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Sample records for rajasthan-1 reactor

  1. Production and quality control of concrete for the Rajasthan Atomic Power Station - [Part 2

    International Nuclear Information System (INIS)

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

    1975-01-01

    The following aspects of the production and quality control of concrete and concrete materials used in the construction of twin-reactor Rajasthan Atomic Power Station are discussed : (1) relationship between strength of cubes and cylinders made of concrete used for the prestressed dome (2) temperature control during pouring of concrete (3) thermal conductivity of heavy concrete (4) various types of grouting procedures used for different structures forming part of reactors (5) quality control of normal and heavy concrete and (6) leakage through form ties. Typical concrete mixes used for grouts are also given. (M.G.B.)

  2. Flow and pressure profiles for the primary heat transport system of Rajasthan Atomic Power Station for the operation with few isolated reactor channels near the end shield cracks

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A J; Chaki, S K; Sehgal, R L; Venkat Raj, V [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    The RAPS (Rajasthan Atomic Power Station) unit-1 is now operating at reduced power due to the removal of fifteen fuel channels for repair of south end shield cracks. The power level is restricted to 50% of the full power capacity as a precautionary measure. The relative difference that operation at 50% power and higher power would make to the end shield structure is being currently analysed with a view to operate this reactor at higher power levels. As a prerequisite, a detailed thermal hydraulic analysis is essential to assess the effect of reactor operation with isolated channels on the primary heat transport (PHT) system pressure, flow, temperature. The adequacy of the existing trip set points for the plant operation under this mode is also required to be assessed. In the present study, analysis of the PHT system has been carried out to determine the flow and pressure profiles for the RAPS heat transport system for operation of the reactor with isolated channels. (author). 5 refs., 1 fig., 1 tab.

  3. Prospecting Organic production of spices in Rajasthan

    OpenAIRE

    Sharma, Arun K.

    2015-01-01

    Low rainfall and low atmospheric humidity favours spices production in Rajasthan, the largest state of India.Details on possibilities of organic production of spices in Rajasthan are described in the paper considering the future challenges of soil, climate,social and market environment.

  4. Interlinking feasibility of five river basins of Rajasthan in India

    OpenAIRE

    Vyas, Sunil Kumar; Sharma, Gunwant; Mathur, Y.P.; Chandwani, Vinay

    2016-01-01

    The increasing population and large scale growth with the development of modern science and technology has indicated very high stress on water sector in Rajasthan in India. Availability of water and uniformity of rainfall distribution is changing day by day due to shifting of monsoon in Rajasthan. The spatial and temporal variations in the rainfall in different river basins in Rajasthan are drastic due to which flood situation arises in the tributaries of Chambal river basin every year. Simul...

  5. Experience from the construction and operation of Tarapur and Rajasthan Nuclear Power Stations

    International Nuclear Information System (INIS)

    Shah, J.C.; Pardiwala, T.F.; Kothare, V.V.; Rao, M.H.P.; Nanjundeswaran, K.

    1977-01-01

    India's experience in construction and operation of nuclear power stations so far covers two BWR and four PHWR Units in three power stations. Two more PHWR units are at an early stage of construction. The twin unit Tarapur Station (2x210 MWe BWR) was built as a turnkey project which restricted participation of Indian engineers in design and construction considerably. The contrasting approach adopted for Rajasthan Station (2x220 MWe PHWR) involved Indian personnel and contractors fully in construction and commissioning, with Canadians providing supervisory assistance in Rajasthan I and essentially consultative help for Rajasthan II. Subsequent stations are wholly Indian efforts. Tarapur went into commercial operation in 1969, 60 months after breaking of ground. Construction was essentially uneventful, major problems faced being stress corrosion induced cracks in the reactor lining and complete change of steam generator tubes. In its seven years of operations, Tarapur has faced several problems mainly arising from rather early designs, indifferent fuel performance, constraints of twin-unit approach and operations in an inadequately developed grid system apart from those generally stemming from assimilation of an advanced technology in a developing country. The Station has undergone six refuellings during this period. Most of the problems have been overcome by design changes, system augmentations and experience and the Station operation since mid 1974 have generally been steady at around 90% of the rated capacity. Construction of Rajasthan I at a remote and isolated site proceeded relatively slowly. Local availability of skilled and semi-skilled manpower was poor, affecting construction. Inadequate roads impeded movements of overdimensioned components. Observing strict Quality Assurance standards required several major rectifications at site. Rajasthan I went on line in 1973 after overcoming major turbine bearing problems during commissioning. Since then, while

  6. Performance of on-power fuelling equipment at Rajasthan Atomic Power Station

    International Nuclear Information System (INIS)

    Jayabarathan, S.; Gopalakrishnan, S.

    1977-01-01

    Natural uranium reactors on account of their intrinsically low reactivity need frequent refuelling. The Rajasthan Atomic Power Station based on natural uranium reactors has, therefore, been provided with on-power fuel handling system which was installed in 1972. Its performance has met the design intent and operational objectives which are enumerated. However, continuous fuelling 7 to 10 days has not been possible because frequent maintenance of refuelling system is needed on account of certain deficiencies major of which is the heavy water leakage. For better performance, installation of a programmable logic controller is suggested. Mention has also been made of inadequate number of skilled man-power required for maintenance which leads to quick depletion of man-rem of all the available personnel trained for maintenance work. (M.G.B.)

  7. Capturing Energy-Saving Opportunities: Improving Building Efficiency in Rajasthan through Energy Code Implementation

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Qing [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Yu, Sha [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Evans, Meredydd [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mathur, Jyotirmay [Malaviya National Institute of Technology, Jaipur (India); Vu, Linh D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-01

    India adopted the Energy Conservation Building Code (ECBC) in 2007. Rajasthan is the first state to make ECBC mandatory at the state level. In collaboration with Malaviya National Institute of Technology (MNIT) Jaipur, Pacific Northwest National Laboratory (PNNL) has been working with Rajasthan to facilitate the implementation of ECBC. This report summarizes milestones made in Rajasthan and PNNL's contribution in institutional set-ups, capacity building, compliance enforcement and pilot building construction.

  8. First record of Galeodes indicus Pocock, 1900 (Arachnida: Solifugae: Galeodidae from Rajasthan, India

    Directory of Open Access Journals (Sweden)

    Ruquaeya Bano

    2016-03-01

    Full Text Available During a regular survey to collect soil arthropods in Lasiurus sindicus Henrard grassland by pitfall methods at Chandan Village near Jaisalmer City, Rajasthan, we found a dead specimen of Galeodes indicus in a sample.  Galeodes indicus (Pocock, 1900 has been reported from Madhya Pradesh, Maharashtra, Andhra Pradesh and Telangana but so far was unknown to Rajasthan, India.  In this communication, we report Galeodes indicus from Jaisalmer District, Rajasthan, India. 

  9. Real-time Extremely Heavy Rainfall Forecast and Warning over Rajasthan During the Monsoon Season (2016)

    Science.gov (United States)

    Srivastava, Kuldeep; Pradhan, D.

    2018-01-01

    Two events of extremely heavy rainfall occurred over Rajasthan during 7-9 August 2016 and 19-21 August 2016. Due to these events, flooding occurred over east Rajasthan and affected the normal life of people. A low-pressure area lying over northwest Madhya Pradesh on 7 August 2016 moved north-westward and lay near east Rajasthan and adjoining northwest Madhya Pradesh on 8 and 9 August 2016. Under the influence of this low-pressure system, Chittorgarh district and adjoining areas of Rajasthan received extremely heavy rainfall of 23 cm till 0300 UTC of 8 August 2016 and 34 cm on 0300 UTC of 9 August 2016. A deep depression lying over extreme south Uttar Pradesh and adjoining northeast Madhya Pradesh on 19 August 2016 moved westward and gradually weakened into a depression on 20 August 2016. It further weakened into a low-pressure area and lay over east Rajasthan on 21 and 22 August 2016. Under the influence of this deep depression, Jhalawar received 31 cm and Baran received 25 cm on 19 August. On 20 August 2016, extremely heavy rainfall (EHR) occurred over Banswara (23.5 cm), Pratapgarh (23.2 cm) and Chittorgarh (22.7 cm) districts. In this paper, the performance of the National Centers for Environmental Prediction (NCEP) global forecast system (GFS) model for real-time forecast and warning of heavy to very heavy/EHR that occurred over Rajasthan during 7-9 August 2016 and 19-21 August 2016 has been examined. The NCEP GFS forecast rainfall (Day 1, Day 2 and Day 3) was compared with the corresponding observed gridded rainfall. Based on the predictions given by the NCEP GFS model for heavy rainfall and with their application in real-time rainfall forecast and warnings issued by the Regional Weather Forecasting Center in New Delhi, it is concluded that the model has predicted the wind pattern and EHR event associated with the low-pressure system very accurately on day 1 and day 2 forecasts and with small errors in intensity and space for day 3.

  10. The Gypsum: White gold of Rajasthan, introduction, uses and future prospective

    Science.gov (United States)

    Sharma, Gayatri

    2013-06-01

    Rajasthan is mineral based state and Bikaner and its surrounding district have been gifted with Gypsum. Mt of Gypsum is available in these districts. Gypsum has multiple uses including basic raw material for POP industry, addition in cement and a natural fertilizer. This mineral has changes the economic scenario in the remote areas of Bikaner, Nagaur, Hanumangarh, Sanchore, Shriganganagar etc. Gypsum and selenite are mined about 3.0 million tons per year. There is huge demand from cement industry as Gypsum is added for improving setting time of cement. Gypsum is a natural fertilizer for alkaline land and it role is vital in state like India where alkaline land is major role. Its high use as fertilizer has potential to change millions of poor farmer families and improving in crop production. Cement Industry has started importing Gypsum from Thailand, Bankong, Pakistan, Iran etc. The mining of gypsum of purity of 70% CaSO4.2H2O is cooperative effort between the land owners and Rajasthan State Mines and Minerals Limited. Gypsum fulfills the demand of POP and Cement industry in Rajasthan and powder gypsum used in agriculture for recon dining of alkaline soil. This paper deals with multiple uses, availability, and future prospective of Gypsum, a white gold of Rajasthan.

  11. 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)

  12. Interlinking feasibility of five river basins of Rajasthan in India

    Directory of Open Access Journals (Sweden)

    Sunil Kumar Vyas

    2016-09-01

    Annual surplus water of about 1437 MCM in the river Chambal is going waste and ultimately reaches to sea after creating flood situations in various places in India including Rajasthan, while on the other hand 1077 MCM water is a requirement in the four other basins in Rajasthan i.e. Banas, Banganga, Gambhir and Parbati at 75% dependability. Interlinking and water transfer from Chambal to these four river basins is the prime solution for which 372 km link channel including 9 km tunnel of design capacity of 300 cumec with 64 m lift is required.

  13. 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.)

  14. Nosocomial infection of CCHF among health care workers in Rajasthan, India.

    Science.gov (United States)

    Yadav, Pragya D; Patil, Deepak Y; Shete, Anita M; Kokate, Prasad; Goyal, Pulkit; Jadhav, Santosh; Sinha, Sanjeev; Zawar, Divya; Sharma, Surendra K; Kapil, Arti; Sharma, D K; Upadhyay, Kamlesh J; Mourya, Devendra T

    2016-11-03

    Ever since Crimean-Congo hemorrhagic fever [CCHF] discovered in India, several outbreaks of this disease have been recorded in Gujarat State, India. During the year 2011 to 2015 several districts of Gujarat and Rajasthan state (Sirohi) found to be affected with CCHF including the positivity among ticks and livestock. During these years many infected individuals succumbed to this disease; which subsequently led to nosocomial infections. Herein, we report CCHF cases recorded from Rajasthan state during January 2015. This has affected four individuals apparently associated with one suspected CCHF case admitted in a private hospital in Jodhpur, Rajasthan. A 30-year-old male was hospitalized in a private hospital in Jodhpur, Rajasthan State, who subsequently had developed thrombocytopenia and showed hemorrhagic manifestations and died in the hospital. Later on, four nursing staff from the same hospital also developed the similar symptoms (Index case and Case A, B, C). Index case succumbed to the disease in the hospital at Jodhpur followed by the death of the case A that was shifted to AIIMS hospital, Delhi due to clinical deterioration. Blood samples of the index case and Case A, B, C were referred to the National institute of Virology, Pune, India for CCHF diagnosis from the different hospitals in Rajasthan, Delhi and Gujarat. However, a sample of deceased suspected CCHF case was not referred. Subsequently, blood samples of 5 nursing staff and 37 contacts (Case D was one of them) from Pokhran area, Jaisalmer district were referred to NIV, Pune. It clearly indicated that nursing staff acquired a nosocomial infection while attending the suspected CCHF case in an Intensive Care Unit of a private hospital in Jodhpur. However, one case was confirmed from the Pokhran area where the suspected CCHF case was residing. This case might have got the infection from suspected CCHF case or through other routes. CCHF strain associated with these nosocomial infections shares the

  15. Study On Harmful Effects Of Opium On Liver And Lungs In Chronic Opium Addicts Of Western Rajasthan

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    M Kamla

    2011-12-01

    Full Text Available Background: Today opium dependence is widely prevalent in certain states of India, especially Rajasthan, Punjab, Haryana, Madhya Pradesh (MP etc. In rural areas of western Rajasthan crude opium is consumed with a social acceptance by a notable proportion (8.0% of adult male population. Later on they become addicted to it.Objective: to observe the changes in some liver and lung function parameters in opium addicted subjects of Barmer city of Western Rajasthan.Methods: The present study was conducted in district hospital of Barmer, Rajasthan. Total fifty (50 adult male subjects with age ranged from 30 to 50 years were participated in this study. Among them 25 were opium addicted and were considered as study group (Group B and another 25 apparently healthy adult male of same age group were designated as control group (Group A. Opium addicts were consuming about 5–11 gm/day opium for >2 years. Then liver function tests were evaluated by estimating serum aspartate amino transferase (AST, alanine amino transferase (ALT, alkaline phosphatase and lung function tests by measuring FVC, FEV1 , FEV1/FVC% , PEF, FEF 25-75% of both the groups.Results: In this study AST, ALT and alkaline phosphatase levels were found significantly (p<.05 higher in group B as compared to those of group A. Again, FVC, FEV1, FEV1/ FVC were significantly (p<.05 lower in group B as compared to those of group A. PEF (L/sec and FEF 25-75% were also significantly (p<0.001 lower in group B as compared to those of group A.Conclusion: it is concluded that chronic long term use of opium, increases the risk of hepatic and pulmonary damage.

  16. 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

  17. 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).

  18. 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.

  19. 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.

  20. 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.

  1. Spatial distribution of tritium in the Rawatbhata Rajasthan site environment

    International Nuclear Information System (INIS)

    GilI, Rajpal; Tiwari, S.N.; Gocher, A.K.; Ravi, P.M.; Tripathi, R.M.

    2014-01-01

    Tritium is one of the most environmentally mobile radionuclides and hence has high potential for migration into the different compartments of environment. Tritium from nuclear facilities at RAPS site is released into the environment through 93 m and 100 m high stack mainly as tritiated water (HTO). The released tritium undergoes dilution and dispersion and then follows the ecological pathway of water molecule. Environmental Survey Laboratory of Health Physics Division, Bhabha Atomic Research Centre (BARC), located at Rajasthan Atomic Power Station (RAPS) site is continuously monitoring the concentration of tritium in the environment to ensure the public safety. Atmospheric tritium activity during the period (2009-2013) was measured regularly around Rajasthan Atomic Power Station (RAPS). Data collected showed a large variation of H-3 concentration in air fluctuating in the range of 0.43 - 5.80 Bq.m -3 at site boundary of 1.6 km. This paper presents the result of analyses of tritium in atmospheric environment covering an area up to 20 km radius around RAPS site. Large number of air moisture samples were collected around the RAPS site, for estimating tritium in atmospheric environment to ascertain the atmospheric dispersion and computation of radiation dose to the public

  2. Deformation of footwall rock of Phulad Shear Zone, Rajasthan ...

    Indian Academy of Sciences (India)

    Phulad Shear Zone (PSZ) of Delhi Fold Belt in Rajasthan is a northeasterly striking ductile shear zone with a well developed mylonitic foliation (035/70E) and a downdip stretching lineation. The deformation in the PSZ has developed in a transpressional regime with thrusting sense of movement. The northeastern unit, i.e. ...

  3. Antimicrobial Screening of Some Exotic Tree Species of Rajasthan Desert

    OpenAIRE

    B.B.S. Kapoor* and Shelja Pandita

    2013-01-01

    Antimicrobial screening of ethyl ether and alcoholic extracts of leaves of four selected exotic tree species growing inRajasthan Desert was carried out. Colophospermum mopane, Holoptelea integrifolia, Kigelia pinnata andPutranjiva roxburghii showed positive reactions against bacterial pathogens i.e. Staphylococcus aureus, Escherichiacoli and a fungal pathogen Candida albicans.

  4. Assessment of Fluoride Level in Groundwater and Prevalence of Dental Fluorosis in Didwana Block of Nagaur District, Central Rajasthan, India

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    M Arif

    2013-10-01

    Full Text Available Background: In India, for the high concentration of fluoride in groundwater, people are at risk of dental fluorosis. The problem is common in various states of India. The condition in Rajasthan is worse where all districts have such a problem. Objective: To study the fluoride concentration in groundwater and prevalence of dental fluorosis in Didwana block of Nagaur district, Central Rajasthan, India. Methods: The fluoride concentration in water of 54 villages was measured electrochemically, using fluoride ion selective electrode. Dental fluorosis was assessed in 1136 people residing in study area by Dean's classification for dental fluorosis. Results: The fluoride concentration in groundwater in studied sites ranged from 0.5 to 8.5 mg/L. The concentration of fluoride was more than the maximum permissible limit set by WHO and Bureau of Indian Standards (1 mg/L in 48 groundwater sources. Of 1136 people studied, 788 (69.4%; 95% CI: 66.7%–72.1% had dental fluorosis—252 had mild and 74 had severe dental fluorosis. Conclusion: High level of fluoride in drinking water of Didwana block of Nagaur district, Central Rajasthan, India, causes dental fluorosis in most people in the region and is an important health problem that needs prompt attention.

  5. Maternal Factors Associated with Nutritional Status of 1-5 years Children Residing in Field Practice Area of Rural Health Training Centre Naila, Jaipur (Rajasthan) India

    OpenAIRE

    Lokesh Sonkaria, Afifa Zafer, Kusum Lata Gaur, Ravindra Kumar Manohar

    2014-01-01

    Background: Good nutrition bene-fits families, their communities and the world as a whole. Maternal factors are important in maintaining the nutrition of 1-5 year children. Objective: To ascertain the association of maternal factors with nutrition of 1-5 year children. Materials and Methods: A community based cross sectional descriptive type of observational study was carried out in the field practice area of RHTC Naila in Jaipur district of Rajasthan. 30 Cluster sampling technique was ...

  6. 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)

  7. 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)

  8. 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...

