemia.

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing