Operating limits for subassembly deformation in EBR-II

International Nuclear Information System (INIS)

Bottcher, J.H.

1977-01-01

The deformation of a subassembly in response to the core environment is frequently the life limiting factor for that component in an LMFBR. Deformation can occur as diametral and axial growth or bowing of the subassembly. Such deformation has caused several handling problems in both the core and the storage basket of EBR-II and may also have contributed to reactivity anomalies during reactor operation. These problems generally affect plant availability but the reactivity anomalies could lead to a potential safety hazard. Because of these effects the deformation mechanisms must be understood and modeled. Diametral and axial growth of subassembly ducts in EBR-II is due to swelling and creep and is a function of temperature, neutron fluence and stress. The source of stress in a duct is the hydraulic pressure difference across the wall. By coupling the calculated subassembly growth rate to the available clearance in the core or storage basket a limiting neutron fluence, or exposure, can be established

Therapeutic efficacy and dosimetric aspects of Rhenium-188-HEDP in bone pain palliation

International Nuclear Information System (INIS)

Liepe, Knut

2005-01-01

Full text: Bone metastases are a frequent complication of cancer, occurring in up 70% of patients suffering from advanced breast or prostate cancer and often present with severe bone pain. In this purpose the radionuclide therapy is a useful option for cancer patients. Different radionuclides are described, such as 89 Sr, 32 P, 153 Sm-EDTMP, 186 Re-HEDP, 131 I-BDP3, 90 Y, 117mSn-DTPA, 188 Re-HEDP and 188 Re-DMSA. The most experiences are available for 89 Sr. An indication for the treatment are patients with osteoblastic metastases, bone pain, sufficient bone marrow function and at least of three bone metastases visualized in bone scan. A bisphosphonate therapy, a chemotherapy with lower bone marrow toxicity or a local field external beam radiotherapy represent no contraindications, especially because the reported synergistic effects to the systemic radionuclide therapy. In 33 treated patients (breast and prostate cancer) we investigated the effect of 188 Re-HEDP on pain relief, analgesic intake and impairment of bone marrow function. There were an improvement on the Karnofsky performance scale from 74 7% to 85 9% 12 weeks after therapy (p= 0.001). The pain score showed a maximum decrease from 44 ± 18% to 27 ± 20% in the 3rd to the 8th week after therapy (p = .009) and 76% had a pain relief (20% were pain free). The maximal differences between the values of platelets and leukocytes before and after therapy were not statistically significant (p = 0.021 and p = 0.094). In 105 investigated patients treated with different radionuclides ( 89 Sr, 153 Sm-EDTMP, 186 Re-HEDP, 188 Re-HEDP and 89 Sr in combination with chemotherapy) no different therapeutic efficacy of the treatments were observed. In dose calculation of 188 Re-HEDP a radiation dose of 3.83 ± 2.01 mGy/MBq (12.6 Gy for 3300 MBq) for bone metastases and 0.61 ± 0.21 mGy/MBq (2 Gy for 3300 MBq) were found. With the introducing of radionuclide treatments with chemotherapy and repeated treatments, the

Fuel cell sub-assembly

Science.gov (United States)

Chi, Chang V.

1983-01-01

A fuel cell sub-assembly comprising a plurality of fuel cells, a first section of a cooling means disposed at an end of the assembly and means for connecting the fuel cells and first section together to form a unitary structure.

Fuel sub-assembly

International Nuclear Information System (INIS)

Jolly, R.

1982-01-01

A fuel sub-assembly for a liquid metal cooled nuclear reactor is described in which the bundle of fuel pins are braced apart by a series of spaced grids. The grids at the lower end are capable of yielding, thus allowing pins swollen by irradiation to be withdrawn with a reduced risk of damage. (U.K.)

Develoment of pressure drop calculation modules for a wire-wrapped LMR subassembly

International Nuclear Information System (INIS)

Kim, Young Gyun; Lim, Hyun Jin; Kim, Won Seok; Kim, Young Il

2000-06-01

Pressure drop calculation modules for a wire-wrapped LMR subassembly was been developed. This report summarizes present information on pressure drop calculation modules for inlet hole, lower part and upper part of a wire-wrapped LMR subassembly which was developed using simple formulas of sudden expansion and sudden contraction. A case calculation study was done using design data of a KALIMER driver fuel subassembly. And the total pressure drop in the driver fuel subassembly, except for the bundle part, was calculated as 0.13 MPa, which is in the reasonable pressure drop range. The developed modules will be integrated in the total subassembly pressure drop calculation code with further improvements

LMFBR subassembly response to local pressure loadings: an experimental approach

International Nuclear Information System (INIS)

Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

1975-01-01

An experimental program to determine the response of LMFBR-type subassemblies to local subassembly accidents caused by pressure loadings is described. Some results are presented and compared with computer calculations

Nuclear reactor fuel sub-assemblies

International Nuclear Information System (INIS)

Ford, J.; Bishop, J.F.W.

1981-01-01

An improved fuel sub-assembly for liquid metal cooled fast breeder nuclear reactors is described which facilitates dismantling operations for reprocessing purposes. The method of dismantling is described. (U.K.)

CURRENT-VOLTAGE CURVES FOR TREATING EFFLUENT CONTAINING HEDP: DETERMINATION OF THE LIMITING CURRENT

Directory of Open Access Journals (Sweden)

T. Scarazzato

2015-12-01

Full Text Available Abstract Membrane separation techniques have been explored for treating industrial effluents to allow water reuse and component recovery. In an electrodialysis system, concentration polarization causes undesirable alterations in the ionic transportation mechanism. The graphic construction of the current voltage curve is proposed for establishing the value of the limiting current density applied to the cell. The aim of this work was to determine the limiting current density in an electrodialysis bench stack, the function of which was the treatment of an electroplating effluent containing HEDP. For this, a system with five compartments was used with a working solution simulating the rinse waters of HEDP-based baths. The results demonstrated correlation between the regions defined by theory and the experimental data.

Diphosphonic Acid (HEDP) Complex As A, Bone Pain Palliative Agent

International Nuclear Information System (INIS)

H G, Adang; Mutalib, A; Bagiawati, Sri; S, Evi; Aguawarini, Sri; Abidin

2003-01-01

Bone pain is a common complication for patient with bone metastases from prostate, breasts, lung and renal cancers. The systemic treatment of metastatic bone cancers can be done by using analgesic drug therapy, hormonal therapy, chemotherapy, narcotic (morphine) and radiopharmaceuticals. Samarium-153 EDTMP is one of the most widely used radiopharmaceutical for the treatment of metallics bone pain. Preparation and quality control of 186 Re-HEDP have been carried out. Radiochemical purity was analysed using paper chromatography and resulted in maximum yields more than 90 % . Complexes quite were stable for 3 days when stored at 4 o C. Rhenium-186 HEDP complex contents in the blood reach optimum activity after 5 minutes and decrease drastically at 24 hours post injection. The complex showed major renal clearance up to 41 % as perrhenate ion within 24 hours after injection, Biodistribution pattern of the injected complex in mice indicates that the accumulated optimum activity in the bone was obtained between 2 - 24 hours post injection, Sterility and pyrogenicity test indicated that the complex were sterile and pyrogen free

Status of Conceptual Design Progress for ITER Sector Sub-assembly Tools

Energy Technology Data Exchange (ETDEWEB)

Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Jae Hyuk; Kim, Kyung Kyu [SFA Engineering Corp., Changwon (Korea, Republic of); Im, Ki Hak; Robert, Shaw [ITER Organization, Paul lez Durance (France)

2010-05-15

The ITER (International Thermonuclear Experimental Reactor) Tokamak assembly tools are purpose-built tools to complete the ITER Tokamak machine which includes the cryostat and the components contained therein. Based on the design description document prepared by the ITER organization, Korea has carried out the conceptual design of assembly tools. The 40 .deg. sector assemblies sub-assembled at assembly hall are transferred to Tokamak hall using the lifting tool operated by Tokamak main cranes. In-pit assembly tools are the purpose-built assembly tools for the completion of final sector assembly at Tokamak hall. The 40 .deg. sector sub-assembly tools are composed of the upending tool, the sector sub-assembly tool, the sector lifting tool and the vacuum vessel support and bracing tools. The process of the ITER sector sub-assembly at assembly hall and status of research and development are described in this paper. The ITER Tokamak device is composed of 9 vacuum vessel (VV)/toroidal field coils (TFCs)/vacuum vessel thermal shields (VVTS) 40 .deg. sectors. Each VV/TFCs/VVTS 40 .deg. sector is made up of one 40 .deg. VV, two 20 .deg. TFCs and associated VVTS segments. The 40 .deg. sectors are sub-assembled at assembly hall respectively and then 9 sectors which sub-assembled at assembly hall are finally assembled at Tokamak hall. As a basic assembly component, the assembly strategy and tools for the 40 .deg. sector sub-assembly and final assembly at inpit should be developed to satisfy the basic assembly requirements of the ITER Tokamak device. Accordingly, the purpose-built assembly tools should be designed and manufactured considering assembly plan, available space, safety, easy operation, efficient maintenance, and so on. The 40 .deg. sector assembly tools are classified into 2 groups. One group is the sub-assembly tools including upending tool, lifting tool, sub-assembly tool, VV supports and bracing tools used at assembly hall and the other group is the in

Formation of imploding plasma liners for fundamental HEDP studies and MIF Standoff Driver Concept

Energy Technology Data Exchange (ETDEWEB)

Cassibry, Jason [Univ. of AL in Huntsville; Hatcher, Richard [Univ. of AL in Huntsville; Stanic, Milos [Univ. of AL in Huntsville

2013-08-17

The disciplines of High Energy Density Physics (HEDP) and Inertial Confinement Fusion (ICF) are characterized by hypervelocity implosions and strong shocks. The Plasma Liner Experiment (PLX) is focused on reaching HEDP and/or ICF relevant regimes in excess of 1 Mbar peak pressure by the merging and implosion of discrete plasma jets, as a potentially efficient path towards these extreme conditions in a laboratory. In this work we have presented the first 3D simulations of plasma liner, formation, and implosion by the merging of discrete plasma jets in which ionization, thermal conduction, and radiation are all included in the physics model. The study was conducted by utilizing a smoothed particle hydrodynamics code (SPHC) and was a part of the plasma liner experiment (PLX). The salient physics processes of liner formation and implosion are studied, namely vacuum propagation of plasma jets, merging of the jets (liner forming), implosion (liner collapsing), stagnation (peak pressure), and expansion (rarefaction wave disassembling the target). Radiative transport was found to significantly reduce the temperature of the liner during implosion, thus reducing the thermal leaving more pronounced gradients in the plasma liner during the implosion compared with ideal hydrodynamic simulations. These pronounced gradients lead to a greater sensitivity of initial jet geometry and symmetry on peak pressures obtained. Accounting for ionization and transport, many cases gave higher peak pressures than the ideal hydrodynamic simulations. Scaling laws were developed accordingly, creating a non-dimensional parameter space in which performance of an imploding plasma jet liner can be estimated. It is shown that HEDP regimes could be reached with ~ 5 MJ of liner energy, which would translate to roughly 10 to 20 MJ of stored (capacitor) energy. This is a potentially significant improvement over the currently available means via ICF of achieving HEDP and nuclear fusion relevant parameters.

186Re-HEDP for metastatic bone pain in breast cancer patients

International Nuclear Information System (INIS)

Lam, Marnix G.E.H.; Rijk, Peter P. van; Klerk, John M.H. de

2004-01-01

Two-thirds of patients with metastatic cancer suffer from pain. Pain originating from skeletal metastases is the most common form of cancer-related pain. Bone pain, often exacerbated by pressure or movement, limits the patient's autonomy and social life. Pain palliation with bone-seeking radiopharmaceuticals has proven to be an effective treatment modality in patients with metastatic bone pain. These bone-seeking radiopharmaceuticals are extremely powerful in treating scattered painful bone metastases, for which external beam radiotherapy is impossible because of the large field of irradiation. 186 Re-hydroxyethylidene diphosphonate (HEDP) is a potentially useful radiopharmaceutical for this purpose, having numerous advantageous characteristics. Bone marrow toxicity is limited and reversible, which makes repetitive treatment safe. Studies have shown encouraging clinical results of palliative therapy using 186 Re-HEDP, with an overall response rate of ca. 70% in painful bone metastases. It is effective for fast palliation of painful bone metastases from various tumours and the effect tends to last longer if patients are treated early in the course of their disease. 186 Re-HEDP is at least as effective in breast cancer patients with painful bone metastases as in patients with metastatic prostate cancer. It is to be preferred to radiopharmaceuticals with a long physical half-life in this group of patients, who tend to have more extensive haematological toxicity since they have frequently been pretreated with bone marrow suppressive chemotherapy. This systemic form of radionuclide therapy is simple to administer and complements other treatment options. It has been associated with marked pain reduction, improved mobility in many patients, reduced dependence on analgesics, and improved performance status and quality of life. (orig.)

Development of a 186Re-HEDP formulation and radio pharmacokinetics comparison with 153Sm-EDTMP

International Nuclear Information System (INIS)

Bribiesca C, A.I.

1998-01-01

Because of the growing interest in the use of the beta emitters radiopharmaceuticals applied to therapy in different cancer cases, we developed a formulation of 186 Re-HEDP (hydroxy ethylidene diphosphonate) as a pain palliative in osseous metastases. Besides serving like therapeutic agent, together with the 153 Sm-EDTMP (ethylene diamine tetra methylene phosphonate), which has already been synthesized and proved, labels EHDP could be very useful like a diagnostic agent in the pursuit of the illness. The irradiation conditions for Rhenium-186 were established by ORIGIN 2 codes for TRIGA reactors. A pharmaceutical formulation was developed employing a factorial experimental design obtaining a complex with a radiochemical purity over 90 %. The complexes 186 Re-HEDP 153 Sm-EDTMP were intravenous administered in BALB-C mice sacrifying them in several intervals of time in order to determine the cumulated activity in each organ to perform absorbed dose calculation by MIRD methodology (Medical Internal Radiation Dose). Radio pharmacokinetic data demonstrated that both complexes follow a biexponential kinetic of first order behavior. In the case of the 186 Re-HEDP the value of the α constant was 0.2789 and β 0.0006 with an effective dose of 2.56 (mSv)/MBq , while for the complex 153 Sm-EDTMP the values of α to and β were 0.9012 and and 0.616 respectively and the effective dose was 0.262 (mSv)/MBq. In conclusion, radiopharmaceutical 153 Sm-EDTMP, showed a greater bone uptake and a minor effective dose, for which it is a better radiopharmaceutical, respect to with the formulation of 186 Re-HEDP. (Author)

Development of a {sup 186}Re-HEDP formulation and radio pharmacokinetics comparison with {sup 153}Sm-EDTMP; Desarrollo de una formulacion de {sup 186}Re-HEDP y comparacion radiofarmacocinetica con el {sup 153}Sm-EDTMP

Energy Technology Data Exchange (ETDEWEB)

Bribiesca C, A I

1998-12-01

Because of the growing interest in the use of the beta emitters radiopharmaceuticals applied to therapy in different cancer cases, we developed a formulation of {sup 186} Re-HEDP (hydroxy ethylidene diphosphonate) as a pain palliative in osseous metastases. Besides serving like therapeutic agent, together with the {sup 153} Sm-EDTMP (ethylene diamine tetra methylene phosphonate), which has already been synthesized and proved, labels EHDP could be very useful like a diagnostic agent in the pursuit of the illness. The irradiation conditions for Rhenium-186 were established by ORIGIN 2 codes for TRIGA reactors. A pharmaceutical formulation was developed employing a factorial experimental design obtaining a complex with a radiochemical purity over 90 %. The complexes {sup 186} Re-HEDP {sup 153} Sm-EDTMP were intravenous administered in BALB-C mice sacrifying them in several intervals of time in order to determine the cumulated activity in each organ to perform absorbed dose calculation by MIRD methodology (Medical Internal Radiation Dose). Radio pharmacokinetic data demonstrated that both complexes follow a biexponential kinetic of first order behavior. In the case of the {sup 186} Re-HEDP the value of the {alpha} constant was 0.2789 and {beta} 0.0006 with an effective dose of 2.56 (mSv)/MBq , while for the complex {sup 153} Sm-EDTMP the values of {alpha} to and {beta} were 0.9012 and and 0.616 respectively and the effective dose was 0.262 (mSv)/MBq. In conclusion, radiopharmaceutical {sup 153} Sm-EDTMP, showed a greater bone uptake and a minor effective dose, for which it is a better radiopharmaceutical, respect to with the formulation of {sup 186} Re-HEDP. (Author)

Thermal-hydraulic of partially blocked fuel subassembly with porous media

International Nuclear Information System (INIS)

Nagata, Takemitsu; Ohshima, Hiroyuki

2000-10-01

The analysis code for investigations of local subassembly phenomena, which has been recognized as an issue of local subassembly accidents, has been required and developed at JNC. It is desirable for the analysis code to be applicable to various blockage conditions and random position of the blockage formation and to evaluate conservatively on the safety assessment with high accuracy. In this study, for the purpose of verifying the application and issues of the subchannel analysis code ASFRE-IV which evaluates thermal hydraulic phenomena in the porous blockage regions, the ASFRE-IV validation analysis was carried out on the basis of the data of an experiment investigation on a local porous blockage in a fuel subassembly performed by Reactor Engineering Groop, O-arai Engineering Center, JNC. Calculational results indicated that ASFRE-IV could reproduce the coolant temperature profile in a fuel subassembly and the peak temperature in the local subchannel conservatively. (author)

Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

Energy Technology Data Exchange (ETDEWEB)

Dunn, F.

1992-09-01

This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

Energy Technology Data Exchange (ETDEWEB)

Dunn, F.

1992-01-01

This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

STRUCTURE FOR SUB-ASSEMBLIES OF ELECTRONIC EQUIPMENT

Science.gov (United States)

Bell, P.R.; Harris, C.C.

1959-03-31

Sub-assemblies for electronic systems, particularly a unit which is self- contained and which may be adapted for quick application to and detachment from a chassis or panel, are discussed. The disclosed structure serves the dual purpose of a cover or enclosure for a subassembly comprising a base plate and also acts as a clamp for retaining the base plate in position on a chassis. The clamping action is provided by flexible fingers projecting from the side walls of the cover and extending through grooves in the base plate to engage with the opposite side of the chassis.

Coincidence measurements of FFTF breeder fuel subassemblies

International Nuclear Information System (INIS)

Eccleston, G.W.; Foley, J.E.; Krick, M.; Menlove, H.O.; Goris, P.; Ramalho, A.

1984-04-01

A prototype coincidence counter developed to assay fast breeder reactor fuel was used to measure four fast-flux test facility subassemblies at the Hanford Engineering Development Laboratory in Richland, Washington. Plutonium contents in the four subassemblies ranged between 7.4 and 9.7 kg with corresponding 240 Pu-effective contents between 0.9 and 1.2 kg. Large count rates were observed from the measurements, and plots of the data showed significant multiplication in the fuel. The measured data were corrected for deadtime and multiplication effects using established formulas. These corrections require accurate knowledge of the plutonium isotopics and 241 Am content in the fuel. Multiplication-corrected coincidence count rates agreed with the expected count rates based on spontaneous fission-neutron emission rates. These measurements indicate that breeder fuel subassemblies with 240 Pu-effective contents up to 1.2 kg can be nondestructively assayed using the shift-register electronics with the prototype counters. Measurements using the standard Los Alamos National Laboratory shift-register coincidence electronics unit can produce an assay value accurate to +-1% in 1000 s. The uncertainty results from counting statistics and deadtime-correction errors. 3 references, 8 figures, 8 tables

Systemic application of rhenium-186 hydroxyethylidenediphosphonate ({sup 186}Re HEDP) as an option for the treatment of chronic arthritis and arthropathy[Radiosynoviorthesis]; Systemische Applikation von Rhenium-186 Hydroxyethylidendiphosphonat ({sup 186}Re HEDP) als Therapieoption bei chronischen Arthritiden und Arthropathien

Energy Technology Data Exchange (ETDEWEB)

Bucerius, J.; Biersack, H.J.; Palmedo, H. [Klinik und Poliklinik fuer Nuklearmedizin, Universitaetsklinikum Bonn (Germany); Wallny, T. [Orthopaedische Klinik l, Klinik fuer Orthopaedische Chirurgie, St. Bernhard-Hospital Kamp-Lintfort (Germany); Klinik und Poliklinik fuer Orthopaedie, Universitaetsklinikum Bonn (Germany); Brackmann, H.H. [Inst. fuer experimentelle Haematologie und Transfusionsmedizin, Universitaetsklinikum Bonn (Germany)

2006-03-15

Chronic arthritis is very common and is associated with a variety of systemic diseases whereas hemophilic arthropathy is one of the most common clinical manifestations of hemophilia, mainly of hemophilia type A. All of these polyarticular diseases are associated with progressive pain and increasing lack of mobility. Therapy is based on conservative treatment such as medication with non-steroidal anti-inflammatory drugs, selective therapy strategies such as intraarticular injections of e.g. radioactive substances (radiosynoviorthesis) or surgical interventions. However, in some cases, the disease does not respond to one of these treatment options or cannot be continued due to important side-effects. Systemic application of radioisotopes like {sup 186}Re HEDP has been successfully administered for pain palliation of osseous metastases. Today, only few data exist for systemic therapy with {sup 186}Re HEDP in patients suffering from benign polyarticular disease. In a prospective study with patients suffering from chronic arthritis a single systemic application of {sup 186}Re HEDP led to a reduction of disease activity in six of eight and to a reduction of the number of painful or swollen joints in five of eight included patients. In a further prospective study with 12 patients with hemophilic arthropathy, 19 of 36 (52.7%) most painful joints could be successfully treated with one systemic {sup 186}Re HEDP therapy. Furthermore, a reduction of global pain could be observed in those patients. However, further randomized studies with larger study populations are necessary in order to confirm this promising results. (orig.)

Reliability of wind turbine subassemblies

NARCIS (Netherlands)

Spinato, F.; Tavner, P.J.; Bussel, van G.J.W.; Koutoulakos, E.

2009-01-01

We have investigated the reliability of more than 6000 modern onshore wind turbines and their subassemblies in Denmark and Germany over 11 years and particularly changes in reliability of generators, gearboxes and converters in a subset of 650 turbines in Schleswig Holstein, Germany. We first start

Fabrication Process for Machined and Shrink-Fitted Impactor-Type Liners for the LOS Alamos Hedp Program

Science.gov (United States)

Randolph, B.

2004-11-01

Composite liners have been fabricated for the Los Alamos liner-driven High Energy Density Physics (HEDP) experiments using impactors formed by physical vapor deposition, and by machining and shrink fitting. Chemical vapor deposition has been proposed for some ATLAS liner applications. This paper describes the processes used to fabricate machined and shrink-fitted impactors; these processes have been used for copper impactors in 1100 aluminum liners and for 6061 T-6 aluminum impactors in 1100 aluminum liners. The most successful processes have been largely empirically developed and rely upon a combination of shrink-fitting and light press fitting. The processes used to date will be described along with some considerations for future composite liners for the HEDP Program.

Bone marrow adsorbed dose of rhenium-186-HEDP and the relationship with decreased platelet counts

International Nuclear Information System (INIS)

Klerk, J.M.H. de; Dieren, E.B. van; Schip, A.D. van het

1996-01-01

Rhenium-186(Sn)-1,1-hydroxyethylidene diphosphonate ( 186 Re-HEDP) has been used for palliation of metastatic bone pain. The purpose of this study was to find a relationship between the bone marrow absorbed dose and the toxicity, expressed as the percentage decrease in the peripheral blood platelet count. The bone marrow absorbed dose was calculated according to the MIRD model using data obtained from ten treatments of patients suffering from metastatic prostate cancer; noninvasive and pharmacokinetic method were used. The bone marrow doses were related to toxicity using the pharmacodynamic sigmoid E max model. The mean bone marrow absorbed doses using the noninvasive and pharmacokinetic methods were in a close range to each other (1.07 mGy/MBq and 1.02 mGy/MBq, respectively). There was a good relationship between the toxicity and the bone marrow absorbed dose (r = 0.80). Furthermore, the EDrm 50 (i.e., the bone marrow absorbed dose producing a 50% platelet decrease) to bone marrow for 186 Re-HEDP was on the order of 2 Gy. Although the function of normal bone marrow is affected by metastases in patients with metastatic bone disease, the MIRD model can be used to relate toxicity to the bone marrow absorbed dose after a therapeutic dosage of 186 Re-HEDP. 33 refs., 1 fig., 1 tab

Nondestructive assay of subassemblies of various spent or fresh fuels by active neutron interrogation

International Nuclear Information System (INIS)

Ragan, G.L.; Ricker, C.W.; Chiles, M.M.; Ingersoll, D.T.; Slaughter, G.G.

1979-01-01

Recent studies show that subassemblies containing various spent fuels could be assayed rapidly and accurately by a nondestructive assay system using active neutron interrogation and prompt-neutron detection. Subassembly penetration is achieved by 24-keV (Sb--Be) interrogation neutrons; the spent-fuel neutron background is overridden by using strong interrogating sources and prompt-neutron signals, and background gammas are absorbed by lead. Experiments have demonstrated the potential for assaying with better than 5% accuracy, three spent plutonium-fueled subassemblies per hour. Calculations, validated by experiments, predict even better performance for fresh or uranium-fueled subassemblies; several performance estimates are given

Nuclear fuel sub-assemblies

International Nuclear Information System (INIS)

Dodd, J.A.; Butterfield, C.E.; Waite, E.

1979-01-01

A fast reactor fuel sub-assembly has honeycomb grids for laterally supporting the fuel pins. The grids are of two series and are arranged alternately along the bundle. The grids of a first series provide a discrete cell for each pin but the grids of the second series have a peripheral group of cells only. The grids of the second series provide intermediate support of the edge pins to restrain bow. (author)

Studies to single subassembly flow monitoring with a complete 7 element array under sodium

International Nuclear Information System (INIS)

Hess, B.; Ruppert, E.; Stehle, H.; Vinzens, K.

1975-01-01

A core restraint system in a fast reactor serves to limit fuel element movement leading to reactivity changes and misalignment of control rod drives and instrumentation. To guarantee proper control rod function the upper ring of the passive restraint system for the SNR-300 should keep subassembly displacement below 20 mm, whereas a free bowing up to 25 mm does not impair subassembly handling. With respect to single subassembly instrumentation the influences of subassembly displacement on temperature and flow monitoring were not exactly known. As part of the SNR-300 R and D programme a complete clamped array, consisting of 4 full size fuel elements and 3 blanket elements was tested for more than 4000 hours at 600 0 C in the AKB sodium loop at Interatom, Bensberg. The test was split into two phases and the total cluster was prestrained in the second phase to simulate 15 mm subassembly displacement at the level of the upper pads. Although this test was mainly considered as an endurance test to demonstrate the integrity of prestrained core elements, effort were made to study the feasibility of single subassembly flow monitoring with this full size model of a core section. (Auth.)

Gas centrifuge power supplies (inverters): Key components and subassemblies

International Nuclear Information System (INIS)

1987-08-01

This document was prepared to serve as a guide for export control officials in their interpretation, understanding, and implementation of exports laws that relate to the international trigger list entry for gas centrifuge power supplies (also known as frequency changers, convertors, or inverters) and parts, components, and subassemblies of such power supplies. Particular emphasis is placed on descriptions of the key parts, components, and subassemblies of such power supplies, which were previously unspecified, so as to clarify the intent of the international trigger list entry

Fuel and fuel pin behaviour in a high burnup fast breeder fuel subassembly: Results of destructive post-irradiation examinations of the KNK II/1 fuel subassembly NY-205

International Nuclear Information System (INIS)

Patzer, G.

1991-05-01

The report gives a summarizing overview of the design characteristics, of the irradiation history and of the results of the destructive post-irradiation examinations of the fuel pins of the high-burnup fuel subassembly NY-205 of the KNK II first core. This element was operated for about 10 years and reached a maximum local burnup of 175 MWd/kg(HM) and a maximum neutron dose of 67 dpa-NRT. The main design data of this subassembly agree with those of the SNR 300 Mark-Ia, and it reached more than twice of the burnup and a similar neutron dose as foreseen for the SNR 300 fuel subassemblies [de

Systemic application of rhenium-186 hydroxyethylidenediphosphonate (186Re HEDP) as an option for the treatment of chronic arthritis and arthropathy

International Nuclear Information System (INIS)

Bucerius, J.; Biersack, H.J.; Palmedo, H.; Wallny, T.; Brackmann, H.H.

2006-01-01

Chronic arthritis is very common and is associated with a variety of systemic diseases whereas hemophilic arthropathy is one of the most common clinical manifestations of hemophilia, mainly of hemophilia type A. All of these polyarticular diseases are associated with progressive pain and increasing lack of mobility. Therapy is based on conservative treatment such as medication with non-steroidal anti-inflammatory drugs, selective therapy strategies such as intraarticular injections of e.g. radioactive substances (radiosynoviorthesis) or surgical interventions. However, in some cases, the disease does not respond to one of these treatment options or cannot be continued due to important side-effects. Systemic application of radioisotopes like 186 Re HEDP has been successfully administered for pain palliation of osseous metastases. Today, only few data exist for systemic therapy with 186 Re HEDP in patients suffering from benign polyarticular disease. In a prospective study with patients suffering from chronic arthritis a single systemic application of 186 Re HEDP led to a reduction of disease activity in six of eight and to a reduction of the number of painful or swollen joints in five of eight included patients. In a further prospective study with 12 patients with hemophilic arthropathy, 19 of 36 (52.7%) most painful joints could be successfully treated with one systemic 186 Re HEDP therapy. Furthermore, a reduction of global pain could be observed in those patients. However, further randomized studies with larger study populations are necessary in order to confirm this promising results. (orig.)

Nuclear reactor fuel element sub-assemblies

International Nuclear Information System (INIS)

Hill, G.D.; Trevalion, P.A.

1977-01-01

A fuel element sub-assembly for a liquid metal cooled fast reactor is described. It comprises a bundle of fuel pins enclosed by a tubular wrapper having a lower end journal for plugging into an upper aperture in a core supporting structure and a spike bar with an articulated bush for engaging a lower aperture in the core supporting structure. The articulated bush is retained on a spherical end portion of the spike bar by a pair of parallel retaining pins arranged transversely and disposed one each side of the spike bar. The pins are tubular and collapsible at a predetermined loading to enable the spherical end portion to pass between them. The articulated bush has an internal groove for engagement by a lifting grab, this groove being formed in a bore for receiving the spherical end portion of the spike bar. The construction lessens liability to rattling of the fuel element sub-assemblies and aids removal for replacement. (U.K.)

Response of subassembly model with internals

International Nuclear Information System (INIS)

Kennedy, J.M.; Belytschko, T.

1977-01-01

For the purpose of predicting the structural response in such accident environments, a program STRAW has been developed. This is a finite element program which can treat the structure-fluid system consisting of the coolant and the subassembly walls. Both material nonlinearities due to elastic-plastic response and geometric nonlinearities due to large displacements can be treated. The energy source can be represented either by a pressure-time history or an equation of state. Because of the lack of any simplifying symmetry in the geometry of the subassembly the program uses a quasi-three dimensional model. The cross section of the accident hexcan and the adjacent hexcan are modelled by a two-dimensional finite element mesh which represents the hexcan walls by flexural element and the internals by two-dimensional continuum elements. This mesh is coupled to a series of one-dimensional elements which represent the axial flow of the coolant and the longitudinal stiffness of the fuel pins and hexcan. The latter is of importance in the adjacent hexcan, for its lateral displacement is resisted entirely by this flexural behavior and its inertia. The adequacy of such quasi-three dimensional models has been examined by comparing the STRAW results against almost complete three-dimensonal analysis performed with the REXCAT program. In this program, the accident hexcan is represented in a true three-dimensional sense by plate-shell elements, whereas the internals are represented as axisymmetric. These comparisons indicate that the quasi-three-dimensional approach employed in STRAW is valid for a large range of pressure time histories; the fidelity of this model suffers primarily when pressure reaches a peak over a very short time, such as 5-10 microseconds

Dynamic response of single hexagonal LMFBR core subassembly wrappers

Energy Technology Data Exchange (ETDEWEB)

Ash, J. E.; Marciniak, T. J.; (Argonne National Lab., IL (United States))

1977-07-01

To analyze the dynamic structural response of the LMFBR core subassembly hexagonal wrappers to postulated local energy releases and the sensitivity of the response to variations in both the pressure loading and the material properties of the stainless steel, a finite-element computer code STRAW has been developed. A series of experiments was performed to study the effects of variations in material properties. The amount of coldworking to which the Type 316 stainless steel is subjected has a strong influence upon the ductility and the elastic yield point. The usual fabrication process produced a nominally 20% coldworking with a yield point of about 680 MPa. By designing a special set of dies for the drawing process, a very low ductility hexcan was produced for which the yield point was raised to 820 MPa. Conversely, the yield point was lowered to 170 MPa by a solution annealing process producing a highly ductile test hexcan. A metallurgical study was conducted to find a representative brittle simulant material for the irradiated end-of-life steel properties. An aging treatment for Type 446 stainless steel was developed which reproduced the expected tensile-flow behavior of the in-pile subassembly. Further study is underway to investigate the fracture properties of the simulant material. The pressure pulses were generated by the controlled expansion of high-pressure detonation poducts from low-density explosives detonated inside a vented steel cannister. The orifice configuration of the cannister and the charge mixture ratio were designed to produce two specified pulse shapes. A charge containing 37,7 g PETN mixed with 35 wt % inert, hollow-glass microballoons developed a pressure pulse peak of 9.5 MPa at 1.0 ms. Increasing the PETN to 41 g resulted in a 14.6 MPa peak pressure, and increasing the explosive concentration to 90 wt % in the mixture increased the burning rate and the pulse risetime, so that the peak occurred at 0.6 ms.