  9. 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)

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Directory of Open Access Journals (Sweden)

    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.

  12. 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

  13. 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.

  14. 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.

  15. 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

  16. 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

  17. 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...

  18. 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

  19. 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.

  20. 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)

  1. 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.

  2. Antenatal care among currently married women in Rajasthan, India

    Directory of Open Access Journals (Sweden)

    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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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...

  8. 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.

  9. 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

  10. 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)

  11. 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.

  12. Proterozoic rifting and major unconformities in Rajasthan, and their implications for uranium mineralisation

    International Nuclear Information System (INIS)

    Sinha-Roy, S.

    2004-01-01

    Evolution of the Precambrian terrain in Rajasthan has taken place via crustal consolidation of the basement at ca. 2.9 Ga, its cratonisation at ca. 2.5 Ga, through protracted tectonostratigraphic evolution of the Proterozoic cover sequences, following repeated rifting and Wilson cycles in the Aravalli and Delhi foldbelts. Consequently, the Proterozoic rift basins are characterised by growth faults and pull-aparts, and multitier volcanose dimentary sequences that contain a number of unconformities and stratigraphic breaks. The Archaean basement of the Mewar terrain that witnessed end-Archaean K-magmatism and ductile shearing, led to the creation of a possible uranium province, namely uranium enriched basement. This province acted as the source of remobilised uranium and its concentration at suitable multilevel structural and stratigraphic traps within the Proterozoic rift basins to give rise to unconformity-related syngenetic uranium mineralisation. Late Neoproterozoic to Pan-African tectonothermal reworking of the basement rocks produced fracture zones and caused Na-metasomatism giving rise to albitite-related uranium mineralisation. Based on an analysis of Proterozoic rift kinematics and lithofacies characteristics, five possible uranium-enriched stratigraphic horizons have been identified in the Aravalli and its equivalent sequences as well as in the North Delhi foldbelt sequences. From a regional synthesis, ten possible uranium metallogenic events, spanning ca. 2.5-0.5 Ga, are recognised in Rajasthan. These uranium events have predictive value for delineation of target areas for exploration. (author)

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Directory of Open Access Journals (Sweden)

    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.

  15. 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.

  16. 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

  17. 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

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

    Directory of Open Access Journals (Sweden)

    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.

  19. 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.

  20. 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

  1. 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...

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

    Directory of Open Access Journals (Sweden)

    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.

  3. 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.

  4. 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.

  5. Physics operating experience and fuel management of RAPS-1

    International Nuclear Information System (INIS)

    Nakra, A.N.; Purandare, H.D.; Srinivasan, K.R.; Rastogi, B.P.

    1976-01-01

    Rajasthan Atomic Power Station Unit-1 achieved criticality on August 11, 1972. Thereafter the reactor was brought to power, in November, 1972. Due to non-availability of the depleted fuel, the loading of which was necessary to obtain full power to begin with, the core was loaded with all natural uranium fuel and only 70% of the full power could be achieved. During the reactor operation for the last three years, the reactor has seen more than one effective full power year and about 1400 fresh fuel bundles have been loaded in the core. The reactor was subjected to about 150 power cycles resulting in more than 30% variation in operating power level and about 10 fuel bundles have failed. For satisfactory fuel management and refuelling decisions, a three dimensional simulator TRIVENI was developed. This was extensively tested during the start-up experiments and was found to be a satisfactory tool for day to day operation of the plant. In this paper, a brief account of analysis of the start-up experiments, approach to full power, power distortions and flux peaking, fuel management service and analysis of the failed fuel data has been given. (author)

  6. 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_{0.098/0.098[]1.90/1

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

    Directory of Open Access Journals (Sweden)

    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.

  8. Training courses at VR-1 reactor

    International Nuclear Information System (INIS)

    Sklenka, L.; Kropik, M.

    2006-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)

  9. Current status of the Thai Research Reactor (TRR-1/M1)

    International Nuclear Information System (INIS)

    Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

    2006-01-01

    The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U 3 O 8 A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (author)

  10. 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...

  11. Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J

    1972-07-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  12. 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.

  13. 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...

  14. 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

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Directory of Open Access Journals (Sweden)

    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.

  17. Annual report on JEN-1 reactor

    International Nuclear Information System (INIS)

    Montes, J.

    1972-01-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

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

    Directory of Open Access Journals (Sweden)

    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

  19. 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)

  20. 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...

  1. 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.

  2. Ageing management studies of RAPS-1

    International Nuclear Information System (INIS)

    Bohra, A.K.; Jain, L.K.; Joshi, K.M.

    2006-01-01

    Unit-l of Rajasthan Atomic Power Station (RAPS-1) is the first nuclear power plant of India with pressurized heavy water reactor. The construction of Unit-l of Rajasthan Atomic Power Station (RAPS-1) was started in the year 1966 in collaboration with Canada. The Unit-1 achieved first criticality on August 1972 and was first synchronized to Grid on November 1972. During initial operation of the Unit, several problems were faced in its various systems and these were addressed by incorporating various engineering changes and procedures. In this unit various major innovative repairs were done like end shield leak repair, OPRD leak repair. Considering the operation of various systems of Unit-1, since year 1971 it was imperative to study ageing degradation mechanisms and mitigating measures were to be taken. Although the ageing management is a continuous process the opportunity of Unit-1 shutdown for upgradations from 30-04-2002 to 08-02-2004 was utilized for inspection and assessment of health of various SSC, which otherwise could not have been done with unit in operational state. This paper contains the following in detail. (1) Ageing management programme, its objectives and scope (2) Methodology of ageing management studies - Replacement and upgradation -Additional inspection programme based on ageing management review - Statistical analysis of ageing degradation occurrence - Estimation of residual life span of cables and relays (3) Criteria for selection of components for ageing management programme (4) Findings of ageing management studies-case studies. The ageing study done for RAPS-1 indicated that appropriate ageing monitoring methods and procedures exist in the station for taking timely mitigating measures. The technological obsoleteness has been overcome by installing new components of latest technology. On overall assessment, the Unit-1 was considered fit for further service. (author)

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Science.gov (United States)

    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.

  5. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    2003-01-01

    Full text: 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 and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kW th . The moderator of neutrons is light demineralized water (H 2 O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included

  6. 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.

  7. 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.

  8. Moving ring reactor 'Karin-1'

    International Nuclear Information System (INIS)

    1983-12-01

    The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

  9. The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

    Energy Technology Data Exchange (ETDEWEB)

    Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

    2008-07-15

    This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

  10. 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

  11. Occupational analysis for the Angra-1 reactor

    International Nuclear Information System (INIS)

    Moraes, A.

    1991-01-01

    Due to several modifications which were imposed to its time schedule during construction, the Angra-1 reactor did not enter to the grid in 1982 as it was initially foreseen. These modifications occurred due to an unforeseen scenario that was verified in steam generators (serie D-3, Westinghouse) of power stations with similar configurations which had been installed in other countries such as Ringhals-3 (Sweden), Almaraz-1 (Spain) and McGuine-1 (USA). So, among the main events that occurred in the Angra-1 reactor, which were of interest from the point of view of radiation protection, it could be pointed out the personnel monitoring, and the occupational exposure measurements at different reactor power, during the reactor fueling and during modification and tests performed at the steam generators and at ducts of the primary coolant circuit. (author)

  12. Reactor operations at SAFARI-1

    International Nuclear Information System (INIS)

    Vlok, J.W.H.

    2003-01-01

    A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

  13. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Karel, Matejka; Lubomir, Sklenka

    2005-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilisation - i.e. extensive educational programme. The educational programme is intended for the training of university students (all technical universities in Czech Republic) and selected nuclear power plant personnel. At the present, students can go through more than 20 different experimental exercises. An attractive programme including demonstration of reactor operation is prepared also for high school students. Moreover, research and development works and information programmes proceed at the VR-1 reactor as well

  14. 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)

  15. 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.

  16. Stability analysis of the Ghana Research Reactor-1 (GHARR-1)

    International Nuclear Information System (INIS)

    Della, R.; Alhassan, E.; Adoo, N.A.; Bansah, C.Y.; Nyarko, B.J.B.; Akaho, E.H.K.

    2013-01-01

    Highlights: • We developed a theoretical model to study the stability of the Ghana Research Reactor-1. • The neutronics transfer function was described by the point kinetics model for a single group of delayed neutrons. • The thermal hydraulics transfer function was based on the modified lumped parameter concept. • A computer code, RESA (REactor Stability Analysis) was developed. • Results show that the closed-loop transfer function was stable and well damped for variable operating power levels. - Abstract: A theoretical model has been developed to study the stability of the Ghana Research Reactor one (GHARR-1). The closed-loop transfer function of GHARR-1 was established based on the model, which involved the neutronics and the thermal hydraulics transfer functions. The reactor kinetics was described by the point kinetics model for a single group of delayed neutrons, whilst the thermal hydraulics transfer function was based on the modified lumped parameter concept. The inherent internal feedback effect due to the fuel and the coolant was represented by the fuel temperature coefficient and the moderator temperature coefficient respectively. A computer code, RESA (REactor Stability Analysis), entirely in Java was developed based on the model for systems analysis. Stability analysis of the open-loop transfer function of GHARR-1 based on the Nyquist criterion and Bode diagrams using RESA, has shown that the closed-loop transfer function was marginally stable for variable operating power levels. The relative stability margins of GHARR-1 were also identified

  17. 1 Observations on the ichnospecies Monomorphichnus multilineatus ...

    Indian Academy of Sciences (India)

    24

    The Ediacaran-Cambrian Marwar Supergroup exposed in the western Rajasthan ... Small Shelly Fauna (SSF) represented by ichnofossils and body fossils ... global distribution of genus Monomorphichnus, mode of preservation, size ...... Ediacaran Jodhpur Sandstone, Marwar Supergroup, western Rajasthan; J. Asian Earth.

  18. Estimation of radioactivity in structural materials of ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

  19. Operation characteristics and conditions of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

    1994-01-01

    The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

  20. Burnup measurements at the RECH-1 research reactor

    International Nuclear Information System (INIS)

    Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

    2002-01-01

    The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

  1. 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.

  2. Training and research on the nuclear reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    1998-01-01

    The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

  3. Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Otra Otra, F; Leira Rey, G

    1971-07-01

    In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

  4. IEA-R1 reactor - Spent fuel management

    International Nuclear Information System (INIS)

    Mattos, J.R.L. De

    1996-01-01

    Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

  5. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  6. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

  7. Modernization and Refurbishment of the RECH-1 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Daie, J. [Nuclear Application Department, Chilean Nuclear Energy Commission (CCHEN), Santiago (Chile)

    2014-08-15

    The Chilean Nuclear Energy Commission (Comisión Chilena de Energía Nuclear, or CCHEN) has operated the RECH-1 research reactor since 1974. This reactor is located at La Reina Nuclear Centre in Santiago, Chile. It is a pool type reactor using LEU MTR fuel assemblies, light water as moderator and coolant, and beryllium as reflector. The reactor has been operated at the nominal power of 5 MW in a continuous shift of 20 hours per week, 48 weeks per year. The main utilizations of the RECH-1 reactor are radioisotope production and neutron activation analysis. Among the most relevant refurbishment and modernization campaigns undertaken at the reactor are: full core conversion to the use of LEU fuel, replacement of the cooling tower, improvement of the containment building by changing the doors and gates and by a better sealant for the penetrations, introduction of an additional source of water by connecting the raw water supply system to the emergency cooling system, improvement of the emergency ventilation system, introduction of a fire detection and alarm system for detection and mitigation to protect the I&C racks, introduction of a radioactive liquid release for those generated at the reactor, introduction of a delay tank degasification system and renewal of the environmental monitoring system. At present we are assessing the possibility of replacing the old analog electronics of control for new digital systems. Detailed descriptions of these diverse activities are presented in the paper. (author)

  8. Extensive utilisation of VR-1 reactor for nuclear education and training

    International Nuclear Information System (INIS)

    Rataj, J.

    2010-01-01

    The paper presents utilisation of the VR-1 reactor for nuclear education and training at national and international level. VR-1 reactor has been operating by the Czech Technical University since December 1990. The reactor is a pool-type light water reactor based on enriched uranium (19.7% 235 U) with maximum thermal power 1kW and for short time period up to 5kW. The moderator of neutrons is light water, which is also used as a reflector, a biological shielding and a coolant. Heat is removed from the core by natural convection. The pool disposition of the reactor facilitates access to the core, setting and removing of various experimental samples and detectors, easy and safe handling of fuel assemblies. The reactor core can contain from 17 to 21 fuel assemblies IRT-4M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The reactor is equipped with several experimental devices; e.g. horizontal, radial and tangential channels used to take out a neutron beam, reactivity oscillator for dynamics study and bubble boiling simulator. The reactor has been used very efficiently especially for education and training of university students and NPP's specialists for more than 18 years. The VR-1 reactor is utilised within various national and international activities such as Czech Nuclear Education Network (CENEN), European Nuclear Education Network and also Eastern European Research Reactor Initiative (EERRI). The reactor is well equipped for education and training not only by the experimental facility itself but also by incessant development of training methods and improvement of education experiments. The education experiments can be combined into training courses attended by students according to their study specialization and knowledge level. The training programme is aimed to the reactor and neutron physics, dosimetry, nuclear safety, and control of nuclear installations. Every year, approximately 250 university students undergo

  9. 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)

  10. Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

    Energy Technology Data Exchange (ETDEWEB)

    Montes Ponce de Leon, J.

    1974-07-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  11. Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

    Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

  12. 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.

  13. Use of the VR-1 ''Vrabec'' training reactor

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Krops, S.; Polach, S.; Sklenka, L.

    1994-01-01

    An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

  14. Electrical system regulations of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2013-01-01

    The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

  15. Modernization of control instrumentation and security of reactor IAN - R1

    International Nuclear Information System (INIS)

    Gonzalez, J. M.

    1993-01-01

    The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

  16. Operation and maintenance of 1MW PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    Adnan Bokhari; Mohammad Suhaimi Kassim

    2006-01-01

    The Malaysian Research Reactor, Reactor TRIGA PUSPATI (RTP) has been successfully operated for 22 years for various experiments. Since its commissioning in June 1982 until December 2004, the 1MW pool-type reactor has accumulated more than 21143 hours of operation, corresponding to cumulative thermal energy release of about 14083 MW-hours. The reactor is currently in operation and normally operates on demand, which is normally up to 6 hours a day. Presently the reactor core is made up of standard TRIAGA fuel element consists of 8.5 wt%, 12 wt% and 20 wt% types; 20%-enriched and stainless steel clad. Several measures such as routine preventive maintenance and improving the reactor support systems have been taken toward achieving this long successful operation. Besides normal routine utilization like other TRIGA reactors, new strategies are implemented for effective increase in utilization. (author)

  17. 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.

  18. Perspectives on reactor safety. Revision 1

    International Nuclear Information System (INIS)

    Haskin, F.E.; Hodge, S.A.

    1997-11-01

    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) the development of safety concepts; (2) severe accident perspectives; (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

  19. Perspectives on reactor safety. Revision 1

    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); Hodge, S.A. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

    1997-11-01

    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) the development of safety concepts; (2) severe accident perspectives; (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.

  20. Annual report on JEN-1 and JEN-2 Reactors

    International Nuclear Information System (INIS)

    Montes Ponce de Leon, J.

    1974-01-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  1. Advances in Reactor Physics, Mathematics and Computation. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

  2. 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.)

  3. Stationary low power reactor No. 1 (SL-1) accident site decontamination ampersand dismantlement project

    International Nuclear Information System (INIS)

    Perry, E.F.

    1995-01-01

    The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft 2 two story, cinder block administrative building; two 4,000 ft 2 single story, steel frame office buildings; a 850 ft 2 steel framed, metal sided PL condenser building, and a 550 ft 2 steel framed decontamination and laydown building

  4. 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)

  5. 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

  6. Reactor utilization, Part 1

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1981-01-01

    The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

  7. 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.

  8. Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

    International Nuclear Information System (INIS)

    Salahuddin, A.; Israr, M.; Hussain, M.

    2002-01-01

    Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

  9. RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1985-01-01

    According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

  10. New instrumentation for the IPR-R1 reactor of CDTN

    International Nuclear Information System (INIS)

    Carvalho, P.V.R. de.

    1992-01-01

    The Nuclear Engineering Institute reactor instrumentation area has developed systems and equipment for reactor operation and safety. In such way, the new I and C for IEN Argonauta reactor and the nuclear instrumentation for IPEN critical facility were built. This paper describes our real work, the new I and C systems for IPR-R1, a Triga type reactor, located at CDTN (Belo Horizonte - MG). (author)

  11. Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

    Energy Technology Data Exchange (ETDEWEB)

    Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

    2013-07-01

    Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

  12. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

    International Nuclear Information System (INIS)

    Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

    2002-01-01

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  13. New human machine interface for VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Sklenka, L.; Chab, V.

    2002-01-01

    The contribution describes a new human machine interface that was installed at the VR-1 training reactor. The human machine interface update was completed in the summer 2001. The human machine interface enables to operate the training reactor. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between a reactor operator and the control and safety system of the nuclear reactor. Messages appear from the control system, the operator can write commands and send them there. The second display is a graphical one. It is possible to represent there the status of the reactor, principle parameters (as power, period), control rods' positions, the course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human machine interface was produced in the InTouch developing environment of the WonderWare Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors at the reactor. The former operator's desk was completely removed and superseded with a new one. Besides of the computer and the two displays, there are control buttons, indicators and individual numerical displays of instrumentation there. Utilised components guarantee high quality of the new equipment. Microcomputer based communication units with proper software were developed to connect the contemporary control and safety system with the personal computer of the human machine interface and the individual displays. New human machine interface at the VR-1 training reactor improves the safety and comfort of the reactor utilisation, facilitates experiments and training, and provides better support of foreign visitors.(author)

  14. Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

    International Nuclear Information System (INIS)

    1988-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)

  15. Department of Reactor Technology annual progress report 1 January -31 December 1977

    International Nuclear Information System (INIS)

    1978-04-01

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, reactor operation, structural reliability, system reliability, reactor physics, fuel management, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, experimental activation measurements and neutron radiography at the DR 1 reactor, underground storage of gas, solar heating and underground heat storage, wind power. (author)

  16. 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)

  17. Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

    International Nuclear Information System (INIS)

    1989-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  18. 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.

  19. Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

    International Nuclear Information System (INIS)

    Ratanatongchai, Wichian

    2009-07-01

    Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

  20. 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.

  1. 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.

  2. New measuring and protection system at VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Jurickova, M.

    2006-01-01

    The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)

  3. Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

    International Nuclear Information System (INIS)

    1991-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  4. Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

    International Nuclear Information System (INIS)

    1992-08-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1991 (April 1, 1991-March 31, 1992). The major Department's programs promoted in the year are assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researchers on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative work to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  5. 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.

  6. Thermohydraulic analysis for power increase of IEAR-1 reactor

    International Nuclear Information System (INIS)

    Umbehaun, Pedro E.; Bastos, Jose L.F.

    1996-01-01

    In this work has been presented the reactor core thermohydraulic model of IEAR-1, aiming its power operation increase from 2MW to 5MW. The design criteria adopted have been established in Safety Series 35. Three configurations of reactor core were analysed: fuel elements 20, 25 and 30

  7. Decommissioning and decontrolling the R1-reactor

    International Nuclear Information System (INIS)

    Bergman, C.; Holmberg, B.T.

    1985-01-01

    Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

  8. Upgrading of the research reactors FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

    In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

  9. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Milosevic, M.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.

    1981-12-01

    The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

  10. Department of Reactor Technology: annual progress report 1 January - 31 December 1976

    International Nuclear Information System (INIS)

    1977-06-01

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, structural reliability, system reliability, radiation fiels in nuclear power plants, reactor physics, fuel management, fission product decay analysis, steady-state thermo-hydraulics, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, control rod ejection accident analysis, economic studies for power plants, experimental activation measurements and neutron radiography at the DR 1 reactor. (author)

  11. 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.

  12. Irradiation routine in the IPR-R1 Triga reactor

    International Nuclear Information System (INIS)

    Maretti Junior, F.

    1980-01-01

    Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

  13. Neutron density optimal control of A-1 reactor analoque model

    International Nuclear Information System (INIS)

    Grof, V.

    1975-01-01

    Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)

  14. 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.

  15. Department of Reactor Technology annual progress report 1 January - 31 December 1978

    International Nuclear Information System (INIS)

    1979-04-01

    The activities of the department of reactor technology at Risoe during 1978 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  16. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    International Nuclear Information System (INIS)

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  17. Reactor engineering department annual report. April 1, 1993-March 31, 1994

    International Nuclear Information System (INIS)

    1994-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1993 (April 1, 1993-March 31, 1994). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees organized by the Department are also summarized in this report. (author)

  18. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  19. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    International Nuclear Information System (INIS)

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  20. Modification of the IAN-R1 reactor

    International Nuclear Information System (INIS)

    Jaime, J.; Ahumada, S.; Spin, R.A.

    1990-01-01

    The IAN-R1 reactor is the only nuclear reactor operating in Colombia; it is installed at the Institute of Nuclear Affairs (AIN) in Bogota, which is an official body coming under the Ministry of Mining and Energy. This reactor started operation in January 1965 with a rated power of 10 kW and was modified a year later to operate at 20 kW, which has been its rated power up to the present. Given its importance for the application of nuclear technology in Columbia for various purposes, principally in the areas of neutron activation analysis, determination of uranium content in minerals using the delayed neutron counting method, production of certain radioisotopes such as 198 Au and 82 Br for engineering applications, and production of radioactive material for teaching and research purposes, research has been in progress for some years into ways of increasing its power. The study on experimental requirements and on the demand for locally produced radioisotopes came to the conclusion that its power should be increased to 1000 kW, which would allow the facility to remain on the same site. The modification includes conversion of the core to low-enriched fuel, operation up to 1 MW, modification of the shielding, renovation of instrumentation and installation of a radioisotope processing plant. When the reactor is modified we will be able to produce other radioisotopes for applications in nuclear medicine, industry and engineering; at the same time, the safety of the facility will be optimized and the experimental facilities improved

  1. Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

    International Nuclear Information System (INIS)

    Moreira, J.M.L.; Kassar, E.

    1986-01-01

    The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

  2. 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)

  3. 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.

  4. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  5. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    International Nuclear Information System (INIS)

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  6. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

  7. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department`s programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  8. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    International Nuclear Information System (INIS)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department's programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  9. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  10. 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.

  11. Present and future activities of TRIGA RC-1 Reactor

    International Nuclear Information System (INIS)

    Festinesi, A.

    1986-01-01

    A summary of reactor activities is presented and discussed. The RC-1 reactor is used by ENEA's laboratories, research institutes and national industries for different aims: research, analysis materials behaviour under neutron flux, etc. To satisfy the requests increase it is important to signalize: - the realization of a new radiochemical laboratory for radioisotopes production, to be used in a medical and/or diagnostic field in general; - the realization of a tritium handling laboratory, to study tritium solubility, release and diffusion in different material (particularly in ceramic breeder as lithium aluminate) to support Italian programs on fusion technology; - a research activity on the reactors computerized control by a console of advanced conception. The aim of this activity is the development of an ergonomic control room that could be a reference point for the planning of the power reactor control rooms

  12. Neutronic studies in the enrichment reduction of research reactor IEAR-1

    International Nuclear Information System (INIS)

    Maiorino, J.R.; Fanaro, L.C.B.; Mai, L.A.; Ferreira, P.S.B.; Garone, J.G.M.

    1987-01-01

    In the present work the codes used by the Reactor Physics Division of IPEN-CNEN-SP in calculations for plate-type reactors are described analyzing research reactor IEAR-1. The IAEA model problem for a plate-type reactor 10 MW with high, medium and low enrichment is solved through different methodologies now in use at the RTF/IPEN-CNEN-SP (HAMMER and HAMMER-TECH-CITATION and LEO4-2DBP-UM) looking into the calculation capability for high to low enrichment conversion within the contract held with the IAEA (BRA-4661). Finally, present reactor configuration calculations are compared with experimental measurements with the aim to validate the calculation method. (Author)

  13. 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

  14. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

    1982-12-01

    Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

  15. Reliabitity study of the accumulator system for Angra-1 reactor

    International Nuclear Information System (INIS)

    Santos Maciel, C.C.R.

    1980-01-01

    The realibility of the Accumulator System of Angra 1 reactor is studied. The fault tree techniques is use for identification and evaluation of the probability of occurrence of the possible failure modes of the system. The study has as a guide the report WASH 1400 in which the analysis of the reliability of a Tipical PWR reactor of USA. Comparisons between results obtained for Accumulator System of Angra 1 and that published in the report WASH 1400 for the Accumulator System of the Typical Reactor are done. Critiques to the methodology used in the reportd WASH 1400 and an analysis of the sensitivity of the system in relation with its components are also done. (author) [pt

  16. Design of a new research reactor : 1st year conceptual design

    International Nuclear Information System (INIS)

    Park, Cheol; Lee, B. C.; Chae, H. T.

    2004-01-01

    A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

  17. Education and research at the VR-1 Vrabec training reactor facility

    International Nuclear Information System (INIS)

    Matejka, K.

    1993-01-01

    The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

  18. Graphite stack corrosion of BUGEY-1 reactor (synthesis)

    International Nuclear Information System (INIS)

    Petit, A.; Brie, M.

    1996-01-01

    The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH 4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs

  19. Sandia reactor kinetics codes: SAK and PK1D

    International Nuclear Information System (INIS)

    Pickard, P.S.; Odom, J.P.

    1978-01-01

    The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time

  20. 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

  1. Ageing problems and renovation programme of ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Khattab, M.S.; Sultan, M.A.

    1995-01-01

    Based on Practical Experience gained from interfacing ageing systems in addition to operating new systems, current problems could be deduced whenever in-service inspection are carried out. This paper summarizes the in-service inspection made, and the proposed programme of rehabilitation of mechanical system in the ET-RR-1 research reactor at Inshass. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of such rehabilitation programme. The paper summarizes also the modernization of control, measuring and radiation monitoring system already carried out at the reactor. (orig.)

  2. 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.

  3. Generating the flux map of Nigeria Research Reactor-1 for efficient ...

    African Journals Online (AJOL)

    One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

  4. Reactor protection system. Revision 1

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

    1975-04-01

    The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

  5. RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1987-01-01

    RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction [sr

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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.)

  11. 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.

  12. Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Kuruc, J.; Matel, L.

    2007-01-01

    The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

  13. Demolition of the FRJ-1 research reactor (MERLIN)

    International Nuclear Information System (INIS)

    Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2003-01-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [de

  14. Evaporation rate measurement in the pool of IEAR-1 reactor

    International Nuclear Information System (INIS)

    Torres, Walmir Maximo; Cegalla, Miriam A.; Baptista Filho, Benedito Dias

    2000-01-01

    The surface water evaporation in pool type reactors affects the ventilation system operation and the ambient conditions and dose rates in the operation room. This paper shows the results of evaporation rate experiment in the pool of IEA-R1 research reactor. The experiment is based on the demineralized water mass variation inside cylindrical metallic recipients during a time interval. Other parameters were measured, such as: barometric pressure, relative humidity, environmental temperature, water temperature inside the recipients and water temperature in the reactor pool. The pool level variation due to water contraction/expansion was calculated. (author)

  15. New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

    Energy Technology Data Exchange (ETDEWEB)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  16. 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

  17. IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

    International Nuclear Information System (INIS)

    Frajndlich, Roberto

    2009-01-01

    The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

  18. 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.

  19. 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

  20. RA reactor operation and maintenance in 1992, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Tanaskovic, M.

    1992-01-01

    During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems [sr

  1. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Bokhari, I.H.; Israr, M.; Pervez, S.

    1999-01-01

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  2. Spent fuel management - two alternatives at the FiR 1 reactor

    International Nuclear Information System (INIS)

    Salmenhaara, S.E.J.

    2001-01-01

    The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The reactor with its subsystems has experienced a large renovation work in 1996-97. The main purpose of the upgrading was to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. The BNCT work dominates the current utilization of the reactor: four days per week for BNCT purposes and only one day per week for neutron activation analysis and isotope production. The Council of State (government) granted for the reactor a new operating license for twelve years starting from the beginning of the year 2000. There is however a special condition in the new license. We have to achieve a binding agreement between our Research Centre and the domestic Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel, if we want to continue the reactor operation beyond the year 2006. In addition to the choosing of one of the spent fuel management alternatives the future of the reactor will also depend strongly on the development of the BNCT irradiations. If the number of patients per year increases fast enough and the irradiations of the patients will be economically justified, the operation of the reactor will continue independently of the closing of the USDOE alternative in 2006. Otherwise, if the number of patients will be low, the funding of the reactor will be probably stopped and the reactor will be shut down. (author)

  3. Measurement of the thermal flux distribution in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

    1986-01-01

    The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

  4. Computer-aided testing and operational aids for PARR-1 nuclear reactor

    International Nuclear Information System (INIS)

    Ansari, S.A.

    1990-01-01

    The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination

  5. Properties of autoregressive model in reactor noise analysis, 1

    International Nuclear Information System (INIS)

    Yamada, Sumasu; Kishida, Kuniharu; Bekki, Keisuke.

    1987-01-01

    Under appropriate conditions, stochastic processes are described by the ARMA model, however, the AR model is popularly used in reactor noise analysis. Hence, the properties of AR model as an approximate representation of the ARMA model should be made clear. Here, convergence of AR-parameters and PSD of AR model were studied through numerical analysis on specific examples such as the neutron noise in subcritical reactors, and it was found that : (1) The convergence of AR-parameters and AR model PSD is governed by the ''zero nearest to the unit circle in the complex plane'' (μ -1 ,|μ| M . (3) The AR model of the neutron noise of subcritical reactors needs a large model order because of an ARMA-zero very close to unity corresponding to the decay constant of the 6-th group of delayed neutron precursors. (4) In applying AR model for system identification, much attention has to be paid to a priori unknown error as an approximate representation of the ARMA model in addition to the statistical errors. (author)

  6. SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1995-03-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

  7. Spent fuel management plans for the FiR 1 Reactor

    International Nuclear Information System (INIS)

    Salmenhaara, S. E. J.

    2002-01-01

    The FiR 1-reactor, a 250 kW TRIGA reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. The final disposal site is situated in Olkiluoto, on the western coast of Finland. Olkiluoto is also one of the two nuclear power plant sites in Finland. In the new operating license of our reactor there is a special condition. We have to achieve a binding agreement between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or US DOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006. the domestic final disposal is the only possibility. At the moment it seems to be reasonable to prepare to both possibilities: the domestic final disposal and the return to the USA offered by US DOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will decide, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNCT treatments will cover the costs. If the BNCT and other irradiations develop satisfactorily, the reactor can be kept in operation beyond the year 2006 and the domestic final disposal will be implemented. If, however, there is still lack of money, there is no reason to continue the operation of the reactor and the choice of US DOE alternative is natural. (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  9. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  10. RA reactor operation and maintenance in 1996, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

    1996-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. Since the RA reactor is shutdown since 1984, it is high time for decision making of its future status. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown. Control and maintenance of the reactor instrumentation and devices was done regularly but dependent on the availability of the spare parts and financial means. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  11. 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.

  12. Possible future roles for fast breeder reactors Part 1 and 2

    International Nuclear Information System (INIS)

    1978-06-01

    Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks

  13. 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)

  14. Flux distribution measurements in the Bruce A unit 1 reactor

    International Nuclear Information System (INIS)

    Okazaki, A.; Kettner, D.A.; Mohindra, V.K.

    1977-07-01

    Flux distribution measurements were made by copper wire activation during low power commissioning of the unit 1 reactor of the Bruce A generating station. The distribution was measured along one diameter near the axial and horizontal midplanes of the reactor core. The activity distribution along the copper wire was measured by wire scanners with NaI detectors. The experiments were made for five configurations of reactivity control mechanisms. (author)

  15. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor; Analise termo-hidraulica do reator TRIGA IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Veloso, Marcelo Antonio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Fortini, Maria Auxiliadora [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2002-07-01

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  16. Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

    Science.gov (United States)

    El-Genk, Mohamed S.

    1991-01-01

    The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept.

  17. Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

    International Nuclear Information System (INIS)

    El-genk, M.S.

    1991-01-01

    The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept

  18. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lightwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. The research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology, are presented. (Author) [pt

  19. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W. de.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lighwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. It is also presented the research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology. (Author) [pt

  20. CAC-RA1 1958-1998. The first years of the Constituyentes Atomic Center (CAC). History of the first Argentine nuclear reactor (RA-1); CAC-RA-1 1958-1998. Los primeros anios del CAC. Historia del primer reactor nuclear argentino (RA-1)

    Energy Technology Data Exchange (ETDEWEB)

    Forlerer, Elena; Palacios, Tulio A [comps.

    1998-07-01

    After giving the milestones of the development of the Constituyentes Atomic Center since 1957, the history of the construction of the first nuclear reactor (RA-1) in Argentina, including the local fabrication of its fuel elements, is surveyed. The RA-1 reached criticality on January 17, 1958. The booklet commemorates the 40th year of the reactor operation.

  1. Benchmark testing of Canadol-2.1 for heavy water reactor

    International Nuclear Information System (INIS)

    Liu Ping

    1999-01-01

    The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

  2. Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

    Energy Technology Data Exchange (ETDEWEB)

    Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

    2017-05-15

    Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

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

    International Nuclear Information System (INIS)

    Galo Rocha, F. del.

    1989-01-01

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

  4. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    Energy Technology Data Exchange (ETDEWEB)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn [Physics Division, Office of Atomic Energy for Peace, Vibhavadi Rangsit Road, Chatuchak, Bangkok (Thailand)

    1999-08-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10{sup 7} n.cm{sup 2}.sec{sup -1} at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  5. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    International Nuclear Information System (INIS)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn

    1999-01-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10 7 n.cm 2 .sec -1 at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  6. Reactor inventory monitoring system for Angra-1 reactor

    International Nuclear Information System (INIS)

    S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M.; Soares, Milton; Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A.

    1996-01-01

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  7. Modifications done in the IPR-R1 reactor and their auxiliary systems

    International Nuclear Information System (INIS)

    Maretti Junior, F.; Amorim, V.A. de; Coura, J.G.

    1986-01-01

    The improvements done in the IPR-R1 reactor for adequateness of operation conditions and increase of irradiation sample capability. The cooling systems, reactor pool, system of control rods were substituted. The optimization of transfer pneumatic system was done. (M.C.K.) [pt

  8. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  9. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  10. Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Blom, K.H.; Falck, K.; Krull, W.

    1990-01-01

    The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

  11. Instrumentation renewal at the FIR 1 research reactor in Finland

    International Nuclear Information System (INIS)

    Bars, Bruno; Kall, Leif

    1982-01-01

    The Finnish TRIGA Mark II reactor (FIR 1 100 kW, later 250 kW steady state power and pulsing capability up to 250 MW) has been in operation for 20 years. The reactor is the only research reactor in Finland and is an important research training and service facility, which obviously will be operated for 10...20 years ahead. The mechanical parts of the reactor are in good shape. Some minor modifications have previously been made in the instrumentation. However, the original instrumentation could hardly have been used for 10...20 years ahead without extensive modifications and modernization. After a careful evaluation and planning process the whole reactor instrumentation was renewed in 1981 at a cost of about 400 000 dollar. The renewal was carried out in cooperation with the Central Research Institute for Physics (KFKI) at the Hungarian Academy of Sciences, which delivered the nuclear part of the instrumentation and with the Finnish company Valmet Oy Instrument Works, which delivered the conventional instrumentation, including the automatic power control system and the control console. The instrumentation, which is located in-a new isolated control room is based on modern industrial standard modular units with standardized signal ranges, electronic testing possibilities, galvanically isolated outputs etc. The instrument renewal project was brought successfully to completion in November 1981 after only about 10 working days of shut down time. The reactor is now in routine operation and the experiences gained from the new instrumentation are excellent. (author)

  12. Dose measurements in controlled area of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Alvarenga, F.L.; Junior, F.M.

    2005-01-01

    The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers

  13. Demolition of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklearservice, Essen (Germany); Cremer, J. [SNT Siempelkamp Nukleartechnik, Heidelberg (Germany)

    2003-06-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [German] Der mit Leichtwasser gekuehlte und moderierte Schwimmbad-Forschungsreaktor FRJ-2 (MERLIN) wurde von 1958 bis 1962 fuer die damalige Kernforschungsanlage Juelich (KFA) errichtet. Von 1964 bis 1985 wurde er fuer Experimente mit zunaechst 5 MW und spaeter 10 MW thermischer Leistung bei einem maximalen thermischen Neutronenfluss von 1,1.10{sup 14} n/cm{sup 2}s genutzt. Im Jahr 1985 stellte der Reaktor seinen Betrieb ein. Die Brennelemente wurden aus der Anlage entfernt und in die USA und nach Grossbritannien verbracht. Seit 1996 erfolgen die wesentlichen Abbautaetigkeiten unter Leitung eines verantwortlichen Projektteams. Bis Ende 1998 wurde das komplette Sekundaerkuehlsystem entfernt. Dem Abbau der Kuehlkreislaeufe und Experimentiereinrichtungen folgte im Jahr 2000 der Ausbau der

  14. Equipment for neutron measurements at VR-1 Sparrow training reactor

    International Nuclear Information System (INIS)

    Kolros, Antonin; Huml, Ondrej; Kos, Josef

    2008-01-01

    Full text: The VR-1 Sparrow training reactor is the experimental nuclear facility especially employed for education and teaching of students from different technical universities in the Czech Republic and other countries. Since 2005 the uniform all-purpose devices EMK310 have been used for measurement at reactor laboratory with different type of gas filled neutron detectors. The neutron detection system are employed for reactivity measurement, control rod calibration, critical experiment, study of delayed neutrons, study of nuclear reactor dynamics and study of detection systems dead time. The small dimension isotropic detectors are especially used for measurement of thermal neutron flux distribution inside the reactor core. The EMK-310 is a high performance, portable, three-channel fast amplitude analyzer designed for counting applications. It was developed for nuclear applications and made in close co-operation with firm TEMA Ltd. The precise rack eliminates electromagnetic disturbance and contains the control unit and four modules. The modules of high voltage supply and amplifier for gas filled detectors or scintillation probes are used in basic configuration. Software is tailored specifically to the reactor measurement and allows full online control. For applications involving the study of signals that may vary with the time, example study of delayed neutrons or nuclear reactor dynamics, the EMK-310 provides a Multichannel Scaling (MCS) acquisition mode. MCS dwell time can be set from 2 ms. Now, the new generation of digital multichannel analyzers DA310 is introduced. They have similarly attributes as EMK310 but the output information of unipolar signals from detector is more complete. The pipeline A/D converter with field programmable gate array (FPGA) is the hearth of the DA310 device. The resolution is 12 bits (4096 channels); the sample frequency is 80 MHz. The application for the neutron noise analysis is supposed. The correction method for non linearity

  15. Report on safety related occurrences and reactor trips July 1, 1979 - December 31, 1979

    International Nuclear Information System (INIS)

    Olsson, S.; Andermo, L.