RF-Based Accelerators for HEDP Research

CERN Document Server

Staples, John W; Keller, Roderich; Ostroumov, Peter; Sessler, Andrew M

2005-01-01

Accelerator-driven High-Energy Density Physics experiments require typically 1 nanosecond, 1 microcoulomb pulses of mass 20 ions accelerated to several MeV to produce eV-level excitations in thin targets, the "warm dense matter" regime. Traditionally the province of induction linacs, RF-based acceleration may be a viable alternative with recent breakthroughs in accelerating structures and high-field superconducting solenoids. A reference design for an RF-based accelerator for HEDP research is presented using 15 T solenoids and multiple-gap RF structures configured with either multiple parallel beams (combined at the target) or a single beam and a small stacking ring that accumulates 1 microcoulomb of charge. In either case, the beam is ballistically compressed with an induction linac core providing the necessary energy sweep and injected into a plasma-neutralized drift compression channel resulting in a 1 mm radius beam spot 1 nanosecond long at a thin foil or low-density target.

Predictive implications of bone turnover markers after palliative treatment with 186Re-HEDP in hormone-refractory prostate cancer patients with painful osseous metastases

International Nuclear Information System (INIS)

Zafeirakis, Athanasios; Papatheodorou, Georgios; Arhontakis, Athanasios; Gouliamos, Athanasios; Vlahos, Lambros; Limouris, Georgios S.

2010-01-01

To prospectively evaluate the predictive value of various bone formation and resorption markers in patients with bone metastases from prostate cancer after palliative treatment with 186 Re-1,1-hydroxyethylidene diphosphonate ( 186 Re-HEDP). Included in the study were 36 men with prostate cancer, suffering from painful osseous metastases and treated with 186 Re-HEDP. None had received any treatment that would have interfered with bone metabolism before 186 Re-HEDP treatment or throughout the follow-up period. For each patient, pretreatment and posttreatment serum levels of osteocalcin (OC), bone alkaline phosphatase (BALP), aminoterminal (PINP) and carboxyterminal (PICP) propeptides of type I collagen, amino-terminal (NTx) and carboxyterminal (CTx) telopeptides of type I collagen and their combinations were compared with the level and duration of pain response to radionuclide treatment. Pain response was correlated only with pretreatment ΝΤx/PINP, PICP/PINP and NTx/CTx ratios and posttreatment decrease in baseline NTx and PICP values (p=0.0025-0.035). According to multivariate and ROC analyses, the best marker-derived predictors of better and longer duration of response to 186 Re-HEDP treatment were a posttreatment decrease in NTx of ≥20% (RR=3.44, p=0.0005) and a pretreatment NTx/PINP ratio of ≥1.2 (RR=3.04, p=0.036) NTx, a potent collagenous marker of bone resorption, along with the novel NTx/PINP ratio provide useful cut-off values for identifying a group of patients suffering from painful osseous metastases from hormone-refractory prostatic carcinoma who do not respond to palliative treatment with 186 Re-HEDP. This information could help avoid an inefficient and expensive radionuclide treatment. Also, in the cohort of patients who will eventually undergo such treatment, the medium-term posttreatment changes in NTx offer valuable predictive information regarding long-term palliative response. (orig.)

Time constants and feedback transfer functions of EBR-II subassembly types

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1986-01-01

Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel

Time constants and feedback transfer functions of EBR-II subassembly types

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1987-01-01

Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel. (author)

Fuel or irradiation subassembly

International Nuclear Information System (INIS)

Seim, O.S.; Hutter, E.

1975-01-01

A subassembly for use in a nuclear reactor is described which incorporates a loose bundle of fuel or irradiation pins enclosed within an inner tube which in turn is enclosed within an outer coolant tube and includes a locking comb consisting of a head extending through one side of the inner sleeve and a plurality of teeth which extend through the other side of the inner sleeve while engaging annular undercut portions in the bottom portion of the fuel or irradiation pins to prevent movement of the pins

Proof tests of irradiated and unirradiated EBR-II subassembly ducts

International Nuclear Information System (INIS)

Ruther, W.E.; Chopra, P.S.; Lambert, J.D.B.

1977-01-01

A series of dynamic pressure tests have been conducted within EBR-II subassembly ducts. The tests were designed to simulate bursting of a driver-fuel element in a cluster of such elements at their burnup limit during off-normal conditions in EBR-II. The major objective of the tests was to assure that such failure, which might cause rapid release of stored fission gas, would not deform or otherwise damage subassembly ducts in a way that would hinder movement of a control rod. The test results are described

Possible new basis for fast reactor subassembly instrumentation

Energy Technology Data Exchange (ETDEWEB)

Edwards, A G

1977-03-01

This is a digest of a paper presented to the Risley Engineering Society. The theme is a speculation that the core instrumentation problem for a liquid metal fast breeder reactor might be transformed by developments in the realm of infrared television and in pattern recognition by computer. There is a possible need to measure coolant flow and cooled exit temperature for each subassembly, with familiar fail-to-safety characteristics. Present methods use electrical devices, for example thermocouples, but this gives rise to cabling problems. It might be possible, however, to instal at the top of each subassembly a mechanical device that gives a direct indication of temperature and flow visible to an infrared television camera. Signal transmission by cable would then be replaced by direct observation. A possible arrangement for such a system is described and is shown in schematic form. It includes pattern recognition by computer. It may also be possible to infer coolant temperature directly from the characteristics of the infrared radiation emitted by a thin stainless steel sheet in contact with the sodium, and an arrangement for this is shown. The type of pattern produced for on-line interpretation by computer is also shown. It is thought that this new approach to the problem of subassembly instrumentation is sufficiently attractive to justify a close study of the problems involved.

Fuel sub-assembly

International Nuclear Information System (INIS)

Jolly, R.

1988-01-01

A nuclear fuel sub-assembly includes a hexagonal bundle of parallel, spaced apart fuel pins coupled at one end to an end-holding grid comprising a number of transverse spaced apart rails to each of which is connected a series of pin-receiving cells which render the pins axially captive with the rails. The series of cells are defined by a pair of metal strips each of which has a series of pocket formations such that when the pocket formations are in registry they define cylindrical shaped cells provided with internal projections which engage annular recesses in the end caps of the fuel pins to effect axial constraint of the pins. (author)

Predictive implications of bone turnover markers after palliative treatment with {sup 186}Re-HEDP in hormone-refractory prostate cancer patients with painful osseous metastases

Energy Technology Data Exchange (ETDEWEB)

Zafeirakis, Athanasios [401 Army Hospital of Athens, Department of Nuclear Medicine, Athens (Greece); Papatheodorou, Georgios [401 Army Hospital of Athens, Clinical Research Unit, Athens (Greece); Arhontakis, Athanasios [401 Army Hospital of Athens, Department of Urology, Athens (Greece); Gouliamos, Athanasios; Vlahos, Lambros [Aretaieion University Hospital, Athens Medical School, Department of Radiology, Athens (Greece); Limouris, Georgios S. [Aretaieion University Hospital, Athens Medical School, Department of Nuclear Medicine, Athens (Greece)

2010-01-15

To prospectively evaluate the predictive value of various bone formation and resorption markers in patients with bone metastases from prostate cancer after palliative treatment with {sup 186}Re-1,1-hydroxyethylidene diphosphonate ({sup 186}Re-HEDP). Included in the study were 36 men with prostate cancer, suffering from painful osseous metastases and treated with {sup 186}Re-HEDP. None had received any treatment that would have interfered with bone metabolism before {sup 186}Re-HEDP treatment or throughout the follow-up period. For each patient, pretreatment and posttreatment serum levels of osteocalcin (OC), bone alkaline phosphatase (BALP), aminoterminal (PINP) and carboxyterminal (PICP) propeptides of type I collagen, amino-terminal (NTx) and carboxyterminal (CTx) telopeptides of type I collagen and their combinations were compared with the level and duration of pain response to radionuclide treatment. Pain response was correlated only with pretreatment {nu}{tau}x/PINP, PICP/PINP and NTx/CTx ratios and posttreatment decrease in baseline NTx and PICP values (p=0.0025-0.035). According to multivariate and ROC analyses, the best marker-derived predictors of better and longer duration of response to {sup 186}Re-HEDP treatment were a posttreatment decrease in NTx of {>=}20% (RR=3.44, p=0.0005) and a pretreatment NTx/PINP ratio of {>=}1.2 (RR=3.04, p=0.036) NTx, a potent collagenous marker of bone resorption, along with the novel NTx/PINP ratio provide useful cut-off values for identifying a group of patients suffering from painful osseous metastases from hormone-refractory prostatic carcinoma who do not respond to palliative treatment with {sup 186}Re-HEDP. This information could help avoid an inefficient and expensive radionuclide treatment. Also, in the cohort of patients who will eventually undergo such treatment, the medium-term posttreatment changes in NTx offer valuable predictive information regarding long-term palliative response. (orig.)

UKAEA fast reactor project research and development programme on fuel element cladding and sub-assembly wrapper materials

International Nuclear Information System (INIS)

Harries, D.R.

1977-01-01

Research and development work on fuel element component (cladding, subassembly wrappers, etc.) materials for the U.K. sodium cooled fast reactor programme has been conducted at the United Kingdom Atomic Energy Authority (UKAEA) establishments at Dounreay, Harwell, Risley, and Springfields during the past fifteen years or so. This work has formed an integral part of, and has been co-ordinated by, the UKAEA Fast Reactor Project and has involved close liaison with the Nuclear Power Company (NPC) and the Central Electricity Generating Board (CEGB). The research and development were initially related to the Prototype Fast Reactor (PFR) but the scope has now been extended to cover the first Civil Fast Reactor (CFR1), which has recently been re-designated the Civil Demonstration Fast Reactor (CDFR). The paper outlines the present status of the development of sodium cooled fast reactors in the U.K. and proceeds to summarize the principal PFR and CDFR core and fuel element parameters which have determined the planning and direction of the fuel element materials programme. The current position on the fuel element cladding and wrapper research and development programme is reviewed, and the facilities and future irradiation programme to be carried out in PFR are described

Desain Sistem Pendeteksi untuk Citra Base Sub-assembly dengan Algoritma Backpropagation

Directory of Open Access Journals (Sweden)

Kasdianto Kasdianto

2017-04-01

Full Text Available Object identification technique using machine vision has been implemented in industrial of electronic manufacturers for years. This technique is commonly used for reject detection (for disqualified product based on existing standard or defect detection. This research aims to build a reject detector of sub-assembly condition which is differed by two conditions that are missing screw and wrong position screw using neural network backpropagation. The image taken using camera will be converted into grayscale before it is processed in backpropagation methods to generate a weight value. The experiment result shows that the network architecture with two layers has the most excellent accuracy level. Using learning rate of 0.5, target error 0.015%, and the number of node 1 of 100 and node 2 of 50, the successive rate for sub-assembly detection in right condition reached 99.02% while no error occurs in detecting the wrong condition of Sub-assembly (missing screw and wrong position screw.

System for nondestructive assay of spent fuel subassemblies: comparison of calculations and measurements

International Nuclear Information System (INIS)

Ragan, G.L; Ricker, C.W.; Chiles, M.M.; Ingersoll, D.T.; Slaughter, G.G.; Williams, L.R.

1979-01-01

A nondestructive assay system was developed for determining the total fissile content of spent fuel subassemblies at the head end of a reprocessing plant. The system can perform an assay in 20 min with an uncertainty of <5%. Antimony-beryllium neutrons interrogate the subassemblies, and proton recoil counters detect the resulting fission neutrons. Pulse-height discrimination differentiates between the low-energy interrogation neutrons and the higher-energy fission neutrons. Calculated and measured results were compared for (1) interrogation-neutron penetrability, (2) fission-neutron detectability, (3) radial variation of assay sensitivity, (4) axial variation of assay sensitivity, and (5) the variation of detector count rate as a function of the number of fuel rods in a special 61-rod, LMFBR-type subassembly

High 240Pu FTR/EMC experiments and analysis: Carbide fuel and UO2 blanket subassembly worths

International Nuclear Information System (INIS)

Ombrellaro, P.A.

1977-06-01

Carbide-plutonium fuel and UO 2 blanket subassembly worth measurements performed at ANL in the EMC/LWR were analyzed. Composition exchange worth calculations were performed for: (a) the replacement of high- 240 Pu fuel composition for low- 240 Pu fuel composition and carbide-plutonium fuel composition, successively, in the center subassembly of the core; (b) the replacement of low- 240 Pu fuel composition for carbide--plutonium fuel composition in one outer driver subassembly; and (c) the replacement of the radial reflector composition with UO 2 blanket composition in one subassembly of the radial reflector. The composition exchange worth calculations were performed in two-dimensional x,y geometry, using diffusion theory and perturbation theory. Each method produces about the same calculated-to-experimental bias factors

Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies

International Nuclear Information System (INIS)

Sorokin, G.A.; Avdeev, E.F.; Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P.

2000-01-01

An investigation into the reactor core accident cooling, which are associated with the power grow up or switch off circulation pumps in the event of the protective equipment comes into action, results in the problem of liquid metal boiling heat transfer. Considerable study has been given over the last 30 years to alkaline metal boiling including researches of heat transfer, boiling patterns, hydraulic resistance, crisis of heat transfer, initial heating up, boiling onset and instability of boiling. The results of these investigations have shown that the process of liquid metal boiling has substantial features in comparison with water boiling. Mathematical modeling of two phase flows in fast reactor fuel subassemblies have been developed intensively. Significant success has been achieved in formulation of two phase flow through the pin bundle and in their numerical realization. Currently a set of codes for thermohydraulic analysis of two phase flows in fast reactor subassembly have been developed with 3D macrotransfer governing equations. These codes are used for analysis of boiling onset and liquid metals boiling in fuel subassemblies during loss-of-coolant accidents, of warming up of reactor core, of blockage of some part of flow cross section in fuel subassembly. (author)

A possible new basis for fast reactor subassembly instrumentation

International Nuclear Information System (INIS)

Edwards, A.G.

1977-01-01

This is a digest of a paper presented to the Risley Engineering Society. The theme is a speculation that the core instrumentation problem for a liquid metal fast breeder reactor might be transformed by developments in the realm of infrared television and in pattern recognition by computer. There is a possible need to measure coolant flow and cooled exit temperature for each subassembly, with familiar fail-to-safety characteristics. Present methods use electrical devices, for example thermocouples, but this gives rise to cabling problems. It might be possible, however, to instal at the top of each subassembly a mechanical device that gives a direct indication of temperature and flow visible to an infrared television camera. Signal transmission by cable would then be replaced by direct observation. A possible arrangement for such a system is described and is shown in schematic form. It includes pattern recognition by computer. It may also be possible to infer coolant temperature directly from the characteristics of the infrared radiation emitted by a thin stainless steel sheet in contact with the sodium, and an arrangement for this is shown. The type of pattern produced for on-line interpretation by computer is also shown. It is thought that this new approach to the problem of subassembly instrumentation is sufficiently attractive to justify a close study of the problems involved. (U.K.)

Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

International Nuclear Information System (INIS)

Kamide, H.; Ieda, Y.; Toda, S.; Isozaki, T.; Sugawara, S.

1993-01-01

Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor core during natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

Energy Technology Data Exchange (ETDEWEB)

Kamide, H; Ieda, Y; Toda, S; Isozaki, T; Sugawara, S [Reactor Engineering Section, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, Narita, O-arai, Ibaraki-ken (Japan)

1993-02-01

Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor coreduring natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

Recommended connections between the main ionizing radiation sensors and their electronic sub-assemblies

International Nuclear Information System (INIS)

Lefevre, Roger; Roquefort, Henri

1970-02-01

The authors report the study of several typical and simple connections which are present between an ionizing radiation detector and the electronic sub-assembly, and can be adequate in most of the cases. They also study recommended outputs of the different types of detectors and their possible connections with electronic functional elements. Thus, they address connections for general use, detector outputs, inputs and outputs of amplifiers and of sub-assemblies, amplifier inputs, commonly used connectors

Some radiation protection problems connected with the use of 186Re-HEDP and 153Sm-EDTMP for palliative therapy of of bone metastases

International Nuclear Information System (INIS)

Husak, V.; Myslivecek, M.

1995-01-01

The aim of this paper was to assess whether the ambulatory (outpatient) therapy with 186 Re-HEDP and 153 Sm-EDTMP is possible in the Czech Republic. Physical characteristics, administered activity, biokinetics of radiopharmaceuticals, radiation protection characteristics, irradiation of patients relatives as well as comparison with limits for rhenium-186 and samarium-153 radiopharmaceuticals are given. The outpatient administration of 186 Re-HEDP and 153 Sm-EDTMP with the subsequent keeping the patient for 6 hours in a department of nuclear medicine appears to be in compliance with regulations proposed in the Czech Republic as well as ICRP Recommendations. (J.K.) 1 tab., 12 refs

Some radiation protection problems connected with the use of 186Re-HEDP and 153Sm-EDTMP for palliative therapy of of bone metastases

Energy Technology Data Exchange (ETDEWEB)

Husak, V; Myslivecek, M [Univ. Hospital, Olomouc (Czech Republic). Department of Nuclear Medicine

1996-12-31

The aim of this paper was to assess whether the ambulatory (outpatient) therapy with {sup 186}Re-HEDP and {sup 153}Sm-EDTMP is possible in the Czech Republic. Physical characteristics, administered activity, biokinetics of radiopharmaceuticals, radiation protection characteristics, irradiation of patients relatives as well as comparison with limits for rhenium-186 and samarium-153 radiopharmaceuticals are given. The outpatient administration of {sup 186}Re-HEDP and {sup 153}Sm-EDTMP with the subsequent keeping the patient for 6 hours in a department of nuclear medicine appears to be in compliance with regulations proposed in the Czech Republic as well as ICRP Recommendations. (J.K.) 1 tab., 12 refs.

Universal Fast Breeder Reactor Subassembly Counter manual

International Nuclear Information System (INIS)

Menlove, H.O.; Eccleston, G.W.; Swansen, J.E.; Goris, P.; Abedin-Zadeh, R.; Ramalho, A.

1984-08-01

A neutron coincidence counter has been designed for the measurement of fast breeder reactor fuel assemblies. This assay system can accommodate the full range of geometries and masses found in fast breeder subassemblies under IAEA safeguards. The system's high-performance capability accommodates high plutonium loadings of up to 16 kg. This manual describes the system and its operation and gives performance and calibration parameters for typical applications

Dynamic structural response of reactor-core subassemblies (hexcans) due to accident overpressurization

International Nuclear Information System (INIS)

Pfeiffer, P.A.; Kulak, R.F.

1993-01-01

This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall. (author)

Dynamic structural response of reactor-core subassemblies (hexcans) due to accident overpressurization

International Nuclear Information System (INIS)

Pfeiffer, P.A.; Kulak, R.F.

1993-01-01

This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall

Calculation of the mechanical equilibrium in a lattice of deformed hexagonal subassemblies

International Nuclear Information System (INIS)

Bernard, A.

1979-01-01

Stainless steel swelling and irradiation creep in the hexagonal wrappers of fast breeder cores induce deformations (mostly bowing), hence mutual interaction (displacements, forces and stresses, which must be calculated). The HARMONIE code was developed to meet these requirements. In this three dimensional code, one minimizes the elastic potential bending energy (quadratic form), with given linear conditions (no overlapping between adjacent subassemblies). The convergence of this function is obtained through a numerical method (parallel gradient). The free bowing of the subassemblies are given as input datas; the output gives the equilibrium displacements and forces while stresses are calculated in a classical manner

Study on mixed convective flow penetration into subassembly from reactor hot plenum in FBRs

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, J.; Ohshima, H.; Kamide, H.; Ieda, Y. [Power Reactor and Nuclear Fuel Development Corporation, Ibaraki (Japan)

1995-09-01

Fundamental experiments using water were carried out in order to reveal the phenomenon of mixed convective flow penetration into subassemblies from a reactor`s upper plenum of fast breeder reactors. This phenomenon appears under a certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system for decay heat removal and might influence the natural circulation head which determines the core flow rate and therefore affects the core coolability. In the experiment, a simplified model which simulates an upper plenum and a subassembly was used and the ultrasonic velocity profile monitor as well as thermocouples were applied for the simultaneous measurement of velocity and temperature distributions in the subassembly. From the measured data, empirical equations related to the penetration flow onset condition and the penetration depth were obtained using relevant parameters which were derived from dimensional analysis.

Universal Fast Breeder Reactor Subassembly Counter manual

Energy Technology Data Exchange (ETDEWEB)

Menlove, H.O.; Eccleston, G.W.; Swansen, J.E.; Goris, P.; Abedin-Zadeh, R.; Ramalho, A.

1984-08-01

A neutron coincidence counter has been designed for the measurement of fast breeder reactor fuel assemblies. This assay system can accommodate the full range of geometries and masses found in fast breeder subassemblies under IAEA safeguards. The system's high-performance capability accommodates high plutonium loadings of up to 16 kg. This manual describes the system and its operation and gives performance and calibration parameters for typical applications.

Fabrication and quality assurance of some important components and sub-assemblies for Prototype Fast Breeder Reactor (PFBR) project

International Nuclear Information System (INIS)

Dutta, N.G.; More, S.S.

2010-01-01

Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500 MWe prototype fast breeder reactor (PFBR) at Kalpakkam, Chennai. In this very important and prestigious national programmed M/s Kay Bouvet Engg. Pvt. Ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies. M/s KBEPL is engaged in manufacturing, quality assurance and supply of many subassemblies of PFBR like under water trolley (UWT), shielding door, container and container storage rack (CSR), vessel in fuel transfer cell (FTC), personnel air lock (PAL), emergency air lock (EAL) and material air lock (MAL), absorber rod drive mechanism (ARDM) flask assembly and carriage in MAL etc. Two partition doors and four nos. of embedded parts (SS 304L) have already been supplied to Bhavini. The paper deals with manufacturing and Q.A. activities being carried out for supply of these important assemblies to PFBR projects. (author)

Time constants and feedback transfer functions of EBR-II [Experimental Breeder Reactor] subassembly types

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1986-09-01

Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel

Assessment of MATRA-LMR-FB with the SHRT-17 Core Subassembly Data

Energy Technology Data Exchange (ETDEWEB)

Chang, Won-Pyo; Yoo, Jin Yoo; Lee, Seung Won; Seong, Seung Hwan; Ahn, Sang June; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Jeong, Taekyeong; Ha, Kwi-Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Since the MATRA-LMRFB code is scheduled to be applied to a partial flow blockage analysis in a PGSFR (Prototype Generation IV Sodium-cooled Fast Reactor) subassembly, code verification is an essential part for the design review. Most of its verification efforts have been devoted to local sub-channel blockages, and thus the predictions were compared to those of other codes as well as experimental data. Verifications using pin bundles with a wire-wrap spacer had to be concentrated to 19-pin bundles, because available experimental data for such a bundle were relatively affluent in world-wide literatures. Therefore, more assessments with diverse pin numbers are necessary for MATRA-LMR-FB to be a more reliable code. Thus far, MATRA-LMR-FB has been applied to a 37-pin subassembly with wire-wrap spacers at most. In this regard, the present comparative study using data produced from the SHRT-17 which was carried out in a 61-pin test subassembly (XX09) placed in the EBR-II (Experimental Breeder Reactor II) core will be a meaningful demonstration for its extensive applicability. The power operation of the EBR-II was begun by ANL (Argonne National Lab.) in 1964 and the SHRT program was carried out in EBR-II between 1984 and 1986 in order to provide not only test data for validation of the computer codes but also demonstration of a passive reactor shutdown and decay heat removal in response of the protected and unprotected transients. The EBR-II SHRT-17 test data were used to demonstrate the prediction capability of MATRA-LMRFB on a radial distribution of the subassembly outlet temperatures during the steady state.

Assessment of MATRA-LMR-FB with the SHRT-17 Core Subassembly Data

International Nuclear Information System (INIS)

Chang, Won-Pyo; Yoo, Jin Yoo; Lee, Seung Won; Seong, Seung Hwan; Ahn, Sang June; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Jeong, Taekyeong; Ha, Kwi-Seok

2015-01-01

Since the MATRA-LMRFB code is scheduled to be applied to a partial flow blockage analysis in a PGSFR (Prototype Generation IV Sodium-cooled Fast Reactor) subassembly, code verification is an essential part for the design review. Most of its verification efforts have been devoted to local sub-channel blockages, and thus the predictions were compared to those of other codes as well as experimental data. Verifications using pin bundles with a wire-wrap spacer had to be concentrated to 19-pin bundles, because available experimental data for such a bundle were relatively affluent in world-wide literatures. Therefore, more assessments with diverse pin numbers are necessary for MATRA-LMR-FB to be a more reliable code. Thus far, MATRA-LMR-FB has been applied to a 37-pin subassembly with wire-wrap spacers at most. In this regard, the present comparative study using data produced from the SHRT-17 which was carried out in a 61-pin test subassembly (XX09) placed in the EBR-II (Experimental Breeder Reactor II) core will be a meaningful demonstration for its extensive applicability. The power operation of the EBR-II was begun by ANL (Argonne National Lab.) in 1964 and the SHRT program was carried out in EBR-II between 1984 and 1986 in order to provide not only test data for validation of the computer codes but also demonstration of a passive reactor shutdown and decay heat removal in response of the protected and unprotected transients. The EBR-II SHRT-17 test data were used to demonstrate the prediction capability of MATRA-LMRFB on a radial distribution of the subassembly outlet temperatures during the steady state

Thermal hydraulic behavior of sub-assembly local blockage in China experiment fast reactor

International Nuclear Information System (INIS)

Yang Zhimin

2000-01-01

The geometrical parameter ratio of pitch to diameter of China Experiment Fast Reactor (CEFR) subassembly is 1,167. To address the thermal hydraulic behavior of subassembly local blockage which may be caused by deformation of cladding due to severe swelling and thermal stresses and by space swirl loosening etc., the porous numerical model and SIMPLE-P code used to solve Navier-Stokes and energy equations in porous medium was developed, and the bundle experiment with 19 pins with 24 subchannels blocked in the sodium coolant was carried on in China Institute of Atomic Energy (CIAE). The comparison of code predictions against experiments (including non-blockage and ten blockage conditions) seems well. The thermal hydraulic behavior of fuel subassembly with 61 fuel pins blockage of CEFR is calculated with SIMPLE-P code. The results indicate that the maximum temperature is 815 deg. C when the blockage area is about 37% (54 central subchannels are blocked). In this case the cladding won't be damaged and no sodium coolant boiling takes place. (author)

Experimental confirmation of the design to minimize vibration and wear in 61-pin wire-spaced EBR-II subassemblies

International Nuclear Information System (INIS)

Fukuda, S.K.

1978-05-01

Examinations of HEDL 61-pin subassemblies comprised of 5.84 mm (0.230) inch diameter mixed-oxide fuel pins with 1.02 mm (0.040'') diameter spacer wire (PNL-9, -10, -11, HEDL-N-E, -N-F), showed severe cladding and spacer wire wear after irradiation in EBR-II. A comparison of a large number of design, fabrication, and irradiation parameters for all of the HEDL subassemblies indicated that the porosity per ring of fuel pins correlated significantly with the occurrence of wear on the fuel pins. The porosity per ring is the clearance between the flat-to-flat pin bundle dimension and the inner hex can dimension divided by the number of hexagonal fuel pin rings in the subassembly. The porosity per ring for PNL-9, -10, -11 and HEDL-N-E was 0.15 mm/ring (6 mils/ring) and 0.18 mm/ring (7 mils/ring) for the HEDL-N-F subassembly. Since the original FTR subassembly design had a porosity/ring spread of 0.04 mm/ring to 0.16 mm/ring (1.67 to 6.11 mils/ring) an additional series of irradiation tests was conducted to confirm that a tighter fuel pin bundle would eliminate the wear

Analysis of effectiveness of the palliative treatment of metastatic bone's pain with 188Re-HEDP

International Nuclear Information System (INIS)

Savio, E.; Zeledon, P.; Paolino, A.; De Marco, E.; Gaudino, J.

2003-01-01

The objective of the study was to evaluate the treatment effectiveness with 188Re-HEDP in a group of 27 patients, who had received 36 doses. A pharmaceutical care programme was also added in order to improve drug follow-up after treatment. Two levels of doses were administered: 30 or 60 mCi. Initially a trace dose was given in order to estimate the therapeutic dose, which was individualise according to bone uptake of the radiopharmaceutical. Bone uptake was determined measuring radioactivity in urine samples (0, 1, 2, 4 and 6 hs), because the radiopharmaceutical showed only renal elimination. Multiple dose schedules with with 3 months between both doses were also tried. Seventy two percent showed an algesic effect during the first week post-treatment, with was kept during one month, while seven tenn (17%) percent of the patients the effect was kept for two of more months. Opioid analgesic (third level of OMS scale) were diminished in eighty two percent of the patients and AINES drugs in seventy one percent. The pharmaceutical care programme also showed the importance of the radio pharmacist role to improve treatment outcomes. 188Re-HEDP effectiveness was achieved in 100% of the patients, but with different pain palliation response in time and/or drug intake, with a suitable radiological safety

Structural response of reactor-core hexcan subassemblies subjected to dynamic overpressurization under accident conditions

International Nuclear Information System (INIS)

Pfeiffer, P.A.; Kulak, R.F.

1993-01-01

This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall

Analytical work on local faults in LMFBR subassembly

International Nuclear Information System (INIS)

Yoshikawa, H.; Miyaguchi, K.; Hirata, N.; Kasahara, F.

1979-01-01

Analytical codes have been developed for evaluating various severe but highly unlikely events of local faults in the LMFBR subassembly (S/A). These include: (1) local flow blockage, (2) two-phase thermohydraulics under fission gas release, and (3) inter-S/A failure propagation. A simple inter-S/A thermal failure propagation analysis code, FUMES, is described that allows an easy parametric study of propagation potential of fuel fog in a S/A. 7 refs

Nuclear sub-assembly for liquid metal fast breeder reactor

International Nuclear Information System (INIS)

1978-01-01

The description is given of a nuclear sub-assembly comprising several spaced out fuel pins in a tubular shroud, the characteristic being that the section of the shroud forms a closed figure with six main straight sides in hexagonal shape, the main sides being joined by subsidiary sides which are either straight or convex towards the centre of the figure [fr

Prediction of the pressure-time history due to fuel-sodium interaction in a subassembly

International Nuclear Information System (INIS)

Jacobs, H.

1975-01-01

A local cooling disturbance may lead to complete voiding of a subassembly and melt down of the fuel pins. Thus molten fuel may be accumulated and mixed with liquid sodium returning accidentally into the subassembly. The resulting fuel-sodium interaction (FSI) produces a pressure load on the surrounding core structures. It is necessary to prove that the corresponding core deformation neither initiates a nuclear excursion nor renders the shut down system inoperable. This requires the knowledge of the initiating FSI pressure time history. In this paper a theoretical pressure time history is presented which differs completely from all calculations known so far. (Auth.)

Identifying subassemblies by ultrasound to prevent fuel handling error in sodium fast reactors: First test performed in water

International Nuclear Information System (INIS)

Paumel, Kevin; Lhuillier, Christian

2015-01-01

Identifying subassemblies by ultrasound is a method that is being considered to prevent handling errors in sodium fast reactors. It is based on the reading of a code (aligned notches) engraved on the subassembly head by an emitting/receiving ultrasonic sensor. This reading is carried out in sodium with high temperature transducers. The resulting one-dimensional C-scan can be likened to a binary code expressing the subassembly type and number. The first test performed in water investigated two parameters: width and depth of the notches. The code remained legible for notches as thin as 1.6 mm wide. The impact of the depth seems minor in the range under investigation. (authors)

Direct current linear measurement sub-assembly data and test methods. Nuclear electronic equipment for control and monitoring panel

International Nuclear Information System (INIS)

1977-12-01

The M.C.H./M.E.N.T.3 document is concerned with sub-assemblies intended for measuring on a linear scale the neutron fluence rate or radiation dose rate when connected with nuclear detectors working in current. The symbols used are described. Some definitions and a bibliography are given. The main characteristics of direct current linear measurement sub-assemblies are then described together with corresponding test methods. This type of instrument indicates on a linear scale the level of a direct current applied to its input. The document reviews linear sub-assemblies for general purpose applications, difference amplifiers for monitoring, and averaging amplifiers. The document is intended for electronics manufacturers, designers, persons participating in acceptance trials and plant operators [fr

The swelling behavior of Ti-stabilized austenitic steels used as structural materials of fissile subassemblies in Phenix

International Nuclear Information System (INIS)

Seran, J.L.; Touron, H.; Maillard, A.; Dubuisson, P.; Hugot, J.P.; Blanchard, P.; Pelletier, M.

1988-06-01

In this paper we analyse the main results obained on pressurized tubes, fissile pins and hexagonal cans, allowing us to characterize the swelling and irradiation creep resistance of Ti-Mod. austenitic steels, used as reference materials for the fast breeder subassembly. After having compared the global behavior of 316Ti and 15-15Ti steels irradiated as fissile pins we examine in more detail the leading variables acting on swelling and irradiation creep resistance of CW 316Ti clads and wrappers. The irradiation creep associated to the principal mechanical stresses (sodium pressure for the wrapper, fission gas pressure for the clad) explain the plastic deformation observed on the wrappers not on the clads. Fissile pins swell more and the scatter of the results is larger than for wrappers or samples. It does not seem possible to invoque flux or primary stress differences to explain this fact. On the opposite the thermal gradient in the thickness of the components appears to be a significant parameter. In fissile pins it gives rise to a swelling gradient observed by electron microscopy that must be taken into account when comparing to the wrapper. As compared to CW 316Ti, CW 15-15Ti is an important improvement since its incubation dose for swelling is far beyond 100 dpa. Further more since it swelling temperature dependence does not seem to be as important as for 316Ti, it should be less sensitive to the effect of thermal gradients

Development of multi-dimensional thermal-hydraulic modeling using mixing factors for wire wrapped fuel pin bundles in fast reactors. Validation through a sodium experiment of 169-pin fuel subassembly

International Nuclear Information System (INIS)

Nishimura, M.; Kamide, H.; Miyake, Y.