    1980-01-01

    This is a report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1979 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 76 safety related occurrences and 27 reactor trips have been reported to the Nuclear Power Inspectorate. It is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 4.5 trips/unit. Approximetely one half of the reactor trips happened at zero or very low power operation. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)

  16. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1986-01-01

    In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  17. 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.

  18. 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.

  19. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  20. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1988-01-01

    According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  1. IPR-R1 TRIGA research reactor decommissioning plan

    International Nuclear Information System (INIS)

    Andrade Grossi, Pablo; Oliveira de Tello, Cledola Cassia; Mesquita, Amir Zacarias

    2008-01-01

    The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

  2. The Chernobyl reactor accident. Pt. 1 and 2

    International Nuclear Information System (INIS)

    1986-06-01

    The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

  3. Current activities at the FiR 1 TRIGA reactor

    International Nuclear Information System (INIS)

    Salmenhaara, Seppo

    2002-01-01

    The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The epithermal neutrons needed for the irradiation of brain tumor patients are produced from the fast fission neutrons by a moderator block consisting of Al+AlF 3 (FLUENTAL), which showed to be the optimum material for this purpose. Twenty-one patients have been treated since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. In the near future the back end solutions of the spent fuel management will have a very important role in our activities. The Finnish Parliament ratified in May 2001 the Decision in Principle on the final disposal facility for spent fuel in Olkiluoto, on the western coast of Finland. There is a special condition in our operating license. We have now about two years' time to achieve a binding agreement between VTT and the Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel. If this will not happen, we have to make the agreement with the USDOE with the well-known time limits. At the moment it seems to be reasonable to prepare for both spent fuel management possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will determine, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNC treatments will cover

  4. 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.)

  5. Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

    International Nuclear Information System (INIS)

    Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

    2013-01-01

    The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

  6. Expanding the storage capability at ET-RR-1 research reactor at Inshass

    International Nuclear Information System (INIS)

    Sultan, Mariy M.; Khattab, M.

    1999-01-01

    Storing of spent fuel from Test Reactor in developing countries has become a big dilemma for the following reasons: The transportation of spent fuel is very expensive; There are no reprocessing plants in most developing countries; The expanding of existing storage facilities in reactor building require experience that most of developing countries lack; Some political motivations from Nuclear Developed countries intervene which makes the transportation procedures and logistics to those countries difficult. This paper gives the conceptual design of a new spent fuel storage now under construction at Inshass research reactor (ET-RR-1). The location of the new storage facility is chosen to be within the premises of the reactor facility so that both reactor and the new storage are one Material Balance Area. The paper also proposes some ideas that can enhance the transportation and storage of spent fuel of test reactors, such as: Intensifying the role of IAEA in helping countries to get rid of the spent fuel; The initiation of regional spent fuel storage facilities in some developing countries. (author)

  7. Licensing of the first reload of Angra-1 reactor

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1985-01-01

    The historical aspects related to the licensing of the first reload of Angra-1 reactor are presented. The dates, the institutions, the experts, as well as the documents generated during that process are presented. (M.I.)

  8. Evaluation of the trial design studies for an advanced marine reactor, (1)

    International Nuclear Information System (INIS)

    1988-03-01

    The trial design of three type reactors, semi-integrated, integrated and integrated (self-pressurized) type, was carried out in order to clarify the reactor type for the advanced marine reactor that would be developed for its realization in future and in order to extract its research and development theme. The trial design was carried and finished as for the three type reactors in same specifications in order to improve the following characteristics, small in size, light in weight, high in safety and reliability, and economic. In this report, a comparison and review of the following items are described as for the above three type reactors, (1) specifications, (2) shielding, (3) refueling, (4) in-service inspection, (5) analysis of the transients and accidents, (6) piping systems, (7) control systems, (8) dynamic analysis, (9) overall comparison, (10) research and development theme and theme for study in future. (author)

  9. Transformation of 1,1,1-trichloroethane in an anaerobic packed-bed reactor at various concentrations of 1,1,1-trichloroethane, acetate and sulfate

    NARCIS (Netherlands)

    deBest, JH; Jongema, H; Weijling, A; Doddema, HJ; Janssen, DB; Harder, W

    Biotransformation of 1,1,1-trichloroethane (CH3CCl3) was observed in an anaerobic packed-bed reactor under conditions of both sulfate reduction and methanogenesis. Acetate (1 mM) served as an electron donor. CH3CCl3 was completely converted up to the highest investigated concentration of 10 mu M.

  10. Evaluation of power behavior during startup and shutdown procedures of the IPR-R1 Triga Reactor

    International Nuclear Information System (INIS)

    Zangirolami, Dante M.; Mesquita, Amir Z.; Ferreira, Andrea V.

    2009-01-01

    The IPR-R1 nuclear reactor of Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN is a TRIGA Mark I pool type reactor cooled by natural circulation of light water. In the IPR-R1, the power is measured by four nuclear channels, neutron-sensitive chambers, which are mounted around the reactor core: the Startup Channel for power indication during reactor startup; the Logarithmic Wide Range Power Monitoring Channel; the Linear Multi-Range Power Monitoring Channel and the Percent Power Safety Channel. A data acquisition system automatically does the monitoring and storage of all the reactor operational parameters including the reactor power. The startup procedure is manual and the time to reach the desired reactor power level is different on each irradiation which may introduces differences in induced activity of samples irradiated in different irradiations. In this work, the power evolution during startup and shutdown periods of IPR-R1 operation was evaluated and the mean values of reactor energy production in these operational phases were obtained. The analyses were performed on basis of the Linear Multi-Range Channel data. The results show that the sum of startup and shutdown periods corresponds to 1% of released energy for irradiations during 1h at 100kW. This value may be useful to correct experimental data in neutron activation experiments. (author)

  11. Modernization of Safety and Control Instrumentation of the IEA-R1 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    De Carvalho, P.V., E-mail: paulov@ien.gov.br [Institute of Nuclear Engineering (IEN), National Nuclear Energy Commission (CNEN), Rio de Janeiro (Brazil)

    2014-08-15

    The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring

  12. Operation and maintenance of the RA Reactor in 1985, Part 1, Annex A - Reactor applications

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1985-01-01

    This document describes reactor operation from 1981 to 1985, including data about short term (shorter than 24 hours) and long term operation interruptions, as well as safety shutdown and reactor applications. During 1982, 1983 until July 1984 reactor was operated at 2 MW power according to the plan. Plan was not fulfilled in 1983 because deposits were noticed again, at the end of 1982, on the surface of fuel elements. Reactor was mainly used for neutron activation purposes and isotope production as source of neutrons for experimental purposes [sr

  13. Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

    International Nuclear Information System (INIS)

    Andermo, L.; Sundman, B.

    1974-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1977 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 48 safety related occurrences and 49 reactor trips have been reported to the Nuclear Power Inspectorate. Included is also one incident June 21 in Barsebaeck 2 which was not included in the last compilation of occurrences. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips have increased nearly 30% since the last period. Those occurred during power operation however, were less. More than 50% of the reactor trips happened in the shutdown condition. The fact that even small deviations from prescribed operation result in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences withou02068NRM 0000169 450

  14. New digital control system for the operation of the Colombian research reactor IAN-R1

    International Nuclear Information System (INIS)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J.

    2015-09-01

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  15. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias

    2005-01-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  16. Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

    International Nuclear Information System (INIS)

    Andermo, L.

    1977-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1, 1976 to December 31, 1976 inclusive. The facilities involved are Oskarshamn 1 and 2, Ringhals 1 and 2 and Barsebaeck 1. During this period of the 6 months 37 safety related occurrences and 34 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The fact that even small deviations from prescribed operation results in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The number of reactor trips are almost as low as during the last period, which is a drastic reduction compared to earlier time periods. The greatest outages are caused by occurrences without safety significance.(author)

  17. Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

    Energy Technology Data Exchange (ETDEWEB)

    El-Messeiry, A M [National Center for Nuclear Safety and Radiation Control, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs.

  18. Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

    International Nuclear Information System (INIS)

    El-Messeiry, A.M.

    1996-01-01

    In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs

  19. COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons

    International Nuclear Information System (INIS)

    Agazzi, A.; Gavazzi, C.; Vincenti, E.; Monterosso, R.

    1964-01-01

    1 - Nature of physical problem solved: The programme studies the spatial dynamics of reactor TESI, in the two group and one space dimension approximation. Only one group of delayed neutrons is considered. The programme simulates the vertical movement of the control rods according to any given movement law. The programme calculates the evolution of the fluxes and temperature and precursor concentration in space and time during the power excursion. 2 - Restrictions on the complexity of the problem: The maximum number of lattice points is 100

  20. Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Jonah, S.A. [Reactor Engineering Section, Centre for Energy Research and Training, Ahmadu Bello University, Zaria, P.M.B. 1014 (Nigeria)], E-mail: jonahsa2001@yahoo.com; Liaw, J.R.; Matos, J.E. [RERTR Program, Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2007-12-15

    The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity ({rho}{sub ex}), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

  1. Welding of the A1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Becka, J.

    1975-01-01

    As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

  2. Determination flux in the Reactor JEN-1

    International Nuclear Information System (INIS)

    Manas Diaz, L.; Montes Ponce de leon, J.

    1960-01-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs

  3. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy's (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors

  4. Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

    Science.gov (United States)

    Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

    2017-11-01

    The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

  5. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    West, C D [comp.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN{sub 2} test, Source LH2-H{sub 2}O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface.

  6. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    West, C.D.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN 2 test, Source LH2-H 2 O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

  7. Nuclear reactor and materials science research: Technical report, May 1, 1985-September 30, 1986

    International Nuclear Information System (INIS)

    1987-01-01

    Throughout the 17-month period of its grant, May 1, 1985-September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The reactor was operated 4115 hours during FY 1986 and for 6080 hours during the entire 17-month period, an average of 82 hours per week. Utilization of the reactor during that period may be classified as follows: neutron beam tube research; nuclear materials research and development; radiochemistry and trace analysis; nuclear medicine; radiation health physics; computer control of reactors; dose reduction in nuclear power reactors; reactor irradiations and services for groups outside MIT; MIT Research Reactor. Data on the above utilization for FY 1986 show that the MIT Nuclear Reactor Laboratory (NRL) engaged in joint activities with nine academic departments and interdepartmental laboratories at MIT, the Charles Stark Draper Laboratory in Cambridge, and 22 other universities and nonprofit research institutions, such as teaching hospitals

  8. Baikal-1 stand complex. Preparation and carrying out of the first energy start-up of the IVG-1 reactor

    International Nuclear Information System (INIS)

    Tikhomirov, L.N.

    1995-01-01

    The IVG-1 reactor was a first ground prototype of nuclear rocket engine. The reactor was built on the site 10 of the Semipalatinsk test site. Since the first energy start-up in 1975 the reactor was exploited 14 years till its modernization in 1989. The Bajkal-1 stand complex was designed and built for the carrying out of tests for fuel assemblies of different modifications. The energy start-up has been sum of long creative work of different research and constructive staffs on creation of high-temperature gas-cooled IVG-1 reactor. The history of construction, project and assembling of the stand complex is presented. Complex start and put works were carried out in the December 1974. Control physical start-up was carried out in the January 1975. Cold start-up by hydrogen was in the February 1975. Hot start-up was in the March 1975. The result of the hot start-up was experimental confirmation of metodics of thermohydrovlical estimations. 2 figs., 3 tabs

  9. RA reactor operation and maintenance in 1994, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

    1994-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. The spent fuel elements used from the very beginning of reactor operation are stored in the existing pools. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988 and was fulfilled in 1990. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  10. 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.

  11. Determination flux in the Reactor JEN-1; Medida de flujos de neutrones en el nucleo del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Manas Diaz, L; Montes Ponce de leon, J.

    1960-07-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 {mu} gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs.

  12. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    Odoi, H.C.

    2014-06-01

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm 2 .s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm 2 .s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U0 2 -12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO 2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at

  13. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

    2015-01-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  14. Reactor Engineering Department annual report, April 1, 1985 - March 31, 1986

    International Nuclear Information System (INIS)

    1986-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1985 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor, High Conversion Light Water Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, reactor decommissioning technology, and activities of the Committee on Reactor Physics. (author)

  15. Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

    International Nuclear Information System (INIS)

    1978-05-01

    An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-29

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

  17. 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

    Directory of Open Access Journals (Sweden)

    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.

  18. Problems of nuclear reactor safety. Vol. 1

    International Nuclear Information System (INIS)

    Shal'nov, A.V.

    1995-01-01

    Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed

  19. Experiment on continuous operation of the Brazilian IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Freitas Pintaud, M. de

    1994-01-01

    In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

  20. Auxiliary control system of the safety parameters for IPR-R1 reactor

    International Nuclear Information System (INIS)

    Coura, J.G.

    1986-01-01

    This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author) [pt

  1. 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.

  2. 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

  3. OSCAR-4 Code System Application to the SAFARI-1 Reactor

    International Nuclear Information System (INIS)

    Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

    2008-01-01

    The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

  4. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    International Nuclear Information System (INIS)

    Novelli, A.

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer. (author)

  5. Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

    International Nuclear Information System (INIS)

    Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

    2008-01-01

    The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

  6. 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.

  7. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2000-01-01

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k eff values within about 1% Δk/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models

  8. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor; Calculo de vairaciones de reactividad en algunos periodos regulares de operacion del reactor JEN-1 Mod.

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F

    1973-07-01

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

  9. Reactor Engineering Department annual report (April 1, 1986 - March 31, 1987)

    International Nuclear Information System (INIS)

    1987-08-01

    Research and development activities in the Department of Reactor Engineering in the fiscal year 1986 are described. The major activities of the Department are closely related to the reactor physics of very high temperature gas-cooled reactor, high conversion light water reactor and liquid metal fast breeder reactor and to blanket neutronics of fusion reactor. Contents of this report are divided into the activities on nuclear data and group constants, theoretical methods and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control, diagnosis and robotics. The activity of the Research Committee on Reactor Physics is also included. (author)

  10. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  11. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  12. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  13. Modelling of the RA-1 reactor using a Monte Carlo code; Modelado del reactor RA-1 utilizando un codigo Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Quinteiro, Guillermo F; Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Reactores y Centrales Nucleares

    2000-07-01

    It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)

  14. Neutronics and thermohydraulics of the reactor C.E.N.E. Pt. 1

    International Nuclear Information System (INIS)

    Caro, R.; Ahnert, C.; Esteban Naudin, A.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-01-01

    The analysis of neutronics (both statics and kinetics), of the 10 Mwt swimming pool reactor C.E.N.E. is included. A short description of the theoretical model used, along with the theoretical versus experimental cheking, carried out, whenever possible, with the reactors JEN-1 and JEN-2 of Junta de Energia Nuclear, is given in each of these chapters. (author) [es

  15. Applied research into direct numerical control of A-1 reactor temperature

    International Nuclear Information System (INIS)

    Karpeta, C.; Volf, K.

    1974-01-01

    Partial results of research efforts aimed at applying modern control theory in the control of the reactor of the A-1 nuclear power station are presented. A mathematical model of the process dynamics was developed. Some parameters of the model were determined using the results of an experimentally performed reactor scram. The optimal stochastic discrete regulator was determined and closed-loop transients were studied. The possibilities of implementing control routines were investigated using the RPP-16 computer. (author)

  16. 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.

  17. RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2002-01-01

    The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

  18. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    1990-01-01

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  19. 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.

  20. Research reactor core conversion guidebook. V.1: Summary

    International Nuclear Information System (INIS)

    1992-04-01

    In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this guidebook has been prepared to assist research reactor operators in addressing the safety and licensing issues for conversion of their reactor cores from the use of HEU fuel to the use of low enriched uranium fuel. This Guidebook, in five volumes, addresses the effects of changes in the safety-related parameters of mixed cores and the converted core. It provides an information base which should enable the appropriate approvals processes for implementation of a specific conversion proposal, whether for a light or for a heavy water moderated research reactor. Refs, figs, bibliographies and tabs

  1. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    International Nuclear Information System (INIS)

    1997-01-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits

  2. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits.

  3. NBR ISO 9001 Certification for activities carried out in IEA-R1 reactor

    International Nuclear Information System (INIS)

    Paiva, Rosemeire P.; Salvetti, Tereza C.

    2005-01-01

    Since its inauguration in 1957, the IEA-R1 research reactor has been used mainly for research, development and teaching by scientific community. In the last years, with the increase of the commercial radiopharmaceutical production by Radiopharmacy Center of IPEN, the IEA-R1 reactor was recognized as a service supplier for that center and has received a treatment more commercial from IPEN Management. In 1999 the radiopharmaceutical production obtained the NBR ISO 9002 Certification, since that the IPEN Management considered convenient to invest in the certification of its internal suppliers. In this context, in 2001 the Research Reactor Center (CRPq) began the implantation of a Quality Management System (QMS) based on NBR 9001: 2000 standard, for activities related to the operation and maintenance of the IEA-R1 research reactor and irradiation services. This QMS was structured to incorporate tools already implemented in order to complain the requirements related to nuclear and radiological safe for a nuclear installation established by the regulatory organism. The QMS is supported by a documentation system composed of approximately 150 documents including quality manual, business and action plans, operational procedures and work instruction. Carlos Alberto Vanzolini Foundation (FCAV), an INMETRO certified organism, certified the 'Operation and Maintenance of the IEA-R1 Research Reactor and Irradiation Services' in December 2002. In 2003 and 2004, the QMS was audited by FCAV that determined the maintenance of the certification. This work presents the main steps of the QMS implementation, including the difficulties found and results obtained in the process. (author)

  4. 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.