1997-04-01

Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature. The prediction of temperature distribution in the subassembly is, therefore one of the important issues for the reactor safety assessment. Mixing factors were applied to multi-dimensional thermal-hydraulic code AQUA to enhance the predictive capability of simulating maximum cladding temperature in the fuel subassemblies. In the previous studies, this analytical method had been validated through the calculations of the sodium experiments using driver subassembly test rig PLANDTL-DHX with 37-pin bundle and blanket subassembly test rig CCTL-CFR with 61-pin bundle. The error of the analyses were comparable to the error of instrumentation's. Thus the modeling was capable of predicting thermal-hydraulic field in the middle scale subassemblies. Before the application to large scale real subassemblies with more than 217 pins, accuracy of the analytical method have to be inspected through calculations of sodium tests in a large scale pin bundle. Therefore, computations were performed on sodium experiments in the relatively large 169-pin subassembly which had heater pins sparsely within the bundle. The analysis succeeded to predict the experimental temperature distributions. The errors of temperature rise from inlet to maximum values were reduced to half magnitudes by using mixing factors, compared to those of analyses without mixing factors. Thus the modeling is capable of predicting the large scale real subassemblies. (author)

Finite element analysis of irradiation-induced dilation of the fuel subassembly duct in LMFBR

International Nuclear Information System (INIS)

Gao Fuhai; Fu Hao; Li Nan; Yang Kongli; Wang Mingzhen

2013-01-01

Background: The calculation of irradiation-induced dilation of the fuel subassembly duct in LMFBR is important for fast reactor core design.. Purpose: To investigate how to calculate the dilation by using finite element method (FEM). Methods: First, irradiation-induced creep and swelling material models are introduced. Then, a theoretical solution based on a simplified bending plate model is briefly given. Finally, a stress update scheme for the adopted material models is presented and furthermore embedded into ABAQUS user interface UMAT to conduct finite element analysis. Both solutions are compared and discussed. Results: FEM successfully predicts the duct dilation and its solution agrees well with theoretical one in small deformation. Conclusions: The proposed stress update scheme is effective, The accuracy of the theory solution declines when dilation becomes larger. The maximum stress occurs at the duct corner point, and the location has stress relaxation effect. (authors)

EBR-II: search for the lost subassembly

International Nuclear Information System (INIS)

King, R.W.; Buschman, H.W.; Poloncsik, J.; Remsburg, J.S.; Sine, H.W.

1983-01-01

Experimental Breeder Reactor II (EBR-II) has been operating for nearly 20 years as part of the foundation of the US Department of Energy's LMFBR development program. During that time, the EBR-II fuel-handling system has performed extremely well, especially considering the conditions under which much of the system operates and the reliability required to maintain the high plant factor routinely demonstrated by EBR-II. Since EBR-II is a pool-type reactor, much of the fuel handling is done remotely within the sodium-filled primary tank at 371 0 C. Activities involved in locating a misplaced fuel subassembly in the primary tank are described

Studies to single subassembly flow monitoring with a complete 7 element array under sodium

International Nuclear Information System (INIS)

Hess, B.; Ruppert, E.; Stehle, H.; Vinzens, K.

1975-01-01

As part of the SNR-300 R and D programme a complete clamped array, consisting of 4 full size fuel elements and 3 blanket elements was tested for more than 4000 hours at 600 deg C in the AKB sodium loop at Interatom, Bensberg. The test was split into two phases and the total cluster was prestrained in the second phase to simulate 15 mm subassembly displacement at the level of the upper bearing pads. Although this test was mainly considered as an endurance test to demonstrate the integrity of prestrained core elements, efforts were made to study the feasibility of single subassembly flow monitoring with this full size model of a core section. The results of these investigations are presented and discussed in this paper

188Rhenium-HEDP in the treatment of pain in bone metastases

International Nuclear Information System (INIS)

Gaudiano, J.; Martinez, G.; Hermida, J.C.; Savio, E.; Verdera, S.; Robles, A.; Muniz, S.; Leon, A.; Knapp, F.F.

2001-01-01

Systemic use of radiopharmaceuticals is a recognised alternative method for the treatment of pain in patients with multiple bone metastases. A new option, 188 Re-HEDP is proposed, using generator-obtained 188 Rhenium (β energy = 2.1 MeV, γ energy = 155 keV, half-life = 16.9 hours). After establishing parameters of biodistribution, dosimetry and image acquisition in mice, rats and rabbits, Phase I and II studies were conducted on 12 patients with multiple metastases from carcinomas, with pain surpassing other analgesic options. More than 50% pain relief was found in 91% of the patients, with total relief during a variable period in 41% of them allowing opiate and other analgesic drugs to be decreased or withdrawn, and showing a lower bone marrow contribution to total absorbed dose than that reported for other similar radiopharmaceuticals. Further study of this option is recommended in order to determine higher dose protocols without toxic bone marrow reaction possibilities. (author)

188Rhenium-HEDP in the Treatment of Pain in Bone Metastases

International Nuclear Information System (INIS)

Gaudiano, J.; Savio, E.; Robles, A.; Muniz, S.; Leon, A.; Verdera, S.; Martinez, G.; Hermida, J.C.; Knapp, F.F. Jr.

1999-01-01

Systemic use of radiopharmaceuticals is a recognized alternative method for the treatment of pain in patients with multiple bone metastasis. A new option, 188 Re-HEDP is proposed, using generator-obtained 188 Rhenium (β energy = 2.1 MeV, γ energy = 155 keV, half-life = 16.9 hours). After establishing parameters of biodistribution, dosimetry and image acquisition in mice, rats and rabbits, Phase I and II studies were conducted on 12 patients with multiple metastasis from carcinomas, with pain surpassing other analgesic options. More than 50% pain relief was found in 91% of the patients, with total relief during a variable period in 41% of them allowing opiate and other analgesic drugs to be decreased or withdrawn, and showing a lower bone marrow contribution to total absorbed dose than that reported for other similar radiopharmaceuticals. Further study of this option is recommended in order to determine higher dose protocols without toxic bone marrow reaction possibilities

Reconstitutable nuclear reactor fuel assembly with unitary removable top nozzle subassembly

International Nuclear Information System (INIS)

Shallenberger, J.M.

1987-01-01

A reconstitutable fuel assembly is described having at least one control rod guide thimble and a top nozzle, the guide thimble including an upper extension, the top nozzle including at least one hold-down spring, an upper hold-down plate and a lower adapter plate, an improved attaching structure removably mounting the top nozzle as a unitary subassembly on the guide thimble. The attaching structure comprises: (a) a coupling member interfitting the lower adapter plate, the upper hold-down plate and the hold-down spring disposed between the plates so as to capture and retain the plates and spring together as a unitary subassembly in which the upper plate is slidably moveable along the coupling member relative to the lower plate with the spring biasing the upper plate away from the lower plate. The coupling member has spaced apart upper and lower portions with a central passageway extending for slidably receiving the upper extension of the guide thimble in a nonattached relationship in which the coupling member is slidably movable relative to the guide thimble extension for respectively inserting and removing the coupling member on and from the guide thimble extension

Fuel cell subassemblies incorporating subgasketed thrifted membranes

Science.gov (United States)

Iverson, Eric J.; Pierpont, Daniel M.; Yandrasits, Michael A.; Hamrock, Steven J.; Obradovich, Stephan J.; Peterson, Donald G.

2016-03-01

A fuel cell roll good subassembly is described that includes a plurality of individual electrolyte membranes. One or more first subgaskets are attached to the individual electrolyte membranes. Each of the first subgaskets has at least one aperture and the first subgaskets are arranged so the center regions of the individual electrolyte membranes are exposed through the apertures of the first subgaskets. A second subgasket comprises a web having a plurality of apertures. The second subgasket web is attached to the one or more first subgaskets so the center regions of the individual electrolyte membranes are exposed through the apertures of the second subgasket web. The second subgasket web may have little or no adhesive on the subgasket surface facing the electrolyte membrane.

Evaluation of copper for divider subassembly in MCO Mark IA and Mark IV scrap fuel baskets

International Nuclear Information System (INIS)

Graves, C.E.

1997-01-01

The K Basin Spent Nuclear Fuel (SNF) Project Multi-Canister Overpack (MCO) subprojection eludes the design and fabrication of a canister that will be used to confine, contain, and maintain fuel in a critically safe array to enable its removal from the K Basins, vacuum drying, transport, staging, hot conditioning, and interim storage (Goldinann 1997). Each MCO consists of a shell, shield plug, fuel baskets (Mark IA or Mark IV), and other incidental equipment. The Mark IA intact and scrap fuel baskets are a safety class item for criticality control and components necessary for criticality control will be constructed from 304L stainless steel. It is proposed that a copper divider subassembly be used in both Mark IA and Mark IV scrap baskets to increase the safety basis margin during cold vacuum drying. The use of copper would increase the heat conducted away from hot areas in the baskets out to the wall of the MCO by both radiative and conductive heat transfer means. Thus copper subassembly will likely be a safety significant component of the scrap fuel baskets. This report examines the structural, cost and corrosion consequences associated with using a copper subassembly in the stainless steel MCO scrap fuel baskets

A phase 2 study of high-activity {sup 186}Re-HEDP with autologous peripheral blood stem cell transplant in progressive hormone-refractory prostate cancer metastatic to bone

Energy Technology Data Exchange (ETDEWEB)

O' Sullivan, J.M. [Queen' s University Belfast/Belfast City Hospital, Department of Oncology, Belfast (United Kingdom); Norman, A.R. [Royal Marsden Foundation NHS Trust, Department of Computing, Sutton, Surrey (United Kingdom); McCready, V.R.; Flux, G.; Buffa, F.M. [Royal Marsden Foundation NHS Trust, Department of Physics, Sutton, Surrey (United Kingdom); Johnson, B. [Royal Marsden Foundation NHS Trust, Bob Champion Unit, Sutton, Surrey (United Kingdom); Coffey, J.; Horwich, A.; Huddart, R.A.; Parker, C.C.; Dearnaley, D.P. [Royal Marsden Foundation NHS Trust, Academic Unit of Urology, Sutton, Surrey (United Kingdom); Cook, G. [Royal Marsden Foundation NHS Trust, Department of Nuclear Medicine, Sutton, Surrey (United Kingdom); Treleaven, J. [Royal Marsden Foundation NHS Trust, Department of Haematology, Sutton, Surrey (United Kingdom)

2006-09-15

We investigated the potential for improvement in disease control by use of autologous peripheral blood stem cell transplant (PBSCT) to permit administration of high activities of {sup 186}Re-hydroxyethylidene diphosphonate (HEDP) in patients with progressive hormone-refractory prostate cancer (HRPC). Eligible patients had progressive HRPC metastatic to bone, good performance status and minimal soft tissue disease. Patients received 5,000 MBq of {sup 186}Re-HEDP i.v., followed 14 days later by PBSCT. Response was assessed using PSA, survival, pain scores and quality of life. Thirty-eight patients with a median age of 67 years (range 50-77) and a median PSA of 57 ng/ml (range 4-3,628) received a median activity of 4,978 MBq {sup 186}Re-HEDP (range 4,770-5,100 MBq). The most serious toxicity was short-lived grade 3 thrombocytopenia in 8 (21%) patients. The median survival of the group is 21 months (95%CI 18-24 months) with Kaplan-Meier estimated 1- and 2-year survival rates of 83% and 40% respectively. Thirty-one patients (81%, 95% CI 66-90%) had stable or reduced PSA levels 3 months post therapy while 11 (29%, 95% CI 15-49%) had PSA reductions of >50% lasting >4 weeks. Quality of life measures were stable or improved in 27 (66%) at 3 months. We have shown that it is feasible and safe to deliver high-activity radioisotope therapy with PBSCT to men with metastatic HRPC. Response rates and survival data are encouraging; however, further research is needed to define optimal role of this treatment approach. (orig.)

{sup 188}Rhenium-HEDP in the Treatment of Pain in Bone Metastases

Energy Technology Data Exchange (ETDEWEB)

Gaudiano, J.; Savio, E.; Robles, A.; Muniz, S.; Leon, A.; Verdera, S.; Martinez, G.; Hermida, J.C.; Knapp, F.F., Jr.

1999-01-18

Systemic use of radiopharmaceuticals is a recognized alternative method for the treatment of pain in patients with multiple bone metastasis. A new option, {sup 188}Re-HEDP is proposed, using generator-obtained {sup 188}Rhenium ({beta} energy = 2.1 MeV, {gamma} energy = 155 keV, half-life = 16.9 hours). After establishing parameters of biodistribution, dosimetry and image acquisition in mice, rats and rabbits, Phase I and II studies were conducted on 12 patients with multiple metastasis from carcinomas, with pain surpassing other analgesic options. More than 50% pain relief was found in 91% of the patients, with total relief during a variable period in 41% of them allowing opiate and other analgesic drugs to be decreased or withdrawn, and showing a lower bone marrow contribution to total absorbed dose than that reported for other similar radiopharmaceuticals. Further study of this option is recommended in order to determine higher dose protocols without toxic bone marrow reaction possibilities.

Development of automatic gamma and neutron monitoring system for PFBR fuel subassemblies at IFSB

International Nuclear Information System (INIS)

Krishnakumar, D.N.; Dhanasekaran, A.; Ajoy, K.C.; Jose, M.T.; Baskaran, R.; Sureshkumar, K.V.

2018-01-01

Health physics surveillance during PFBR fuel pin assembling operation at Interim Fuel Storage Building (IFSB) mandates scanning of the fuel assembly using Telector and Rem counter to find out the maximum gamma and neutron dose rates respectively. Throughout the process health physicist involved in the operation must hold the survey meter at a constant distance from the subassembly and simultaneously should make a note of dose rate values displayed. This practice might lead to the occupational exposures and also might induce human errors during measurements. To make this process more simple, faultless and effortless, an automatic Gamma Neutron Monitoring System (AGNMS) is designed and developed at RSD to measure, store and visualize instantaneous gamma and neutron dose rates of PFBR fuel subassembly. Development of the system, calibration and deployment of the system at IFSB and preliminary results obtained using the system is depicted in this paper

Seismic analysis and design of spent subassembly storage bay (SSSB) pool

International Nuclear Information System (INIS)

Abdul Gani, H.I.; Ramanjaneyulu, K.V.S.; Pillai, C.S.; Chetal, S.C.

2003-01-01

Fuel bundles, after their specified stay in reactor core, are replaced by fresh fuel for sustaining power generation at rated levels. The irradiated fuel subassembly, removed fresh from core, known as spent fuel sub assembly, is radioactive and decay heat generating. It needs to be cooled before it becomes amenable for handling, either for reprocessing or for immobilisation. For this purpose, it is immersed in a pool of water, retained in a concrete structure referred as Spent Subassembly Storage Bay (SSSB) pool. The height of water column above fuel bundles is arrived from shielding considerations. SSSB pool is one of the nuclear safety related structures and warrants rigorous analysis and design. The SSSB pool, in case of PFBR 500 MW(e) is located in fuel building. It is a stainless steel lined. water retaining rectangular R.C.C. open tank of size 7.5 X 29.0 m, with a height of 11.0 m. This structure is analysed for two levels of site specific earthquakes taking in to account liquid structure interactions as per ASCE-4, 1998. The design of walls and bottom slab is carried out satisfying the AERB code for nuclear safety related structures. Analysis and design of SSSB pool of PFBR is presented in the following paper. (author)

Field test and evaluation of the passive neutron coincidence collar for prototype fast reactor fuel subassemblies

International Nuclear Information System (INIS)

Menlove, H.O.; Keddar, A.

1982-08-01

The passive neutron Coincidence Collar, which was developed for the verification of plutonium content in fast reactor fuel subassemblies, has been field tested using Prototype Fast Reactor fuel. For passive applications, the system measures the 240 Pu-effective mass from the spontaneous fission rate, and in addition, a self-interrogation technique is used to determine the fissile content in the subassembly. Both the passive and active modes were evaluated at the Windscale Works in the United Kingdom. The results of the tests gave a standard deviation 0.75% for the passive count and 3 to 7% for the active measurement for a 1000-s counting time. The unit will be used in the future for the verification of plutonium in fresh fuel assemblies

Thermal experiments with LMFBR subassembly models in sodium flow

International Nuclear Information System (INIS)

Moeller, R.; Tschoeke, H.

1982-01-01

Within the framework of the Fast Breeder Project research work has been undertaken at the Karlsruhe Nuclear Research Center on the thermal and fluid dynamics of nominal and distorted core subassemblies. In 19-rod bundle models (P/D=1.30, W/R=1.38) three-dimensional temperature distributions were measured in the cladding tubes exposed to sodium flow. Results of measurements of the azimuthal temperature profiles of rotated rods in the duct wall zone are indicated for different operating conditions 80 2 , evenly distributed load and oblique load; different axial positions of the spacer grids; and different positions of one bowed rod

Absorbed dose distributions in patients with bone metastases from hormone refractory prostate cancer treated with Re-186 HEDP

International Nuclear Information System (INIS)

Denis Bacelar, A.M.; Dearnaley, D.P.; Divoli, A.; Chittenden, S.; Du, Y.; Flux, G.D.; O'Sullivan, J.M.

2015-01-01

Full text of publication follows. Aim: intravenous administration of Re-186 hydroxyethylidene-diphosphonate (HEDP) is used for metastatic bone pain palliation in hormone refractory prostate cancer patients. Dosimetry for bone seeking radionuclides is challenging due to the complex structure with osteoblastic, osteolytic and mixed lesions. The aim of this study was to perform image-based patient-specific 3D convolution dosimetry to obtain a distribution of the absorbed doses to each lesion and estimate inter- and intra-patient variations. Materials and methods: 28 patients received a fixed 5 GBq activity of Re-186 HEDP followed by peripheral blood stem cell rescue at 14 days in a phase II trial. A FORTE dual-headed gamma camera was used to acquire sequential Single-Photon-Emission Computed Tomography (SPECT) data of the thorax and pelvis area at 1, 4, 24, 48 and 72 hours following administration. The projection data were reconstructed using filtered-back projection and were corrected for attenuation and scatter. Voxelised cumulated activity distributions were obtained with two different methods. First, the scans were co-registered and the time-activity curves were obtained on a voxel-by-voxel basis. Second, the clearance curve was obtained from the mean number of counts in each individual lesion and used to scale the uptake distribution taken at 24 hours. The calibration factors required for image quantification were obtained from a phantom experiment. An in-house developed EGSnrc Monte Carlo code was used for the calculation of dose voxel kernels for soft-tissue and cortical/trabecular bone used to perform convolution dosimetry. Cumulative dose-volume histograms were produced and mean absorbed doses calculated for each spinal and pelvic lesion. Results: preliminary results show that the lesion mean absorbed doses ranged from 25 to 55 Gy when the medium was soft tissue and decreased by 40% if bone was considered. The use of the cumulated activity distribution

French studies on local blockages in LMFBR fuel subassemblies

International Nuclear Information System (INIS)

Girard, C.; Jolas, P.; Seiler, J.M.

1979-08-01

This paper reviews experimental and theoretical studies done in FRANCE on the problem of partial subassembly blockages. The priorities are defined and the development of the French program in the European context is presented. Results of the out of pile experiments performed at CEA and EDF in single and two phases flow are given. A description of the main codes used to interpret these experiments is then shortly reviewed. It is found that the thermal behavior in single phase may be calculated with good precision, and that a simple semi-empirical formula can predict with good accuracy the number of channels blocked that lead to sodium boiling

Flow of ideal fluid through a central region of a nuclear reactor wire-spaced fuel subassembly

International Nuclear Information System (INIS)

Schmid, J.

1991-04-01

The results are given of calculations of the flow of an ideal fluid through the central region of a nuclear reactor wire-spaced fuel subassembly. The computer code used is briefly described. (author). 10 figs., 4 refs

Unique capabilities for ICF and HEDP research with the KrF laser

Science.gov (United States)

Obenschain, Stephen; Bates, Jason; Chan, Lop-Yung; Karasik, Max; Kehne, David; Sethian, John; Serlin, Victor; Weaver, James; Oh, Jaechul; Jenkins, Bruce; Lehmberg, Robert; Hegeler, Frank; Terrell, Stephen; Aglitskiy, Yefim; Schmitt, Andrew

2014-10-01

The krypton-fluoride (KrF) laser provides the shortest wavelength, broadest bandwidth and most uniform target illumination of all developed high-energy lasers. For directly driven targets these characteristics result in higher and more uniform ablation pressures as well as higher intensity thresholds for laser-plasma instability. The ISI beam smoothing scheme implemented on the NRL Nike KrF facility allows easy implementation of focal zooming where the laser radial profile is varied during the laser pulse. The capability for near continuous zooming with KrF would be valuable towards minimizing the effects of cross beam energy transport (CBET) in directly driven capsule implosions. The broad bandwidth ISI beam smoothing that is utilized with the Nike KrF facility may further inhibit certain laser plasma instability. In this presentation we will summarize our current understanding of laser target interaction with the KrF laser and the benefits it provides for ICF and certain HEDP experiments. Status and progress in high-energy KrF laser technology will also be discussed. Work supported by the Deparment of Energy, NNSA.

A randomised, phase II study of repeated rhenium-188-HEDP combined with docetaxel and prednisone versus docetaxel and prednisone alone in castration-resistant prostate cancer (CRPC) metastatic to bone; the Taxium II trial

Energy Technology Data Exchange (ETDEWEB)

Dodewaard-de Jong, Joyce M. van [VU University Medical Centre, Department of Medical Oncology, Amsterdam (Netherlands); Meander Medical Centre, Department of Medical Oncology, Amersfoort (Netherlands); Klerk, John M.H. de [Meander Medical Centre, Department of Nuclear Medicine, Amersfoort (Netherlands); Bloemendal, Haiko J. [Meander Medical Centre, Department of Medical Oncology, Amersfoort (Netherlands); University Medical Centre Utrecht, Department of Medical Oncology, Utrecht (Netherlands); Oprea-Lager, Daniela E.; Hoekstra, Otto S. [VU University Medical Centre, Department of Radiology and Nuclear Medicine, Amsterdam (Netherlands); Berg, H.P. van den [Tergooi Medical Hospital, Department of Medical Oncology, Hilversum (Netherlands); Los, Maartje [St Antonius Hospital Utrecht, Department of Medical Oncology, Utrecht (Netherlands); Beeker, Aart [Spaarne Gasthuis, Department of Medical Oncology, Hoofddorp (Netherlands); Jonker, Marianne A. [VU University Medical Centre, Department of Epidemiology and Biostatistics, Amsterdam (Netherlands); O' Sullivan, Joe M. [Queen' s University Belfast, Centre for Cancer Research and Cell Biology, Belfast, Northern Ireland (United Kingdom); Verheul, Henk M.W.; Eertwegh, Alfons J.M. van den [VU University Medical Centre, Department of Medical Oncology, Amsterdam (Netherlands)

2017-08-15

Rhenium-188-HEDP is a beta-emitting radiopharmaceutical used for palliation of metastatic bone pain. We investigated whether the addition of rhenium-188-HEDP to docetaxel/prednisone improved efficacy of chemotherapy in patients with CRPC. Patients with progressive CRPC and osteoblastic bone metastases were randomised for first-line docetaxel 75 mg/m{sup 2} 3-weekly plus prednisone with or without 2 injections of rhenium-188-HEDP after the third (40 MBq/kg) and after the sixth (20 MBq/kg) cycle of docetaxel. Primary endpoint was progression-free survival (PFS), defined as either PSA, radiographic or clinical progression. Patients were stratified by extent of bone metastases and hospital. Forty-two patients were randomised for standard treatment and 46 patients for combination therapy. Median number of cycles of docetaxel was 9 in the control group and 8 in the experimental group. Median follow-up was 18.4 months. Two patients from the experimental group did not start treatment after randomisation. In the intention to treat analysis no differences in PFS, survival and PSA became apparent between the two groups. In an exploratory per-protocol analysis median overall survival was significantly longer in the experimental group (33.8 months (95%CI 31.75-35.85)) than in the control group (21.0 months (95%CI 13.61-28.39); p 0.012). Also median PFS in patients with a baseline phosphatase >220U/L was significantly better with combination treatment (9.0 months (95%CI 3.92-14.08) versus 6.2 months (95%CI 3.08-9.32); log rank p 0.005). As expected, thrombocytopenia (grade I/II) was reported more frequently in the experimental group (25% versus 0%). Combined treatment with rhenium-188-HEDP and docetaxel did not prolong PFS in patients with CRPC. The observed survival benefit in the per-protocol analysis warrants further studies in the combined treatment of chemotherapy and radiopharmaceuticals. (orig.)

Subassembly faults diagnostic of an LMFBR type reactor by the measurement of temperature noise

International Nuclear Information System (INIS)

Kokorev, B.V.; Palkin, I.I.; Turchin, N.M.; Pallagi, D.; Horanyi, S.

1979-09-01

The subassembly faults detection possibility by temperature noise analysis of an LMFBR is described. The paper contains the results of diagnostical examinations obtained on electrically heated NaK test rigs. On the basis of these results the measurement of temperature noise RMS value seems to be a practicable method to detect local blockages in an early phase. (author)

Improvement of computer programs 'BAMBOO' and 'ASFRE-IV' for coupling analysis of deformation and thermal-hydraulics in a high burn-up fuel subassembly of fast reactor

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ohshima, Hiroyuki; Imai, Yasutomo

2003-04-01

A simulation system of a deformed fuel subassembly is being developed for the structure integrity of high burn-up wire-spacer-type fuel subassemblies of sodium-cooled fast breeder reactors. This report describes a computer program improvement work for coupling analyses of deformation and thermal-hydraulics in a fuel subassembly as part of the simulation system development. In this work, a function of data conversion as an interface between a bundle deformation analysis program BAMBOO and a thermal hydraulic analysis program ASFRE-IV was incorporated to each program. BAMBOO was improved to accept the coolant temperature data from ASFRE-IV and to offer bundle deformation data to ASFRE-IV. ASFRE-IV was also improved to offer the coolant temperature data to BAMBOO and to obtain the bundle deformation data from BAMBOO. Improved BAMBOO and ASFRE-IV were applied to an analysis of 169-pin bundle for the program verification. It was confirmed that the coupling analysis gave the physically reasonable results on both deformation and thermal hydraulic behaviors in the fuel subassembly. (author)

Development of a FBR fuel bundle-duct interaction analysis code-BAMBOO. Analysis model and verification by Phenix high burn-up fuel subassemblies

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ito, Masahiro; Ukai, Shigeharu

2005-01-01

The bundle-duct interaction analysis code ''BAMBOO'' has been developed for the purpose of predicting deformation of a wire-wrapped fuel pin bundle of a fast breeder reactor (FBR). The BAMBOO code calculates helical bowing and oval-distortion of all the fuel pins in a fuel subassembly. We developed deformation models in order to precisely analyze the irradiation induced deformation by the code: a model to analyze fuel pin self-bowing induced by circumferential gradient of void swelling as well as thermal expansion, and a model to analyze dispersion of the orderly arrangement of a fuel pin bundle. We made deformation analyses of high burn-up fuel subassemblies in Phenix reactor and compared the calculated results with the post irradiation examination data of these subassemblies for the verification of these models. From the comparison we confirmed that the calculated values of the oval-distortion and bowing reasonably agreed with the PIE results if these models were used in the analysis of the code. (author)

Preliminary validation of the MATRA-LMR-FB code for the flow blockage in a subassembly

International Nuclear Information System (INIS)

Jeong, H. Y.; Ha, K. S.; Kwon, Y. M.; Chang, W. P.; Lee, Y. B.; Heo, S.

2005-01-01

To analyze the flow blockage in a subassembly of a Liquid Metal-cooled Reactor (LMR), the MATRA-LMR-FB code has been developed and validated for the existing experimental data. Compared to the MATRA-LMR code, which had been successfully applied for the core thermal-hydraulic design of KALIMER, the MATRA-LMR-FB code includes some advanced modeling features. Firstly, the Distributed Resistance Model (DRM), which enables a very accurate description of the effects of wire-wrap and blockage in a flow path, is developed for the MATRA-LMR-FB code. Secondly, the hybrid difference method is used to minimize the numerical diffusion especially at the low flow region such as recirculating wakes after blockage. In addition, the code is equipped with various turbulent mixing models to describe the active mixing due to the turbulent motions as accurate as possible. For the validation of the MATRA-LMR-FB code the ORNL THORS test and KOS 169-pin test are analyzed. Based on the analysis results for the temperature data, the accuracy of the code is evaluated quantitatively. The MATRA-LMR-FB code predicts very accurately the exit temperatures measured in the subassembly with wire-wrap. However, the predicted temperatures for the experiment with spacer grid show some deviations from the measured. To enhance the accuracy of the MATRA-LMR-FB for the flow path with grid spacers, it is suggested to improve the models for pressure loss due to spacer grid and the modeling method for blockage itself. The developed MATRA-LMR-FB code is evaluated to be applied to the flow blockage analysis of KALIMER-600 which adopts the wire-wrapped subassemblies

Device and method for unfastening and lifting a top nozzle subassembly from a reconstitutable fuel assembly

International Nuclear Information System (INIS)

Wilson, J.F.

1987-01-01

This patent describes a reconstitutable fuel assembly including at least one guide thimble having an upper end portion and a top nozzle subassembly having a lower adapter plate with at least one opening, and an upper hold-down plate with at least one passageway positioned above and aligned with the lower adapter plate opening. At least one hold-down spring is disposed and extends between the upper and lower plates and at least one elongated tubular hollow sleeve is disposed and extends between the upper and lower plates. The upper end portion of the guide thimble extends upwardly through the opening in the lower adapter plate and has a threaded terminal end disposed above the adapter plate. The threaded terminal end of the guide thimble and an upper end extend upwardly through the passageway of the upper hold-down plate. A device is described for unfastening and lifting the top nozzle subassembly from the guide thimble of the fuel assembly, comprising: (a) at least one hollow gripper tube, the tube having an open lower end; (b) means mounting the gripper tube for vertical alignment with and insertion of its lower end portion into the elongated sleeve of the top nozzle subassembly to a position therein located above and adjacent to the threaded lower end of the sleeve; (c) force-generating means disposed within the gripper tube for rotatable movement and concurrent axial movement upwardly and downwardly within the tube and also disposed at the open lower end of the gripper tube for extension into and from the gripper tube open lower end upon axial movement upwardly and downwardly within the gripper tube

Irradiation experience with KNK II Fast Breeder Fuel Subassemblies

International Nuclear Information System (INIS)

Hess, B.

1993-02-01

During the operation of the second core of KNK II fuel pin failures occurred, which were caused by local cladding weakening due to mechanical interaction between fuel pins and pin spacers. The present report gives a summarizing presentation of the consequences of these interactions, of the experimental and theoretical investigations to clarify the reason for the interactions and of measures to reduce their consequences in the extended residence time of the second core of KNK II. This type of interaction is caused by thermo-elastic instabilities of the fuel pin bundle, and its strength depends sensitively on the geometry of the pin bundle and the pin power. Finally, measures are described, which were taken for the fuel subassemblies of the third core of KNK II to avoid the wear causing instabilities [de

Probabilistic distributions of pin gaps within a wire-spaced fuel subassembly and sensitivities of the related uncertainties to pin gap

International Nuclear Information System (INIS)

Sakai, K.; Hishida, H.

1978-01-01

Probabilistic fuel pin gap distributions within a wire-spaced fuel subassembly and sensitivities of the related uncertainties to fuel pin gaps are discussed. The analyses consist mainly of expressing a local fuel pin gap in terms of sensitivity functions of the related uncertainties and calculating the corresponding probabilistic distribution through taking all the possible combinations of the distribution of uncertainties. The results of illustrative calculations show that with the reliability level of 0.9987, the maximum deviation of the pin gap at the cladding hot spot of a center fuel subassembly is 8.05% from its nominal value and the corresponding probabilistic pin gap distribution is shifted to the narrower side due to the external confinement of a pin bundle with a wrapper tube. (Auth.)

Application of 1-hydroxyethylidene-1, 1-diphosphonic acid in boiler water for industrial boilers.

Science.gov (United States)

Zeng, Bin; Li, Mao-Dong; Zhu, Zhi-Ping; Zhao, Jun-Ming; Zhang, Hui

2013-01-01

The primary method used for boiler water treatment is the addition of chemicals to industrial boilers to prevent corrosion and scaling. The static scale inhibition method was used to evaluate the scale inhibition performance of 1-hydroxyethylidene-1, 1-diphosphonic acid (HEDP). Autoclave static experiments were used to study the corrosion inhibition properties of the main material for industrial boilers (20# carbon steel) with an HEDP additive in the industrial boiler water medium. The electrochemical behavior of HEDP on carbon steel corrosion control was investigated using electrochemical impedance spectroscopy and Tafel polarization techniques. Experimental results indicate that HEDP can have a good scale inhibition effect when added at a quantity of 5 to 7 mg/L at a test temperature of not more than 100 °C. To achieve a high scale inhibition rate, the HEDP dosage must be increased when the test temperature exceeds 100 °C. Electrochemical and autoclave static experimental results suggest that HEDP has a good corrosion inhibition effect on 20# carbon steel at a concentration of 25 mg/L. HEDP is an excellent water treatment agent.

Quasi-steady state boiling downstream of a central blockage in a 19-rod simulated LMFBR subassembly (FFM bundle 3B)

International Nuclear Information System (INIS)

Hanus, N.; Fontana, M.H.; Gnadt, P.A.; MacPherson, R.E.; Smith, C.M.; Wantland, J.L.

1976-01-01

Results of sodium boiling tests in a centrally blocked 19-rod simulated LMFBR subassembly are discussed. The tests were part of the experimental series conducted with bundle 3B in the Fuel Failure Mockup (FFM) at ORNL

Computation of turbulent flow and heat transfer in subassemblies

International Nuclear Information System (INIS)

Slagter, W.

1979-01-01

This research is carried out in order to provide information on the thermohydraulic behaviour of fast reactor subassemblies. The research work involves the development of versatile computation methods and the evaluation of combined theoretical and experimental work on fluid flow and heat transfer in fuel rod bundles. The computation method described here rests on the application of the distributed parameter approach. The conditions considered cover steady, turbulent flow and heat transfer of incompressible fluids in bundles of bare rods. Throughout 1978 main efforts were given to the development of the VITESSE program and to the validation of the hydrodynamic part of the code. In its present version the VITESSE program is applicable to predict the fully developed turbulent flow and heat transfer in the subchannels of a bundle with bare rods. In this paper the main features of the code are described as well as the present status of development

EXPEL - a computing module for molten fuel/coolant interactions in fast reactor sub-assemblies

International Nuclear Information System (INIS)

Fishlock, T.P.

1975-10-01

This report describes a module for computing the effects of a molten fuel/coolant interaction in a fast reactor subassembly. The module is to be incorporated into the FRAX code which calculates the consequences of hypothetical whole core accidents. Details of the interaction are unknown and in consequence the model contains a large number of parameters which must be set by assumption. By variation of these parameters the interaction may be made mild or explosive. Results of a parametric survey are included. (author)

The EBR-II materials-surveillance program. 4: Results of SURV-4 and SURV-6

International Nuclear Information System (INIS)

Ruther, W.E.; Hayner, G.O.; Carlson, B.G.; Ebersole, E.R.; Allen, T.R.

1998-01-01

In March of 1965, a set of surveillance (SURV) samples was placed in the EBR-II reactor to determine the effect of irradiation, thermal aging, and sodium corrosion on reactor materials. Eight subassemblies were placed into row 12 positions of EBR-II to determine the effect of irradiation at 370 C. Two subassemblies were placed into the primary sodium basket to determine the effect of thermal aging at 370 C. For both the irradiated and thermally aged samples, one half of all samples were exposed to primary system sodium while one half were sealed in capsules with a helium atmosphere. Fifteen different structural materials were tested in the SURV program. In addition to the fifteen types of metal samples, graphite blocks were irradiated in the SURV subassemblies to determine the effect of irradiation on the graphite neutron shield. In this report, the properties of these materials irradiated at 370 C to a total fluence of 2.2 x 10 22 n/cm 2 (over 2,994 days) are compared with those of similar specimens thermally aged at 370 C for 2,994 days in the storage basket of the reactor. The properties analyzed were weight, density, microstructure, hardness, tensile and yield strength, impact strength, and creep

Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

International Nuclear Information System (INIS)

Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

2005-01-01

Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

Large scale FCI experiments in subassembly geometry. Test facility and model experiments

International Nuclear Information System (INIS)

Beutel, H.; Gast, K.

A program is outlined for the study of fuel/coolant interaction under SNR conditions. The program consists of a) under water explosion experiments with full size models of the SNR-core, in which the fuel/coolant system is simulated by a pyrotechnic mixture. b) large scale fuel/coolant interaction experiments with up to 5kg of molten UO 2 interacting with liquid sodium at 300 deg C to 600 deg C in a highly instrumented test facility simulating an SNR subassembly. The experimental results will be compared to theoretical models under development at Karlsruhe. Commencement of the experiments is expected for the beginning of 1975

Analyses of deformation and thermal-hydraulics within a wire-wrapped fuel subassembly in a liquid metal fast reactor by the coupled code system

Energy Technology Data Exchange (ETDEWEB)

Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp; Ohshima, Hiroyuki; Ito, Masahiro

2017-06-15

Highlights: • The coupled computational code system allowed for mechanical and thermal-hydraulic analyses in a fast reactor fuel subassembly. • In this system interactive calculations between flow area deformations and coolant temperature changes are repeated to their convergence state. • Effects on bundle-duct interaction on coolant temperature distributions were investigated by using the code system. - Abstract: The coupled numerical analysis of mechanical and thermal-hydraulic behaviors was performed for a wire-wrapped fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal-hydraulic analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that the radial distribution of coolant temperature in the subassembly tended to flatten as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such flattening of temperature distribution was slightly observed as a result of fuel pin bowings due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal-hydraulics was also investigated in this study.

The thermalhydraulics of a pin bundle with a helical wire wrap spacer. Modeling and qualification for a new sub-assembly concept

International Nuclear Information System (INIS)

Valentin, B.

2000-01-01

For the sub-assembly composed by an hexcan and a pin bundle with an helical wire wrap spacer, the calculation of the maximum clad temperatures, with the design code CADET, imposed to correctly evaluate the heat and mass transfers due to the helical wire. The models use theoretical and experimental arguments which are presented after a brief description of the hydraulic behavior of a such bundle. The design of a new sub-assembly concept, in the framework of high plutonium consumption in fast reactor projects needs to qualify tile models from RAPSODIE, PHENIX and SUPER-PHENIX programs. The qualification program, which could be used, is described. the approach is notably comparative for the hydraulic fields and the past experimental results will be useful. Another approach is briefly presented. It uses a multidimensional code (TRIO) which solves Navier-Stokes equations. The utility and the limits of a such method are described. (author)

3He(d,p)4He reaction calculation with three-body Faddeev equations

International Nuclear Information System (INIS)

Oryu, S.; Uzu, E.; Sunahara, H.; Yamada, T.; Tabaru, G.; Hino, T.

1998-01-01

In order to investigate the 3 He-n-p system as a three-body problem, we have formulated 3 He-n and 3 H-p effective potentials using both a microscopic treatment and a phenomenological approach. In the microscopic treatment, potentials are generated by means of the resonating group method (RGM) based on the Minnesota nucleon-nucleon potential. These potentials are converted into separable form by means of the microscopic Pauli correct (MPC) method. The MPC potentials are properly formulated to avoid Pauli forbidden states. The phenomenological potentials are obtained by modifying parameters of the EST approximation to the Paris nucleon-nucleon potential, such that they fit the low-energy 3 He-n, 3 H-p, and 3 He-p phase shifts. Therefore, they describe the 3 He-n differential cross section, the polarization observables, and the energy levels of 4 He. The 3 He-n-p Faddeev equations are solved numerically. We reproduce correctly the ground state and the first excited state of 5 Li. Furthermore, the Paris-type potential is used to investigate the 3 He(d,p) 4 He reaction at a deuteron bombarding energy of 270 MeV, where the system is treated as a three-body problem. Results for the polarized and unpolarized differential cross sections demonstrate convergence of the Born series. (orig.)

The MIT HEDP Accelerator Facility for Diagnostic Development for OMEGA, Z, and the NIF

Science.gov (United States)

Sio, H.; Gatu Johnson, M.; Birkel, A.; Doeg, E.; Frankel, R.; Kabadi, N. V.; Lahmann, B.; Manzin, M.; Simpson, R. A.; Parker, C. E.; Sutcliffe, G. D.; Wink, C.; Frenje, J. A.; Li, C. K.; Seguin, F. H.; Petrasso, R. D.; Leeper, R.; Hahn, K.; Ruiz, C. L.; Sangster, T. C.; Hilsabeck, T.

2017-10-01

The MIT HEDP Accelerator Facility utilizes a 135-keV, linear electrostatic ion accelerator; DT and DD neutron sources; and two x-ray sources for development and characterization of nuclear diagnostics for OMEGA, Z, and the NIF. The accelerator generates DD and D3He fusion products through the acceleration of D+ ions onto a 3He-doped Erbium-Deuteride target. Accurately characterized fusion product rates of around 106 s- 1 are routinely achieved. The DT and DD neutron sources generate up to 6×108 and 1×107 neutrons/s, respectively. One x-ray generator is a thick-target W source with a peak energy of 225 keV and a maximum dose rate of 12 Gy/min; the other uses Cu, Mo, or Ti elemental tubes to generate x-rays with a maximum energy of 40 keV. Diagnostics developed and calibrated at this facility include CR-39-based charged-particle spectrometers, neutron detectors, and the particle Time-Of-Flight (pTOF) and Magnetic PTOF CVD-diamond-based bang time detectors. The accelerator is also a valuable hands-on tool for graduate and undergraduate education at MIT. This work was supported in part by the U.S. DoE, SNL, LLE and LLNL.

Thermal-hydraulic numerical simulation of fuel sub-assembly for Sodium-cooled Fast Reactor

International Nuclear Information System (INIS)

Saxena, Aakanksha

2014-01-01

The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR). First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry. The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (∼20 C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (∼1 C-2 C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10 Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures. (author) [fr

Development of multi-dimensional analysis method for porous blockage in fuel subassembly. Numerical simulation for 4 subchannel geometry water test

International Nuclear Information System (INIS)

Tanaka, Masa-aki; Kamide, Hideki

2001-02-01

This investigation deals with the porous blockage in a wire spacer type fuel subassembly in Fast Breeder Reactors (FBR's). Multi-dimensional analysis method for a porous blockage in a fuel subassembly is developed using the standard k-ε turbulence model with the typical correlations in handbooks. The purpose of this analysis method is to evaluate the position and the magnitude of the maximum temperature, and to investigate the thermo-hydraulic phenomena in the porous blockage. Verification of this analysis method was conducted based on the results of 4-subchannel geometry water test. It was revealed that the evaluation of the porosity distribution and the particle diameter in a porous blockage was important to predict the temperature distribution. This analysis method could simulate the spatial characteristic of velocity and temperature distributions in the blockage and evaluate the pin surface temperature inside the porous blockage. Through the verification of this analysis method, it is shown that this multi-dimensional analysis method is useful to predict the thermo-hydraulic field and the highest temperature in a porous blockage. (author)

Sodium-fuel interaction: dropping experiments and subassembly test

International Nuclear Information System (INIS)

Holtbecker, H.; Schins, H.; Jorzik, E.; Klein, K.

1978-01-01

Nine dropping tests, which bring together 2 to 4 kg of molten UO 2 with 150 l sodium, showed the incoherency and non-violence of these thermal interactions. The pressures can be described by sodium incipient boiling and bubble collapse; the UO 2 fragmentation by thermal stress and bubble collapse impact forces. The mildness of the interaction is principally due to the slowness and incoherency of UO 2 fragmentation. This means that parametric models which assume instantaneous mixing and fragmentation are of no use for the interpretation of dropping experiments. One parametric model, the Caldarola Fuel Coolant Interaction Variable Mass model, is being coupled to the two dimensional time dependent hydrodynamic REXCO-H code. In a first step the coupling is applicated to a monodimensional geometry. A subassembly test is proposed to validate the model. In this test rapid mixing between UO 2 and sodium has to be obtained. Dispersed molten UO 2 fuel is obtained by flashing injected sodium drops inside a UO 2 melt. This flashing is theoretically explained and modelled as a superheat limited explosion. The measured sodium drop dwell times of two experiments are compared to results obtained from the mentioned theory, which is the basis of the Press 2 Code

The EBR-II materials-surveillance program. 5: Results of SURV-5

International Nuclear Information System (INIS)

Ruther, W.E.; Staffon, J.D.; Carlson, B.G.; Allen, T.R.

1998-01-01

In March of 1965, a set of surveillance (SURV) samples was placed in the EBR-II reactor to determine the effect of irradiation, thermal aging, and sodium corrosion on reactor materials. Eight subassemblies were placed into row 12 positions of EBR-II to determine the effect of irradiation at 370 C. Two subassemblies were placed into the primary sodium basket to determine the effect of thermal aging at 370 C. One half of all samples were exposed to primary system sodium while one half were sealed in capsules with a helium atmosphere. Fifteen different structural materials were tested in the SURV program. In this work, the properties of these materials irradiated at 370 C to a total fluence of 3.2 x 10 22 n/cm 2 were determined. These materials are the fifth set of irradiated subassemblies to be examined as part of the SURV program (SURV-5). The properties analyzed were weight, density, microstructure, hardness, tensile and yield strength, and fracture resistance. Of all the alloys examined in SURV-5, only Berylco-25 showed any significant weight loss. Stainless steel (both 304 and 347) had the largest density decrease, although the density decrease from irradiation for all alloys was less than 0.4 percent. The microstructure of both Berylco-25 and the aluminum-bronze alloy was altered significantly. Iron- and nickel-base alloys showed little change in microstructure. Austenitic steels (304 and 347) harden with irradiation. The hardness of Inconel X750 did not change significantly with irradiation. The ultimate tensile strength of Inconel X750, 304 stainless steel, 420 stainless steel and welded 304 changed little due to a fluence increase from 2.2 x 10 22 n/cm 2 (the maximum fluence of the SURV-4 samples) to 3.2 x 10 22 n/cm 2

Specialists' meeting on thermodynamics of FBR fuel subassemblies under nominal and non-nominal operating conditions. Summary report

Energy Technology Data Exchange (ETDEWEB)

1979-01-01

The purpose of the meeting was to provide a forum for exchange of information on thermo- and fluiddynamic investigations of LMFBR-subassembly. Special emphasis was placed on nominal and non-nominal conditions. The technical part of the meeting was divided into four sessions, as follows: status of the thermo- and fluiddynamic activities; physical and mathematical modelling of single phase; rod bundle thermohydraulics; experimental investigations; and future R and D. Separate abstracts are included for each of the papers.

Design and adjustment on test bed of replacing subassembly machine control system for China experimental fast reactor

International Nuclear Information System (INIS)

Dong Shengguo; Ma Hongsheng; Zhao Lixia

2008-01-01

The present research concerns in the design and adjustment of replacing sub- assembly machine control system of China Experimental Fast Reactor. The design of replacing subassembly machine control system adopts some electric equipments, such as programmable controllers, digital DC drivers. The designed control system was adjusted on the test bed. The results indicate that the operation of the control system is steady and reliable, and designed control system can meet the needs of the design specification. (authors)

Influence of leakage flow on the behaviour of gas behind a blockage in LMFBR subassembly geometry

International Nuclear Information System (INIS)

Fukuzawa, Y.

1980-07-01

Observations were made of the behaviour of gas behind a uniform porous 21% corner blockage within a pin-bundle of LMFBR subassembly geometry. The main parameter of the experiment was the leakage flow rate through the blockage. The behaviour of gas is significantly influenced by the leakage flow rate. The measured size and residence time of a gas cavity formed behind the blockage are shown and the mechanisms of the gas cavity dispersion by the leakage flow discussed by using a simple model of the liquid flow distribution behind the blockage. (orig.) [de

Specialists' meeting on thermodynamics of FBR fuel subassemblies under nominal and non-nominal operating conditions. Summary report

Energy Technology Data Exchange (ETDEWEB)

1979-01-01

The purpose of the meeting was to provide a forum for exchange of information on thermo- and fluiddynamic investigations of LMFBR-subassembly. Special emphasis was placed on nominal and non-nominal conditions. The technical part of the meeting was divided into four sessions, as follows: status of the thermo- and fluiddynamic activities; physical and mathematical modelling of single phase; rod bundle thermohydraulics; experimental investigations; and future R and D. Separate abstracts are included for each of the papers.

The MIT HEDP Accelerator Facility for education and advanced diagnostics development for OMEGA, Z and the NIF

Science.gov (United States)

Petrasso, R.; Gatu Johnson, M.; Armstrong, E.; Han, H. W.; Kabadi, N.; Lahmann, B.; Orozco, D.; Rojas Herrera, J.; Sio, H.; Sutcliffe, G.; Frenje, J.; Li, C. K.; Séguin, F. H.; Leeper, R.; Ruiz, C. L.; Sangster, T. C.

2015-11-01

The MIT HEDP Accelerator Facility utilizes a 135-keV linear electrostatic ion accelerator, a D-T neutron source and two x-ray sources for development and characterization of nuclear diagnostics for OMEGA, Z, and the NIF. The ion accelerator generates D-D and D-3He fusion products through acceleration of D ions onto a 3He-doped Erbium-Deuteride target. Fusion reaction rates around 106 s-1 are routinely achieved, and fluence and energy of the fusion products have been accurately characterized. The D-T neutron source generates up to 6 × 108 neutrons/s. The two x-ray generators produce spectra with peak energies of 35 keV and 225 keV and maximum dose rates of 0.5 Gy/min and 12 Gy/min, respectively. Diagnostics developed and calibrated at this facility include CR-39 based charged-particle spectrometers, neutron detectors, and the particle Time-Of-Flight (pTOF) and Magnetic PTOF CVD-diamond-based bang time detectors. The accelerator is also a vital tool in the education of graduate and undergraduate students at MIT. This work was supported in part by SNL, DOE, LLE and LLNL.

An Assessment of the International Space Station's Trace Contaminant Control Subassembly Process Economics

Science.gov (United States)

Perry J. L.; Cole, H. E.; El-Lessy, H. N.

2005-01-01

The International Space Station (ISS) Environmental Control and Life Support System includes equipment speci.cally designed to actively remove trace chemical contamination from the cabin atmosphere. In the U.S. on-orbit segment, this function is provided by the trace contaminant control subassembly (TCCS) located in the atmosphere revitalization subsystem rack housed in the laboratory module, Destiny. The TCCS employs expendable adsorbent beds to accomplish its function leading to a potentially signi.cant life cycle cost over the life of the ISS. Because maintaining the TCCSs proper can be logistically intensive, its performance in .ight has been studied in detail to determine where savings may be achieved. Details of these studies and recommendations for improving the TCCS s process economics without compromising its performance or crew health and safety are presented and discussed.

Irradiation environment and materials behavior

International Nuclear Information System (INIS)

Ishino, Shiori

1992-01-01

Irradiation environment is unique for materials used in a nuclear energy system. Material itself as well as irradiation and environmental conditions determine the material behaviour. In this review, general directions of research and development of materials in an irradiation environment together with the role of materials science are discussed first, and then recent materials problems are described for energy systems which are already existing (LWR), under development (FBR) and to be realized in the future (CTR). Topics selected are (1) irradiation embrittlement of pressure vessel steels for LWRs, (2) high fluence performance of cladding and wrapper materials for fuel subassemblies of FBRs and (3) high fluence irradiation effects in the first wall and blanket structural materials of a fusion reactor. Several common topics in those materials issues are selected and discussed. Suggestions are made on some elements of radiation effects which might be purposely utilized in the process of preparing innovative materials. (J.P.N.) 69 refs

Using Modeling and Simulation to Complement Testing for Increased Understanding of Weapon Subassembly Response.

Energy Technology Data Exchange (ETDEWEB)

Wong, Michael K. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Davidson, Megan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-09-01

As part of Sandia’s nuclear deterrence mission, the B61-12 Life Extension Program (LEP) aims to modernize the aging weapon system. Modernization requires requalification and Sandia is using high performance computing to perform advanced computational simulations to better understand, evaluate, and verify weapon system performance in conjunction with limited physical testing. The Nose Bomb Subassembly (NBSA) of the B61-12 is responsible for producing a fuzing signal upon ground impact. The fuzing signal is dependent upon electromechanical impact sensors producing valid electrical fuzing signals at impact. Computer generated models were used to assess the timing between the impact sensor’s response to the deceleration of impact and damage to major components and system subassemblies. The modeling and simulation team worked alongside the physical test team to design a large-scale reverse ballistic test to not only assess system performance, but to also validate their computational models. The reverse ballistic test conducted at Sandia’s sled test facility sent a rocket sled with a representative target into a stationary B61-12 (NBSA) to characterize the nose crush and functional response of NBSA components. Data obtained from data recorders and high-speed photometrics were integrated with previously generated computer models in order to refine and validate the model’s ability to reliably simulate real-world effects. Large-scale tests are impractical to conduct for every single impact scenario. By creating reliable computer models, we can perform simulations that identify trends and produce estimates of outcomes over the entire range of required impact conditions. Sandia’s HPCs enable geometric resolution that was unachievable before, allowing for more fidelity and detail, and creating simulations that can provide insight to support evaluation of requirements and performance margins. As computing resources continue to improve, researchers at Sandia are hoping

NABUB a non-saturated model of coolant boiling in a fast reactor sub-assembly

International Nuclear Information System (INIS)

Brook, A.J.; Mills, D.S.

1975-08-01

A theoretical model is described of sodium boiling in a fast reactor sub-assembly in which the usual assumptions of a saturated vapour are not made. Instead, vapour pressure is calculated in a perfect gas basis, which enables some allowance to be made for the possible presence of non-condensables, which may inhibit the condensation f the vapour. Indications are given of the circumstances under which such inhibition might be expected to show the most marked effects, and some sample results ontained by the code are presented. These show that the coolant voiding pattern is most sensitive to restrictions on the condensing flux in the 100 to 200w/cm 2 range. If unrestricted condensation is assumed, the results of the code are in excellent agreement with more conventional saturation models. (author)

Pump and Flow Control Subassembly of Thermal Control Subsystem for Photovoltaic Power Module

Science.gov (United States)

Motil, Brian; Santen, Mark A.

1993-01-01

The pump and flow control subassembly (PFCS) is an orbital replacement unit (ORU) on the Space Station Freedom photovoltaic power module (PVM). The PFCS pumps liquid ammonia at a constant rate of approximately 1170 kg/hr while providing temperature control by flow regulation between the radiator and the bypass loop. Also, housed within the ORU is an accumulator to compensate for fluid volumetric changes as well as the electronics and firmware for monitoring and control of the photovoltaic thermal control system (PVTCS). Major electronic functions include signal conditioning, data interfacing and motor control. This paper will provide a description of each major component within the PFCS along with performance test data. In addition, this paper will discuss the flow control algorithm and describe how the nickel hydrogen batteries and associated power electronics will be thermally controlled through regulation of coolant flow to the radiator.

Manufacture of core sub-assemblies and fertile fuel assemblies for Indian fast breeder programme

International Nuclear Information System (INIS)

Jayaraj, R.N.

2009-01-01

sintered ThO 2 . Different varieties of stainless steel are employed for the manufacture of intricate components required for Fast Breeder sub-assemblies which are not easily machinable. Large number of precision machined components are fabricated through specialized machining, forming and welding techniques and finally assembled with the help of special jigs and fixtures. Several fabrication techniques were developed like Clad Tube Crimping, Button Forming of Hexagonal Tube, Bead Forming of Spacer Wire, Welding of Clad Tubes to End Plugs and Hexagonal Tube to Foot and Handling Head. Specialized Joining Techniques like Pulsed Current GTAW are employed for the fabrication of thin- walled Fuel Elements. Developmental works are also undertaken for standardizing manufacturing techniques for Oxide Dispersion Strengthened (ODS) alloys for clad tubes of Fast Breeder Reactors, which will have an edge over conventional materials with respect to excellent resistance to void swelling and irradiation embrittlement and also capable of operating under severe conditions for extended periods. The paper highlights various developmental activities carried out for the manufacture of core sub-assemblies for the Fast Breeder Test Reactor (FBTR) and the forthcoming Prototype Fast Breeder Reactor (PFBR). (author)

Fundamental water experiment on subassembly with porous blockage in 4 sub-channel geometry. Influence of flow on temperature distribution in the porous blockage

International Nuclear Information System (INIS)

Tanaka, Masa-aki; Kobayashi, Jun; Isozaki, Tadasi; Nishimura, Motohiko; Kamide, Hideki

1998-03-01

In the liquid metal cooled Fast Breeder Reactor, Local Fault incident is recognized as a key issue of the local subassembly accident. In terms of the reactor safety assessment, it is important to predict the velocity and temperature distributions not only in the fuel subassembly but also in the blockage accurately to evaluate the location of the hottest point and the maximum temperature. In this study, the experiment was performed with the 4 sub-channel geometry water test facility. Dimension is five times larger than that of a real FBR. The porous blockage is located at the center sub-channel in the test section and surrounded with three unplugged sub-channels. The blockages used in this study were, the solid metal, the porous medium consisted of metal spheres, the porous blockage with end plates covering the side or top faces of the blockage to prevent the horizontal and axial flows into the blockage. The experimental parameters were the heater output provided by the electrical heater in the simulated fuel pins and the flow rate. Temperature of the fluid was measured inside/outside the blockage and velocity profiles outside the blockage were measured. (J.P.N.)

Development of computer code models for analysis of subassembly voiding in the LMFBR

International Nuclear Information System (INIS)

Hinkle, W.

1979-12-01

The research program discussed in this report was started in FY1979 under the combined sponsorship of the US Department of Energy (DOE), General Electric (GE) and Hanford Engineering Development Laboratory (HEDL). The objective of the program is to develop multi-dimensional computer codes which can be used for the analysis of subassembly voiding incoherence under postulated accident conditions in the LMFBR. Two codes are being developed in parallel. The first will use a two fluid (6 equation) model which is more difficult to develop but has the potential for providing a code with the utmost in flexibility and physical consistency for use in the long term. The other will use a mixture (< 6 equation) model which is less general but may be more amenable to interpretation and use of experimental data and therefore, easier to develop for use in the near term. To assure that the models developed are not design dependent, geometries and transient conditions typical of both foreign and US designs are being considered

A model of gas cavity breakup behind a blockage in fast breeder reactor subassembly geometry

International Nuclear Information System (INIS)

Fukuzawa, Y.

1980-05-01

A semi-empirical model has been developed to describe the transient behaviour of a gas cavity due to breakup behind a blockage in Liquid Metal Fast Breeder Reactor subassembly geometry. The main mechanisms assumed for gas cavity breakup in the present model are as follows: The gas cavity is broken up by the pressure fluctuation at the interface due to turbulence in the liquid. The centrifugal force on the liquid opposes breakup. The model is able to describe experimental results on the transient behaviour of a gas cavity due to breakup after the termination of gas injection. On the basis of the present model the residence time of a gas cavity behind a blockage in sodium is predicted and the dependence of the residence time on blockage size is discussed. (orig.) [de

Development of materials and manufacturing technologies for Indian fast reactor programme

International Nuclear Information System (INIS)

Raj, Baldev; Jayakumar, T.; Bhaduri, A.K.; Mandal, Sumantra

2010-01-01

Fast Breeder Reactors (FBRs) are vital towards meeting security and sustainability of energy for the growing economy of India. The development of FBRs necessitates extensive research and development in domains of materials and manufacturing technologies in association with a wide spectrum of disciplines and their inter-twining to meet the challenging technology. The paper highlight the work and the approaches adopted for the successful deployment of materials, manufacturing and inspection technologies for the in-core and structural components of current and future Indian Fast Breeder Reactor Programme. Indigenous development of in-core materials viz. Titanium modified austenitic stainless steel (Alloy D9) and its variants, ferritic/martensitic oxide-dispersion strengthened (ODS) steels as well as structural materials viz. 316L(N) stainless steel and modified 9Cr-1Mo have been achieved through synergistic interactions between Indira Gandhi Centre for Atomic Research (IGCAR), education and research institutes and industries. Robust manufacturing technology has been established for forming and joining of various components of 500 MWe Prototype Fast Breeder Reactor (PFBR) through 'science-based technology' approach. To achieve the strict quality standards of formed parts in terms of geometrical tolerances, residual stresses and microstructural defects, FEM-based modelling and experimental validation was carried out for estimation of spring-back during forming of multiple curvature thick plantes. Optimization of grain boundary character distribution in Alloy D9 was carried out by adopting the grain boundary engineering approach to reduce radiation induced segregation. Extensive welding is involved in the fabrication of reactor vessels, piping, steam generators, fuel sub-assemblies etc. Activated Tungsten Inert Gas Welding process along with activated flux developed at IGCAR has been successfully used in fabrication of dummy fuel subassemblies (DFSA) required for testing

Development of materials and manufacturing technologies for Indian fast reactor programme

Energy Technology Data Exchange (ETDEWEB)

Raj, Baldev; Jayakumar, T.; Bhaduri, A.K.; Mandal, Sumantra [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

2010-07-01

Fast Breeder Reactors (FBRs) are vital towards meeting security and sustainability of energy for the growing economy of India. The development of FBRs necessitates extensive research and development in domains of materials and manufacturing technologies in association with a wide spectrum of disciplines and their inter-twining to meet the challenging technology. The paper highlight the work and the approaches adopted for the successful deployment of materials, manufacturing and inspection technologies for the in-core and structural components of current and future Indian Fast Breeder Reactor Programme. Indigenous development of in-core materials viz. Titanium modified austenitic stainless steel (Alloy D9) and its variants, ferritic/martensitic oxide-dispersion strengthened (ODS) steels as well as structural materials viz. 316L(N) stainless steel and modified 9Cr-1Mo have been achieved through synergistic interactions between Indira Gandhi Centre for Atomic Research (IGCAR), education and research institutes and industries. Robust manufacturing technology has been established for forming and joining of various components of 500 MWe Prototype Fast Breeder Reactor (PFBR) through 'science-based technology' approach. To achieve the strict quality standards of formed parts in terms of geometrical tolerances, residual stresses and microstructural defects, FEM-based modelling and experimental validation was carried out for estimation of spring-back during forming of multiple curvature thick plantes. Optimization of grain boundary character distribution in Alloy D9 was carried out by adopting the grain boundary engineering approach to reduce radiation induced segregation. Extensive welding is involved in the fabrication of reactor vessels, piping, steam generators, fuel sub-assemblies etc. Activated Tungsten Inert Gas Welding process along with activated flux developed at IGCAR has been successfully used in fabrication of dummy fuel subassemblies (DFSA) required

Specification of steam generator, condenser and regenerative heat exchanger materials for nuclear applications

International Nuclear Information System (INIS)

Jovasevic, J.V.; Stefanovic, V.M.; Spasic, Z.LJ.

1977-01-01

The basic standards specifications of materials for nuclear applications are selected. Seamless Ni-Cr-Fe alloy Tubes (Inconel-600) for steam generators, condensers and other heat exchangers can be employed instead of austenitic stainless steal or copper alloys tubes; supplementary requirements for these materials are given. Specifications of Ni-Cr-Fe alloy plate, sheet and strip for steam generator lower sub-assembly, U-bend seamless copper-alloy tubes for heat exchanger and condensers are also presented. At the end, steam generator channel head material is proposed in the specification for carbon-steel castings suitable for welding

Summary and implications of out-of-pile investigations of local cooling disturbances in LMFBR subassembly geometry under single-phase and boiling conditions

International Nuclear Information System (INIS)

Huber, F.; Peppler, W.

1985-05-01

The consequences of local cooling disturbances in subassemblies of LMFBRs have been investigated out-of-pile at KfK. Flow and temperature distributions in the disturbed region as well as cooling under boiling conditions up to loss of cooling were investigated. Fission gas release was simulated by gas injection. A total of 16 different blockages in 20 test set-ups were used, four of them under sodium and the rest under water conditions. Mainly planar plates of different sizes and arrangements were used as blockages. In some of the experiments performed in water also porous blockages were investigated. The test sections consisted of electrically heated pin bundles with a thermal-hydraulic characteristic corresponding to that of an SNR 300 subassembly. With different parameter settings the single-phase tests in water furnished a multitude of test results on flow and temperature fields and on the behaviour of gas in the recirculation zone. In the experiments involving boiling two boiling patterns were observed: steady-state boiling and oscillating boiling. With increasing boiling intensity the boiling region grew to some extent, but it remained always confined to the blocked zone because of the relatively cold sodium flow around this zone. In the experiments simulating fission gas release it was found that under certain conditions gas accumulates in the reverse flow region behind a blockage and leads to loss of cooling. (orig./GL) [de

Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

International Nuclear Information System (INIS)

Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

2009-01-01

The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

Observations of the behaviour of gas in the wake behind a corner blockage in fast breeder reactor subassembly geometry

International Nuclear Information System (INIS)

Fukuzawa, Y.

1979-07-01

Observations were made of gas behaviour in the wake behind a 21% corner blockage in the subassembly geometry of a liquid metal fast breeder reactor. The test section used represented one half of the reactor fuel subassembly, divided along the vertical plane of symmetry through the blockage. A glass wall occupied the position of this plane. Water was allowed to flow between glass rods simulating fuel pins, the velocity being changed from 1.2 to 4.5 m/s. Argon was injected into the wake or into the flow upstream of the blockage, the injection rate being changed from 1 to 230 Ncm 3 /s (standard temperature and pressure). From the present experiment, the following is evident: The gas is accumulated in the wake behind the blockage, forming a gas cavity. The flow patterns of the two-phase mixture in the wake are classified into three types, depending on the liquid velocity. In the lower velocity range, a gas cavity cannot be present at rest, rising up through the wake as a single bubble due to buoyancy. In the higher velocity range, the gas cavity is broken up by the liquid flow forces, only small gas bubbles circulating in the wake. In the velocity range in between, the gas cavity is present in the wake. The cavity size depends on the gas injection rate and on the liquid velocity. From the results, the possibility of fuel failure caused by fission gas release at a blockage in the fast breeder reactor can be considered to depend on the operating conditions of the reactor, specially on the coolant velocity. (orig.) [de

Development of process route for production of tubing for various core sub-assemblies and heat exchangers for 500 MWe Indian PFBR

International Nuclear Information System (INIS)

Lakshminarayana, B.; Phani Babu, C.; Dubey, A.K.; Surender, A.; Deshpande, K.V.K.; Maity, P.K.

2009-01-01

India's three stage Nuclear Power Program has entered its second stage on commercial scale with the commencement of construction of 500 MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam. Nuclear Fuel Complex (NFC), Hyderabad is playing a crucial role in the manufacture of all the critical sub-assemblies and control elements for this reactor. The challenging task of process development and production of the various critical tubing for these sub assemblies for PFBR has been taken up by Stainless Steel Tubes Plant (SSTP), NFC with indigenous development of the equipment and technology

Liquid metal cooled fast breeder nuclear reactors

International Nuclear Information System (INIS)

Duncombe, E.; Thatcher, G.

1979-01-01

The invention described relates to a liquid metal cooled fast breeder nuclear reactor in which the fuel assembly has an inner zone comprised mainly of fissile material and a surrounding outer zone comprised mainly of breeder material. According to the invention the sub-assemblies in the outer zone include electro-magnetic braking devices (magnets, pole pieces and armature) for regulating the flow of coolant through the sub-assemblies. The magnetic fields of the electro-magnetic breaking devices are temperature sensitive so that as the power output of the breeder sub-assemblies increases the electro-magnetic resistance to coolant flow is reduced thereby maintaining the temperature of the coolant outlets from the sub-assemblies substantially constant. (UK)

Integrated circuit manufacture and tuning of subassemblies of a statistical analyzer of voltage oscillations (AKON). Izgotovleniye na integral'nykh skhemakh i nastroyka uzlov statisticheskogo analizatora kolebaniy napryazheniya (AKON)

Energy Technology Data Exchange (ETDEWEB)

Yermakov, V.F.; Oleynik, V.I.; Sambarov, Yu.M.