  5. Life extension of the St. Lucie unit 1 reactor vessel

    International Nuclear Information System (INIS)

    Rowan, G.A.; Sun, J.B.; Mott, S.L.

    1991-01-01

    In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program

  6. Reactor calculations in aid of isotope production at SAFARI-1

    International Nuclear Information System (INIS)

    Ball, G.

    2003-01-01

    Varying levels of reactor physics support is given to the isotope production industry. As the pressures on both the safety limits and economical production of reactor produced isotopes mount, reactor physics calculational support is playing an ever increasing role. Detailed modelling of the reactor, irradiation rigs and target material enables isotope production in reactors to be maximised with respect to yields and quality. NECSA's methodology in this field is described and some examples are given. (author)

  7. Shielding assessment of the ETRR-1 Reactor Under power upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs.

  8. Recuperation of the energy released in the G-1, an air-cooled graphite reactor core

    International Nuclear Information System (INIS)

    Chambadal, P.; Pascal, M.

    1955-01-01

    The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [fr

  9. Development of a training simulator to operators of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Carvalho, Ricardo Pinto de

    2006-01-01

    This work reports the development of a Simulator for the IEA-R1 Research Reactor. The Simulator was developed with Visual C++ in two stages: construction of the mathematics models and development and configuration of graphics interfaces in a Windows XP executable. A simplified modeling was used for main physics phenomena, using a point kinetics model for the nuclear process and the energy and mass conservation laws in the average channel of the reactor for the thermal hydraulic process. The dynamics differential equations were solved by using finite differences through the 4th order Runge- Kutta method. The reactivity control, reactor cooling, and reactor protection systems were also modeled. The process variables are stored in ASCII files. The Simulator allows navigating by screens of the systems and monitoring tendencies of the operational transients, being an interactive tool for teaching and training of IEA-R1 operators. It also can be used by students, professors, and researchers in teaching activities in reactor and thermal hydraulics theory. The Simulator allows simulations of operations of start up, power maneuver, and shut down. (author)

  10. An overview of the RECH-1 reactor conversion

    International Nuclear Information System (INIS)

    Klein, J.; Medel, J.; Daie, J.; Torres, H.

    2000-01-01

    The RECH-l research reactor achieved the first criticality on October 13, 1974 using HEU MTR type fuel elements, which were fabricated by the UKAEA at Dounreay, Scotland. In 1979, the conversion of the reactor to use LEU fuel was decided; however, a rough estimate of the uranium density needed to convert the reactor gave 3.7 g/cm 3 . This density was not available, and to maintain the overall fuel element geometry it was necessary to convert the reactor to use 45% enriched uranium fuel. In 1985, the conversion of the reactor to use medium enriched uranium was achieved. Some years later, the Chilean Nuclear Energy Commission developed the capability to produce fuel elements based on U 3 Si 2 -Al dispersion fuel. Once the plant and the manufacturing and quality control procedures were commissioned to permit the production of fuel elements, a fabrication program starts to produce LEU fuel elements with a uranium density of 3.4 g/cm 3 . A fabrication qualification period that extended to the required fuel plates for the assembly of two fuel elements started. In November 1998, the first four LEU fuel elements manufactured by the Chilean Fuel Fabrication Plant were delivered to the reactor. When the first two fuel elements were introduced into the core a LEU fuel element qualification program began. While those fuel elements remain in the core, an evaluation program is being applied to observe its performance under irradiation condition. (author)

  11. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  12. Verification of the linearity of the IPR-R1 TRIGA reactor power channels

    International Nuclear Information System (INIS)

    Souza, Rose Mary Gomes do Prado; Campolina, Daniel de Almeida Magalhaes

    2013-01-01

    The aim of this paper is to verify the linearity of the three power channels of the IPR-R1 TRIGA reactor. Located at Nuclear Technology Development Center-CDTN in Belo Horizonte, the IPR-R1 reactor is a typical 100 kW Mark I light-water reactor cooled by natural convection. When the experiments were performed, the reactor core had 59 fuel elements, containing 8% by weight of uranium enriched to 20% in 235 U. The core has cylindrical configuration with an annular graphite reflector. The responses of the detectors of the Linear, Log N and Percent Power channels were compared with the responses of detectors which only depend on the overall neutron flux within the reactor. Gold and cobalt foils were activated at low and high powers, respectively, and the specific count results were compared with measurements performed, simultaneously, with a fission chamber, and with the power registered by the three channels. The results show that the Linear channel responds linearly up to 100 kW, and the Log N channel responses are linear at low powers. In the range of high power, the Log N and the Percent Power channels exhibit linearity only from 10 kW to 50 kW. (author)

  13. Conceptual design study for the demonstration reactor of JSFR. (1) Current status of JSFR development

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Ohshima, Jun; Ito, Takaya

    2011-01-01

    JAEA is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. This paper describes development status of key technologies and comparison between 750 MWe and 500 MWe plants with the view points of demonstration ability for commercial JSFR plant. (author)

  14. An economic analysis of stretch-out for Angra-1 reactor

    International Nuclear Information System (INIS)

    Sakai, M.

    1989-01-01

    An application of NUCOST code for calculating nuclear energy cost is presented. Ann optimization of stretch-out for Angra-1 reactor based on international costs of nuclear fuel, operation and maintenance is done. (M.C.K.)

  15. The analysis for inventory of experimental reactor high temperature gas reactor type

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

    Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

  16. Welding electrode for peripheral welds of A-1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Lakatos, L.

    1975-01-01

    The properties are outlined of the VUZ-AC1-52 welding electrode used in welding the Bohunice A-1 reactor pressure vessel. The mechanical properties of welded joints after the final thermal treatment are summed up. (J.K.)

  17. 1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Kim, Jae Cheol

    2007-02-01

    During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

  18. The reactor kinetics code tank: a validation against selected SPERT-1b experiments

    International Nuclear Information System (INIS)

    Ellis, R.J.

    1990-01-01

    The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

  19. Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Cremer, J. [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany)

    2003-07-01

    By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 10{sup 11} Bq. (orig.)

  20. . Effects of extended shutdown on the control rod drive mechanism of nigeria research reactor-1(NIRR-1)

    International Nuclear Information System (INIS)

    Yusuf, I; Mati, A. A.

    2010-01-01

    The control rod drive mechanism of the Nigeria Research Reactor-1 is being driven by a servo motor, type SDE-45 through a mechanical gear system. The servo motor ensures the position control of the control rod, and hence the stability of the neutron-flux of the nuclear research reactor. The control rod drive mechanism assembly is mounted on top of the reactor vessel, about 0.6m above 30m 3 volume of reactor pool water. The top of the pool is covered with a Perspex material to protect the water in the pool from environmental contamination and to reduce evaporation. Although most of the materials in the control rod drive mechanism assembly are made of stainless steel, the servo motor however contains corrodible materials. The paper reveals a practical experience of failure of the control rod drive mechanism as a result of corrosion growth between the rotor of the servo motor and its stator windings, due to an extended shutdown of the facility.

  1. VR-1 training reactor in use for twelve years to train experts for the Czech nuclear power sector

    International Nuclear Information System (INIS)

    Matejka, K.; Sklenka, L.

    2003-01-01

    The VR-1 training reactor has been serving students of the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, for more than 12 years now. The operation history of the reactor is highlighted. The major changes made at the VR-1 reactor are outlined and the main experimentally verified core configurations are shown. Some components of the new equipment installed on the VR-1 reactor are described in detail. The fields of application are shown: the reactor serves not only the training of university students within whole Czech Republic but also the training of specialists, research activities, and information programmes in the nuclear power domain. (P.A.)

  2. FiR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

    International Nuclear Information System (INIS)

    Auterinen, I.; Salmenhaara, S.E.J. . Author

    2004-01-01

    The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT), but FiR 1 has also an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics etc. with isotope production and activation analysis services. In the 1990's a BNCT treatment facility was built at the FiR 1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material Fluental TM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality after a major renovation of the whole reactor building in 1997. Recently the lithiated polyethylene neutron shielding of the beam aperture was modified to ease the positioning of the patient close to the beam aperture. Increasing the reactor power to 500 kW would allow positioning of the patient further away from the beam aperture. Possibilities to accomplish a safety analysis for this is currently under considerations. Over thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well as in the head and neck region are recruiting patients. (author)

  3. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  4. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  5. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  6. Nuclear material control at IEA-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    1988-01-01

    The control measurements system and verification of physical inventory for fuel elements used in the operation of IEA-R1 nuclear research reactor are described. The computer code used for burn-up calculation are shown. (E.G.) [pt

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

    Directory of Open Access Journals (Sweden)

    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.

  8. Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor; Determinacion del espectro neutronico en distintas posiciones del reactor RA-1

    Energy Technology Data Exchange (ETDEWEB)

    Lerner, A M; Madariaga, M R [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

    1999-12-31

    Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm{sup 2}sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm{sup 2}.sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm{sup 2}.sec). ((1) According to the reaction Au{sup 197}(n,{gamma})Au{sup 198}, having a cross section of {sigma}{sub 0}=98.8b for thermal neutrons. (2) According to the reaction In{sup 115}(n,n`)In{sup 115m}, with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [Espanol] Texto completo: Como se sabe, el reactor RA1 se utiliza para irradiar con neutrones distintos tipos de materiales. El grupo de

  9. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    1997-01-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits

  10. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits.

  11. 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.

  12. Reliability database of IEA-R1 Brazilian research reactor: Applications to the improvement of installation safety

    International Nuclear Information System (INIS)

    Oliveira, P.S.P.; Tondin, J.B.M.; Martins, M.O.; Yovanovich, M.; Ricci Filho, W.

    2010-01-01

    In this paper the main features of the reliability database being developed at Ipen-Cnen/SP for IEA-R1 reactor are briefly described. Besides that, the process for collection and updating of data regarding operation, failure and maintenance of IEA-R1 reactor components is presented. These activities have been conducted by the reactor personnel under the supervision of specialists in Probabilistic Safety Analysis (PSA). The compilation of data and subsequent calculation are based on the procedures defined during an IAEA Coordinated Research Project which Brazil took part in the period from 2001 to 2004. In addition to component reliability data, the database stores data on accident initiating events and human errors. Furthermore, this work discusses the experience acquired through the development of the reliability database covering aspects like improvements in the reactor records as well as the application of the results to the optimization of operation and maintenance procedures and to the PSA carried out for IEA-R1 reactor. (author)

  13. 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.

  14. 1DB, a one-dimensional diffusion code for nuclear reactor analysis

    International Nuclear Information System (INIS)

    Little, W.W. Jr.

    1991-09-01

    1DB is a multipurpose, one-dimensional (plane, cylinder, sphere) diffusion theory code for use in reactor analysis. The code is designed to do the following: To compute k eff and perform criticality searches on time absorption, reactor composition, reactor dimensions, and buckling by means of either a flux or an adjoint model; to compute collapsed microscopic and macroscopic cross sections averaged over the spectrum in any specified zone; to compute resonance-shielded cross sections using data in the shielding factor formnd to compute isotopic burnup using decay chains specified by the user. All programming is in FORTRAN. Because variable dimensioning is employed, no simple restrictions on problem complexity can be stated. The number of spatial mesh points, energy groups, upscattering terms, etc. is limited only by the available memory. The source file contains about 3000 cards. 4 refs

  15. Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-08-15

    The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

  16. Dominant accident sequences in Oconee-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

    1985-04-01

    A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

  17. RA reactor operation and maintenance in 1989, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Sanovic, V.

    1989-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in July 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The following major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the power supply system. Project concerned with renewal of RA reactor complete instrumentation was started at the end of 1988. Contract was signed between the IAEA and Soviet Atomenergoexport for supplying the new instrumentation for the RA reactor. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988. In 1989, device for water purification designed by the reactor staff started operation and spent fuel handling equipment is being mounted. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  18. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators 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 objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  19. Tensile properties of quadruple melted Zr-2.5Nb pressure tubes evaluated from pressure tube offcuts

    International Nuclear Information System (INIS)

    Shah, Priti Kotak; Dubey, J.S.; Anantharaman, S.

    2013-12-01

    Rajasthan Atomic Power Station-2 (RAPS-2) is the first Pressurised Heavy Water Reactor (PHWR) in India having quadruple melted Zr-2.5Nb pressure tubes. Front-end and back-end off-cuts of sixteen pressure tubes were selected for studying the mechanical properties in axial and transverse directions of the tube. Tension tests were carried out at room temperature and at 300℃ using miniature tensile test specimens. The report presents the experimental details and discusses the base line tensile property data for the quadruple melted pressure tubes of RAPS-2. This data will be useful for the reactor life management. (author)

  20. 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.)

  1. System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

    International Nuclear Information System (INIS)

    Ellis, R.J.

    2000-01-01

    The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors

  2. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

  3. Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

    International Nuclear Information System (INIS)

    Ferreira, P.S.B.

    1988-01-01

    Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt

  4. Core calculations for the upgrading of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E.

    1998-01-01

    The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

  5. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  6. Dose measurements in controlled area and laboratory of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto; Alvarenga, Frederico Ladeia

    2005-01-01

    The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers. (author)

  7. Ageing Management Programme for the IEA-R1 Reactor in São Paulo, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ramanathan, L. V. [Institute of Energy and Nuclear Research (IPEN), National Nuclear Energy Commission (CNEN), São Paulo (Brazil)

    2014-08-15

    IEA-R1 is a swimming pool type reactor. It is moderated and cooled by light water and uses graphite and beryllium as reflector elements. First criticality was achieved on 16 September 1957, and the reactor is currently operating at 4.0 MW on a 64 h per week cycle. In 1996, a reactor ageing study was established to determine general deterioration of systems and components such as cooling towers, secondary cooling system, piping, pumps, specimen irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation, and safety system. The basic structure of the reactor from the original design has been maintained, but several improvements and modifications have been made over the years to various components, systems and structures. During the period 1996–2005 the reactor power was increased from 2 MW to 5 MW and the operational cycle from 8 h per day for 5 days a week to 120 h continuous per week, mainly to increase production of {sup 99}Mo. Prior to increasing reactor power, several modifications were made to the reactor system and its components. Simultaneously, a vigorous ageing management, inspection and modernization programme was put in place.

  8. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  9. 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

  10. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    International Nuclear Information System (INIS)

    Faghihi, F.; Ramezani, E.; Yousefpour, F.; Mirvakili, S.M.

    2008-01-01

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation

  11. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of); Nuclear Safety Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Ramezani, E. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Yousefpour, F. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of); Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of)

    2008-10-15

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation.

  12. Source term determination from subcritical multiplication measurements at Koral-1 reactor

    International Nuclear Information System (INIS)

    Blazquez, J.B.; Barrado, J.M.

    1978-01-01

    By using an AmBe neutron source two independent procedures have been settled for the zero-power experimental fast-reactor Coral-1 in order to measure the source term which appears in the point kinetical equations. In the first one, the source term is measured when the reactor is just critical with source by taking advantage of the wide range of the linear approach to critical for Coral-1. In the second one, the measurement is made in subcritical state by making use of the previous calibrated control rods. Several applications are also included such as the measurement of the detector dead time, the determinations of the reactivity of small samples and the shape of the neutron importance of the source. (author)

  13. Studies in fusion reactor technology. Final report, September 1, 1974--August 31, 1977

    International Nuclear Information System (INIS)

    Axtmann, R.C.; Perkins, H.K.

    1977-08-01

    Two independent measurements of hydrogen permeation through stainless steel at driving pressures in the range from 10 -6 to 1 Pa indicate that most extant predictions of tritium permeation through fusion reactors are probably overestimated grossly. A comprehensive analysis demonstrates that, given available structural materials, the prospects are negligible for the economic production of synthetic fuels via radiolytic reactions in fusion reactor systems

  14. 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)

  15. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  16. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    International Nuclear Information System (INIS)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

    2015-01-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  17. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

    2015-07-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  18. Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

    International Nuclear Information System (INIS)

    D'Utra Bitelli, U.

    1993-01-01

    This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)

  19. AKR-1 nuclear training reactor of Dresden Technical University turns twenty-five

    International Nuclear Information System (INIS)

    Hansen, W.

    2003-01-01

    Twenty-five years ago, in the night of July 27 to 28, 1978, the AKR-1 nuclear training reactor of the Dresden Technical University went critical for the first time and was commissioned. On the occasion of this anniversary, a colloquy was arranged with representatives from science, politics and industry, at which the reactor's history, the excellent achievements in research and training with the reactor, and the status and perspectives of this research facility were described. The AKR-1 had been built within the framework of the Nuclear Development Program of the then German Democratic Republic (GDR). The Nuclear Power Scientific Division of the Dresden Technical University had been entrusted with the responsibility, among other things, to train university personnel for the GDR Nuclear Power Program. The review by an expert group in 1996 of this plant had resulted in a recommendation in favor of long-term plant operation. A nuclear licensing procedure to this effect was initiated, and the necessary technical backfitting measures were implemented. The AKR-1 plant now equally serves for the specialized training of students and for research. (orig.) [de

  20. Measures aimed at enhancing safe operation of the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    Balogun, G.I.; Jonah, S.A.; Umar, I.M.

    2005-01-01

    Safety culture has been defined as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. This paper briefly highlights efforts being made at the Centre for Energy Research and Training (CERT) towards realizing this broad objective as far as possible. To this end CERT realizes the need for instituted safety measures to reflect significant, site-specific peculiar characteristics of any generic reactor types. Consequently, standard procedures for pre-startup, startup and shutdown of NIRR-1 (a miniature neutron source reactor - MNSR) have been reviewed to reflect our local conditions and peculiarities. The review has revealed the need to incorporate important steps that impact on overall safety of the facility. For instance an interlocking system is being considered between NIRR-1 startup on the one hand and mandatory pre-startup measures on the other. Also a procedure has been put in place that would facilitate rapid response in the event of a rod-stuck-at-full-withdrawal incident. Furthermore, a program of automation of important analysis and design calculations of MNSRs is going on. Emphases are also placed, and deliberate efforts are being made, to ensure that a working atmosphere prevails that would foster the correct attitudinal approach to matters of reactor safety. A regime of constant dialogue and discussions amongst operating personnel has been factored into the overall operational program. (author)

  1. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    1995-01-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy's Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period

  2. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy`s Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period.

  3. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: joseluis.gonzalez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  4. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    International Nuclear Information System (INIS)

    Block, R.C.; Feiner, F.

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  5. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  6. Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Adib, M.; Naguib, K.; Morcos, H.N

    2001-09-01

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% {sup 235}U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% {sup 235}U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.

  7. Calculation of the main neutron parameters of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Ojima, Mario Katsuhiko

    1977-01-01

    The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined

  8. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  9. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    International Nuclear Information System (INIS)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy's (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher's workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead

  10. Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors; Validacion del codigo AZTRAN 1.1 con problemas Benchmark de reactores LWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M., E-mail: amhed.jvq@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S{sub N}, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO{sub 2} cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

  11. Fusion reactor technology studies. Final report for period August 1, 1972 - October 31, 1978

    International Nuclear Information System (INIS)

    Kulcinski, G.L.; Maynard, C.W.