1982-01-01

The basic circuits and instructions for tuning subassemblies of a statistical analyzer of voltage oscillation are described. The device is intended for monitoring quality of voltage in electric networks in accordance with GOST13109-67. The component base of the device includes integrated circuits of the series 140, 155, 218 and 228.

Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

International Nuclear Information System (INIS)

Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S.

2003-01-01

During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- ε turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential

Criticality safety assessment of FBTR fuel sub-assemblies using WIMS cross section set

International Nuclear Information System (INIS)

Gupta, H.C.; Chakraborty, B.

2002-01-01

Full text: FBTR's irradiated fuel sub-assemblies (FSAs) are sent to RML at Indira Gandhi Centre for Atomic Research for post irradiation examination. The FSAs are cut open and the fuel pins are separated for examination in the hot cells. It was required to evaluate the criticality safety in handling the FSAs in the hot cells. Criticality safety studies for handling two as well as three irradiated FSAs in the hot cells under dry conditions were carried out by the Safety Group at IGCAR, Kalpakkam. Monte Carlo code KENO (Version Va) which uses 16-group Hansen-Roach cross-section set was used for the calculations. Subsequently, during the safety review of the proposition by the Safety Review Committee (SARCOP) of AERB, it was stipulated to carry out the criticality safety studies under flooded condition also. We carried out the criticality safety studies for these fuel sub assemblies in different configurations under dry (buried in concrete) as well as wet condition (flooded with light water) using Monte Carlo codes MONALI (developed at BARC) and KENO4 using WlMS-69 group cross section set. Results of our analyses under various conditions are presented in this paper

RESEARCH ON THE STUDY OF MATERIAL DEFECTS AND SOMECOAL MILLS SUBASSEMBLIES LIFE TIME

Directory of Open Access Journals (Sweden)

Cristina LAPADUSI

2013-05-01

Full Text Available The defectsfrom the structureof metallic materials of whichare manufactured the pieces, canbeputoutbyNDT. One ofNDTmethods, commonly usedin practiceisultrasonicmethod.In this paper are rendered the results of the determinations by the effects of coal mills bars by type DGS 100,obtained with ultrasound devices by type PHASOR XS.

Electrochemical preparation of technetium hydroxyethylidene diphosphonate radiopharmaceuticals

International Nuclear Information System (INIS)

Scott, R.B.

1984-01-01

This work describes the liquid chromatographic and electrochemical analysis of electrogenerated technetium hydroxyethylidene diphosphonate (HEDP) complexes, and studies the effectiveness of the resulting bone imaging agents. Anion exchange High Performance Liquid Chromatography is used to separate components, and γ emission is used as the detection mode. The reaction mixtures were prepared at a series of reduction potentials and pH values, at both carrier added and no carrier added technetium levels. The results indicate that all three parameters affect the final complex composition to varying degrees. By optimizing the conditions, a preparation was made which results in a high percentage of a Tc-HEDP complex thought to be a very good home imager. This component was isolated chromatographically and injected into female Sprague-Dawley rats. Comparisons were run on the uptake for seven tissue types at two incubation times. Mercury and Reticulated Vitreous Carbon were used as the working electrode materials, and it is shown how reduced technetium will significantly alter the electrode characteristics, where a conditioned electrode will produce different complexes from those produced at fresh electrode material. By employing coulometric analysis as the preparation was reduced, an n value of 4 was calculated for a particular complex. This procedure involved tracking the radioactive technetium species carefully to account for all electrons used in the system. Finally, an electrochemical detection method for HEDP was explored, utilizing the property of mercury complexation. Anodic sweep Differential Pulse Polarography gives an analytical signal for HEDP at +0.250 V vs Ag/AgCl

MISER-I: a computer code for JOYO fuel management

International Nuclear Information System (INIS)

Yamashita, Yoshioki

1976-06-01

A computer code ''MISER-I'' is for a nuclear fuel management of Japan Experimental Fast Breeder Reactor JOYO. The nuclear fuel management in JOYO can be regarded as a fuel assembly management because a handling unit of fuel in JOYO plant is a fuel subassembly (core and blanket subassembly), and so the recording of material balance in computer code is made with each subassembly. The input information into computer code is given with each subassembly for a transfer operation, or with one reactor cycle and every one month for a burn-up in reactor core. The output information of MISER-I code is the fuel assembly storage record, fuel storage weight record in each material balance subarea at any specified day, and fuel subassembly transfer history record. Change of nuclear fuel composition and weight due to a burn-up is calculated with JOYO-Monitoring Code by off-line computation system. MISER-I code is written in FORTRAN-IV language for FACOM 230-48 computer. (auth.)

Method of Joining Graphite Fibers to a Substrate

Science.gov (United States)

Beringer, Durwood M. (Inventor); Caron, Mark E. (Inventor); Taddey, Edmund P. (Inventor); Gleason, Brian P. (Inventor)

2014-01-01

A method of assembling a metallic-graphite structure includes forming a wetted graphite subassembly by arranging one or more layers of graphite fiber material including a plurality of graphite fibers and applying a layer of metallization material to ends of the plurality of graphite fibers. At least one metallic substrate is secured to the wetted graphite subassembly via the layer of metallization material.

Methods for nuclear material control used in the basic production of a typical radiochemical plant

International Nuclear Information System (INIS)

Kositsyn, V.F.; Mukhortov, N.F.; Korovin, Yu.I.; Rudenko, V.S.; Petrov, A.M.

1999-01-01

Techniques for destructive and non-destructive assay of the component and isotopic composition of nuclear materials are described, namely gravimetric, titrimetric, coulometric, mass spectrometry, as well as those based on registration of neutron and γ radiations. Their metrologic characteristics are described. The techniques described are suggested to be used for nuclear material (NM) control and accounting purposes at the model radiochemical plant for processing irradiated fuel subassemblies from power reactors. The measurement control program is also described. This program is intended for the measurement quality assurance in the framework of NM control and accountancy system [ru

Fieldable Nuclear Material Identification System

International Nuclear Information System (INIS)

Radle, James E.; Archer, Daniel E.; Carter, Robert J.; Mullens, James Allen; Mihalczo, John T.; Britton, Charles L. Jr.; Lind, Randall F.; Wright, Michael C.

2010-01-01

The Fieldable Nuclear Material Identification System (FNMIS), funded by the NA-241 Office of Dismantlement and Transparency, provides information to determine the material attributes and identity of heavily shielded nuclear objects. This information will provide future treaty participants with verifiable information required by the treaty regime. The neutron interrogation technology uses a combination of information from induced fission neutron radiation and transmitted neutron imaging information to provide high confidence that the shielded item is consistent with the host's declaration. The combination of material identification information and the shape and configuration of the item are very difficult to spoof. When used at various points in the warhead dismantlement sequence, the information complimented by tags and seals can be used to track subassembly and piece part information as the disassembly occurs. The neutron transmission imaging has been developed during the last seven years and the signature analysis over the last several decades. The FNMIS is the culmination of the effort to put the technology in a usable configuration for potential treaty verification purposes.

In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

International Nuclear Information System (INIS)

Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

1983-01-01

Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

Phase I/II trials of {sup 186}Re-HEDP in metastatic castration-resistant prostate cancer: post-hoc analysis of the impact of administered activity and dosimetry on survival

Energy Technology Data Exchange (ETDEWEB)

Denis-Bacelar, Ana M.; Chittenden, Sarah J.; Divoli, Antigoni; Flux, Glenn D. [The Institute of Cancer Research and The Royal Marsden Hospital NHS Foundation Trust, Joint Department of Physics, London (United Kingdom); Dearnaley, David P.; Johnson, Bernadette [The Institute of Cancer Research and The Royal Marsden Hospital NHS Foundation Trust, Division of Radiotherapy and Imaging, London (United Kingdom); O' Sullivan, Joe M. [Queen' s University Belfast, Centre for Cancer Research and Cell Biology, Belfast (United Kingdom); McCready, V.R. [Brighton and Sussex University Hospitals NHS Trust, Department of Nuclear Medicine, Brighton (United Kingdom); Du, Yong [The Royal Marsden Hospital NHS Foundation Trust, Department of Nuclear Medicine and PET/CT, London (United Kingdom)

2017-04-15

To investigate the role of patient-specific dosimetry as a predictive marker of survival and as a potential tool for individualised molecular radiotherapy treatment planning of bone metastases from castration-resistant prostate cancer, and to assess whether higher administered levels of activity are associated with a survival benefit. Clinical data from 57 patients who received 2.5-5.1 GBq of {sup 186}Re-HEDP as part of NIH-funded phase I/II clinical trials were analysed. Whole-body and SPECT-based absorbed doses to the whole body and bone lesions were calculated for 22 patients receiving 5 GBq. The patient mean absorbed dose was defined as the mean of all bone lesion-absorbed doses in any given patient. Kaplan-Meier curves, log-rank tests, Cox's proportional hazards model and Pearson's correlation coefficients were used for overall survival (OS) and correlation analyses. A statistically significantly longer OS was associated with administered activities above 3.5 GBq in the 57 patients (20.1 vs 7.1 months, hazard ratio: 0.39, 95 % CI: 0.10-0.58, P = 0.002). A total of 379 bone lesions were identified in 22 patients. The mean of the patient mean absorbed dose was 19 (±6) Gy and the mean of the whole-body absorbed dose was 0.33 (±0.11) Gy for the 22 patients. The patient mean absorbed dose (r = 0.65, P = 0.001) and the whole-body absorbed dose (r = 0.63, P = 0.002) showed a positive correlation with disease volume. Significant differences in OS were observed for the univariate group analyses according to disease volume as measured from SPECT imaging of {sup 186}Re-HEDP (P = 0.03) and patient mean absorbed dose (P = 0.01), whilst only the disease volume remained significant in a multivariable analysis (P = 0.004). This study demonstrated that higher administered activities led to prolonged survival and that for a fixed administered activity, the whole-body and patient mean absorbed doses correlated with the extent of disease, which, in turn, correlated

Evaluation of safe use of 188Re-HEDP comparing urine data and whole body counting in gamma camera

International Nuclear Information System (INIS)

Paolino, Andrea; Teran, Mariella; Savio, Eduardo; Coppe, Fatima; Lopez, Andrea; Hermida, Juan C.; Gaudiano, Javier

2008-01-01

Cancer is the second more frequent cause of death, after cardiovascular disease, in developing countries. Most of adult patients with neoplasms will develop skeletal metastases that can lead to progressive pain. 188 Re emits both beta particles suitable for therapy and a gamma ray (155 keV), adequate for diagnostic imaging in order to verify localization in the pain areas associated to metastatic process. The aim of this work was to correlate 188 Re-HEDP dose estimations using biological samples and direct measures. All the patients had breast or prostate cancer, with bone metastases. Each patient received a tracer dose of 185 MBq of radiopharmaceutical. Urine samples were collected at 0-1, 1-2, 2-4 and, 4-6 hours post administration, and measured in dose calibrator. Whole body counts were acquired using a camera without collimator, window centered at 155 KeV, matrix 256 x 256, during 60 seconds. Data were obtained at 1 and 6 hours post administration with the patient in sitting position at 2 meter from the detector. Percentage of injected dose was calculated both for urine samples and image for each patient. The number of disintegrations was determined for organs in which higher concentration of activity was observed: those involved in the excretion, red marrow and the reminder of the body. Total doses were estimated using OLINDA/EXM software. Conclusions: Data showed that the organs chosen as more compromised during the tracer dose procedure received very low effective doses. A good correlation between calculations performed both for image and urine samples was obtained. Safety of the radiopharmaceutical was also verified using this method. (author)

Nuclear fuel string assembly

International Nuclear Information System (INIS)

Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

1976-01-01

A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

Formation of Imploding Plasma Liners for HEDP and MIF Application

Energy Technology Data Exchange (ETDEWEB)

Witherspoon, F. Douglas [HyperV Technologies Corp., Chantilly, VA (United States); Case, Andrew [HyperV Technologies Corp., Chantilly, VA (United States); Brockington, Samuel [HyperV Technologies Corp., Chantilly, VA (United States); Messer, Sarah [HyperV Technologies Corp., Chantilly, VA (United States); Bomgardner, Richard [HyperV Technologies Corp., Chantilly, VA (United States); Phillips, Mike [HyperV Technologies Corp., Chantilly, VA (United States); Wu, Linchun [HyperV Technologies Corp., Chantilly, VA (United States); Elton, Ray [Univ. of Maryland, College Park, MD (United States)

2014-11-11

Plasma jets with high density and velocity have a number of important applications in fusion energy and elsewhere, including plasma refueling, disruption mitigation in tokamaks, magnetized target fusion, injection of momentum into centrifugally confined mirrors, plasma thrusters, and high energy density plasmas (HEDP). In Magneto-Inertial Fusion (MIF), for example, an imploding material liner is used to compress a magnetized plasma to fusion conditions and to confine the resulting burning plasma inertially to obtain the necessary energy gain. The imploding shell may be solid, liquid, gaseous, or a combination of these states. The presence of the magnetic field in the target plasma suppresses thermal transport to the plasma shell, thus lowering the imploding power needed to compress the target to fusion conditions. This allows the required imploding momentum flux to be generated electromagnetically using off-the-shelf pulsed power technology. Practical schemes for standoff delivery of the imploding momentum flux are required and are open topics for research. One approach for accomplishing this, called plasma jet driven magneto-inertial fusion (PJMIF), uses a spherical array of pulsed plasma guns to create a spherically imploding shell of very high velocity, high momentum flux plasma. This approach requires development of plasma jet accelerators capable of achieving velocities of 50-200 km/s with very precise timing and density profiles, and with high total mass and density. Low-Z plasma jets would require the higher velocities, whereas very dense high-Z plasma shells could achieve the goal at velocities of only 50-100 km/s. In this report, we describe our work to develop the pulsed plasma gun technology needed for an experimental scientific exploration of the PJMIF concept, and also for the other applications mentioned earlier. The initial goal of a few hundred of hydrogen at 200 km/s was eventually replaced with accelerating 8000 μg of argon or xenon to 50 km

Some observations on phosphate based corrosion inhibitors in preventing carbon steel corrosion

International Nuclear Information System (INIS)

Anupkumar, B.; Satpathy, K.K.

2000-01-01

Among the various types of phosphonic acid based inhibitors assayed, namely HEDP, ATMP and a commercial corrosion inhibitor (code named Betz), it was found that Betz has the maximum amount of organic phosphate followed by HEDP and ATMP. The corrosion rate studies show that Betz gives the highest inhibition efficiency followed by HEDP and ATMP. This shows that organic phosphate plays a significant role in corrosion protection. However, it was observed that due to synergestic effect, HEDP in the presence of Zn 2+ gave a better corrosion protection than Betz. The results are discussed in the light of available literature. (author)

Development of more efficacious Tc-99m organ imaging agents for use in nuclear medicine by analytical characterization of radiopharmaceutical mixtures. Progress report, May 1, 1981-April 30, 1982

International Nuclear Information System (INIS)

Heineman, W.R.; Deutsch, E.A.

1981-12-01

The objectives of this year's research were to develop a method for rapidly determining TcO 4 - in 99 Mo//sup 99m/Tc generator eluates, to improve the ability to chromatographically determine individual Tc-HEDP complexes in radiopharmaceuticals, and to investigate the effects of TcO 4 - concentration and electrochemical reduction on the types and relative amounts of Tc-HEDP complexes present in a radiopharmaceutical formulation. A rapid and sensitive high performance liquid chromatographic (HPLC) method for the quantitative determination of pertechnetate (TcO 4 - ) was developed. This HPLC-based analysis may be of considerable utility in assessing the history and function of 99 MO/sup 99m/Tc generators as well as in the routine analysis of reduced technetium radiopharmaceuticals for the presence of undesired TcO 4 - . Encouraging results were obtained on a dimethyl amine column using aqueous (NH 4 ) 2 SO 4 as the mobile phase. The preparation of Tc(NaBH 4 ) HEDP radiopharmaceutical analogues using varying concentrations of total TcO 4 - shows a dramatic effect in the number and distribution of Tc-HEDP complexes over a TcO 4 - concentration range of 10 -2 to 10 -8 M. These results suggest that total TcO 4 - concentration is an important parameter to be considered in the preparation of a specific Tc-HEDP complex to improve skeletal imaging. The preparation of Tc(electrode) HEDP radiopharmaceutical analogues by using electrochemical reduction was explored. The resulting solutions contain Tc-HEDP complexes that are tentatively identified as being the same complexes formed by NaBH 4 reduction, although the relative concentrations of these complexes are quite different with the two modes of reduction. Thus, electrochemical reduction shows promise as a viable route to the preparation of specific Tc-HEDP complexes for improved skeletal imaging

Materials requirements for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Bennett, J.W.; Horton, K.E.

1978-01-01

Materials requirements for Liquid Metal Fast Breeder Reactors (LMFBRs) are quite varied with requisite applications ranging from ex-reactor components such as piping, pumps, steam generators and heat exchangers to in-reactor components such as heavy section reactor vessels, core structurals, fuel pin cladding and subassembly flow ducts. Requirements for ex-reactor component materials include: good high temperature tensile, creep and fatigue properties; compatibility with high temperature flowing sodium; resistance to wear, stress corrosion cracking, and crack propagation; and good weldability. Requirements for in-reactor components include most of those cited above for ex-reactor components as supplemented by the following: resistance to radiation embrittlement, swelling and radiation enhanced creep; good neutronics; compatibility with fuel and fission product materials; and resistance to mass transfer via flowing sodium. Extensive programs are currently in place in a number of national laboratories and industrial contractors to address the materials requirements for LMFBRs. These programs are focused on meeting the near term requirements of early LMFBRs such as the Fast Flux Test Facility and the Clinch River Breeder Reactor as well as the longer term requirements of larger near-commercial and fully-commercial reactors

R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

International Nuclear Information System (INIS)

Vanderhaegen, M.; Paumel, K.; Seiler, J. M.; Tourin, A.; Jeannot, J. P.; Rodriguez, G.

2011-01-01

In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boiling intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)

Fine 3D neutronic characterization of a gas-cooled fast reactor based on plate-type sub-assemblies

International Nuclear Information System (INIS)

Bosq, J. C.; Peneliau, Y.; Rimpault, G.; Vanier, M.

2006-01-01

CEA neutronic studies have allowed the definition of a first 2400 MWth reference gas-cooled fast reactor core using plate-type sub-assemblies, for which the main neutronic characteristics were calculated by the so-called ERANOS 'design calculation scheme' relying on several method approximations. The last stage has consisted in a new refine characterization, using the reference calculation scheme, in order to confirm the impact of the approximations of the design route. A first core lay-out taking into account control rods was proposed and the reactivity penalty due to the control rod introduction in this hexagonal core lay-out was quantified. A new adjusted core was defined with an increase of the plutonium content. This leads to a significant decrease of the breeding gain which needs to be recovered in future design evolutions in order to achieve the self breeding goal. Finally, the safety criteria associated to the control rods were calculated with a first estimation of the uncertainties. All these criteria are respected, even if the safety analysis of GFR concepts and the determination of these uncertainties should be further studied and improved. (authors)

Calculations of the possible consequences of molten fuel sodium interactions in subassembly and whole core geometries

International Nuclear Information System (INIS)

Coddington, P.; Fishlock, T.P.; Jakeman, D.

1976-01-01

The possible consequences of molten fuel sodium interactions are calculated using various modelling assumptions and key parameters. And the significance of the choice of assumptions and parameters are discussed. As for subassembly geometry, the results of one-dimensional code EXPEL are compared with the solutions of the one-dimensional Lagrangian equations of a compressible fluid (TOPAL was used). The adequacy of acoustic approximation used in EXPEL is discussed here. The effects of heat transfer time constant on the behaviour of peak pressure are also analyzed by parametric surveys. Other items investigated are the length and position of the interacting zone, the existence of a non-condensable gas volume, and the vapour condensation on cold clad. As for whole core geometry, a simple dynamical model of arc expanding spherical interacting zone immersed in a semi-infinite sea of cold liquid was used (SHORE code). Within the interacting zone a simple heat transfer model (including a heat transfer time and a fragmentation time) was adopted. Vapour blanketing was considered in a number of ways. Representative results of the calculations are given in a table. Containment studies were also performed for ''ducted'' design and ''open pool'' design. The development of new codes in the U.K. for these analysis are also briefly described. (Aoki, K.)

Stress analysis of glass-ceramic insulator and molybdenum cylinders in vacuum tube subassembly

International Nuclear Information System (INIS)

Spears, R.K.

1980-01-01

This study determined the state of stress between molybdenum cylinders and a glass-ceramic insulator of a vacuum tube during cooling when the glass-ceramic coefficient of expansion differed from molybdenum by +-2 x 10 -7 / 0 C. A thermoelastic stress analysis was performed on the vacuum tube subassembly using the finite element method. Two cases, which examined the effect of cooling over a 700 0 C range, were considered. In Case One, the expansion coefficient of the glass-ceramic was 2 x 10 -7 / 0 C less than that of molybdenum while for Case Two, it was 2 x 10 -7 / 0 C greater. For Case One, it was found that the tangential stresses in the insulator were entirely compressive but the maximum principal stresses in the r-z plane were mainly tensile. For Case Two, the tangential stresses were tensile in the insulator as were most of the maximum principal stresses in the r-z plane except for stress in the upper regions of the insulator. The magnitude of the stress at the maximum principal stress location appears to be substantially lower than what has been observed in practice (i.e., cracking of this design had never been a major problem, but it has been observed that if the coefficient of expansion of the glass-ceramic was 2 x 10 -7 / 0 C lower than molybdenum, cracking usually resulted). This analysis showed that the expansion coefficient of the glass-ceramic could be varied quite liberally from molybdenum before the ultimate strength (13,000 lb/in. 2 ) of the glass-ceramic was exceeded

X-ray doppler velocimetry for diagnosis of fluid motion in ICF implosions

Science.gov (United States)

Koch, J. A.; King, J. A.; Huffman, E.; Freeman, R. R.; Dutra, E. C.; Field, J. E.; Kilkenny, J. D.; Hall, G. N.; Harding, E.; Rochau, G. A.; Porter, J. L.; Covington, A. M.; Beg, F. N.

2017-08-01

We are developing a novel diagnostic for measurement of bulk fluid motion in materials, that is particularly applicable to very hot, x-ray emitting plasmas in the High Energy Density Physics (HEDP) regime. The X-ray Doppler Velocimetry (XDV) technique relies on monochromatic imaging in multiple x-ray energy bands near the center of an x-ray emission line in a plasma, and utilizes bent imaging crystals. Higher energy bands are preferentially sensitive to plasma moving towards the viewer, while lower energy bands are preferentially sensitive to plasma moving away from the viewer. Combining multiple images in different energy bands allows for a reconstruction of the fluid velocity field integrated along the line of sight. We review the technique, and we discuss progress towards benchmarking the technique with proof-of-principle HEDP experiments.

Individualized FAC on bottom tab subassemblies to minimize adhesive gap between emitter and optics

Science.gov (United States)

Sauer, Sebastian; Müller, Tobias; Haag, Sebastian; Beleke, Andreas; Zontar, Daniel; Baum, Christoph; Brecher, Christian

2017-02-01

High Power Diode Laser (HPDL) systems with short focal length fast-axis collimators (FAC) require submicron assembly precision. Conventional FAC-Lens assembly processes require adhesive gaps of 50 microns or more in order to compensate for component tolerances (e.g. deviation of back focal length) and previous assembly steps. In order to control volumetric shrinkage of fast-curing UV-adhesives shrinkage compensation is mandatory. The novel approach described in this paper aims to minimize the impact of volumetric shrinkage due to the adhesive gap between HPDL edge emitters and FAC-Lens. Firstly, the FAC is actively aligned to the edge emitter without adhesives or bottom tab. The relative position and orientation of FAC to emitter are measured and stored. Consecutively, an individual subassembly of FAC and bottom tab is assembled on Fraunhofer IPT's mounting station with a precision of +/-1 micron. Translational and lateral offsets can be compensated, so that a narrow and uniform glue gap for the consecutive bonding process of bottom tab to heatsink applies (Figure 4). Accordingly, FAC and bottom tab are mounted to the heatsink without major shrinkage compensation. Fraunhofer IPT's department assembly of optical systems and automation has made several publications regarding active alignment of FAC lenses [SPIE LASE 8241-12], volumetric shrinkage compensation [SPIE LASE 9730-28] and FAC on bottom tab assembly [SPIE LASE 9727-31] in automated production environments. The approach described in this paper combines these and is the logical continuation of that work towards higher quality of HPDLs.

Calculations of the Possible Consequences of Molten Fuel Sodium Interactions in Subassembly and Whole Core Geometries

International Nuclear Information System (INIS)

Coddington, P.; Fishlock, T.P.; Jakeman, D.

1976-01-01

In making assessments of fast reactor safety a number of accident sequences can be postulated in which molten fuel contacts sodium in a number of possible modes. In the absence of an understanding of the way in which reactor materials interact for these contact modes it is necessary to make assessments over a range of plausible conditions and assumptions. This enables those areas where an interaction might cause a new stage in the escalation of the accident to be identified and at the same time to establish what characteristics of the interaction may be important. Whether in real situations interaction of molten reactor materials can have such characteristics can then be considered from both a theoretical and experimental viewpoint. It is suggested that although high efficiency vapour explosions involving large amounts of fuel in which there is rapid and coherent fragmentation are a main source of concern in many accident sequences, interactions with other characteristics may also be important. Two areas which have been identified are: (i) the interactions of low efficiency which need only involve small fractions of the fuel or possibly could include molten clad but which can accelerate sodium and fuel sufficiently to give rise to large reactivity changes. The recent incident at a steel plant in the U.K. in which 100 tons of molten steel was ejected to a height of 10 m from a torpedo ladle when water accidentally poured into it is a particularly striking illustration of such movement; and (ii) interactions giving rise to a much slower and less coherent heat transfer which may require some degree of fragmentation but not the extensive fragmentation by the specific mechanisms associated with vapour explosions but which nevertheless on the reactor scale could lead to high slug impacts on the containment. Accident codes are being constructed in the U.K. to investigate a series of hypothetical incidents. Modules are required for these codes which enable the consequences

Comparison of measured and calculated composition of irradiated EBR-II blanket assemblies

International Nuclear Information System (INIS)

Grimm, K. N.

1998-01-01

In anticipation of processing irradiated EBR-II depleted uranium blanket subassemblies in the Fuel Conditioning Facility (FCF) at ANL-West, it has been possible to obtain a limited set of destructive chemical analyses of samples from a single EBR-II blanket subassembly. Comparison of calculated values with these measurements is being used to validate a depletion methodology based on a limited number of generic models of EBR-II to simulate the irradiation history of these subassemblies. Initial comparisons indicate these methods are adequate to meet the operations and material control and accountancy (MC and A) requirements for the FCF, but also indicate several shortcomings which may be corrected or improved

Rapsodie first core manufacture. 1. part: processing plant

International Nuclear Information System (INIS)

Masselot, Y.; Bataller, S.; Ganivet, M.; Guillet, H.; Robillard, A.; Stosskopf, F.

1968-01-01

This report is the first in a series of three describing the processes, results and peculiar technical problems related to the manufacture of the first core of the fast reactor Rapsodie. A detailed study of manufacturing processes(pellets, pins, fissile sub-assemblies), the associated testings (raw materials, processed pellets and pins, sub-assemblies before delivery), manufacturing facilities and improvements for a second campaign are described. (author) [fr

Design, construction and operating experience of demonstration LMFBRs. The application of core and fuel performance experience in British reactors to commercial fast reactor design

International Nuclear Information System (INIS)

Bagley, K.Q.

1978-01-01

The Prototype Fast Reactor (PFR) sub-assembly design is described with particular emphasis on the choice of factors that are important in determining satisfactory performance. Reasons for the adoption of specific clad and fuel design details are given in their historical context, and irradiation experience - mostly from the Dounreay Fast Reactor (DFR) - in support of the choices is described. The implications of factors that are now better understood than when the PFR fuel was designed, notably neutron-induced void swelling and irradiation creep, are then considered. It is shown that the 'free-standing' core design used in PFR, in which the sub-assembly is unsupported above the level of the lower axial breeder, relies on the availability of low-swelling, preferably irradiation-creep-resistant alloys as sub-assembly structural materials in order to achieve the prescribed burn-up target. The advantages of a 'restrained core', which makes use of irradiation creep to redress the effects of material swelling, are noted briefly, and the application of this concept to the Commercial Demonstration Fast Reactor (CDFR) core design is described. Probable future trends in pin and sub-assembly design are reviewed and the scope of associated irradiation testing programmes defined. Arrangements for monitoring and evaluating fuel performance, both in reactor and post-irradiation, are outlined and the provisions for endorsement of CDFR pin, sub-assembly and core design details in PFR are indicated. (author)

The impact of duct-to-duct interaction on the hex duct dilation

International Nuclear Information System (INIS)

Lee, M.J.; Chang, L.K.; Lahm, C.E.; Porter, D.L.

1992-01-01

Dilation of the hex duct is an important factor in the operational lifetime of fuel subassemblies in liquid metal fast reactors. It is caused primarily by the irradiation-enhanced creep and void swelling of the hex duct material. Excessive dilation may jeopardize subassembly removal from the core or cause a subassembly storage problem where the grid size of the storage basket is limited. Dilation of the hex duct in Experimental Breeder Reactor II (EBR-II) limits useful lifetime because of these storage basket limitations. It is, therefore, important to understand the hex duct dilation behavior to guide the design and in-core management of fuel subassemblies in a way that excessive duct deformation can be avoided. To investigate the dilation phenomena, finite-element models of the hex duct have been developed. The inelastic analyses were performed using the structural analysis code, ANSYS. Both Type 316 and D9 austenitic stainless steel ducts are considered. The calculated dilations are in good agreement with profilometry measurements made after irradiation. The analysis indicates that subassembly interaction is an important parameter in addition to neutron fluence and temperature in determining hex duct dilation. 5 refs

An IC-MS/MS Method for the Determination of 1-Hydroxyethylidene-1,1-diphosphonic Acid on Uncooked Foods Treated with Peracetic Acid-Based Sanitizers.

Science.gov (United States)

Suzuki, Ippei; Kubota, Hiroki; Ohtsuki, Takashi; Tatebe, Chiye; Tada, Atsuko; Yano, Takeo; Akiyama, Hiroshi; Sato, Kyoko

2016-01-01

A rapid, sensitive, and specific analytical method for the determination of 1-hydroxyethylidene-1,1-diphosphonic acid (HEDP) on uncooked foods after treatment with a peracetic acid-based sanitizer (PAS) was developed. The method involves simple sample preparation steps and analysis using ion chromatography (IC) coupled with tandem mass spectrometry (MS/MS). The quantification limits of HEDP on uncooked foods are 0.007 mg/kg for vegetables and fruits and 0.2 mg/kg for meats. The recovery and relative standard deviation (RSD) of HEDP analyses of uncooked foods ranged from 73.9 to 103.8% and 1.9 to 12.6%, respectively. The method's accuracy and precision were evaluated by inter-day recovery tests. The recovery for all samples ranged from 93.6 to 101.2%, and the within-laboratory repeatability and reproducibility were evaluated based on RSD values, which were less than 6.9 and 11.5%, respectively. Analyses of PAS-treated fruits and vegetables using the developed method indicated levels of HEDP ranging from 0.008 to 0.351 mg/kg. Therefore, the results of the present study suggest that the proposed method is an accurate, precise, and reliable way to determine residual HEDP levels on PAS-treated uncooked foods.

Experimental Studies of the Transport Parameters of Warm Dense Matter

Energy Technology Data Exchange (ETDEWEB)

Chouffani, Khalid [Idaho State Univ., Pocatello, ID (United States)

2014-12-01

There is a need to establish fundamental properties of matter and energy under extreme physical conditions. Although high energy density physics (HEDP) research spans a wide range of plasma conditions, there is one unifying regime that is of particular importance and complexity: that of warm dense matter, the transitional state between solid state condensed matter and energetic plasmas. Most laboratory experimental conditions, including inertial confinement implosion, fall into this regime. Because all aspects of laboratory-created high-energy-density plasmas transition through the warm dense matter regime, understanding the fundamental properties to determine how matter and energy interact in this regime is an important aspect of major research efforts in HEDP. Improved understanding of warm dense matter would have significant and wide-ranging impact on HEDP science, from helping to explain wire initiation studies on the Sandia Z machine to increasing the predictive power of inertial confinement fusion modeling. The central goal or objective of our proposed research is to experimentally determine the electrical resistivity, temperature, density, and average ionization state of a variety of materials in the warm dense matter regime, without the use of theoretical calculations. Since the lack of an accurate energy of state (EOS) model is primarily due to the lack of experimental data, we propose an experimental study of the transport coefficients of warm dense matter.

Comparative study of skeletal dosimetry methods in therapeutic schemes with Re186 HEDP and Sm153 EDTMP

International Nuclear Information System (INIS)

Papanikolos, G.; Lyra, M.; Kontogeorgakos, D.; Jordanou, J.; Vlahos, L.; Limouris, G.

2005-01-01

Full text: Optimum therapeutic management of patients suffering from metastatic bone pain, requires accurate calculations concerning absorbed dose by metastatic lesions and other critical organs, such as red marrow. Mean absorbed dose, which is the current parameter used to predict the efficacy of the treatment, can either overestimate or underestimate, actual doses delivered in these organs/tissues of interest (TOIs). This study presents differences in dosimetric calculations derived utilizing parameters from different sources (MIRDOSE3, MIRD Pamphlet No 11 and S values published by Bouchet et al.), in therapeutic schemes with Re186HEDP and Sm153EDTMP. A set of planar scintigraphic images for 2 groups of patients (1 for Re 186 patients and the other for Sm153 patients) were obtained in the following sequence: 2 during the first 24h post injection (the last of which at 24h post injection) and 2 more from 24h 7d post injection. Processing of the obtained images utilizing ROI quantitative methods, previously calibrated with waterphantom measurements, determine residence times and radionuclide uptakes not only by TOIs but by specific skeletal sites as well. Dosimetric calculations were performed using MIRDOSE3 computer code, S values from MIRD Pamphlet No 11 and site specific Re 186 and Sm 153 S values for several source target combinations within trabecular and cortical bone, reported by Bouchet et al. (J Nucl Med 2000; 41:189 212), along with cumulative site specific activities derived from values obtained by image processing. Skeletal averaged Re 186 and Sm 153 S values were also used from the aforementioned study by Bouchet et al. Time activity curves for various skeletal sites were generated for both groups of patients. Absorbed dose distributions along with time dose rate curves were derived for both red marrow and different regions of the skeleton. Comparisons are made between these parameters and mean absorbed doses calculated using skeletal averaged S values

Multi-scale modelling for HEDP experiments on Orion

Science.gov (United States)

Sircombe, N. J.; Ramsay, M. G.; Hughes, S. J.; Hoarty, D. J.

2016-05-01

The Orion laser at AWE couples high energy long-pulse lasers with high intensity short-pulses, allowing material to be compressed beyond solid density and heated isochorically. This experimental capability has been demonstrated as a platform for conducting High Energy Density Physics material properties experiments. A clear understanding of the physics in experiments at this scale, combined with a robust, flexible and predictive modelling capability, is an important step towards more complex experimental platforms and ICF schemes which rely on high power lasers to achieve ignition. These experiments present a significant modelling challenge, the system is characterised by hydrodynamic effects over nanoseconds, driven by long-pulse lasers or the pre-pulse of the petawatt beams, and fast electron generation, transport, and heating effects over picoseconds, driven by short-pulse high intensity lasers. We describe the approach taken at AWE; to integrate a number of codes which capture the detailed physics for each spatial and temporal scale. Simulations of the heating of buried aluminium microdot targets are discussed and we consider the role such tools can play in understanding the impact of changes to the laser parameters, such as frequency and pre-pulse, as well as understanding effects which are difficult to observe experimentally.

Laser Welding of Sub-assemblies before Forming

DEFF Research Database (Denmark)

Rasmussen, Mads; Olsen, Flemmming Ove; Pecas, Paulo

1996-01-01

This paper describes some experimental investigations of the formability of CO2-laser-welded 0.75 mm and 1.25 mm low carbon steel. There will be a description of how the laser welded blanks behave in different forming tests, and the influene of misalignment and undercut on the formability....... The quality is evalutated by measuring the imit strain and the limit effective strain for the laser welded sheets and the base material. These strains will be presented in a forming limit diagram (FLD). Finally the formability of the laser sheets is compared to that of the base materials....

Effect of carrier on labeling and biodistribution of Re-188-Hydroxyethylidene diphosphonate

International Nuclear Information System (INIS)

Chang, Young Soo; Jeong, Jae Min; Kim, Bo Kwang; Cho, Jung Hyuk; Lee, Dong Soo; Chung, June Key; Lee, Myung Chul; Lee, Seung Jin; Jin, Ren Jie; Lee, Sang Eun

2000-01-01

Re-188-Hydroxyethylidene diphosphonate (HEDP) is a new cost-effective agent for systemic radioisotope therapy of metastatic bone pain. We investigated the influence of carrier for labeling and biodistribution of Re-188-HEDP using HEDP kit with or without carrier (KReO 4 ). The kits (HEDP 15 mg, gentisic acid 4 mg and SnC1 2 .2H 2 O 4.5 mg) with or without carrier (KReO 4 0.1 mg) were labeled with Re-188 solution, made available from an in-house generator by boiling for 15 min. We compared the labeling efficiency and stability of carrier-added and carrier-free preparations of Re-188-HEDP. Biodistribution and imaging studies of each preparation were performed in ICR mice (1.85-3.7 MBq/0.1 ml) and SD rats (74.1-85.2 MBq/0.5 ml). The carrier-added preparation showed high labeling efficiency (95% at pH 5) and high stability in serum (88%, 3hr). However, the carrier-free preparation showed low labeling efficiency (59% at pH 5) and low stability (43%, 3 hr). The carrier-added preparation showed high uptake in bone and low uptake in stomach and kidneys. However, the carrier-free preparation showed lower uptake in bone and higher uptake in both stomach and kidneys, which is supposed to be due to released perrhenate. The carrier-added preparation also showed better images with higher skeletal accumulation, lower uptake in other organs and lower soft tissue uptake than the carrier-free preparation. The results of these studies clearly demonstrate that addition of carrier perrhenate is required for high labeling efficiency, stability, bone uptake and good image quality of Re-188-HEDP.=20

Decoloration studies of some fluorescent dye solutions

International Nuclear Information System (INIS)

Zafar-uz-Zaman, M.; Ditta, A.

1997-01-01

Rhenium-186-(Sn)-l, l hydroxy ethylene diphosphonate (/sup 186/Re-HEDP) has been used for the palliation of metastatic bone pain. /sup 186/Re- has excellent physical properties that may be useful for the formulation of radiotherapeutic agents. It has a short half-life (90.6 hrs) with moderate energy particles (E /sub max/=1.07 MeV) that penetrate over a short range of tissue and gamma ray of 137 keV which is well suited to image. A number of samples of natural rhenium (metal) power were irradiated in PARR-I research reactor at a thermal neutrons flux of the order lx10/sup 14/ n.cm /sup -2/.s/sup -1/ for various time intervals in order to optimize the production yield of /sup 186/Re. The data indicated that 60 mCi/mg of radioactivity could be obtained for an irradiation time of 24 hours. The irradiated target was converted to its ammonium salt which was used for preparation of /sup 186/Re-HEDP complex. Labeling studies of dissolution salt of HEDP with /sup 186/Re were performed by varying the amounts of rhenium, HEDP and Sn. These studies were also carried out at different pH of the solutions. The quality control of /sup 186/Re-HEDP complex was checked by radio chromatographic techniques. These investigations indicated that the complex of optimum yield (approx. 95%) could be obtained by using amounts of Re (0.15 mg), HEDP (10 mg), stannous chloride dihydrate (4mg) and pH range of 4-6. The effect of antioxidant genetic acid was studied on the stability of the complex which was found to be stable up to five days in the presence of 3 mg of genetic acid. The biodistribution studies in rats showed optimum uptake by bone after 2.5 hours. (author)

Consequences of the improvement of fast reactor material behavior under irradiation on fuel element performance

International Nuclear Information System (INIS)

Leclere, J.; Dupouy, J.M.; Marcon, J.P.

1979-01-01

The most important problems in fast reactor fuel element come from the excessive swelling of the structural materials used. The limitations of irradiation time for a given reactor result from the cladding or hexagonal wrapper deformations. Irradiation creep plays a major role, either in inducing additional deformations, or in providing possible ways of accommodation of bending stresses. Progress has been made in designing swelling resistant and/or low irradiation creep modulus materials. For instance in FRANCE, annealed 316 SS has been eliminated from pin and subassembly, and replaced by cold worked 316; we are now considering introduction of stabilizing elements in 316 SS as a further improvement and studying different alloys (nickel alloys, or ferritic steels). It has to be checked that the improvement of irradiation characteristic is not counterbalanced by losses on other properties (embrittlement for instance). Considering that pushing off or eliminating a limit may lead to the onset of a new one, it is porposed to make a review of the consequences of substantial improvement of structural material behavior

Flowing and freezing of molten core materials during unprotected loss of flow accidents in sodium cooled fast reactors

International Nuclear Information System (INIS)

Maschek, W.; Royl, P.

1988-09-01

Flowing and freezing of mobile core materials change the fissile material distribution and core-inventory under hypothetical accident conditions and determine the path to permanent shutdown of the neutronic events and the energetic potentials. The report classifies the bondary conditions for such flowing and freezing processes by going through the different situations under which these processes can occur in the scenario of the unprotected loss of flow (ULOF) accident. The classification is based on ULOF-accident simulations for a homogeneous reactor core concept of a 300 MWe LMFBR (e. g. SNR-300), but many boundary conditions are also characteristic for other core designs. A review of the relevant experiments is then made to correlate the available experimental information with these classified boundary conditions and to look at the resulting flowing and freezing processes. Boundary conditions that have been experimentally shown to be important are assigned high priorities. The data are specifically valued in relation to these boundary conditions of high priorities. The review includes the major experimental programs with published results. The discussion shows that the results from most clean condition tests for melt relocations are valuable for a better understanding of basic phenomena and analytical model development, but are not directly applicable to real accident conditions. The database for relevant boundary conditions from the ULOF scenario is limited and largely included in integral sequence tests from which quantitative information for modelling is difficult to obtain. Needs for additional investigations are identified. The suggestions are mainly restricted to investigations of the early phase of fuel removal. They are given with reference to candidate facilities and include relocations in the subassemblies and in the inter-subassembly gaps. Particular emphasis is put on the leading edge properties and possible driving forces to which more attention

Diamond and Diamond-Like Materials as Hydrogen Isotope Barriers

International Nuclear Information System (INIS)

Foreman, L.R.; Barbero, R.S.; Carroll, D.W.; Archuleta, T.; Baker, J.; Devlin, D.; Duke, J.; Loemier, D.; Trukla, M.

1999-01-01

This is the final report of a two-year, Laboratory Directed Research and Development (LDRD) project at Los Alamos National Laboratory (LANL). The purpose of this project was to develop diamond and diamond-like thin-films as hydrogen isotope permeation barriers. Hydrogen embrittlement limits the life of boost systems which otherwise might be increased to 25 years with a successful non-reactive barrier. Applications in tritium processing such as bottle filling processes, tritium recovery processes, and target filling processes could benefit from an effective barrier. Diamond-like films used for low permeability shells for ICF and HEDP targets were also investigated. Unacceptable high permeabilities for hydrogen were obtained for plasma-CVD diamond-like-carbon films

Complexation of 188Re-phosphonates: in vitro and in vivo studies

International Nuclear Information System (INIS)

Faintuch, B.L.; Muramoto, E.; Faintuch, S.

2003-01-01

MDP (methylenediphosphonate) and HEDP (hydroxyethylidene diphosphonate), both disphosphonates, and EDTMP (ethylenediamine tetramethylene phosphonic acid), a tetraphosphonate ligand, have been previously labeled with 188 Re for use in metastatic bone-pain palliation. The aim of this study was a comparison between the three complexes 188 Re-MDP, 188 Re-HEDP and 188 Re-EDTMP concerning the complexation conditions, in order to achieve maximum yield, stability and bone uptake. Methods: MDP was dissolved in water and HEDP and EDTMP were dissolved in NaOH 1 N followed by reduction of pH with HCl 1 N. To all mixtures stannous chloride and 188 ReO 4 - were added in a nitrogen atmosphere. The preparations were heated in boiling water bath for 15 min. Yield as well as radiochemical stability was estimated by ITLC. Different concentrations of phosphonates and stannous chloride were evaluated. Biodistribution studies in Swiss mice were done for the three 188 Re-phosphonates that presented the best radiochemical yield. The optimal ligand concentration for maximum complexation was 85.2 mM for MDP, 72.8 mM for HEDP and 45.8 mM for EDTMP. The best amount of SnCl 2 .2H 2 O was 1.5 mg/mL for 188 Re-HEDP and 1 mg/mL for both 188 Re-MDP and 188 Re-EDTMP. In these conditions the three complexes showed a complexation rate above 95%. Reasonable radiochemical stability for 24 hours was achieved by 188 Re-EDTMP when employing ascorbic acid. All products showed a great uptake by the kidneys. 188 Re-EDTMP had the greatest uptake by femur (3.1 ± 0.2% ID/g) followed by 188 Re-MDP (1.2 ± 0.1% ID/g) and 188 Re-HEDP (1.0 ± 0.1% ID/g), 4 hours post injection. 188 Re-EDTMP displayed a femur bone/muscle ratio of 28.5, 188 Re-MDP 4.9 and 188 Re-HEDP 4.9. In conclusion 188 Re-EDTMP demonstrated the best potential as a radiopharmaceutical for bone cancer pain relief, encouraging further dosimetric studies and clinical trials. (orig.)

Bone uptake by di and tetraphosphonates labeled with Rhenium-188

International Nuclear Information System (INIS)

Faintuch, B.L.; Osso, J.A. Jr.; Muramoto, E.; Faintuch, S.

2002-01-01

MDP (methylenediphosphonate) and HEDP (hydroxyethylidenediphosphonate), both diphosphonates, and EDTMP (ethylenediamine tetramethylene phosphonic acid) a tetraphosphonate ligand, have been labeled with 188 Re for use in metastatic bone-pain palliation. The aim of this study was a comparison between the three complexes 188 Re-MDP, 188 Re-HEDP and 188 Re-EDTMP concerning the complexation conditions, in order to achieve maximum yield, stability and bone uptake. Methods: MDP was dissolved in water and HEDP and EDTMP were dissolved in NaOH 1N followed by decreasing pH with HCl 1N. To all mixtures stannous chloride and 188 ReO 4 were added in a nitrogen atmosphere. The preparations were heated in a boiling water bath for 15 min. The yields as well as the radiochemical stability were estimated by ITLC. Different concentrations of phosphonates and stannous chloride were evaluated. Biodistribution studies in swiss mice were done for the three 188 Re-phosphonates that presented the best radiochemical yield. Results: For 188 Re-MDP and 188 Re-HEDP the optimal ligand concentration for maximum complexation was 30 mg whereas for 188 Re-EDTMP, it was 40 mg. The best amount of SnCl 2 .2H 2 O was 2 mg/mL for MDP, 3 mg/mL for HEDP and 1 mg/mL for EDTMP. In these conditions the three complexes showed a complexation yield above 95%. All of them presented 4-hour radiochemical stability without the need for ascorbic acid solution, but for 24 hours this stability existed only in the presence of that substance otherwise re-oxidation of 188 Re occurred. All products showed a great uptake by the kidneys. 188 Re-EDTMP had the greatest uptake by the bone (3.13 ± 0.18% ID/g) followed by 188 Re-MDP (1.18 ± 0.05%ID/g) and 188 Re-HEDP (1.03 ± 0.12 %ID/g), 4 hour postinjection. 188 Re-EDTMP displayed a bone/muscle ratio of 28.5, 188 Re-MDP 4.9 and 188 Re-HEDP 4.9. Conclusion: 188 Re-EDTMP demonstrated the best potential as a radiopharmaceutical for bone cancer pain relief, encouraging further

The use of boiling noise detection as a protection against faults in sub-assemblies in LMFBRs. Status report of work in the United Kingdom

International Nuclear Information System (INIS)

Burton, E.J.; MacLeod, I.D.

1982-01-01

The development of acoustic techniques for the surveillance of LMFBRs has the objective of providing a monitoring system on-line to give an early warning of incipient failures whilst the reactor is at power at present in the UK. Most attention is being given to safety protection to meet the design proposals for the Commercial Demonstration Fast Reactor (CDFR). One concern in the safety analysis is the hypothetical possibility that a local fault in a subassembly, if undetected could spread to its neighbours, eventually involving the whole core. An early warning of such a potentially propagating event would be given by detecting the boiling of the sodium. The specification of the acoustic technique, and therefore of the development programme, is set by the requirements of the safety analysis and the important features are outlined in the first section of the paper. This is followed by a description of the signal strength from boiling, based on out-of-pile experiments. This signal hat to be discriminated against the background noise arising from thc coolant pumps and the subassembly gag and flow noise. The detection of the acoustic signal may now be made by transducers rather than waveguides provided that the transducers are shown to be reliable enough and the recent work is summarised in the next section. The estimate of the signal/noise ratio depends upon the. transmission of the acoustic waves through the core to the sensor position. There is little experience on transmission in the reactor environment, possibilities for experiments are limited and laboratory tests are being used to improve basic knowledge. Modern computers offer the possibility of improving the sensitivity of detection by advanced data processing and the techniques which are being pursued are briefly described. Although acoustic technology has made great improvements in the last decade, especially in the application of acoustic emission techniques in thermal reactors, there is no experience of the

The use of boiling noise detection as a protection against faults in sub-assemblies in LMFBRs. Status report of work in the United Kingdom

Energy Technology Data Exchange (ETDEWEB)

Burton, E J; MacLeod, I D [United Kingdom Atomic Energy Authority, Risley Nuclear Power Development Laboratories, Risley, Warrington (United Kingdom)

1982-01-01

The development of acoustic techniques for the surveillance of LMFBRs has the objective of providing a monitoring system on-line to give an early warning of incipient failures whilst the reactor is at power at present in the UK. Most attention is being given to safety protection to meet the design proposals for the Commercial Demonstration Fast Reactor (CDFR). One concern in the safety analysis is the hypothetical possibility that a local fault in a subassembly, if undetected could spread to its neighbours, eventually involving the whole core. An early warning of such a potentially propagating event would be given by detecting the boiling of the sodium. The specification of the acoustic technique, and therefore of the development programme, is set by the requirements of the safety analysis and the important features are outlined in the first section of the paper. This is followed by a description of the signal strength from boiling, based on out-of-pile experiments. This signal had to be discriminated against the background noise arising from thc coolant pumps and the subassembly gag and flow noise. The detection of the acoustic signal may now be made by transducers rather than waveguides provided that the transducers are shown to be reliable enough and the recent work is summarised in the next section. The estimate of the signal/noise ratio depends upon the transmission of the acoustic waves through the core to the sensor position. There is little experience on transmission in the reactor environment, possibilities for experiments are limited and laboratory tests are being used to improve basic knowledge. Modern computers offer the possibility of improving the sensitivity of detection by advanced data processing and the techniques which are being pursued are briefly described. Although acoustic technology has made great improvements in the last decade, especially in the application of acoustic emission techniques in thermal reactors, there is no experience of the

A conceptual design and structural stabilities of in-pit assembly tools for the completion of final sector assembly at tokamak hall

International Nuclear Information System (INIS)

Nam, K.O.; Park, H.K.; Kim, D.J.; Ahn, H.J.; Kim, K.K.; Im, K.; Shaw, R.

2010-01-01

The final assembly of main components of the International Thermonuclear Experimental Reactor (ITER) tokamak, Vacuum Vessel (VV) and Toroidal Field Coils (TFCs), is achieved by the sequential assembly of the nine sub-assembled 40 o sectors in tokamak pit. Each sub-assembled 40 o sector is composed of one VV 40 o sector, two TFCs, and in-between Vacuum Vessel Thermal Shield (VVTS) segments. Sub-assembly is carried out in the assembly building and then the sub-assembled sectors are transferred into tokamak pit, in sequence, to complete sector assembly. The role of in-pit assembly tool is to support and align the sub-assembled sectors in tokamak pit. It also plays the role of reference datum during assembly until the completion of main components assembly. Korea Domestic Agency (KO DA) has developed the conceptual design of most ITER purpose-built assembly tools under the collaboration with the ITER Organization. Among the conceptual designs carried out, this paper describes the function, the structure, the selected material and the design results of the in-pit assembly tools comprising central column, radial beams and their supports, TF inner supports and in-pit working floor. The results of structural analysis using ANSYS for the various loading cases are given as well. The resultant stresses and deflections turned out to fall within the allowable ranges.

Determination of radiochemical yields of 186Re-labelled complexes using thin layer chromatography

International Nuclear Information System (INIS)

Konirova, R.; Kohlickova, M.; Jedinakova-Krizova, V.

1999-01-01

The reaction conditions for synthesis of three rhenium complexes 186 Re-methylendiphosphonate (MDP), 186 Re-hydroxyethylidendiphosphonate (HEDP) and 186 Re-citrate have been investigated. Radiochemical yield of complexation has been determined by thin layer chromatography and paper chromatography. The rhenium complexation with corresponding ligand is dependent on pH values of reaction mixture, concentration of studied ligand (MDP, HEDP and sodium citrate) and concentration of reducing agent. Stannous chloride with ascorbic acid (as antioxidant) was used for reduction of perrhenate. The labeling yield of 186 Re-MDP was about 90 %, of 186 Re-HEDP more than 80 % and more than 75 % for 186 Re-citrate under optimum conditions. Besides, the possibility of application of porphyrins as organic ligands for complexation with rhenium isotopes is examined. (authors)

Preparation of 186Re complexes of dimercaptosuccinic acid hydroxy ethylidine diphosphonate

International Nuclear Information System (INIS)

Kothari, K.; Pillai, M.R.A.; Unni, P.R.; Mathakar, A.R.; Shimpi, H.H.; Noronha, O.P.D.; Samuel, A.M.

1998-01-01

99m Tc(V)-DMSA and 99m Tc-HEDP are widely used for imaging medullary carcinoma and bone, respectively. 186 Re-HEDP is now well established as a therapeutic radiopharmaceutical for palliation of pain due to bone metastases. It is expected that 186/188 Re(V)-DMSA could find application for treating medullary carcinoma. In the present paper we report the work carried out for the preparation of 186 Re complexes of DMSA and HEDP and their bio-distribution studies in Wistar rats. 186 Re was prepared by irradiation of natural Re metal at a flux of 3x10 13 neutrons/cm 2 /s for seven days and processed after a cooling period of four days. The specific activity of 186 Re formed was ∼35 mCi/mg. Complexes with RC purity >98% could be prepared in both the cases by carefully optimizing the reaction conditions. Bio-distribution studies carried out in rats revealed that pharmacological behaviour of 186 Re(V)-DMSA was similar to that of 99m Tc(V)-DMSA. 186 Re-HEDP showed a bone uptake of ∼ 30% at 3 h post injection which remained almost constant for 48 h. (author)

Correlation of creep and swelling with fuel pin performance

International Nuclear Information System (INIS)

Jackson, R.J.; Washburn, D.F.; Garner, F.A.; Gilbert, E.R.

1975-09-01

The HEDL PNL-11 experiment described was one in a series of fueled subassemblies irradiated in EBR-II to demonstrate the adequacy of the FFTF fuel pin design. The cladding material, dimensions, and fuel density are prototypic of FFTF. Because neutron flux in EBR-II is lower than in FFTF, the uranium enrichment is higher in these experimental fuel pins, irradiated in EBR-II, than the FFTF enrichment for comparable linear heat rates. Some pertinent oprating conditions for the center fuel pin in this experiment are listed. This 37-pin subassembly represents, at 110,000 MWd/MTM, the highest burnup yet attained by a prototypic FFTF subassembly. Similarly, this is the highest fluence presently attained by prototypic fuel pins. A cladding breach occurred in one fuel pin which is presently being examined. Results are presented and discussed

Environmental Assessment for DOE permission for off-loading activities to support the movement of Millstone Unit 2 steam generator sub-assemblies across the Savannah River Site

International Nuclear Information System (INIS)

1992-10-01

The Department of Energy (DOE) has prepared an Environmental Assessment (EA), for the proposed granting of DOE permission of offloading activities to support the movement Millstone Unit 2 steam generator sub-assemblies (SGSAs) across the Savannah River Site (SRS). Based on the analyses in the EA, DOE has determined that the proposed action is not a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, an environmental impact statement is not required, and the Department is issuing this Finding of No Significant Impact. On the basis of the floodplain/wetlands assessment in the EA, DOE has determined that there is no practicable alternative to the proposed activities and that the proposed action has been designed to minimize potential harm to or within the floodplain of the SRS boat ramp. No wetlands on SRS would be affected by the proposed action

Interactions between the Tetrasodium Salts of EDTA and 1-Hydroxyethane 1,1-Diphosphonic Acid with Sodium Hypochlorite Irrigants.

Science.gov (United States)

Biel, Philippe; Mohn, Dirk; Attin, Thomas; Zehnder, Matthias

2017-04-01

A clinically useful all-in-one endodontic irrigant with combined proteolytic and decalcifying properties is still elusive. In this study, the chemical effects of dissolving the tetrasodium salts of 1-hydroxyethane 1,1-diphosphonic acid (Na 4 HEDP) or Na 4 EDTA directly in sodium hypochlorite (NaOCl) irrigants in polypropylene syringes were assessed during the course of 1 hour. The solubility of the salts in water was determined. Their compatibility with 1% and 5% NaOCl was measured by iodometric titration and in a calcium complexation experiment by using a Ca 2+ -selective electrode. The salts dissolved within 1 minute. The dissolution maximum of Na 4 HEDP in water (wt/total wt) was 44.6% ± 1.6%. The corresponding dissolution maximum of Na 4 EDTA was 38.2% ± 0.8%. Na 4 HEDP at 18% in 5% NaOCl caused a mere loss of 16% of the initially available chlorine during 1 hour. In contrast, a corresponding mixture between NaOCl and the Na 4 EDTA salt caused 95% reduction in available chlorine after 1 minute. Mixtures of 3% Na 4 EDTA with 1% NaOCl were more stable, but only for 30 minutes. Na 4 HEDP lost 24% of its calcium complexation capacity after 60 minutes. The corresponding loss for Na 4 EDTA was 34%. The compatibility and solubility of particulate Na 4 HEDP with/in NaOCl solutions are such that these components can be mixed and used for up to 1 hour. In contrast, short-term compatibility of the Na 4 EDTA salt with NaOCl solutions was considerably lower, decreasing at higher concentrations of either compound. Especially for Na 4 HEDP but also for Na 4 EDTA, the NaOCl had little effect on calcium complexation. Copyright © 2017 American Association of Endodontists. Published by Elsevier Inc. All rights reserved.

A case of industrial safety appraisal for extension of service life of GTK-10-4 gas turbines used at gas transmission stations

Science.gov (United States)

Rybnikov, A. I.; Kovalev, A. G.; Kryukov, I. I.; Leont'ev, S. A.; Moshnikov, A. V.

2017-04-01

It is shown that the extended life and enhanced operational reliability of parts and subassemblies of the most popular GTK-10-4 gas transmission plants are determined by the enhanced efficiency of the control over technical condition and operational safety of turbine plants in conformity with industrial safety requirements imposed on gas pipeline compressor stations. It has been established that the materials of parts and subassemblies of gas turbine plants with different, especially with maximal operating time, shall be exposed to NDT for the purpose of determining the actual mechanical characteristics of these materials with different operating time and calculating residual life. The analysis of damageability and operating conditions has helped to identify parts and subassemblies for repair or replacement with the highest frequency of unacceptable defects. These parts and subassemblies have been shown to include base members of the axial compressor (AC), a turbine housing, an axial compressor rotor, high- and low-pressure turbine (HPT and LPT) discs, a 12-part holder, the housing of the holder of HPT and LPT guiding blades, a sealed baffler, and working and guiding AC, LPT and HPT blades. The most typical operational defects have been enumerated and analyzed. It has been determined that the primary task of the industrial safety appraisal for extending the life of GTK-10-4 with limit-exceeding operating time is to thoroughly examine HPT and LPT discs with more than 130,000 hours of operating time and establish by DT methods characteristics of materials for evaluation, taking account of their degradation, and residual life of critical turbine elements. In addition, it has been shown that the service life of HP turbine discs can be extended by replacing the disc material (EP-428 12% chromium steel) with a material with a higher linear expansion factor that somewhat exceeds the expansion factor of EI-893 nickel alloy used to melt out working blades.

MARTINS: A foam/film flow model for molten material relocation in HWRs with U-Al-fueled multi-tube assemblies

International Nuclear Information System (INIS)

Kalimullah.

1994-01-01

Some special purpose heavy-water reactors (EM) are made of assemblies consisting of a number of coaxial aluminum-clad U-Al alloy fuel tubes and an outer Al sleeve surrounding the fuel tubes. The heavy water coolant flows in the annular gaps between the circular tubes. Analysis of severe accidents in such reactors requires a model for predicting the behavior of the fuel tubes as they melt and disrupt. This paper describes a detailed, mechanistic model for fuel tube heatup, melting, freezing, and molten material relocation, called MARTINS (Melting and Relocation of Tubes in Nuclear subassembly). The paper presents the modeling of the phenomena in MARTINS, and an application of the model to analysis of a reactivity insertion accident. Some models are being developed to compute gradual downward relocation of molten material at decay-heat power levels via candling along intact tubes, neglecting coolant vapor hydrodynamic forces on molten material. These models are inadequate for high power accident sequences involving significant hydrodynamic forces. These forces are included in MARTINS

Preparation and quality control of 186Re compounds

International Nuclear Information System (INIS)

Noto, M.G.; Amor, R.A.; Caviglia, D.A.; Ratner, M.T.; Schroder, A.M.; Rocco, J.C.; Mancini, A.C.

1987-01-01

The optimal conditions were investigated in order to label the methylendiphosphonate (MDP), hydroxyethylendiphosphonate (HEDP), pyrophosphonate (PYP) and ethylendiaminotetramethylenphosphoric (EDTMP) with 186 Re. The biodistribution of these compounds in experimental animals were studied to determine the most suitable therapeutic agent for its eventual use as pain palliative in patients with bone metastases. The biodistribution assays were performed in Wistar rats, and the 186 Re HEDP was finally chosen. (M.E.L.) [es

Study of low cost eco-friendly compounds as corrosion inhibitors for cooling systems

Energy Technology Data Exchange (ETDEWEB)

Farooqi, I H; Hussain, A; Saini, P A [AMU, Aligarh (India). Dept. of Civil Engineering; Quraishi, M A [AMU, Aligarh (India). Dept. of Applied Chemistry

1999-07-01

Attempts are made to utilize the aqueous extracts of natural compounds, namely cordia latifolia and curcumin, as corrosion inhibitors for mild steel in cooling systems, and their inhibition efficiencies are compared with that of Hydroxyethylidene 1-1 diphosphonic acid (HEDP). HEDP is also blended with aqueous extracts of natural compounds so as to improve their inhibition efficiency. The blowdown of the cooling system is also analysed for environmental factors. (author)

National Ignition Facility quality assurance plan for laser materials and optical technology

Energy Technology Data Exchange (ETDEWEB)

Wolfe, C.R.

1996-05-01

Quality achievement is the responsibility of the line organizations of the National Ignition Facility (NIF) Project. This subtier Quality Assurance Plan (QAP) applies to activities of the Laser Materials & Optical Technology (LM&OT) organization and its subcontractors. It responds to the NIF Quality Assurance Program Plan (QAPP, L-15958-2, NIF-95-499) and Department of Energy (DOE) Order 5700.6C. This Plan is organized according to 10 Quality Assurance (QA) criteria and subelements of a management system as outlined in the NIF QAPP. This Plan describes how those QA requirements are met. This Plan is authorized by the Associate Project Leader for the LM&OT organization, who has assigned responsibility to the Optics QA engineer to maintain this plan, with the assistance of the NIF QA organization. This Plan governs quality-affecting activities associated with: design; procurement; fabrication; testing and acceptance; handling and storage; and installation of NIF Project optical components into mounts and subassemblies.

National Ignition Facility quality assurance plan for laser materials and optical technology

International Nuclear Information System (INIS)

Wolfe, C.R.

1996-05-01

Quality achievement is the responsibility of the line organizations of the National Ignition Facility (NIF) Project. This subtier Quality Assurance Plan (QAP) applies to activities of the Laser Materials ampersand Optical Technology (LM ampersand OT) organization and its subcontractors. It responds to the NIF Quality Assurance Program Plan (QAPP, L-15958-2, NIF-95-499) and Department of Energy (DOE) Order 5700.6C. This Plan is organized according to 10 Quality Assurance (QA) criteria and subelements of a management system as outlined in the NIF QAPP. This Plan describes how those QA requirements are met. This Plan is authorized by the Associate Project Leader for the LM ampersand OT organization, who has assigned responsibility to the Optics QA engineer to maintain this plan, with the assistance of the NIF QA organization. This Plan governs quality-affecting activities associated with: design; procurement; fabrication; testing and acceptance; handling and storage; and installation of NIF Project optical components into mounts and subassemblies

Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

International Nuclear Information System (INIS)

2005-01-01

This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

EFSA BIOHAZ Panel (EFSA Panel on Biological Hazards), 2014. Scientific Opinion on the evaluation of the safety and efficacy of peroxyacetic acid solutions for reduction of pathogens on poultry carcasses and meat

DEFF Research Database (Denmark)

Hald, Tine; Baggesen, Dorte Lau

no concerns for environmental risk of peroxyacids, acetic acid and octanoic acid. On the basis of a conservative preliminary guideline for surface water quality, the emission of HEDP from a poultry plant into the environment could not be considered safe a priori. It was recommended that HACCP plans should...... include monitoring of the concentration of HEDP and of the decontaminating substance in the working solution and post-marketing surveillance for resistance in both pathogenic and commensal bacteria....

Study of the action of a phosphonate additive on steel scale deposit and corrosion in the hydrodynamic conditions of a channel flow cell; Etude de l'action d'un additif phosphone sur l'entartrage et sur la corrosion de l'acier dans les conditions hydrodynamiques d'une cellule a canal

Energy Technology Data Exchange (ETDEWEB)

Garcia, C.