    1984-04-01

    Major accomplishments for the period August 1, 1972 - October 31, 1978 include the publishing of four comprehensive fusion reactor conceptual design studies; experimental studies in the areas of radiation damage, plasma-wall interactions, superconducting magnets and 14-MeV neutron cross sections; development of the concepts of carbon curtains and ISSEC's for use in fusion reactors; development of a neutron and gamma heating computer code, a radioactivity and afterheat computer code and a neutral transport computer code; and studies in the areas of RF heating for tokamaks and resource assessment for fusion reactors

  12. PR-EDB: Power Reactor Embrittlement Data Base, version 1: Program description

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Kam, F.B.K.; Taylor, B.J.

    1990-06-01

    Data concerning radiation embrittlement of pressure vessel steels in commercial power reactors have been collected form available surveillance reports. The purpose of this NRC-sponsored program is to provide the technical bases for voluntary consensus standards, regulatory guides, standard review plans, and codes. The data can also be used for the exploration and verification of embrittlement prediction models. The data files are given in dBASE 3 Plus format and can be accessed with any personal computer using the DOS operating system. Menu-driven software is provided for easy access to the data including curve fitting and plotting facilities. This software has drastically reduced the time and effort for data processing and evaluation compared to previous data bases. The current compilation of the Power Reactor Embrittlement Data base (PR-EDB, version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points from 110 different irradiated base materials (plates and forgings) and 161 data points from 79 different welds. Results from heat-affected-zone materials are also listed. Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of the PR-EDB and will be supplementing the data base with additional data and documentation. 2 figs., 28 tabs

  13. Project management plan for the 105-C Reactor interim safe storage project. Revision 1

    International Nuclear Information System (INIS)

    Miller, R.L.

    1997-01-01

    In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

  14. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, T L; Hellstrand, E; Londen, S O; Tiren, L I

    1965-08-15

    FR0 is a fast zero power reactor built for experiments in reactor physics. It is a split table machine containing vertical fuel elements. 120 kg of U{sup 235} are available as fuel, which is fabricated into metallic plates of 20 % enrichment. The control system comprises 5 spring-loaded safety elements and 3 + 1 elements for startup operations and power control. The reactor went critical in February 1964. The first assemblies studied were made up of undiluted fuel into a cylindrical and a spherical core, respectively, surrounded by a reflector made of copper. The present report describes some experiments made on these systems. Primarily, critical mass determinations, flux distribution measurements and studies of the conversion ratio are dealt with. The measured quantities have been compared with theoretical predictions using various transport theory programmes (DSN, TDC) and cross section sets. The experimental results show that the neutron spectrum in the copper reflector is softer than predicted, but apart from this discrepancy agreement with theory has generally been obtained.

  15. Neutron radiography in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Pugliesi, R.; Moraes, A.P.V. de; Yamazaki, I.M.; Freitas Acosta, C. de.

    1988-08-01

    Neutronradiography of several materials have been obtained at the IEA-R1 Nuclear Research Reactor (IPEN-CNEN/SP), by means of two conversion techniques: a) (n, α) at the beam-hole n 0 3 where a collimated thermal neutron beam, exposure area 4 cm x 8cm and flux at the sample 10 5 n/s cm 2 is obtained. The film used was the CN-85 cellulose nitrate coated with lithium tetraborate (conversor). The time irradiation of the film was 15 minutes and in following was eteched during 30 minutes in a NaOH(10%) aqueous solution at a constant temperature of 60 0 C.; b) (n,γ) by using an experimental arrangement installed in the botton of the pool of the reactor. The flux of the collimated neutron beam is 10 5 n/s/cm 2 at the sample and the conversion is made by means of a dysprozium sheet. The film used was Kodak T-5. The irradiation and the transfering time was 2 hours and 20 hours respectively. (author) [pt

  16. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients

  17. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  18. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia

    International Nuclear Information System (INIS)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E.

    2014-10-01

    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  19. Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2017-05-15

    Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

  20. Digital Systems Implemented at the IPEN Nuclear Research Reactor (IEA-R1): Results and Necessities

    International Nuclear Information System (INIS)

    Nahuel-Cardenas, Jose-Patricio; Madi-Filho, Tufic; Ricci-Filho, Walter; Rodrigues-de-Carvalho, Marcos; Lima-Benevenuti, Erion-de; Gomes-Neto, Jose

    2013-06-01

    (Nuclear and Energy Research Institute) was founded in 1956 with the main purpose of doing research and development in the field of nuclear energy and its applications. It is located at the campus of University of Sao Paulo (USP), in the city of Sao Paulo, in an area of nearly 500, 000 m2. It has over 1.000 employees and 40% of them have qualification at master or doctor level The institute is recognized as a national leader institution in research and development (R and D) in the areas of radiopharmaceuticals, industrial applications of radiation, basic nuclear research, nuclear reactor operation and nuclear applications, materials science and technology, laser technology and applications. Along with the R and D, it has a strong educational activity, having a graduate program in Nuclear Technology, in association with the University of Sao Paulo, ranked as the best university in the country. The Federal Government Evaluation institution CAPES, granted to this course grade 6, considering it a program of Excellence. This program started at 1976 and has awarded 458 Ph.D. degrees and 937 master degrees since them. The actual graduate enrollment is around 400 students. One of major nuclear installation at IPEN is the IEA-R1 research reactor; it is the only Brazilian research reactor with substantial power level suitable for its utilization in researches concerning physics, chemistry, biology and engineering as well as for producing some useful radioisotopes for medical and other applications. IEA-R1 reactor is a swimming pool type reactor moderated and cooled by light water and uses graphite and beryllium as reflectors. The first criticality was achieved on September 16, 1957. The reactor is currently operating at 4.5 MW power level with an operational schedule of continuous 64 hours a week. In 1996 a Modernization Program was started to establish recommendations in order to mitigate equipment and structures ageing effects in the reactor components, detect and evaluate

  1. Experimental facilities for PEC reactor design central channel test loop: CPC-1 - thermal shocks loop: CEDI

    International Nuclear Information System (INIS)

    Calvaresi, C.; Moreschi, L.F.

    1983-01-01

    PEC (Prova Elementi di Combustibile: Fuel Elements Test) is an experimental fast sodium-cooled reactor with a power of 120 MWt. This reactor aims at studying the behaviour of fuel elements under thermal and neutron conditions comparable with those existing in fast power nuclear facilities. Given the particular structure of the core, the complex operations to be performed in the transfer cell and the strict operating conditions of the central channel, two experimental facilities, CPC-1 and CEDI, have been designed as a support to the construction of the reactor. CPC-1 is a 1:1 scale model of the channel, transfer-cell and loop unit of the channel, whereas CEDI is a sodium-cooled loop which enables to carry out tests of isothermal endurance and thermal shocks on the group of seven forced elements, by simulating the thermo-hydraulic and mechanical conditions existing in the reactor. In this paper some experimental test are briefy discussed and some facilities are listed, both for the CPC-1 and for the CEDI. (Auth.)

  2. 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

  3. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    Energy Technology Data Exchange (ETDEWEB)

    Novelli, A. (Politecnico di Milano (Italy). Centro Studi Nucleari E. Fermi)

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer.

  4. 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

  5. Measurements of reactivity of reactor G1

    International Nuclear Information System (INIS)

    Bernot, J.; Koechlin, J.C.; Portes, L.; Teste du Bailler, A.

    1957-01-01

    The various methods used during the physical study of the reactor G1 to determine the variations of the effective multiplication factor consecutive to a given change in the geometry of the multiplying medium, are presented and discussed. The comparison of the results obtained by these various methods has allowed their validity to be tested and precise conditions of use to be given. In the first part are presented the principles used and their ranges of validity. In the second part the experimental results are given, together with some indications on their comparison with theoretical estimations. (author) [fr

  6. Summary Report of Commercial reactor Criticality Data for Three Mile Island Unit 1

    International Nuclear Information System (INIS)

    Larry B. Wimmer

    2001-01-01

    The objective of the ''Summary Report of Commercial Reactor Criticality Data for Three Mile Island Unit I'' is to present the CRC data for the TMI-1 reactor. Results from the CRC evaluations will support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel. These models and their validation are discussed in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000)

  7. Modifications in the operational conditions of the IEA-R1 reactor under continuous 48 hours operation

    International Nuclear Information System (INIS)

    Moreira, Joao Manoel Losada; Frajndlich, Roberto

    1995-01-01

    This work shows the required changes in the IEA-R1 reactor for operation at 2 Mw, 48 hours continuously. The principal technical change regards the operating conditions of the reactor, namely, the required excess reactivity which now will amount to 4800 pcm in order to compensate the Xe poisoning at equilibrium at 2 Mw. (author). 6 refs, 1 fig, 1 tab

  8. Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

    Energy Technology Data Exchange (ETDEWEB)

    Bond, Leonard J. [Iowa State Univ., Ames, IA (United States); Bowler, John R. [Iowa State Univ., Ames, IA (United States)

    2017-08-30

    The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-service inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO3-xPbTiO3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.

  9. Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

    International Nuclear Information System (INIS)

    Tanabe, Fumiya

    2011-01-01

    In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)

  10. TEM examination of irradiated zircaloy-2 pressure tube material

    International Nuclear Information System (INIS)

    Srivastava, D.; Tewari, R.; Dey, G.K.; Sharma, B.P.; Sah, D.N.; Banerjee, Suparna; Sahoo, K.C.

    2005-09-01

    In the present work, microstructure of the zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water RAPP-1. Reactor (PHWR) has been examined for the first time using transmission electron microscope (TEM). The samples were obtained from a zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit -1, for a period for 6.77 effective full power years (EFPYs) and expected to have a cumulative radiation damage of about 3 dpa. In this study irradiated microstructure has been characterized and compared it with the microstructure of the unirradiated pressure tube samples. The effect of irradiation on the hydriding behaviour is also studied. (author)

  11. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 1

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Tähtinen, S.; Moilanen, P.

    CrZr(HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Special experimental facilities were designed and fabricated for this purpose. A number of in-reactor creep-fatigue experiments were successfully carried out in the BR-2 reactor at Mol...

  12. Reactor of the XXI century

    International Nuclear Information System (INIS)

    Zhotabaev, Zh.R.; Solov'ev, Yu.A.

    2001-01-01

    The advantages of both molten salt reactors (MSR) and homogenous molten salt reactors (HMSR) are illuminated. It is noted that the MSR possess accident probability A=10 -6 1/reactor.years and the HMSR with integral configuration has A=10 -7 1/reactor.years. The methods for these reactors technological problems solution - tritium removal, salt melt circulation and capacity pick up - are discussed

  13. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  14. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  15. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  16. Extension of cycle 8 of Angra-1 reactor, optimization of electric power generation reduction

    International Nuclear Information System (INIS)

    Miranda, Anselmo Ferreira; Moreira, Francisco Jose; Valladares, Gastao Lommez

    2000-01-01

    The main objective of extending fuel cycle length of Angra-1 reactor, is in fact of that each normal refueling are changed about 40 fuel elements of the reactor core. Considering that these elements do not return for the reactor core, this procedure has became possible a more gain of energy of these elements. The extension consists in, after power generation corresponding to a cycle burnup of 13700 MWD/TMU or 363.3 days, to use the reactivity gain by reduction of power and temperature of primary system for power generation in a low energy patamar

  17. Power Trip Set-points of Reactor Protection System for New Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Yang, Soohyung

    2013-01-01

    This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

  18. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  19. The Osiris reactor. Descriptive report - Volume 1 - text

    International Nuclear Information System (INIS)

    1969-05-01

    Osiris is a pool type reactor with a 70 MW thermal power. Its main purpose is to irradiate under high flows of neutrons the materials of which future nuclear power stations are made. This report proposes a description of this pool reactor. A first part describes the functional aspects and general characteristics of all installations which are in principle definitely defined (premises, irradiation and experimentation equipment, water circuits, power supply, venting, controls). The second part addresses elements which are likely to be changed, and more particularly the reactor core: fuel elements and controls (uranium and boron load in different fuel element generations, experimental locations within the core), neutron transport aspects (calculation and experiment), and thermal aspects (power generation and removal) of the pile). The third part addresses the operation: operation cycles, stops, exploitation organisation [fr

  20. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  1. 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)

  2. Monitoring of primary circuit and reactor of NPP A-1

    International Nuclear Information System (INIS)

    Prazska, M.; Majersky, M.; Rezbarik, J.; Sekely, S.; Vozarik, P.; Walthery, R.; Stuller, P.

    2005-01-01

    Nuclear Power Plant A-1 in Jaslovske Bohunice was commissioned in 1972. Heavy water moderated, carbon dioxide cooled channel type reactor was shut down after two accidents in 1977. During more serious second accident, the reduced coolant flow caused local overheating of the fuel and consequent damage/melting of the fuel channel. Both accidents had led to the damage of several fuel assemblies with extensive local damage of fuel claddings. As a consequence, the main cooling circuit was significantly contaminated by fission products and long-life alpha nuclides. The detailed monitoring of dose rates, smearable contamination and sampling of contamination was performed. Extended monitoring in reacto vessel, primary circuit pipes, turbo-compressors, steam generators, main valves, gas tanks and also heavy water system with collectors, coolers, distilling and purification station, pumps and valves was done. Appropriate devices and procedures for the monitoring and examination of the installations were prepared and applied. Obtained results will serve for the future planning of the decontamination and decommissioning works. The 3-D model of the reactor that had been developed as part of this Project proved invaluable for orientation, visualisation, planning and analysis of results. Dose rates were measured in the technological channels from the reactor hall floor to the bottom of the hot gas chamber in decrements of 1 m and 0.5 m. The highest absolute values of dose rates were found in channels located in the middle of the reactor (up to 1900 mGy/h in the active zone region). It is estimated that the total contaminated area of primary circuit equipment (pipework, steam generators and turbo-compressors) is some 48 000 m 2 . It follows that the total gamma contamination is of the order of 10 14 to 10 15 Bq and total alpha contamination 10 11 to 10 13 Bq. The total amount of deposits in the gas circuit is about 14.3 tons. (authors)

  3. Characterisation of reactor control rod drives. Specification 1-6. Reaktorstellstabantriebe. Typenblaetter 1-6

    Energy Technology Data Exchange (ETDEWEB)

    1975-03-01

    The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN).

  4. The FRJ-1 (MERLIN) research reactor: its main activity inventory has been removed by successful demolition of the reactor block

    International Nuclear Information System (INIS)

    Stahn, B.; Printz, R.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2004-01-01

    The FRJ-1 (MERLIN) research reactor was decommissioned in 1985 after twenty-three years of operation. Demolition of the plant was begun in 1996. The article contains a survey of the demolition steps carried out so far within the framework of three partial permits. The main activity is the demolition of the reactor core structures as a precondition for subsequent measures to ensure clearance measurements of the building. The core structures are demolished which were exposed to high neutron fluxes during reactor operation and now show the highest activity and dose rate levels, except for the core internals. For demolition and disassembly of the metal structures in this part of the plant, the tools specially designed and made include a remotely operated sawing system and a pipe cutting system for internal segmentation of the beam lines. The universal demolition tool for use also above and beyond the concrete structures has been found to be a remotely controlled electrohydraulic demolition shovel. Spreading contamination in the course of the demolition work was avoided. One major reason for this success was the fact that no major airborne contamination existed at any time as a consequence of the quality of the material demolished and also of the consistent use of technical tools. While the reactor block was being demolished, an application for clearance measurement of the reactor hall and subsequent release from the scope of the Atomic Energy Act was filed as early as in mid-2003. The fourth partial permit covering these activities is expected to be issued in the spring of 2004. (orig.)

  5. Computer codes for simulation of Angra 1 reactor steam generator

    International Nuclear Information System (INIS)

    Pinto, A.C.

    1978-01-01

    A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt

  6. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  7. 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.

  8. Calculation of radiation heat generation on a graphite reflector side of IAN-R1 Reactor

    International Nuclear Information System (INIS)

    Duque O, J.; Velez A, L.H.

    1987-01-01

    Calculation methods for radiation heat generation in nuclear reactor, based on the point kernel approach are revisited and applied to the graphite reflector of IAN-R1 reactor. A Fortran computer program was written for the determination of total heat generation in the reflector, taking 1155 point in it

  9. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor

    International Nuclear Information System (INIS)

    Alcala Ruiz, F.

    1973-01-01

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

  10. Characterisation of reactor control rod drives. Specification 1-6

    International Nuclear Information System (INIS)

    1975-03-01

    The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN) [de

  11. Developing maintainability in controlled thermonuclear reactors. Progress report, October 1, 1977--April 30, 1978

    International Nuclear Information System (INIS)

    Zahn, H.S.

    1977-05-01

    During the period 1 October 1977 through 30 April 1978 the study has completed work on Task 6, Candidate Reference Systems. Four candidate reference systems have been defined. These are based on the conceptual designs of the UWMAK-III, the General Atomic Company Demonstration Power Reactor, the Oak Ridge National Laboratory Cassette defined in the Demonstration Power Study and the Culham laboratory Mark II Reactors. These reactor concepts are normalized to 3000 MW/sub th/ and near minimum cost of electricity. In addition, designs of four major subsystems have been selected and defined for application to these reactors. These include a primary coolant system, primary and secondary vacuum zone systems, the neutral beam injection system and the magnetic field system. These magnet systems are unique to each reactor. The cases for which maintenance plans are being developed in Task 7 have been selected to allow evaluation of design features, particularly the vacuum wall locations, and the impacts of unscheduled and contact maintenance of subsystems on the cost of electricity

  12. The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto; Prado Fernandes, Marcio; Oliveira, Paulo Fernando; Alves de Amorim, Valter

    1999-01-01

    The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)

  13. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  14. 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

  15. Supervisory system to monitor the neutron flux of the IPR-R1 TRIGA research reactor at CDTN

    International Nuclear Information System (INIS)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Tello, Cledola Cassia Oliveira

    2009-01-01

    The IPR-R1 TRIGA Mark I nuclear research reactor at the Nuclear Technology Development Center - CDTN (Belo Horizonte) is a pool type reactor. It was designed for research, training and radioisotope production. The International Atomic Energy Agency- IAEA - recommends the use of friendly interfaces for monitoring and controlling the operational parameters of nuclear reactors. This paper reports the activities for implementing a supervisory system, using LabVIEW software, with the purpose to provide the IPR-R1 TRIGA research reactor with a modern, safe and reliable system to monitor the time evolution of the power of its core. The use of the LabVIEW will introduce modern techniques, based on electronic processor and visual interface in video monitor, substituting the mechanical strip chart recorders (ink-pen drive and paper) that monitor the current neutrons flux, which is proportional to the thermal power supplied by reactor core. The main objective of the system will be to follow the evolution of the neutronic flux originated in the Linear and Logarithmic channels. A great advantage of the supervisory software nowadays, in relation to computer programs currently used in the facility, is the existence of new resources such as the data transmission and graphical interfaces by net, grid lines display in the graphs, and resources for real time reactor core video recordings. The considered system could also in the future be optimized, not only for data acquisition, but also for the total control of IPR-R1 TRIGA reactor(author)

  16. Probabilistic risk analysis of Angra-1 reactor

    International Nuclear Information System (INIS)

    Spivak, R.C.; Collussi, I.; Silva, M.C. da; Onusic Junior, J.