2000-10-17

In cooling systems, an improved control of scale deposit and corrosion processes is a major challenge and an realistic evaluation tool for water treatments is of the utmost economic importance. In this study, a channel flow cell was used to allow in-situ electrochemical measurements in well defined electrolyte tube flowing conditions. An expression of the mass transfer towards the electrode was established where the diffusion-limited current is a function of Re{sup 1/3} in the laminar regime and was verified experimentally using the redox couples Fe[CN]{sub 6}{sup 4-}/ Fe[CN]{sub 6}{sup 3-} and O{sub 2}/OH{sup -}. This hydrodynamically controlled experimental device was developed to investigate scale deposit processes and to evaluate scale inhibitor efficiency using a electrochemical quartz crystal microbalance. Experiments were performed on three different waters, at various flow rates and temperatures. The efficiency of a well known phosphonate (HEDP) was tested at different concentrations and an optimum concentration could be established (0.7 mg dm{sup -3}). The effect of additive injection during the scale formation as well as the influence of flow rate on the inhibiting efficiency were evaluated. The anti-scale additive was shown to be more effective in the turbulent regime. HEDP has shown a strong effect on inhibiting crystal growth and that affected the morphology of CaCO{sub 3} crystals. The HEDP effect on protecting carbon steel against corrosion was also studied in mineral water containing Ca{sup 2+} ions. It was found that anti-corrosion effect of HEDP is enhanced by the presence of calcium in solution and that is due to the formation of an HEDP-Ca{sup 2+} complex, which adsorbs onto the metallic surface and protects it from dissolution. (author)

76 FR 45845 - Notice of Issuance of Final Determination Concerning a Certain Patient Transport Chair

Science.gov (United States)

2011-08-01

... approximately 481 components. All of the components are of U.S., Chinese, Canadian, or French origin. The... (which includes a French-origin handle circuit board, a control box, a key switch subassembly, and a... battery cable subassemblies, a handle cable subassembly, an emergency stop switch subassembly, a horn...

Increasing the magnetic-field capability of the magneto-inertial fusion electrical discharge system using an inductively coupled coil

Science.gov (United States)

Barnak, D. H.; Davies, J. R.; Fiksel, G.; Chang, P.-Y.; Zabir, E.; Betti, R.

2018-03-01

Magnetized high energy density physics (HEDP) is a very active and relatively unexplored field that has applications in inertial confinement fusion, astrophysical plasma science, and basic plasma physics. A self-contained device, the Magneto-Inertial Fusion Electrical Discharge System, MIFEDS [G. Fiksel et al., Rev. Sci. Instrum. 86, 016105 (2015)], was developed at the Laboratory for Laser Energetics to conduct magnetized HEDP experiments on both the OMEGA [T. R. Boehly et al., Opt. Commun. 133, 495-506 (1997)] and OMEGA EP [J. H. Kelly et al., J. Phys. IV France 133, 75 (2006) and L. J. Waxer et al., Opt. Photonics News 16, 30 (2005)] laser systems. Extremely high magnetic fields are a necessity for magnetized HEDP, and the need for stronger magnetic fields continues to drive the redevelopment of the MIFEDS device. It is proposed in this paper that a magnetic coil that is inductively coupled rather than directly connecting to the MIFEDS device can increase the overall strength of the magnetic field for HEDP experiments by increasing the efficiency of energy transfer while decreasing the effective magnetized volume. A brief explanation of the energy delivery of the MIFEDS device illustrates the benefit of inductive coupling and is compared to that of direct connection for varying coil size and geometry. A prototype was then constructed to demonstrate a 7-fold increase in energy delivery using inductive coupling.

Natural compounds as corrosion inhibitors for highly cycled systems

Energy Technology Data Exchange (ETDEWEB)

Quraishi, M.A.; Farooqi, I.H.; Saini, P.A. [Corrosion Research Lab., Aligarh (India)

1999-11-01

Strict environmental legislations have led to the development of green inhibitors in recent years. In continuation of the authors` research work on development of green inhibitors, they have investigated the aqueous extracts of three plants namely: Azadirachta indica, Punica Granatum and Momordica charantia as corrosion inhibitors for mild steel in 3% NaCl using weight loss and electrochemical methods. All the investigated compounds exhibited excellent corrosion inhibition properties comparable to that of HEDP. Azadirachta showed better scale inhibition effect than HEDP.

INVESTIGATION OF THE PRESENCE OF DRUGSTORE BEETLES WITHIN CELOTEX ASSEMBLIES IN RADIOACTIVE MATERIAL PACKAGINGS

Energy Technology Data Exchange (ETDEWEB)

Loftin, B; Glenn Abramczyk, G

2008-06-04

During normal operations at the Department of Energy's Hanford Site in Hanford, WA, drugstore beetles, (Stegobium paniceum (L.) Coleoptera: Anobiidae), were found within the fiberboard subassemblies of two 9975 Shipping Packages. Initial indications were that the beetles were feeding on the Celotex{trademark} assemblies within the package. Celotex{trademark} fiberboard is used in numerous radioactive material packages serving as both a thermal insulator and an impact absorber for both normal conditions of transport and hypothetical accident conditions. The Department of Energy's Packaging Certification Program (EM-63) directed a thorough investigation to determine if the drugstore beetles were causing damage that would be detrimental to the safety performance of the Celotex{trademark}. The Savannah River National Laboratory is conducting the investigation with entomological expertise provided by Clemson University. The two empty 9975 shipping packages were transferred to the Savannah River National Laboratory in the fall of 2007. This paper will provide details and results of the ongoing investigation.

Structural integrity testing of glass-ceramic/molybdenum vacuum tube frames

International Nuclear Information System (INIS)

Spears, R.K.

1980-01-01

In this study, vacuum tube subassemblies made of glass-ceramic insulators sealed to inner and outer molybdenum frames were loaded in compression to failure with a tensile test machine. Several factors were varied in processing these subassemblies. These factors included etching and nonetching of molybdenum piece parts, annealing and nonannealing of subassemblies, and vapor and non-vapor honing of insulators after sealing. After failure, the subassemblies were examined for fracture patterns. In most cases, fracture started at points near the lower portion of the inner sleeve-insulator interface. More load was carried by subassemblies having molybdenum piece parts that were acid etched. No difference appeared between the strength of subassemblies having annealed and nonannealed glass-ceramic insulators. Parts with vapor-honed insulators failed at substantially lower loads

Managing traceability information in manufacture

NARCIS (Netherlands)

Jansen-Vullers, M.H.; Dorp, van C.A.; Beulens, A.J.M.

2003-01-01

In this paper, an approach to design information systems for traceability is proposed. The paper applies gozinto graph modelling for traceability of the goods flow. A gozinto graph represents a graphical listing of raw materials, parts, intermediates and subassemblies, which a process transforms

Managing traceability information manufacture

NARCIS (Netherlands)

Jansen-Vullers, M.H.; van Dorp, C.A.; Beulens, A.J.M.

2003-01-01

In this paper, an approach to design information systems for traceability is proposed. The paper applies gozinto graph modelling for traceability of the goods flow. A gozinto graph represents a graphical listing of raw materials, parts, intermediates and subassemblies, which a process transforms

Liquid metal cooled fast breeder nuclear reactors

International Nuclear Information System (INIS)

Thatcher, G.; Mitchell, A.J.

1981-01-01

Fuel sub-assemblies for liquid metal-cooled fast breeder reactors are described which each incorporate a fluid flow control valve for regulating the rate of flow through the sub-assembly. These small electro-magnetic valves seek to maintain the outlet coolant temperature of at least some of the breeder sub-assemblies substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (U.K.)

High and rapid hydrogen release from thermolysis of ammonia borane near PEM fuel cell operating temperature

Science.gov (United States)

Varma, Arvind; Hwang, Hyun Tae; Al-Kukhun, Ahmad

2016-11-15

A system for generating and purifying hydrogen. To generate hydrogen, the system includes inlets configured to receive a hydrogen carrier and an inert insulator, a mixing chamber configured to combine the hydrogen carrier and the inert insulator, a heat exchanger configured to apply heat to the mixture of hydrogen carrier and the inert insulator, wherein the applied heat results in the generation of hydrogen from the hydrogen carrier, and an outlet configured to release the generated hydrogen. To purify hydrogen, the system includes a primary inlet to receive a starting material and an ammonia filtration subassembly, which may include an absorption column configured to absorb the ammonia into water for providing purified hydrogen at a first purity level. The ammonia filtration subassembly may also include an adsorbent member configured to adsorb ammonia from the starting material into an adsorbent for providing purified hydrogen at a second purity level.

Design and manufacturing of the CFRP lightweight telescope structure

Science.gov (United States)

Stoeffler, Guenter; Kaindl, Rainer

2000-06-01

Design of earthbound telescopes is normally based on conventional steel constructions. Several years ago thermostable CFRP Telescope and reflector structures were developed and manufacturing for harsh terrestrial environments. The airborne SOFIA TA requires beyond thermostability an excessive stiffness to mass ratio for the structure fulfilling performance and not to exceed mass limitations by the aircraft Boeing 747 SP. Additional integration into A/C drives design of structure subassemblies. Thickness of CFRP Laminates, either filament wound or prepreg manufactured need special attention and techniques to gain high material quality according to aerospace requirements. Sequential shop assembly of the structure subassemblies minimizes risk for assembling TA. Design goals, optimization of layout and manufacturing techniques and results are presented.

Soldadura laser de sub-conjuntos para estampagem (Tailored blanks)

DEFF Research Database (Denmark)

Olsen, Flemming Ove

1998-01-01

Laser welding has an increasing role in the automotive industry, namely on the sub-assemblies manufacturing. Several sheet-shape parts are laser welded, on a dissimilar combination of thicknesses and materials, and are afterwards formed (stamped) being transformed in a vehicle body component. In ...

An assessment of core wide coherency effects in the multichannel modeling of the initiating phase of a severe accident in a sodium fast reactor

International Nuclear Information System (INIS)

Guyot, M.; Gubernatis, P.; Suteau, C.; Le Tellier, R.; Lecerf, J.

2014-01-01

To consolidate the safety assessment for liquid-metal fast breeder reactors (LMFBRs), hypothetical core disruptive accident (HCDA) sequences have been extensively studied over the past decades. Numerous analyses of the so called initiating phase (or primary phase) of a HCDA have been made with the safety analysis system code SAS4A. The SAS4A accident analysis code requires that subassemblies or groups of subassemblies be represented together as independent channels. For simulating a severe accident sequence, a subassembly-to-channel assignment procedure has to be implemented to produce the consistent SAS4A input decks. Generally, one uses imposed criteria over relevant reactor parameters to determine the subassembly to- channel arrangement. The multiple-assembly-per-channel approach introduces core wide coherency effects, which can affect the reactivity balance and therefore the overall accident development. In this paper, a subassembly-to channel assignment procedure based on the subassembly power-to-flow ratio is presented and implemented to generate the SAS4A input decks over a range of parameter values. The corresponding SAS4A calculations have been performed on a large LMFBR. The purpose of the present series of calculations is to investigate the magnitude of errors encountered in the analysis of the initiating phase related to the subassembly-to-channel arrangement selection, by comparison with a one-subassembly-per-channel reference solution. It appears that a refinement in the channel arrangement substantially reduces core wide coherency effects. Analysis of the calculations also suggests that an accurate representation of the scenario requires the number of channels to be on approximately the same order of magnitude as the total number of subassemblies. Numerical results are examined to provide the reader with quantitative measurements of bias related to subassembly to- channel arrangement. (authors)

Liquid kit for preparation of 188rhenium-etidronate

International Nuclear Information System (INIS)

Marczewski, Barbara; Dias, Carla Roberta; Moraes, Vanessa; Osso Junior, Joao Alberto

2005-01-01

The aim of this study was the preparation of a liquid kit for radiolabeling of 188 Re-HEDP (hydroxyethylidene diphosphonate). 188 Re was obtained from alumina based 188 W/ 188 Re generators. This paper reports the efficacy of a cold kit stored for more than two weeks, determined by the dependence of the radiolabeling yields of 188 Re-HEDP on the incubation time, reducing agent concentration, the effects of concentration of ligand, the p H of the reaction and the temperature. The cold kits showed a good stability when carrie-free rhenium-188 was added in the reaction mixture. (author)

Design and synthesis of new poly-phosphorylated upper-rim modified calix[4]arenes as potential and selective chelating agents of uranyl ion

International Nuclear Information System (INIS)

Migianu-Griffoni, E.; Mbemba, C.; Burgada, R.; Lecouvey, M.; Lecercle, D.; Taran, F.

2009-01-01

New upper-rim poly-phosphorylated calix[4]arenes were designed for decorporation of uranium in case of nuclear contamination. A ligand system containing four pre-organized 1-hydroxymethylene-1, 1-bisphosphonic acid moieties anchored onto a calix[4]arene platform has been developed. Three calix[4]-arene-bis-phosphonates were efficiently prepared in multi-step syntheses with a variable carbon chain length between the bis-phosphonate and the calix[4]arene. Affinity constants towards uranyl ion were determined and compared with those of bis(HEDP) and tris(HEDP) phosphonates, known as efficient ligands for uranyl. (authors)

A review of the United Kingdom fast reactor programme

International Nuclear Information System (INIS)

Bramman, J.I.; Hickey, H.B.; Wheeler, R.C.; Gregory, C.V.

1989-01-01

The total electricity generating capacity in the UK is approximately 54 GW. Total electricity generation in 1988 was 288 TW hours, of which just over 20% was nuclear. In Scotland the percentage of electricity generated by nuclear stations was 49% of the total, and will exceed 60% in 1989. The privatization of the Electricity Supply Industry (ESI) in the UK (mentioned in last year's report) is proceeding on schedule. Considerable efforts are being made to ensure that the maximum benefits will be obtained from operating the PFR during the next five years. The main thrust of the UKAEA's programme continues to be towards the requirements of the EFR. Reload 16 included the biennial maintenance and statutory inspection period. It was extended from its original 60 days by the need to carry out modifications aimed at improving the reliability of the protection systems designed to safeguard the components of the secondary circuit, including the IHXs, in the event of a sodium-water reaction in a steam generator unit, and by the need to inspect and repair the vessels of the steam generator units. Good progress was made with the fuel development programme. The leading experimental cluster of PE16-clad 6.6 mm diameter pins is continuing irradiation above 21% burnup and 150 dpa (NRT). The lead subassembly with 5.8 mm pins clad in PE16 has exceeded 17.6% burnup, 130 dpa (NRT). The leading subassembly with pins of the same type to have undergone complete PIE contained fuel at 16% burnup and PE16 clad at 116 dpa; these pins were found to be in very good condition. Radial blanket subassemblies have exceeded 2% burnup without failure. In 1988/89 there was one reprocessing campaign in the PFR Reprocessing Plant lasting from November 1988 to February 1989. Feed material included irradiated fuel from 12 subassemblies irradiated in the PFR, some unirradiated subassemblies and loose pins and residues; in all containing 1.3t of Heavy Metal (HM) containing 242 kg plutonium. The cumulative

Development of JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Tabuchi, Shiro; Aoyama, Takafumi

2000-01-01

The MK-II core of the experimental fast reactor JOYO served as the irradiation bed for testing fuels and materials for FBR development since 1982 for 15 years. During the MK-II operation, extensive data were accumulated from the core management calculations and characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database recorded on CD-ROM for user convenience. The calculated core management data are the text style data. The 'Configuration Data' include the history of the fuel exchange and core arrangement for each cycle. The Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of about 300 fuel subassemblies, and 60 irradiation subassemblies. The 'Output Data' include the neutron fluxes, gamma fluxes, power density, linear heat rates, coolant and fuel temperature distributions of each core position at the beginning and end of each cycle. The measured core characteristics data, such as the excess reactivity, control rod worths, temperature coefficient, power coefficient, and burn-up coefficient are also included along with the measurement conditions. (J.P.N.)

MENT reconstruction and potting comparison of a LMFBR fuel bundle

International Nuclear Information System (INIS)

Betten, P.R.; Tow, D.M.

1984-01-01

Since the advent of computer-assisted-tomography (CAT), the CAT techniques have been rapidly expanded to the nuclear industry. A number of investigators have applied these techniques to reconstruct the fuel bundle configuration inside a subassembly with various degrees of resolution; however, there has been little data available on the accuracy of these reconstructions, and no comparisons have been made with the internal structure of actual irradiated subassemblies. Some efforts have utilized pretest mock-ups to calibrate the CAT algorithms, but the resulting mock-up configurations do not necessarily represent an actual subassembly, so an exact comparison has been lacking. The purpose of this paper is to present the results of a comparison between a CAT reconstruction of an irradiated subassembly and the destructive examination of the same subassembly

Procedures and instrumentation for sodium boiling experiments in EBR-II

International Nuclear Information System (INIS)

Crowe, R.D.

1976-01-01

The development of instrumentation capable of detecting localized coolant boiling in a liquid metal cooled breeder reactor (LMFBR) has a high priority in fast reactor safety. The detection must be rapid enough to allow corrective action to be taken before significant damage occurs to the core. To develop and test a method of boiling detection, it is desirable to produce boiling in a reactor and thereby introduce a condition in the reactor the original design concepts were chosen to preclude. The proposed boiling experiments are designed to safely produce boiling in the subassembly of a fast reactor and provide the information to develop boiling detection instrumentation without core damage or safety compromise. The experiment consists of the operation of two separate subassemblies, first, a gamma heated boiling subassembly which produces non-typical but highly conservative boiling and then a fission heated subassembly which simulates a prototypical boiling event. The two boiling subassemblies are designed to operate in the instrumentation subassembly test facility (INSAT) of Experiment Breeder Reactor II

P3: An installation for high-energy density plasma physics and ultra-high intensity laser–matter interaction at ELI-Beamlines

Directory of Open Access Journals (Sweden)

S. Weber

2017-07-01

Full Text Available ELI-Beamlines (ELI-BL, one of the three pillars of the Extreme Light Infrastructure endeavour, will be in a unique position to perform research in high-energy-density-physics (HEDP, plasma physics and ultra-high intensity (UHI (>1022W/cm2 laser–plasma interaction. Recently the need for HED laboratory physics was identified and the P3 (plasma physics platform installation under construction in ELI-BL will be an answer. The ELI-BL 10 PW laser makes possible fundamental research topics from high-field physics to new extreme states of matter such as radiation-dominated ones, high-pressure quantum ones, warm dense matter (WDM and ultra-relativistic plasmas. HEDP is of fundamental importance for research in the field of laboratory astrophysics and inertial confinement fusion (ICF. Reaching such extreme states of matter now and in the future will depend on the use of plasma optics for amplifying and focusing laser pulses. This article will present the relevant technological infrastructure being built in ELI-BL for HEDP and UHI, and gives a brief overview of some research under way in the field of UHI, laboratory astrophysics, ICF, WDM, and plasma optics.

Liquid kit for preparation of {sup 188}rhenium-etidronate

Energy Technology Data Exchange (ETDEWEB)

Marczewski, Barbara; Dias, Carla Roberta; Moraes, Vanessa; Osso Junior, Joao Alberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), SP (Brazil). Centro de Radiofarmacia]. E-mail: baszot@gmail.com

2005-10-15

The aim of this study was the preparation of a liquid kit for radiolabeling of {sup 188} Re-HEDP (hydroxyethylidene diphosphonate). {sup 188} Re was obtained from alumina based {sup 188} W/{sup 188} Re generators. This paper reports the efficacy of a cold kit stored for more than two weeks, determined by the dependence of the radiolabeling yields of {sup 188} Re-HEDP on the incubation time, reducing agent concentration, the effects of concentration of ligand, the p H of the reaction and the temperature. The cold kits showed a good stability when carrie-free rhenium-188 was added in the reaction mixture. (author)

Finite-element formulations for the thermal stress analysis of two- and three-dimensional thin reactor structures

International Nuclear Information System (INIS)

Kulak, R.F.; Kennedy, J.M.; Belytschko, T.B.; Schoeberle, D.F.

1977-01-01

In several postulated LMFBR subassembly-to-subassembly failure propagation events, it is hypothesized that the duct wall of an accident subassembly fails and deposits molten fuel on the outer wall of an adjacent subassembly. It is therefore necessary to determine if the deposited fuel will fail the adjacent wall and thus propagate the event. This entails a thermal stress analysis, and since at times the adjacent subassembly is internally pressurized, thermomechanical analysis are also of value. Solutions are presented for several elastic plastic thermal problems. Some of these examples are compared to available analytic solutions. In addition, the hypothetical accident of molten fuel deposition on the adjacent hexcan is addressed. Combinations of pressure and thermal loading are considered. It is shown that the principal feature of the response is a large in-plane compressive stress which would undoubtedly cause buckling

The Design and Construction Process of a Test Stand for Casting the Power Steering’S Housing with the Use of the Pdcpd Material

Science.gov (United States)

Sobek, M.; Baier, A.; Grabowski, Ł.

2018-01-01

The use of new technologies and materials in various industries is a natural process that is directly related to the very high rate of development of these technologies. Certain industries decide to much faster introduce new technologies and materials. One of such branches is the automotive industry, whose representatives are very energetically looking for both financial savings and savings resulting from the vehicles mass reduction. An economically justified approach to construction materials is leading the search for new solutions and materials. The use of a modern material such as the two-component PDCPD composite shows hitherto unknown possibilities of producing subassemblies of many different constructions. The possibility of using a modern composite material with parameters comparable to that of metals and significantly lighter, can be an excellent alternative in the selection of materials for many parts of motor vehicles. The potentiality of precise casting of tolerated surfaces will allow to reduce the operations related to machining process, which is an indispensable part of the production process of elements that are cast of metal. This article describes the process of designing and building a test stand for precise positioning of power steering gear components at the stage of casting their housing. The article presents the principle of operation of the test stand and the process of preparation for the casting and the cast itself will be rudely described. Due to the implementation of research as part of a research project with an industrial partner, the article will only describe some operations. This is related to the confidentiality of the project.

Local flow blockage analysis with checkerboard configuration in a wire wrapped fuel subassembly using the ASFRE code

International Nuclear Information System (INIS)

Nishimura, Masahiro; Fukano, Yoshitaka

2014-01-01

Local fault (LF) has been historically considered as one of the possible causes of severe accidents in sodium-cooled fast reactors because fuel pins are generally densely arranged in the fuel subassemblies (FSAs) in this type of reactors. Local flow blockage (LB) has been one of the dominant initiators of LFs. Therefore evaluations were performed on LBs in the past safety licensing assuming a planar and impermeable blockage of 66% of the total flow area at an FSA for the Japanese prototype fast breeder reactor. A conservative evaluation revealed that fuel pin damage propagation would be limited within a restricted area of the reactor core, even assuming such a hypothetical initiating event. In the newly formulated regulatory requirements, however, after the accident at the Fukushima Dai-ichi nuclear power plant, best estimate (BE) safety analyses on the basis of state-of-the-art knowledge are being required for beyond design basis accidents. A deterministic and BE evaluation therefore based on the most-recent knowledge was newly performed in this study for revalidation of the above-mentioned historical background using the ASFRE code, whereas the LF accidents would not be identified as a representative accident sequence from a viewpoint of both its frequencies and consequences. Nominal power and flow rate without safety margins were assumed for the analyses in order to make the accidental conditions to be realistic. A most likely and realistic blockage configuration was newly proposed and employed based on the existing experimental data in accordance with the BE concept mentioned above. The aforementioned blockage configuration was excessively conservative on a state-of-the-art knowledge basis. The most-recent experimental studies clarified that LBs due to foreign substances would be formed by accumulating the steel fragments of certain sizes trapped along the wrapping wires. This leads to an LB in a checkerboard configuration for an FSA of wire spacer type, which

Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

International Nuclear Information System (INIS)

Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

1982-01-01

The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling

Effects of plasma jet parameters, ionization, thermal conduction, and radiation on stagnation conditions of an imploding plasma liner

Science.gov (United States)

Stanic, Milos

The disciplines of High Energy Density Physics (HEDP) and Inertial Confinement Fusion (ICF) are characterized by hypervelocity implosions and strong shocks. The Plasma Liner Experiment (PLX) is focused on reaching HEDP and/or ICF relevant regimes in excess of 1 Mbar peak pressure by the merging and implosion of discrete plasma jets, as a potentially efficient path towards these extreme conditions in a laboratory. In this work we have presented the first 3D simulations of plasma liner, formation, and implosion by the merging of discrete plasma jets in which ionization, thermal conduction, and radiation are all included in the physics model. The study was conducted by utilizing a smoothed particle hydrodynamics code (SPHC) and was a part of the plasma liner experiment (PLX). The salient physics processes of liner formation and implosion are studied, namely vacuum propagation of plasma jets, merging of the jets (liner forming), implosion (liner collapsing), stagnation (peak pressure), and expansion (rarefaction wave disassembling the target). Radiative transport was found to significantly reduce the temperature of the liner during implosion, thus reducing the thermal expansion rates and leaving more pronounced gradients in the plasma liner during the implosion compared with ideal hydrodynamic simulations. These pronounced gradients lead to a greater sensitivity of initial jet geometry and symmetry on peak pressures obtained. Accounting for ionization and transport, many cases gave higher peak pressures than the ideal hydrodynamic simulations. Scaling laws were developed accordingly, creating a non-dimensional parameter space in which performance of an imploding plasma jet liner can be estimated. It is shown that HEDP regimes could be reached with ≈ 5 MJ of liner energy, which would translate to roughly 10 to 20 MJ of stored (capacitor) energy. This is a potentially significant improvement over the currently available means via ICF of achieving HEDP and nuclear

Removable bearing arrangement for a wind turbine generator

Science.gov (United States)

Bagepalli, Bharat Sampathkumaran; Jansen, Patrick Lee; Gadre, Aniruddha Dattatraya

2010-06-15

A wind generator having removable change-out bearings includes a rotor and a stator, locking bolts configured to lock the rotor and stator, a removable bearing sub-assembly having at least one shrunk-on bearing installed, and removable mounting bolts configured to engage the bearing sub-assembly and to allow the removable bearing sub-assembly to be removed when the removable mounting bolts are removed.

R and D on early detection of the Total Instantaneous Blockage for 4. Generation Reactors - Inventory of non-nuclear methods investigated by the CEA

International Nuclear Information System (INIS)

Paumel, K.; Jeannot, J.-P.; Vanderhaegen, M.; Massacret, N.; Jeanne, T.; Laffont, G.

2013-06-01

In the safety analysis for the core of the 4. Generation Reactors, the Total Instantaneous Blockage (TIB) is a hypothetic accident scenario involving the melting of the blocked subassembly with a risk of propagation to the neighbouring subassemblies. To avoid this latter consequence a detection system has to scram the reactor. For Superphenix or EFR project a Delayed Neutron Detection Integrated (DND I) was considered as efficient to limit the melting to the first neighbouring subassemblies. Nonetheless for the CFV core the objective of improving the safety leads to limit the melting to the blocked subassembly. For this purpose, the CEA has launched a program development to find a new detection method. This paper provides a brief review of the feedback of R and D, progress and program on the various early non-nuclear detection methods investigated by the CEA: - Temperature measurement at the subassemblies outlet by thermocouples. The advantage of this method is that it will require no additional instrumentation to that already present for continuous monitoring. - Temperature measurement at the subassemblies outlet by Optical Fibers Bragg Grating (OFBG). This technology has the electromagnetic immunity, compactness and short response time. - Temperature measurement at the subassemblies outlet by ultrasound. The measuring point is located closer to the head subassembly and the response time could be shorter. - Acoustic detection of sodium boiling. Boiling occurs early in the accident progress and the area to be monitored may be covered by few sensors. - Subassemblies loss of flow detection by eddy-current flowmeters. This method seems logically the easiest and the most immediate method to detect a blockage. To date, none of these methods has been fully demonstrated to be feasible. It should be noted that temperature measurement methods will probably consist of the detection of a low increase rate using specific signal processing. These methods have been compared

Sixth Status Report: Testing of Aged Softwood Fiberboard Material for the 9975 Shipping Package

Energy Technology Data Exchange (ETDEWEB)

Daugherty, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-03-31

Samples have been prepared from several 9975 lower fiberboard subassemblies fabricated from softwood fiberboard. Physical, mechanical and thermal properties have been measured following varying periods of conditioning in each of several environments. These tests have been conducted in the same manner as previous testing on cane fiberboard samples. Overall, similar aging trends are observed for softwood and cane fiberboard samples, with a few differences. Some softwood fiberboard properties tend to degrade faster in some environments, while some cane fiberboard properties degrade faster in the two most aggressive environments. As a result, it is premature to assume both materials will age at the same rates, and the preliminary aging models developed for cane fiberboard might not apply to softwood fiberboard. However, it is expected that both cane and softwood fiberboard assemblies will perform satisfactorily in conforming packages stored in a typical KAC storage environment for up to 15 years. Samples from an additional 3 softwood fiberboard assemblies have begun aging during the past year to provide information on the variability of softwood fiberboard behavior. Aging and testing of softwood fiberboard will continue and additional data will be collected to support development of an aging model specific to softwood fiberboard.

Labelling of Re-ABP with 188Re for bone pain palliation

International Nuclear Information System (INIS)

Arteaga de Murphy, Consuelo; Ferro-Flores, Guillermina; Pedraza-Lopez, Martha; Melendez-Alafort, Laura; Croft, B.Y.Barbara Y.; Ramirez, Flor de Maria; Padilla, Juan

2001-01-01

Etidronate and medronate have been labelled with technetium-99m ( 99m Tc-HEDP, 99m Tc-MDP) for bone scanning and, with rhenium-188 ( 188 Re-HEDP) to palliate the pain resulting from bone metastases. The objective of this study was to label alendronate, ABP, a new bisphosphonate, with SnF 2 -reduced- 188 Re. The reagents for the 5 mg ABP kit were SnF 2 , KReO 4 and gentisic acid at acid pH. The chemical, spectroscopic and microscopic characteristics, quality control, rat bone uptake of [ 188 Re]Re-ABP and similarities with 99m Tc-ABP are presented. We conclude that this is a promising new radiopharmaceutical for bone metastases pain palliation

Liquid metal cooled fast breeder nuclear reactors

International Nuclear Information System (INIS)

Gatley, J.A.

1979-01-01

Breeder fuel sub-assemblies with electromagnetic brakes and fluidic valves for liquid metal cooled fast breeder reactors are described. The electromagnetic brakes are of relatively small proportions and the valves are of the controlled vortex type. The outlet coolant temperature of at least some of the breeder sub-assemblies are maintained by these means substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (UK)

ePHM System Development, Hardware-in-the-Loop Testing, Fault Tree, and FMECA Applied to and Integrated on NASA Hybrid Electric Testbeds, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — Hybrid-Electric distributed propulsion (HEDP) is becoming widely accepted and new tools will be required for future development with validation and demonstrations...

Design of the ITER tokamak assembly tools

Energy Technology Data Exchange (ETDEWEB)

Park, Hyunki [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of)], E-mail: hkpark@nfri.re.kr; Lee, Jaehyuk; Kim, Taehyung [SFA Engineering Corp., 42-7 Palyong-dong, Changwon-si, Gyeongsangnam-do 641-847 (Korea, Republic of); Song, Yunju [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of); Im, Kihak [ITER Organization, CEA Cadarasche, 13108 Saint Paul-lez-Durance (France); Kim, Byungchul; Lee, Hyeongon; Jung, Ki-Jung [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of)

2008-12-15

ITER tokamak assembly is mainly composed of lower cryostat activities, sector sub-assembly, sector assembly, in-vessel activities and ex-vessel activities. The main tools for sector sub-assembly procedures consists of upending tool, sector lifting tool, vacuum vessel support and bracing tool and sector sub-assembly tool. Conceptual design of assembly tools for sector sub-assembly procedures is described herein. The basic structure for upending tool has been developed under the assumption that upending is performed with crane which will be installed in Tokamak building. Sector lifting tool is designed to adjust the position of a sector to minimize the difference between the center of the tokamak building crane and the center of gravity of the sector. Sector sub-assembly tool is composed of special frame for the fine adjustment of position control with 6 degrees of freedom. The design of VV support and bracing tool for four kinds of VV 40 deg. sectors has been developed. Also, structural analysis for upending tool, sector sub-assembly tool has been studied using ANSYS for the situation of an applied load with the same dead weight multiplied by 3/4. The results of structural analyses for these tools were below the allowable values.

Cold-crucible melting of hulls and structural materials

International Nuclear Information System (INIS)

Jouan, A.; Jacquet-Francillon, N.; Puyou, M.; Piccinato, R.

1990-01-01

The method currently implemented at the La Hague UP3 reprocessing plant for conditioning of PWR zircaloy hulls is cement embedding. Another promising method, mainly for reducing the waste volume and the available exchange surface area, is melting. A cold-crucible melting process has therefore been developed by the CEA at Marcoule (France) over the last decade. Development work first concentrated on cladding hulls from fast breeder reactors, then from pressurized water reactors. The process can be used for both types of cladding wastes. Subassembly head and foot end-caps are sheared off and should be suitable for surface storage after α decontamination by successive rinsing. If necessary because of their α activity, they could be melted in a larger furnace

Classical and Ablative Richtmyer-Meshkov Instability and Other ICF-Relevant Plasma Flows Diagnosed With Monochromatic X-Ray Imaging

National Research Council Canada - National Science Library

Aglitskiy, Y; Karasik, M; Velikovich, A. L; Metzler, N; Zalesak, S; Schmitt, A. J; Gardner, J. H; Serlin, V; Weaver, J; Obenschain, S. P

2007-01-01

In inertial confinement fusion (ICF) and high-energy density physics (HEDP), the most important manifestations of the hydrodynamic instabilities and other mixing processes involve lateral motion of the accelerated plasmas...

Physics-based Radiator Design, Sizing & Weight Estimation Tool for Conceptual Design of More-, Hybrid-, and All-Electric Next Gen Aircraft, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — Hybrid electric distributed propulsion (HEDP) systems have proven worthy for further consideration by approaching NASA's goals for N+2 and N+3 energy consumption,...

Continued Development of Environmentally COnscious "ECO" Transport Aircraft Concepts as Hybrid Electric Distributed Propulsion Research Platforms, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — ESAero's vast TeDP and HEDP-specific experience, Helden Aerospace's distributed propulsion airframe integration effects (2) Advance the TMS design with a new TMS...

Sodium flow distribution in test fuel assembly P-23B

International Nuclear Information System (INIS)

Taylor, J.P.S.

1978-08-01

Relatively large cladding diametral increases in the exterior fuel pins of HEDL's test fuel subassembly P-23B were successfully explained by a thermal-hydraulic/solid mechanics analysis. This analysis indicates that while at power, the subassembly flow was less than planned and that the fuel pins were considerably displaced and bowed from their nominal position. In accomplishing this analysis, a method was developed to estimate the sodium flow distribution and pin distortions in a fuel subassembly at power

Nuclear reactor instrumentation

International Nuclear Information System (INIS)

Duncombe, E.; McGonigal, G.