    1986-01-01

    The first phase of probabilistic study for safety analysis and operational analysis of Angra-1 reactor is presented. The study objectives and uses are: to support decisions about safety problems; to identify operational and/or project failures; to amplify operator qualification tests to include accidents in addition to project base; to provide informations to be used in development and/or review of operation procedures in emergency, test and maintenance procedures; to obtain experience for data collection about abnormal accurences; utilization of study results for training operators; and training of evaluation and reliability techniques for the personnel of CNEN and FURNAS. (M.C.K.) [pt

  17. 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.

  18. Fusion reactor design and technology 1986. V. 1

    International Nuclear Information System (INIS)

    1987-01-01

    The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

  19. Estimated long lived isotope activities in ET-RR-1 reactor structural materials for decommissioning study

    International Nuclear Information System (INIS)

    Ashoub, N.; Saleh, H.

    1995-01-01

    The first Egyptian research reactor, ET-RR-1 is tank type with light water as a moderator, coolant and reflector. Its nominal power is 2MWt and the average thermal neutron flux is 10 13 n/cm 2 sec -1 . Its criticality was on the fall of 1961. The reactor went through several modifications and updating and is still utilized for experimental research. A plan for decommissioning of ET-RR-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences of decommissioning. This paper presents a conservative calculation to estimate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are presented in significant quantities in the reactor structural materials are aluminum, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from 60 Co and 55 Fe which are presented in aluminium as trace elements and in large quantities in other construction materials. (author)

  20. Integral tightness measurements at the Paks-1 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Taubner, R.; Techy, Z. (Villamosenergiaipari Kutato Intezet, Budapest (Hungary))

    1983-01-01

    The containment system experiments of the Paks-1 nuclear reactor are described. The integrated tightness measurements of the hermetic system were completed in 1982. The principles and methods and the evaluation of the results of the measurements are discussed. Some features of the filtration characteristics are demonstrated using relative values and a method enabling the description of the physical contents of the characteristics by flow technical functions is outlined.

  1. 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

  2. Molten-salt reactor strategies viewed from fuel conservation effect, (1)

    International Nuclear Information System (INIS)

    Furuhashi, Akira

    1976-01-01

    Saving of material requirements in the long-term fuel cycle is studied by introducing molten-salt reactors with good neutron economy into a projection of nuclear generating capacity in Japan. In this first report an examination is made on the effects brought by the introduction of molten-salt converter reactors starting with Pu which are followed by 233 U breeders of the same type. It is shown that the sharing of some Pu in the light water- and fast breeder-reactor system with molten-salt reactors provides a more rapid transition to the self-supporting, breeding cycle than the simple fast breeding system, thus leading to an appreciable fuel conservation. Considerations are presented on the strategic repartition of generating capacity among reactor types and it is shown that all of the converted 233 U should be promptly invested to molten-salt breeders to quickly establish the dual breeding system, instead of recycling to converters themselves. (auth.)

  3. Security devices and experiment facilities at ENEA TRIGA RC-1 reactor

    International Nuclear Information System (INIS)

    Bianchi, P.; Festinesi, A.; Santoro, E.; Tardani, G.; Magli, M.; Reis, G.

    1990-01-01

    RC-1 TRIGA operating exercise staff has produced some auxiliary security devices. These are the neutron source automatic handling device, irradiated samples rabbit connection rotating rack, and auxiliary equipment for transferring hot fuel elements. The reactor electronic control instrumentation system includes various instrumentation channels, the operating capability of which must be verified by the licensee as per Italian regulations. In order to obtain automatic and repeatable operations, TEMAV designed and constructed a remotely-driven source transfer device, based on requirements, performance specifications and technical data supplied by ENEA-TIB. The pneumatic irradiating system for short lived materials allows extraction of radiated samples in a time no longer than 4 seconds. To optimize the system, both as to operability and health protection, a specific rotating rack for the connection of irradiated samples with pneumatic transfer (RABBIT) was produced. To permit 1 MW hot fuel element storage in pits it is necessary to remove hot 100 KW fuel elements and transfer them to a re-treatment plant. Feasibility studies showed the impossibility of using heavy trucks inside the reactor hall. To avoid problems trucks are left outside the reactor hall and only the PEGASO container is removed with a special device that runs on rails. Movement from Rail truck is assured by an electromotor driving pull device and security cable

  4. Qualification process of dispersion fuels in the IEAR1 research reactor

    International Nuclear Information System (INIS)

    Domingos, D.B.; Silva, A.T.; Silva, J.E.R.

    2010-01-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a miniplate irradiation device (MID) to be placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 -Al dispersion fuels, LEU type (19,9% of 235 U) with uranium densities of, respectively, 3.0 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and TWODB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer code LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. This paper also presents a system designed for fuel swelling evaluation. The determination of the fuel swelling will be performed by means of the fuel miniplate thickness measurements along the irradiation time. (author)

  5. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    International Nuclear Information System (INIS)

    Diaz Rizo, O.; Alvarez, I.; Herrera, E.; Lima, L.; Tores, J.; Lopez, M.C.; Ixquiac, M.

    1996-01-01

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K o neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott's formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented

  6. Membrane-aerated biofilm reactor for the removal of 1,2-dichloroethane by Pseudomonas sp strain DCA1

    NARCIS (Netherlands)

    Hage, J.C.; Houten, R.T.; Tramper, J.; Hartmans, S.

    2004-01-01

    A membrane-aerated biofilm reactor (MBR) with a biofilm of Pseudomonas sp. strain DCA1 was studied for the removal of 1,2-dichloroethane (DCA) from water. A hydrophobic membrane was used to create a barrier between the liquid and the gas phase. Inoculation of the MBR with cells of strain DCA1 grown

  7. TRAC-BD1: transient reactor analysis code for boiling-water systems

    International Nuclear Information System (INIS)

    Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

    1981-01-01

    The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented

  8. CAC-RA1 1958-1998. The first years of the Constituyentes Atomic Center (CAC). History of the first Argentine nuclear reactor (RA-1)

    International Nuclear Information System (INIS)

    Forlerer, Elena; Palacios, Tulio A.

    1998-01-01

    After giving the milestones of the development of the Constituyentes Atomic Center since 1957, the history of the construction of the first nuclear reactor (RA-1) in Argentina, including the local fabrication of its fuel elements, is surveyed. The RA-1 reached criticality on January 17, 1958. The booklet commemorates the 40th year of the reactor operation

  9. ATMEA and medium power reactors. The ATMEA joint venture and the ATMEA1 medium power reactor

    International Nuclear Information System (INIS)

    Mathet, Eric; Castello, Gerard

    2012-01-01

    This Power Point presentation presents the ATMEA company (a joint venture of Areva and Mitsubishi), the main features of its medium power reactor (ATMEA1) and its building arrangement, indicates the general safety objectives. It outlines the features of its robust design which aim at protecting, cooling down and containing. It indicates the regulatory and safety frameworks, comments the review of the safety options by the ASN and the results of this assessment

  10. Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report October 1, 1977--July 1, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Conn, R.W.; Kulcinski, G.L.; Maynard, C.W.

    1978-01-01

    Since the last progress report we have concentrated on three main areas of research: (1) the study of the NUWMAK reactor design, (2) the study of rf heating for tokamak reactors, and (3) the initiation of a tandem mirror reactor study. The initial work on the tandem mirror reactor is included as background in the technical proposal. Summaries of our work on recent assessments of lithium reserves and neutral transport codes are included.

  11. Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report October 1, 1977--July 1, 1978

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Maynard, C.W.

    1978-01-01

    Since the last progress report we have concentrated on three main areas of research: (1) the study of the NUWMAK reactor design, (2) the study of rf heating for tokamak reactors, and (3) the initiation of a tandem mirror reactor study. The initial work on the tandem mirror reactor is included as background in the technical proposal. Summaries of our work on recent assessments of lithium reserves and neutral transport codes are included

  12. University Reactor Sharing Program. Period covered: September 1, 1981-August 31, 1982

    International Nuclear Information System (INIS)

    Hajek, B.K.; Myser, R.D.; Miller, D.W.

    1982-12-01

    During the period from September 1, 1981 to August 31, 1982, the Ohio State University Nuclear Reactor Laboratory participated in the Reactor Sharing Program by providing services to eight colleges and universities. A laboratory on Neutron Activation Analysis was developed for students in the program. A summary of services provided and a copy of the laboratory procedure are attached. Services provided in the last funded period were in three major areas. These were neutron activation analysis, nuclear engineering labs, and introductions to nuclear research. One group also performed radiation surveys and produced isotopes for calibration of their own analytical equipment

  13. Reversal of OFI and CHF in Research Reactors Operating at 1 to 50 Bar. Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, A. P. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Matos, J. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-02-28

    The conditions at which the critical heat flux (CHF) and the heat flux at the onset of Ledinegg flow instability (OFI) are equal, are determined for a coolant channel with uniform heat flux as a function of five independent parameters: the channel exit pressure (P), heated length (Lh) , heated diameter (Dh), inlet temperature (Tin), and mass flux (G). A diagram is made by plotting the mass flux and heat flux at the OFI-CHF intersection (reversal from CHF > OFI to CHF < OFI as G increases) as a function of P (1 to 50 bar), for 36 combinations of the remaining three parameters (Lh , Dh , Tin): Lh = 0.28, 0.61, 1.18 m; Dh = 3, 4, 6, 8 mm; Tin = 30, 50, 70 °C. The use of the diagram to scope whether a research reactor is OFI-limited (below the curve) or CHF-limited based on the five parameters of its coolant channel is described. Justification for application of the diagram to research reactors with axially non-uniform heat flux is provided. Due to its limitations (uncertainties not included), the diagram cannot replace the detailed thermal-hydraulic analysis required for a reactor safety analysis. In order to make the OFI-CHF intersection diagram, two world-class CHF prediction methods (the Hall-Mudawar correlation and the extended Groeneveld 2006 table) are compared for 216 combinations of the five independent parameters. The two widely used OFI correlations (the Saha- Zuber and the Whittle-Forgan with η = 32.5) are also compared for the same combinations of the five parameters. The extended Groeneveld table and the Whittle-Forgan OFI correlation are selected for use in making the diagram. Using the above five design parameters, a research reactor can be represented by a point on the reversal diagram, and the diagram can be used to scope, without a thermal-hydraulic calculation, whether the OFI will occur before the CHF, or the CHF will occur before the OFI when the reactor power is increased keeping the five parameters fixed.

  14. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  15. Upgrade of Instrumentation for Purdue Reactor PUR-1

    International Nuclear Information System (INIS)

    Revankar, S.T.; Merritt, E.; Bean, R.

    2000-01-01

    The major objective of this program was to upgrade and replace instruments and equipment that significantly improve the performance, control and operational capability of the Purdue University nuclear reactor (PUR-1). Under this major objective two projects on instrument upgrade were implemented. The first one was to convert the vacuum tube control and safety amplifiers (CSA) to solid state electronics, and the other was to upgrade the electrical and electronic shielding. This report is the annual report and gives the efforts and progress achieved on these two projects from July 1999 to June 2000

  16. Main refurbishment activities on electronic and electrical equipment for the FRG-1 research reactor

    International Nuclear Information System (INIS)

    Blom, K.H.; Krull, W.

    1997-01-01

    As GKSS intends to operate the research reactor FRG-1 safely and reliably for many years to come, the plant is constantly refurbished and upgraded both in the interests of safety and operational reasons. The following electronic and electrical systems have been replaced or improved since 1990: Information and signalling systems; Emergency power plant (permit applied for); External and internal lightning protection system; Reactor protection system (in part); Safety lighting; Alarm and staff locating system; Control room telephone system; Closed-circuit television system; Beam tube controls; Storage plant for radioactive liquid waste; Ambient dose rate measuring system; Meteorological measuring system; Control and measuring system for the primary cooling circuit; Control rod drives; Control rod control system; Soft start for the secondary pumps; Control and switching devices for the emergency power plant; Trailing cable installation for the reactor bridge; Main-voltage distribution systems/cable routes. (author). 13 figs, 1 tab

  17. Modernization of turbine control system and reactor control system in Almaraz 1 and 2; MOdernizacion de los sistemas de control de turbina y del reactor en Almaraz 1 y 2

    Energy Technology Data Exchange (ETDEWEB)

    Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

    2005-07-01

    The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

  18. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  19. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    Maillard, M.L.

    1960-01-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

  20. 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

  1. Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

    International Nuclear Information System (INIS)

    Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

    2010-01-01

    The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  2. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Rizo, O; Alvarez, I; Herrera, E; Lima, L; Tores, J [Secretaria Ejecutiva para Asuntos Nucleares, Holguin (Cuba). Delegacion Territorial; Manso, M V [Centro de Isotopos, La Habana (Cuba); Lopez, M C [Instituto Nacional de Investigaciones Nucleares, Mexico City (Mexico); Ixquiac, M [Universidad de San Carlos de Guatemala, Guatemala City (Guatemala)

    1997-12-31

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K{sub o} neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott`s formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented.

  3. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    International Nuclear Information System (INIS)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo

    2011-01-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  4. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  5. 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.)

  6. Real time monitoring system of the operation variables of the TRIGA IPR-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    Ricardo, Carla Pereira; Mesquita, Amir Zacarias

    2007-01-01

    During the last two years all the operation parameters of the TRIGA IPR-R1 were monitored and real time indicated bu the data acquisition system developed for the reactor. All the information were stored on a rigid disk, at the collection system computer, leaving the information on the reactor performance and behaviour available for consultation in a chronological order. The data acquisition program has been updated and new reactor operation parameters were included for increasing the investigation and experiments possibilities. The register of reactor operation variables are important for the immediate or subsequent safety analyses for reporting the reactor operations to the external organizations. This data acquisition satisfy the IAEA recommendations. (author)

  7. The different generation of nuclear reactors from Generation-1 to Generation-4

    International Nuclear Information System (INIS)

    Cognet, G.

    2010-01-01

    In this work author deals with the history of the development of nuclear reactors from Generation-1 to Generation-4. The fuel cycle and radioactive waste management as well as major accidents are presented, too.

  8. Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)

  9. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

    The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

  10. 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)

  11. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, St.

    2005-01-01

    The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

  12. Feasibility study of application of Prompt Gamma Neutron Activation Analysis (PGNAA) method in TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Guerra, Bruno Teixeira

    2016-01-01

    The TRIGA Mark I IPR-R1 research reactor is located at Nuclear Technology Development Centre (CDTN), Brazilian Commission for Nuclear Energy (CNEN), in Belo Horizonte, Brazil. The reactor operates at 100 kW but the core configuration allows the increasing of the power up to 250 kW. It has been applied research, training and radioisotopes production. The establishment of the Prompt Gamma Neutron Activation Analysis (PGNAA) method at the TRIGA IPR-R1 reactor will significantly increase the types of matrices analysed as well as the number of chemical elements. Additionally it will complement the neutron activation analysis. This work presents a proposed design of a PGNAA facility to be installed at the TRIGA IPR-R1. The proposed design is based on a tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room’s level where shall be located the rack containing the set sample/detector/shielding. Thus, the aim of this study is to verify the feasibility to establish the PGNAA method in IPR-R1 through theoretical study applying the Monte Carlo code. The feasibility of establishing the PGAA method at the IPR-R1 installations was evaluated through of the calculations of neutron flux, radioactive capture reaction rates and detection limits for some isotopes. According to the obtained results, it can be concluded that is possible to establish the PGAA method at the IPR-R1 reactor, even with some restrictions in its theoretical design calculated by MCNP. (author)

  13. Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

    International Nuclear Information System (INIS)

    1989-02-01

    Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

  14. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities.

  15. Application of Cherenkov light observation to reactor measurements (1). Estimation of reactor power from Cherenkov light intensity

    International Nuclear Information System (INIS)

    Yamamoto, Keiichi; Takeuchi, Tomoaki; Kimura, Nobuaki; Ohtsuka, Noriaki; Tsuchiya, Kunihiko; Sano, Tadafumi; Nakajima, Ken; Homma, Ryohei; Kosuge, Fumiaki

    2015-01-01

    Development of the reactor measurement system was started to obtain the real-time in-core nuclear and thermal information, where the quantitative measurement of brightness of Cherenkov light was investigated. The system would be applied as a monitoring system in severe accidents and for the advanced operation management technology in existing LWRs. The calculation and the observation were performed to obtain the quantity of the Cherenkov light caused by the gamma and beta rays emitted from the fuels in the core of Kyoto University Research Reactor. The results indicate that the real-time reactor power can be estimated from the brightness of the Cherenkov light observed by a CCD camera. This method can also work for the estimation of the burn-up of spent fuels at commercial reactors. Since the observed brightness value of the Cherenkov light was influenced by the camera position, the optical observation method should be improved to achieve high accuracy observation. (author)

  16. 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)

  17. Measurements and calculations of reactivity for the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Ferreira, P.S.B.; Maiorino, J.R.; Yamaguchi, M.

    1988-01-01

    This work shows a measurement of reactivity parameters, such as integral and diferential control rod worth, local void coefficient, and moderator temperature coefficient for the research reactor IEA-R1. The measured values were compared with those calculated through HAMMER-CITATION codes, having shown good agreement. (author) [pt

  18. Accident management strategies for VVER-1000 reactors. Part 1: text

    International Nuclear Information System (INIS)

    Sdouz, G.; Sonneck, G.; Pachole, M.

    1994-10-01

    This report describes the effect of different accident management strategies on the onset, development and end of the core-concrete-interaction as well as on the containment integrity for a TMLB'-type sequence in a Pressurized Water Reactor of the type VVER- 1000. Using the computer code MARCH3 the following strategies were investigated: (1) One or more Spray and LP ECC Systems available with and without coolers after 10 hours (2) Inclusion of the reactor pressure vessel testing facility room to the cavity (3) Containment venting (4) External water supply and (5) Different electric power restoration times. The results show that some of these accident management measures will maintain the containment integrity and reduce the source term drastically, others will reduce the source term rate. For some measures final conclusions can only be given after complete source term calculations have been performed. (authors)

  19. Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

    International Nuclear Information System (INIS)

    Moukhamadeev, R.; Suvorov, A.

    2000-01-01

    This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

  20. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling

    International Nuclear Information System (INIS)

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail

  1. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

  2. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-01

    This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  3. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-01

    This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  4. Low enriched uranium UAl{sub X}-Al targets for the production of Molybdenum-99 in the IEA-R1 and RMB reactors

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2011-07-01

    The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralized water and having Beryllium and Graphite as reflectors. In 1997 the reactor received the operating licensing for 5 MW. A new research reactor is being planned in Brazil to replace the IEA-R1 reactor. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Low enriched uranium (LEU) (<20% {sup 235}U) UAl{sub x} dispersed in Al targets are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed, respectively, to compare the production of {sup 99}Mo for these targets in IEA-R1 reactor and RMB and to determine the temperatures achieved in the UAl{sub x}-Al targets during irradiation. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations was utilized the computer code MTRCR-IEAR1. (author)

  5. Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter

    2013-01-01

    The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

  6. Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

  7. Acoustic emission monitoring of preservice testing at Watts Bar Unit 1 Nuclear Reactor

    International Nuclear Information System (INIS)

    Hutton, P.H.; Pappas, R.A.; Friesel, M.A.

    1985-02-01

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Plant in the US during hot functional preservice testing is described. Background, methodology, and results are included. The work discussed here is a major milestone in a program supported by the US NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing to AE monitoring during reactor operation. 3 refs., 6 figs

  8. 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.)

  9. 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

  10. Space reactor electric systems: system integration studies, Phase 1 report

    International Nuclear Information System (INIS)

    Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

    1983-01-01

    This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied

  11. Thermal performance of Egypt's research reactor core (ET-RR-1)

    International Nuclear Information System (INIS)

    Khattab, M.; Mariy, A.

    1986-01-01

    The steady state thermal performance of the ET-RR-1 core system is theoretically investigated by different models describing the heat flux and the coolant mass flow rate. The magnitude of the heat generated by a fuel element depends upon its position in the core. Normal and uniform distributions for heat flux and coolant mass flow rate are considered. The clad and coolant temperatures at different core positions are evaluated and compared with the experimental measurements at different operating conditions. The results indicated large discrepancy between the predicted and the experimental results. Therefore, the previous models and the experimental results are evaluated in order to develop the best model that describes the thermal performance of the ET-RR-1 core. The adapted model gives 99.5% significant confidence limit. The effect of increasing the heat flux or decreasing the mass flow rate by 20% from its maximum recommended operating condition is tested and discussed. Also, the thermal behaviour towards increasing the reactor power more than its maximum operating condition is discussed. The present work could also be used in extending the investigation to other PWR reactor operating conditions

  12. Calibration of new I and C at VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, Martin; Jurickova, Monika

    2011-01-01

    The paper describes a calibration of the new instrumentation and control (I and C) at the VR-1 training reactor in Prague. The I and C uses uncompensated fission chambers for the power measurement that operate in a pulse or a DC current and a Campbell regime, according to the reactor power. The pulse regime uses discrimination for the avoidance of gamma and noise influence of the measurement. The DC current regime employs a logarithmic amplifier to cover the whole reactor DC current power range with only one electronic circuit. The system computer calculates the real power from the logarithmic data. The Campbell regime is based on evaluation of the root mean square (RMS) value of the neutron noise. The calculated power from Campbell range is based on the square value of the RMS neutron noise data. All data for the power calculation are stored in computer flash memories. To set proper data there, it was necessary to carry out the calibration of the I and C. At first, the proper discrimination value was found while examining the spectrum of the neutron signal from the chamber. The constants for the DC current and Campbell calculations were determined from an independent reactor power measurement. The independent power measuring system that was used for the calibration was accomplished by a compensated current chamber with an electrometer. The calculated calibration constants were stored in the computer flash memories, and the calibrated system was again successfully compared with the independent power measuring system. Finally, proper gamma discrimination of the Campbell system was carefully checked.

  13. Neutronics Design of Helical Type DEMO Reactor FFHR-d1

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Sagara, A.; Goto, T.; Yanagi, N.; Masuzaki, S.; Tamura, H.; Miyazawa, J.; Muroga, T., E-mail: teru@nifs.ac.jp [National Institute for Fusion Science, Toki (Japan)

    2012-09-15

    Full text: Neutronics design study has been performed in a newly started conceptual design activity for a helical type DEMO reactor FFHR-d1. Features of the FFHR-d1 design are enlargement of the basic configurations of reactor components and extrapolation of plasma parameters from those of the helical type plasma experimental machine Large Helical Device (LHD) to achieve the highest feasibility. From the neutronics point of view, a blanket space of FFHR-d1 is severely limited at the inboard of the torus. This is due to the core plasma position shifting to the inboard side under the confinement condition extrapolated from LHD. The first step of the neutronics investigation using the MCNP code has been performed with a simple torus model simulating thin inboard blanket space. A Flibe+Be/Ferritic steel breeding blanket showed preferable performances for both tritium breeding and shielding, and has been adapted as a reference blanket system for FFHR-d1. The investigations indicate that a combination of a 15 cm thick breeding blanket, 55 cm thick WC+B4C shield, i.e., the blanket space of 70 cm, could suppress the fast neutron flux and nuclear heating in the helical coils to the design targets for the neutron wall loading of 1.5 MW/m{sup 2}. Since the outboard side can provide a large space for a 60 cm thick breeding blanket, a fully-covered tritium breeding ratio (TBR) of 1.31 has been obtained in the simple torus model. The neutronics design study has proceeded to the second step using a 3-D helical reactor model. The most important issue in the 3-D neutronics design is a compatibility with the helical divertor design. To achieve a higher TBR and shielding performance, the core plasma has to be covered by the breeding blanket layers as possible. However, the dimensions of the blanket layers are limited by magnetic field lines connecting an edge of the core plasma and divertor pumping ports. After repeating modification of the blanket configuration, the global TBR of 1

  14. 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

  15. 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.)

  16. 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

  17. The role of SASSYS-1 in LMR [Liquid Metal Reactor] safety analysis

    International Nuclear Information System (INIS)

    Dunn, F.E.; Wei, T.Y.C.

    1988-01-01

    The SASSYS-1 liquid metal reactor systems analysis computer code is currently being used as the principal tool for analysis of reactor plant transients in LMR development projects. These include the IFR and EBR-II Projects at Argonne National Laboratory, the FFTF project at Westinghouse-Hanford, the PRISM project at General Electric, the SAFR project at Rockwell International, and the LSPB project at EPRI. The SASSYS-1 code features a multiple-channel thermal-hydraulics core representation coupled with a point kinetics neutronics model with reactivity feedback, all combined with detailed one-dimensional thermal-hydraulic models of the primary and intermediate heat transport systems, including pipes, pumps, plena, valves, heat exchangers and steam generators. In addition, SASSYS-1 contains detailed models for active and passive shutdown and emergency heat rejection systems and a generalized plant control system model. With these models, SASSYS-1 provides the capability to analyze a wide range of transients, including normal operational transients, shutdown heat removal transients, and anticipated transients without scram events. 26 refs., 16 figs

  18. Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Souza, Rose Mary Gomes do Prado

    2009-01-01

    Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R1 TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculate as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor. (author))

  19. 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.

  20. 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)

  1. 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)

  2. First fuel re-load of Angra-1 reactor - Inspection and hearing plan

    International Nuclear Information System (INIS)

    Pollis, W.; Alvarenga, M.A.B.; Meldonian, N.L.; Paiva, R.L.C. de; Pollis, R.

    1985-01-01

    The plan of inspection and hearing of the first fuel reload of Angra-1 nuclear reactor is detailed. It consists in five steps: receiving and storage of the fuel; reload preparation; activities during; post-reload activities, and preliminary activities. (M.I.)

  3. Multipurpose RTOF Fourier diffractometer at the ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Maayouf, R.M.A.; Tiitta, A.T.

    1993-09-01

    The present work represents a further study of the basic RTOF Fourier multipurpose diffractometer, to start with, at the ET-RR-1 reactor. The functions of the suggested arrangement are thoroughly discussed and the possibilities if its expansion are also assessed. The flexibility of the arrangement allows its further expansion both for stress measurement at 90 deg. scattering angle with two detector banks at opposite sides of the incident beam and for operation in the transmission diffraction mode. (orig.). (19 refs., 10 figs., 1 tab.)

  4. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

    International Nuclear Information System (INIS)

    Reyes F, M. C.; Del Valle G, E.; Filio L, C.

    2013-10-01

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  5. Supercritical Water Reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

    2014-01-01

    The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

  6. Current utilization and long term strategy of the Finnish TRIGA research reactor FiR 1

    International Nuclear Information System (INIS)

    Auterinen, Iiro; Salmenhaara, Seppo

    2008-01-01

    FiR 1 (TRIGA Mark II, 250 kW) has an important international role in the development of boron neutron capture therapy (BNCT) for cancer. The safety and efficacy of BNCT is studied for several different cancers: - primary glioblastoma, a highly malignant brain tumour (since 1999); - recurrent glioblastoma or anaplastic astrocytoma (since 2001); - recurrent inoperable head and neck carcinoma (since 2003). It is one of the few facilities in the world providing this kind of treatments. The successes in the BNCT development have now created a demand for these treatments, although they are given on an experimental basis. Well over 100 patients treated now since May 1999: - at least 1 patient irradiation / week, often 2 (Tuesday and Thursday) - patients are referred to BNCT-treatments from several hospitals, also outside research protocols; - the hospitals pay for the treatment. The FiR 1 reactor has proven to be a reliable neutron source for the BNCT treatments; no patient irradiations have been cancelled because of a failure of the reactor. The BNCT facility has become a center of extensive academic research especially in medical physics. Nuclear education and training continue to play also a role at FiR 1 in the form of university courses and training of nuclear industry personnel. FiR 1 is one of the two sources in Scandinavia for short lived radioisotopes used in tracer studies in industry. The main isotope produced is Br-82 in the form of either KBr or ethylene bromide. Other typical isotopes are Na-24, Ar-41, La-140. The isotopes are used mainly in tracer studies in industry (Indmeas Inc., Finland). Typical activity of one irradiated Br-sample is 20 - 80 GBq; total activity produced in one year is over 3 TBq; the reactor operating time needed for the isotope production is one or two days per week. Accelerator based neutron sources are developed for BNCT. The prospect is that when BNCT will achieve a status of a fully accepted and efficient treatment modality for

  7. Neutron Spectrum Parameters In Inner Irradiation Channel Of The Nigeria Research Reactor-1 (NIRR-1) For Use In Absolute And KO-NAA Methods

    International Nuclear Information System (INIS)

    Jonah, S.A; Balogun, G.I; Mayaki, M.C.

    2004-01-01

    In Nigeria, the first Nuclear Reactor achieved critically on February 03, 2004 at about 11:35 GMT and has been commissioned or training and research. It is a Miniature Neutron Source Reactor (MNSR), code-named Nigeria Research Reactor-1 (NIRR-1). NIRR-1 has a tan-in-pool structural configuration and a nominal thermal power rating of 30 Kw. With a built-in clean old core excess reactivity of 3.77 mk determined during the on-site zero and critically experimental, the reactor can operate for a n.cm-2 .s-1 in the inner irradiation channels). Under these conditions, the reactor can operate with the same fuel loading for over ten years with a burn-up of <1%. A detailed description of operating characteristics for NIRR-1, measured during the on-site zero-power and criticality experiments has been given elsewhere. In order to extend its utilization to include absolute and ko-NAA methods, the neutron spectrum parameters in the irradiation channels: power and critically experiments has been given elsewhere. In order to extend it's the irradiation channels: thermal-to-epithermal flux ration, F; and epithermal flux shape factor, a in both the inner and outer irradiation channels must be determined experimentally. In this work, we have developed and experimental procedure for monitoring the neutron spectrum parameters in an inner irradiation channel based on irradiation and gamma-ray counting of detector foils via (n,y), (n,p) and (n,a) dosimetry reactions. Results obtained indicate that a thermal neutron flux of (5.14+-0.02) x 1011 n/c m2.s determined by foil activation method in the inner irradiation channel, B2, at a power level of 15.5 kw corresponds to the flux indicators on the control console and the micro-computer control system respectively. Other parameters of the neutron spectrum determined for inner irradiation channel B2, are: a -0.0502+0.003; 18.92+-0.14; F = 3.87=0.23. The method was validated through the comparison of our result with published neutron spectrum

  8. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor; Istrazivacki nuklearni reaktor RA, Deo 1 - Pogon, odrzavanje i eksploatacija reaktora u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Milosevic, M; Martinc, R; Kozomara-Maic, S; Cupac, S; Radivojevic, J; Stamenkovic, D; Skoric, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1981-12-15

    The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

  9. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    Veeraraghaven, N.

    1990-01-01

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  10. Maintenance of reactor recirculation pumps [Paper No.: II-1

    International Nuclear Information System (INIS)

    Ansari, M.A.; Bhat, K.P.

    1981-01-01

    At Tarapur Atomic Power Station (TAPS), two reactor recirculation pumps are provided, one each for the two reactor units. The performance of pumps has been uniformly good; however, leakage through the cartridge type, two stage, mechanical seals which are installed on these pumps was encountered on few occasions. The paper describes the leakage problems, identification of certain design deficiencies and rectification carried out at TAPS for overcoming these problems. (author)

  11. Study on operational aspect of natural circulation HLMC reactor (1)

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Cahalan, J.E.; Spencer, B.W.

    2000-08-01

    The concept of a heavy liquid metal cooled fast reactor that achieves 100% natural circulation heat removal from the core has the potential to attain improved cost competitiveness through extreme simplification, proliferation resistance, and heightened passive safety. The concept offers the potential for simplifications in plant control strategies wherein inherent reactor feedbacks may restore balance between energy release and heat removal from the reactor during operation as well as providing passive reactivity shutdown in the event of transients involving failure to scram. This study was initiated to evaluate the operational characteristics of the 100% natural circulation reactor under normal and transient states using a plant dynamics analysis computer code and to seek design and operational optimization of the concept. In the current Phase I of the project, the stage for the overall study has been prepared. A coupled thermal hydraulics-kinetics plant dynamics analysis code has been developed/modified that has the capabilities to calculate operational and accident transients. Code input has been prepared for the heavy liquid metal cooled natural circulation reactor concept. A preliminary analysis using the plant dynamics code and its input to calculate three illustrative cases relevant to initial startup, shutdown following long-term operation, and change in turbine load demonstrates the capability to analyze typical transient cases. (author)

  12. Design and development of weld inspection manipulator for reactor pressure vessel of TAPS-1

    International Nuclear Information System (INIS)

    Chatterjee, H.; Singh, J.P.; Ranjon, R.; Kulkarni, M.P.; Patel, R.J.

    2013-01-01

    The reactor pressure vessel (RPV) of TAPS-1 BWR contains six longitudinal and four circumferential welds. Periodical in-service inspection of these weld joints has been a regulatory issue pending for long. In the 22 nd refuelling outage in July 2012 the inspection of L1-1, L1-2 longitudinal welds as well as their junctions with C1 circumferential weld were proposed to be done using ultrasonic technique. Approaching these welds from OD side of the RPV is a difficult and tedious task. Therefore it was decided to examine these welds from ID side of the RPV by filling the cavity with water and approaching the RPV from top. No technology was locally available to take the probes at a depth of 10-12 m under water. NPCIL approached RTD, BARC to develop an underwater manipulator to accomplish this task. RTD took up this work as a challenge and came out with the design of manipulator. The weld inspection manipulator (WIM) was fabricated on a war foot basis, tested and successfully implemented in the reactor for the first time in TAPS history. The entire activity was completed in three months time. This article gives the details of design, manufacturing, performance testing, qualification trials and implementation of WIM in the reactor. Ultrasonic testing techniques were developed by QAD, BARC which are not covered in this article. (author)

  13. Investigation on innovative water reactor for flexible fuel cycle (FLWR). (1) Conceptual design

    International Nuclear Information System (INIS)

    Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiko; Ohnuki, Akira; Iwamura, Takamichi

    2005-01-01

    A concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI) in order to ensure sustainable energy supply in the future based on the well-experienced Light Water Reactor (LWR). The concept aims at effective and flexible utilization of uranium and plutonium resources through plutonium multiple recycling by two stages. In the first stage, the FLWR core realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The core in the second stage represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the core concepts in both stages utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of natural uranium resource, or establishment of economical reprocessing technology of MOX spent fuel. The FLWR is essentially a BWR-type reactor, and its core design is characterized by use of hexagonal-shaped fuel assemblies with the triangular-lattice fuel rod configuration of highly enriched MOX fuel, control rods with Y-shaped blades, and a short and flat core design. Detailed investigations have been performed on the core design, in conjunction with the other related studies such as on thermal hydraulics in the tight lattice core including experimental activities, and the results obtained so far have shown the proposed concept is feasible and promising. (author)

  14. Neutronics analysis of Nigerian Research Reactor-1

    International Nuclear Information System (INIS)

    Azande, T.S.; Balogun, G.I.

    2010-01-01

    Feasibility studies for the conversion of the Nigerian Research Reactor-1 (NIRR-1) have been performed using WIMS and CITATION codes (Azande et al, 2009 and Balogun, 2003) at the Centre for Energy Research and Training (CERT), Ahmadu Bello University, Zaria Kaduna State. In this work, the neutronics analysis of NIRR-1 core concerning mass loading of U-235 in the core, shut down margin (SDM), safety reactivity factor (SRF), control rod worth, and control rod critical depth of insertion were investigated at low enrichment. Two fuel types (UAl 4 and UO 2 ) were considered and the uranium densities required for the conversion of NIRR-1 core to low enrichment were computed to be 1201g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1274 g/cc with 15% enrichment, 1448 g/cc with 10% enrichment for UAl 4 fuel type and 1141g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1216 g/cc with 15% enrichment, and 1389 g/cc with 10% enrichment for UO 2 fuel type. Signi ficantly, higher uranium densities are required to convert NIRR-1 from HEU to LEU - indicating a drastic review of the NIRR-1 core.

  15. Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ∼ 25 effective full power years of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M and calculated (C results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE/C ratios of 1.10 for both neutron (E >1.0 MeV flux and iron atom displacement rate.

  16. Development of System Model for Level 1 Probabilistic Safety Assessment of TRIGA PUSPATI Reactor

    International Nuclear Information System (INIS)

    Tom, P.P; Mazleha Maskin; Ahmad Hassan Sallehudin Mohd Sarif; Faizal Mohamed; Mohd Fazli Zakaria; Shaharum Ramli; Muhamad Puad Abu

    2014-01-01

    Nuclear safety is a very big issue in the world. As a consequence of the accident at Fukushima, Japan, most of the reactors in the world have been reviewed their safety of the reactors including also research reactors. To develop Level 1 Probabilistic Safety Assessment (PSA) of TRIGA PUSPATI Reactor (RTP), three organizations are involved; Nuclear Malaysia, AELB and UKM. PSA methodology is a logical, deductive technique which specifies an undesired top event and uses fault trees and event trees to model the various parallel and sequential combinations of failures that might lead to an undesired event. Fault Trees (FT) methodology is use in developing of system models. At the lowest level, the Basic Events (BE) of the fault trees (components failure and human errors) are assigned probability distributions. In this study, Risk Spectrum software used to construct the fault trees and analyze the system models. The results of system models analysis such as core damage frequency (CDF), minimum cut set (MCS) and common cause failure (CCF) uses to support decision making for upgrading or modification of the RTP?s safety system. (author)

  17. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  18. 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

  19. 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

  20. 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.