1975-01-01

A liquid metal cooled nuclear reactor is described which has an equal number of fuel sub-assemblies and sensing instruments. Each instrument senses temperature and rate of coolant flow of a coolant derived from a group of three sub-assemblies so that an abnormal value for one sub-assembly will be indicated on three instruments thereby providing for redundancy of up to two of the three instruments. The abnormal value may be a precurser to unstable boiling of coolant

Experimental validation of the HARMONIE code

International Nuclear Information System (INIS)

Bernard, A.; Dorsselaere, J.P. van

1984-01-01

An experimental program of deformation, in air, of different groups of subassemblies (7 to 41 subassemblies), was performed on a scale 1 mock-up in the SPX1 geometry, in order to achieve a first experimental validation of the code HARMONIE. The agreement between tests and calculations was suitable, qualitatively for all the groups and quantitatively for regular groups of 19 subassemblies at most. The differences come mainly from friction between pads, and secondly from the foot gaps. (author)

An analysis of options available for developing a common laser ray tracing package for Ares and Kull code frameworks

Energy Technology Data Exchange (ETDEWEB)

Weeratunga, S K

2008-11-06

Ares and Kull are mature code frameworks that support ALE hydrodynamics for a variety of HEDP applications at LLNL, using two widely different meshing approaches. While Ares is based on a 2-D/3-D block-structured mesh data base, Kull is designed to support unstructured, arbitrary polygonal/polyhedral meshes. In addition, both frameworks are capable of running applications on large, distributed-memory parallel machines. Currently, both these frameworks separately support assorted collections of physics packages related to HEDP, including one for the energy deposition by laser/ion-beam ray tracing. This study analyzes the options available for developing a common laser/ion-beam ray tracing package that can be easily shared between these two code frameworks and concludes with a set of recommendations for its development.

Toward More Productive Naval Shipbuilding

Science.gov (United States)

1984-01-01

J. Seymour Exxon Production Research Scripps Institution of Oceanography •HeuiLon,.Tleas La Jolla, California William Creelman William H. Silcox...subassemblies move to become finished products. Figure 14 indicates the many organizational functions and physical steps through which information and...supplier control, and in some cases physical material control systems unique to its requirements. Systems developed along organizational linesuse some

Hybrid-Electric Aircraft TOGW Development Tool with Empirically-Based Airframe and Physics-Based Hybrid Propulsion System Component Analysis, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — Hybrid-Electric distributed propulsion (HEDP) is becoming widely accepted and new tools will be required for future development. This Phase I SBIR proposal creates a...

Integrated Computational Materials Engineering Development of Advanced High Strength Steel for Lightweight Vehicles

Energy Technology Data Exchange (ETDEWEB)

Hector, Jr., Louis G. [General Motors, Warren, MI (United States); McCarty, Eric D. [United States Automotive Materials Partnership LLC (USAMP), Southfield, MI (United States)

2017-07-31

The goal of the ICME 3GAHSS project was to successfully demonstrate the applicability of Integrated Computational Materials Engineering (ICME) for the development and deployment of third generation advanced high strength steels (3GAHSS) for immediate weight reduction in passenger vehicles. The ICME approach integrated results from well-established computational and experimental methodologies to develop a suite of material constitutive models (deformation and failure), manufacturing process and performance simulation modules, a properties database, as well as the computational environment linking them together for both performance prediction and material optimization. This is the Final Report for the ICME 3GAHSS project, which achieved the fol-lowing objectives: 1) Developed a 3GAHSS ICME model, which includes atomistic, crystal plasticity, state variable and forming models. The 3GAHSS model was implemented in commercially available LS-DYNA and a user guide was developed to facilitate use of the model. 2) Developed and produced two 3GAHSS alloys using two different chemistries and manufacturing processes, for use in calibrating and validating the 3GAHSS ICME Model. 3) Optimized the design of an automotive subassembly by substituting 3GAHSS for AHSS yielding a design that met or exceeded all baseline performance requirements with a 30% mass savings. A technical cost model was also developed to estimate the cost per pound of weight saved when substituting 3GAHSS for AHSS. The project demonstrated the potential for 3GAHSS to achieve up to 30% weight savings in an automotive structure at a cost penalty of up to $0.32 to $1.26 per pound of weight saved. The 3GAHSS ICME Model enables the user to design 3GAHSS to desired mechanical properties in terms of strength and ductility.

Use of EBR-II as a principal fast breeder reactor irradiation test facility in the U.S

International Nuclear Information System (INIS)

Staker, R.G.; Seim, O.S.; Beck, W.N.; Golden, G.H.; Walters, L.C.

1975-01-01

The EBR-II as originally designed and operated by the Argonne National Laboratory was successful in demonstrating the operation of a sodium-cooled fast breeder power plant with a closed fuel reprocessing cycle. Subsequent operation has been as an experimental facility where thousands of irradiation tests have been performed. Conversion to this application entailed the design and fabrication of special irradiation subassemblies for in-core irradiations, additions to existing facilities for out-of-core irradiations, and additions to existing facilities for out-of-core experiments. Experimental subassemblies now constitute about one third of the core, and changes in the core configuration occur about monthly, requiring neutronic and thermal-hydraulics analyses and monitoring of the reactor dynamic behavior. The surveillance programs provided a wealth of information on irradiation induced swelling and creep, in-reactor fracture behavior, and the compatibility of materials with liquid sodium. (U.S.)

Fuel management codes for fast reactors

International Nuclear Information System (INIS)

Sicard, B.; Coulon, P.; Mougniot, J.C.; Gouriou, A.; Pontier, M.; Skok, J.; Carnoy, M.; Martin, J.

The CAPHE code is used for managing and following up fuel subassemblies in the Phenix fast neutron reactor; the principal experimental results obtained since this reactor was commissioned are analyzed with this code. They are mainly concerned with following up fuel subassembly powers and core reactivity variations observed up to the beginning of the fifth Phenix working cycle (3/75). Characteristics of Phenix irradiated fuel subassemblies calculated by the CAPHE code are detailed as at April 1, 1975 (burn-up steel damage)

Physics-Based Wing Structure Design, Analysis and Weight Estimation Conceptual Design Tool for Hybrid Electric Distributed Propulsion, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — As HEDP systems have proven worthy of further consideration by approaching NASA's goals for N+2 and N+3 energy consumption, noise, emission and field length,...

Methods of statistical calculation of fast reactor core with account of influence of fuel assembly form change in process of campaign and other factors

International Nuclear Information System (INIS)

Sorokin, G.A.; Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P.

2000-01-01

The method of calculation of a temperature field in fast reactor core using criterion equal thermo-technical reliability of subassemblies in various zones throttling taking into account change thermohydraulic characteristics of subassemblies during campaign under influence change form of core, redistribution heat generation, casual any deviation of various parameters is stated. The distribution of the statistical characteristics of a temperature field in subassemblies is calculated on subchannel method with account of an interchannel exchange and feature of influence of deformation on a temperature field in subassemblies using Monte-Carlo method. The results of the calculations show that deformation can have significant influence on a temperature mode of core. It is necessary to make thermohydraulic analysis of core during campaign at a stage of preliminary study of the projects fast reactors. (author)

Development of failed fuel detection and location system in sodium-cooled large reactor. Sampling method of failed fuels under the slit

International Nuclear Information System (INIS)

Aizawa, Kousuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

2010-01-01

A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the 'Fast Reactor Cycle Technology Development (FaCT)' project in Japan. JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for indentifying under-slit failed fuel subassemblies has been demonstrated by water experiments. (author)

A Comparative Performance Evaluation of Some Novel “Green” and Traditional Antiscalants in Calcium Sulfate Scaling

Directory of Open Access Journals (Sweden)

Konstantin Popov

2016-01-01

Full Text Available A relative ability of industrial samples of four phosphorus-free polymers (polyaspartate (PASP; polyepoxysuccinate (PESA; polyacrylic acid sodium salt (PAAS; copolymer of maleic and acrylic acid (MA-AA and of three phosphonates (aminotris(methylenephosphonic acid, ATMP; 1-hydroxyethane-1,1-bis(phosphonic acid, HEDP; phosphonobutane-1,2,4-tricarboxylic acid, PBTC to inhibit calcium sulfate precipitation is studied following the NACE Standard along with dynamic light scattering (DLS, scanning electron microscopy (SEM, and X-ray diffraction (XRD technique. For the 0.5 mg·dm−3 dosage, the following efficiency ranking was found: MA-AA~ATMP>PESA (400–1500 Da>PASP (1000–5000 Da ≫ PAAS (3000–5000 Da~PBTC~HEDP. The isolated crystals are identified as gypsum. SEM images for PESA, PASP, PAAS, and HEDP and for a blank sample indicated the needle-like crystal morphology. Surprisingly, the least effective reagent PBTC revealed quite a different behavior, changing the morphology of gypsum crystals to an irregular shape. The DLS experiments exhibited a formation of 300 to 700 nm diameter particles with negative ζ-potential around −2 mV for all reagents. Although such ζ-potential values are not capable of providing colloidal stability, all three phosphonates demonstrate significant gypsum particles stabilization relative to a blank experiment.

Electrodeposition and Properties of Copper Layer on NdFeB Device

Directory of Open Access Journals (Sweden)

LI Yue

2017-06-01

Full Text Available To decrease the impact of the regular Ni/Cu/Ni coating on the magnetic performance of sintered NdFeB device, alkaline system of HEDP complexing agent was applied to directly electro-deposit copper layer on NdFeB matrix, then nickel layer was electrodeposited on the copper layer and Cu/Ni coating was finally obtained to replace the regular Ni/Cu/Ni coating. The influence of concentration of HEDP complexing agent on deposition course was tested by electrochemical testing; morphology of copper layer was characterized by SEM, XRD and TEM; the binding force of copper layer and the thermal reduction of magnetic of NdFeB caused by electrodeposited coating were respectively explored through the thermal cycle test and thermal demagnetization test. The results show that the concentration of HEDP has great impact on the deposition overpotential of copper. In the initial electrodepositing stage, copper particles precipitate at the grain boundaries of NdFeB magnets with a preferred (111 orientation. The copper layer is compact and has enough binding force with the NdFeB matrix to meet the requirements in SJ 1282-1977. Furthermore, the thermal demagnetization loss rate of the sintered NdFeB with the protection of Cu/Ni coating is significantly less than that with the protection of Ni/Cu/Ni coating.

Experience in quality assurance of alloy D9 clad tubes for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Kapoor, K.; Prahlad, B.

2012-01-01

Stainless Steel Alloy D9 is the material for cladding in various sub-assemblies of Prototype Fast Breeder Reactor (PFBR). The fabrication, inspection, testing and supply of the clad tubes for the first core of PFBR is nearly completed. The paper also compares the specification requirements and the achieved results for some of the critical aspects which is arrived after completing supply against the first core requirement

Control rod testing apparatus

International Nuclear Information System (INIS)

Gaunt, R.R.; Ashman, C.M.

1987-01-01

A control rod testing apparatus is described comprising: a first guide means having a vertical cylindrical opening for grossly guiding a control rod; a second guide means having a vertical cylindrical opening for grossly guiding a control rod. The first and second guide means are supported at axially spaced locations with the openings coaxial; and a substantially cylindrical subassembly having a vertical cylindrical opening therethrough. The subassembly is trapped coaxial with and between the first and second guide means, and the subassembly radially floats with respect to the first and second guide means

Latching device for nuclear reactor housing

International Nuclear Information System (INIS)

Barnes, J.G.

1981-01-01

A latching device for use in liquid metal cooled nuclear reactors is described which is not detached under normal operational loads on the absorber sub-assemblies. The sub-assemblies are however easily detached for repair or replacement. (U.K.)

Costing the OMNIUM-G system 7500

Science.gov (United States)

Fortgang, H. R.

1980-01-01

A complete OMNIUM-G System 7500 was cost analyzed for annual production quantities ranging from 25 to 10,000 units per year. Parts and components were subjected to in-depth scrutiny to determine optimum manufacturing processes, coupled with make or buy decisions on materials and small parts. When production quantities increase both labor and material costs reduce substantially. A redesign of the system that was analyzed could result in lower costs when annual production runs approach 100,000 units/year. Material and labor costs for producing 25, 100, 25,000 and 100,00 units are given for 17 subassembly units.

Rapsodie first core manufacture. 1. part: processing plant; Fabrication du premier coeur de rapsodie. Premiere partie: l'atelier de fabrication

Energy Technology Data Exchange (ETDEWEB)

Masselot, Y; Bataller, S; Ganivet, M; Guillet, H; Robillard, A; Stosskopf, F [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

1968-07-01

This report is the first in a series of three describing the processes, results and peculiar technical problems related to the manufacture of the first core of the fast reactor Rapsodie. A detailed study of manufacturing processes(pellets, pins, fissile sub-assemblies), the associated testings (raw materials, processed pellets and pins, sub-assemblies before delivery), manufacturing facilities and improvements for a second campaign are described. (author) [French] Ce rapport est le premier d'une serie de trois qui decrivent les procedes, les resultats et les problemes techniques particuliers de la fabrication du du premier coeur de la pile a neutrons rapides Rapsodie. Il comporte une etude detaillee des procedes de fabrication (pastilles, aiguilles, assemblages combustibles) et des methodes de controle associees (matieres premieres, pastilles et aiguilles en cours de fabrication, assemblages fissiles avant livraison), ainsi qu'une decription complete des installations de l'atelier de fabrication et les modifications apportees pour une deuxieme campagne. (auteur)

A vibration sieve

Energy Technology Data Exchange (ETDEWEB)

Alekhin, S.A.; Denisenko, V.V.; Dzhalalov, M.G.; Kirichek, F.P.; Pitatel, Yu.A.; Prokopov, L.I.; Tikhonov, Yu.P.

1982-01-01

A vibration sieve is proposed which includes a vibration drive, a body and a screen installed on shock absorbers, a device for washing out the screen, and a subassembly for loading the material. To increase the operational reliability and effectiveness of the vibration sieve by improving the cleaning of the screen, the loading subassembly is equipped with a baffle with a lever which is hinged to it. The device for washing out the screen is made in the form of an electromagnet with a connecting rod, a switch and an eccentric, a friction ratchet mechanism and sprinkling systems. Here, the latter are interconnected, using a connecting rod, while the sprinkling system is installed on rollers under the screen. The electromagnetic switch is installed under the lever. The body is made with grooves for installing the sprinkling system. The vibration sieve is equipped with a switch which interacts with the connecting rod. The friction ratchet mechanism is equipped with a lug.

Measure of the albedo of a warm plasma in the XUV range

Science.gov (United States)

Busquet, Michel; Thais, Frederic; Geoffroy, Ghita; Raffestin, Didier

2009-11-01

It has been shown in a recent experience at PALS [1] that the radiative precursor celerity in front of a strong radiative shock is sensitive to the lateral radiative losses, thus to the albedo of the wall of a ``radiative shock tube.'' In the experiment presented here, we measure the albedo of various materials (Al, Cu, Au) heated by a Xenon gaz at temperature around 30 eV. The Xenon gas was heated by the ALISE laser in CESTA in Bordeaux (France). The emission of Xenon with and without the reflecting samples is measured with a spatially resolving XUV spectrograph in the 30-250 eV range. [4pt] [1] M. Busquet et al, HEDP 3, 8 (2007)

Reliability & availability of wind turbine electrical & electronic components

NARCIS (Netherlands)

Tavner, P.; Faulstich, S.; Hahn, B.; Bussel, van G.J.W.

2010-01-01

Recent analysis of European onshore wind turbine reliability data has shown that whilst wind turbine mechanical subassemblies tend to have relatively low failure rates but long downtimes, electrical and electronic subassemblies have relatively high failure rates and short downtimes. For onshore wind

First experimental validation on the core equilibrium code: HARMONIE

International Nuclear Information System (INIS)

Van Dorsselaere, J.; Cozzani, M.; Gnuffi, M.

1981-08-01

The code HARMONIE calculates the mechanical equilibrium of a fast reactor. An experimental program of deformation, in air, of groups of subassemblies, was performed on a mock-up, in the Super Phenix 1- geometry. This program included three kinds of tests, all performed without and then with grease: on groups of 2 or 3 rings of subassemblies, subjected to a force acting upon flats or angles; on groups of 35 and 41 subassemblies, subjected to a force acting on the first row, then with 1 or 2 empty cells; and on groups with 1 or 2 bowed subassemblies or 1 enlarged one over flats. A preliminary test on the friction coefficient in air between two pads showed some dependance upon the pad surface condition with a scattering factor of 8. Two basic code hypotheses were validated: the rotation of the subassemblies around their axis was negligible after deformation of the group, and the choice of a mean Maxwell coefficient, between those of 1st and 2nd slope, led to very similar results to experimental. The agreement between tests and HARMONIE calculations was suitable, qualitatively for all the groups and quantitatively for regular groups of 3 rings at most. But the difference increased for larger groups of 35 or 41 subassemblies: friction between pads, neglected by HARMONIE, seems to be the main reason. Other reasons for these differences are: the influence of the loading order on the mock-up, and the initial contacts issued from the gap between foot and diagrid-insert, and from manufacture bowings

Ultrasonic testing device

International Nuclear Information System (INIS)

Lawrie, W.E.

1978-01-01

The ultrasonic transmitter made of polarized ferroelectric ceramic material (lead zirconate titanate) is arranged in a strip carrier which allows it to be introduced between the fuel elements of a fuel subassembly in a water cooled nuclear reactor. The ultrasonic transmitter is insulated relative to the carrier. The echo of the ra dal ultrasonic pulse is recorded which changes as faulty water filled fuel elements are detected. (RW) [de

Studies of decay heat removal by natural convection using the SONACO sodium-cooled 37-pin bundle

International Nuclear Information System (INIS)

Wydler, P.; Dury, T.V.; Hudina, M.; Weissenfluh, T. von; Sigg, B.; Dutton, P.

1986-01-01

Natural convection measurements in an electrically heated sodium-cooled rod bundle are being performed with the aim of contributing to a better understanding of natural convection effects in subassemblies with stagnant sodium and providing data for code validation. Measurements include temperature distributions in the bundle for different cooling configurations which simulate heat transfer to the intersubassembly gap and neighbouring subassemblies and possible thermosyphonic interaction between a subassembly and the reactor plenum above. Conditions for which stable natural convection patterns exist are identified, and results are compared with predictions of different computer codes of the porous-medium type. (author)

Fuel penetration of intersubassembly gaps in LMFBRs: a calculational method with the SIMMER-II code

International Nuclear Information System (INIS)

DeVault, G.P.

1983-01-01

Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor (LMFBR) undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. A possible avenue for early fuel removal in heterogeneous core LMFBRs is the failure of duct walls in disrupted driver subassemblies followed by fuel penetration into the gaps between blanket subassemblies. The SIMMER-II code was modified to simulate flow between subassembly gaps. Calculations with the modified SIMMER-II code indicate the capabilities of the method and the potential for fuel mass reduction in the active core

Skeletal and reticuloendothelial imaging in osteopetrosis: case report

International Nuclear Information System (INIS)

Park, H.M.; Lambertus, J.

1977-01-01

Skeletal and reticuloendothelial images, using Tc-99m HEDP and Tc-99m sulfur colloid, respectively, were obtained from two adult patients with osteopetrosis. Skeletal images demonstrated increased activity in multiple fracture sites, in mandibular osteomyelitis, in ends of splayed long bones adjacent to joints, and in the epiphyseal ends of short tubular bones. The remainder of the skeleton involved with osteopetrosis showed no generalized increased uptake of Tc-99m HEDP. These findings indicate that metabolic activity in this disease is abnormally increased in the usual areas of bone growth but appears normal elsewhere. Reticuloendothelial imaging showed an almost total lack of activity in the axial and peripheral skeletal marrow space. Anemia, however, was only moderate in these patients. Skeletal scintigraphy may be useful to evaluate the presence and extent of the frequent complications of osteopetrosis, namely fractures and osteomyelitis

Evaluation method for core thermohydraulics during natural circulation in fast reactors numerical predictions of inter-wrapper flow

International Nuclear Information System (INIS)

Kamide, H.; Kimura, N.; Miyakoshi, H.; Nagasawa, K.

2001-01-01

Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. As a decay heat removal system, direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this design, dumped heat exchanger provides cold sodium and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such natural circulation phenomena in a reactor core has been developed, which models each subassembly as a rectangular duct with gap region between the subassemblies and also the upper plenum in a reactor vessel. This numerical simulation method was verified based on experimental data of a sodium test using 7- subassembly core model and also a water test which simulates IWF using the 1/12 sector model of a reactor core. We applied the estimation method to the natural circulation in a 600 MW class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. It is desired for prediction methods on the natural circulation to simulate these phenomena. (author)

Verification and implications of the multiple pin treatment in the SASSYS-1 LMR systems analysis code

International Nuclear Information System (INIS)

Dunn, F.E.

1994-01-01

As part of a program to obtain realistic, as opposed to excessively conservative, analysis of reactor transients, a multiple pin treatment for the analysis of intra-subassembly thermal hydraulics has been included in the SASSYS-1 liquid metal reactor systems analysis code. This new treatment has made possible a whole new level of verification for the code. The code can now predict the steady-state and transient responses of individual thermocouples within instrumented subassemlies in a reactor, rather than just predicting average temperatures for a subassembly. Very good agreement has been achieved between code predictions and the experimental measurements of steady-state and transient temperatures and flow rates in the Shutdown Heat Removal Tests in the EBR-II Reactor. Detailed multiple pin calculations for blanket subassemblies in the EBR-II reactor demonstrate that the actual steady-state and transient peak temperatures in these subassemblies are significantly lower than those that would be calculated by simpler models

Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

International Nuclear Information System (INIS)

Zheng, Meiyin; Tian, Wenxi; Zhang, Dalin; Qiu, Suizheng; Su, Guanghui

2015-01-01

Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

The TFTR RF Limiter upgrade design and installation

International Nuclear Information System (INIS)

Barnes, G.W.; Fan, H.M.; Ulrickson, M.

1991-01-01

The RF Limiters originally installed at Bays K-L and N-O[1] were upgraded to a new configuration and six new limiters of similar design were added. The RF Limiter upgrade protects the (2) existing RF Launchers and with a minor addition will protect the (2) RF Launchers to be installed in FY92 and will permit 50 Megawatts of auxiliary input power for two seconds during plasma operation. Each of the new RF Limiters are comprised of 18 tiles for a total of 108. The design provides for revised and strengthened supporting mounts because of additional forces induced in the tiles. Tile material is a 2D carbon-carbon composite identical to the original tile material. The channel shaped tile is geometrically the same as the original design. Subassembly of the panels took place outside the vessel in order to minimize exposure levels to the workers. Tooling was designed to replicate the vessel hardpoints and ease the subassembly tasks. Installation of the entire system occurred during the FY 91 opening. Integrated into the design are provisions to eliminate plasma damage to the insulators at the mounts. Detail design philosophy and an overview of the project are addressed by this paper. 2 refs., 2 figs

Optimal base-stock levels in a serial two-echelon system with random yield and rework of orders

NARCIS (Netherlands)

Kiesmüller, G.P.; Kok, de A.G.

2007-01-01

In this paper we consider the manufacturing process of a finished product assembled out of one unit of an expensive subassembly and some other non-expensive parts. The subassembly itself is assembled out of several components delivered just in time from outside suppliers. After the production of the

Core shroud corner joints

Science.gov (United States)

Gilmore, Charles B.; Forsyth, David R.

2013-09-10

A core shroud is provided, which includes a number of planar members, a number of unitary corners, and a number of subassemblies each comprising a combination of the planar members and the unitary corners. Each unitary corner comprises a unitary extrusion including a first planar portion and a second planar portion disposed perpendicularly with respect to the first planar portion. At least one of the subassemblies comprises a plurality of the unitary corners disposed side-by-side in an alternating opposing relationship. A plurality of the subassemblies can be combined to form a quarter perimeter segment of the core shroud. Four quarter perimeter segments join together to form the core shroud.

Safety verdict about the 300-MW prototype nuclear power station with fast sodium cooled reactor in Kalkar

International Nuclear Information System (INIS)

1986-08-01

The safety verdict had been elaborated on the order of the Ministry of Economy, Middle Classes and Technology of the state North Rhine-Westphalia. It covers the behaviour of the so-called target core of the Mark-Ia core as a whole as well as its individual subassemblies, i.e. fuel, diluent, absorber, blanket and reflector subassemblies. The report considers the aspects of the neutron physics, thermal hydraulics and mechanical design together with the quality assurance, the treatment of radiological questions and the determination of the decay heat. The safety authorities come to the conclusion the sufficient provision against damages has been taken in the design of the core subassemblies

Mode of failure of LMFBR fuel pins

International Nuclear Information System (INIS)

Washburn, D.F.

1975-01-01

The objectives of the irradiation test described were to evaluate mixed-oxide fuel performance and to confirm the design adequacy of the FFTF fuel pins. After attainment of the initial objectives the irradiation of several of the original fuel pins was continued until a cladding breach occurred. The consequences of a cladding breach were evaluated by reconstituting the original 37-pin subassembly into two 19-pin subassemblies after a burnup at 50,000 MWd/MTM (5.2 a/o). The original pins were supplemented with fresh pins as necessary. Irradiation of the subassemblies was continued until a cladding breach occurred. Results are presented and discussed

HMI annual report 1990

International Nuclear Information System (INIS)

1991-01-01

This report describes the activities of the Hahn-Meitner-Institut (HMI) on four special subjects: 1. Nuclear physics (nuclear reactions, nuclear structure, VICKSI accelerator development), 2. photochemical energy conversion (basic and material research, radiation chemistry), 3. structure research (theory of many-body systems, solid-state physics, neutron scattering, highly stressed metallic materials, trace nutrients in health and nutrition), and 4. information technics (software, real time systems, semiconductor subassembly). The following is presented for every special subject: a) the topics worked on in 1989, b) selected results and c) publications, lectures, theses submitted for diplomas and doctoral theses. (orig.) [de

Large ceramics for fusion applications

International Nuclear Information System (INIS)

Hauth, W.E.; Stoddard, S.D.

1979-01-01

Prominent ceramic raw materials and products manufacturers were surveyed to determine the state of the art for alumina ceramic fabrication. This survey emphasized current capabilities and limitations for fabrication of large, high-density, high-purity, complex shapes. Some directions are suggested for future needs and development. Ceramic-to-ceramic sealing has applications for several technologies that require large and/or complex vacuum-tight ceramic shapes. Information is provided concerning the assembly of complex monolithic ceramic shapes by bonding of subassemblies at temperatures ranging from 450 to 1500 0 C. Future applications and fabrication techniques for various materials are presented

NUMERICAL THERMAL ANALYSIS OF A CAR BRAKING SYSTEM

Directory of Open Access Journals (Sweden)

Patryk Różyło

2017-06-01

Full Text Available The study involved performing a numerical thermal analysis of selected components in a car braking system. The primary goal of the study was to determine the regions which are the most susceptible to variations in temperature, and to determine the degree of thermal impact upon them. The analysis was performed using the Abaqus environment. The examined components of the braking system were made of materials reflecting the mechanical properties of the real subassemblies. The FEM analysis enabled determination of the distribution of temperature in the system with respect to the properties of the investigated materials and applied boundary conditions.

Hahn-Meitner-Institut Berlin. Annual report 1993

International Nuclear Information System (INIS)

1994-01-01

This report describes the activities of the Hahn-Meitner-Institute (HMI) on four special subjects: 1. Nuclear physics (nuclear reactions, nuclear structure, VICKSI accelerator development), 2. photochemical energy conversion (basic and material research, radiation chemistry), 3. structure research (theory of many-body systems, solid-state physics, neutron scattering, highly stressed metallic materials, trace nutrients in health and nutrition), and 4. information technics (software, real time systems, semiconductor subassembly). The following is presented for every special subject: a) The topics worked on in 1993, b) selected results and c) publications, lectures, theses submitted for diplomas and doctoral theses. (orig.) [de

HMI annual report 1989

International Nuclear Information System (INIS)

1990-01-01

This report describes the activities of the Hahn-Meitner-Institut (HMI) on four special subjects: 1. Nuclear physics (nuclear reactions, nuclear structure, VICKSI accelerator development), 2. photochemical energy conversion (basic and material research, radiation chemistry), 3. structure research (theory of many-body systems, solid-state physics, neutron scattering, highly stressed metallic materials, trace nutrients in health and nutrition, and 4. information technics (software technics, real time systems, semiconductor subassembly). The following is presented for every special subject: a) the topics worked on in 1989, b) selected results and c) publications, lectures, theses submitted for diplomas and doctoral theses. (MM) [de

Neutronics and thermal-hydraulics coupling: some contributions toward an improved methodology to simulate the initiating phase of a severe accident in a sodium fast reactor

International Nuclear Information System (INIS)

Guyot, Maxime

2014-01-01

This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and re-criticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios. During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. In the multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level. In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling. (author) [fr

Dynamic behaviour of FBR fuel pin bundles

International Nuclear Information System (INIS)

Martin, P.H.; Van Dorsselaere, J.P.; Ravenet, A.

1990-01-01

A programme of shock tests on a fast neutron reactor subassembly model (SPX1 geometry) including a complete bundle of fuel pins (dummy elements) is being carried out in the BELIER test facility at Cadarache. The purpose of these tests is: to determine the distribution of dynamic forces applied to the fuel rod clads under the impact conditions encountered in a reactor during a earthquake; to reduce as much as possible the conservatism of the methods presently used for the calculation of those forces. The test programme, now being completed, consists of the following steps: impacts on the mock-up in air with an non-compact bundle (situation of the subassembly at beginning of life (BOL) with clearances within the bundle); impacts under the same conditions but with fluid (water) in the subassembly; impacts on the mock-up in air and with a compacted bundle (simulating the conditions of an end-of-life (EOL) bundle with no clearance within the bundle). The accelerations studied in these tests cover the range encountered in design calculations for the subassembly frequencies in beam mode. (author)

Conceptual design of PFBR core

International Nuclear Information System (INIS)

Lee, S.M.; Govindarajan, S.; Indira, R.; John, T.M.; Mohanakrishnan, P.; Shankar Singh, R.; Bhoje, S.B.

1996-01-01

The design options selected for the core of the 500 MWe Prototype Fast Breeder Reactor are presented. PFBR has a conventional mixed oxide fuel core of homogeneous type with two enrichment zones for power flattening and with radial and axial blankets to make the reactor self-sustaining in fissile material. Pin diameter has been selected for minimization of fissile inventory. Considerations for the choice of number of pins per subassembly, integrated versus separate axial blankets, and other pin and subassembly parameters are discussed. As the core size is moderate, no special schemes for reducing the maximum positive sodium voiding coefficient is envisages. Two independent, diverse fast acting shutdown systems working in fail-safe mode are selected. The number of absorber rods has been minimized by choosing a layout for maximum antishadow effect. Nine control and safety rods are distributed in two rods for power flattening by differential insertion. Three Diverse Safety Rods, are also provided which are normally fully withdrawn. The optimization of layout of radial and axial shielding and adequacy of flux at detector location are also discussed. (author). 2 figs

Thermohydraulic and thermal stress aspects of a porous blockage in an LMFBR fuel assembly

International Nuclear Information System (INIS)

Kuzay, T.M.; Marr, W.W.; Helenberg, H.W.; Ariman, T.; Wilson, R.E.; Pedersen, D.R.

1979-01-01

The current safety scenarios of Liquid Metal Fast Breeder Reactors (LMFBR) under local fault propagation include the study of a hypothetical accident initiated by the formation of an external debris porous blockage in a fuel subassembly. In this preliminary experimental and analytical investigation, a non-heat-generating porous blockage was postulated to cover 18 flow channels of a 37 pin Fast Test Reactor (FTR) type fuel subassembly. The axial extent of the blockage is 50 mm. The blockage material is stainless steel (SS 316) with 30 percent average porosity (percent void volume). The blockage and the pins were modeled with a finite element technique and the thermal field in the blockage was predicted. This thermal field was utilized to do a planar thermal stress analysis of the postulated blockage. To verify the analytical model and also to better understand the thermal-hydraulics of such a porous blockage out-of-pile tests were conducted in a sodium loop. Data from the out-of-pile tests was utilized to calibrate and improve the analytical model

Space station high gain antenna concept definition and technology development

Science.gov (United States)

Wade, W. D.

1972-01-01

The layout of a technology base is reported from which a mechanically gimballed, directional antenna can be developed to support a manned space station proposed for the late 1970's. The effort includes the concept definition for the antenna assembly, an evaluation of available technology, the design of critical subassemblies and the design of critical subassembly tests.

600 A HTc current lead based on BSCCO 2212 rods for LHC magnets

CERN Document Server

García-Tabarés, L; Abramian, P; Toral, F; Angurel, L A; Diez, J C; Burriel, R; Natividad, E; Iturbe, R; Etxeandia, J

2000-01-01

A 600 A current lead using 2212 BSCCO bulk material is now under construction and will be soon delivered to CERN. Present paper describes the main steps of its design and fabrication, including the two main parts in which it is divided, the superconducting module with the BSCCO rods and the conventional resistive part. An important role in this design was played by previous experimental measurements on subassemblies, which are also described along the paper. (5 refs).

Method for the determination of technical specifications limiting temperature in EBR-II operation

International Nuclear Information System (INIS)

Chang, L.K.; Hill, D.J.; Ku, J.Y.

2004-01-01

The methodology and analysis procedure to qualify the Mark-V and Mark-VA fuels for the Experimental Breeder Reactor II are summarized in this paper. Fuel performance data and design safety criteria are essential for thermal-hydraulic analyses and safety evaluations. Normal and off-normal operation duty cycles and transient classifications are required for the safety assessment of the fuels. Design safety criteria for steady-state normal and transient off-normal operations were developed to ensure structural integrity of the fuel pin. The maximum allowable coolant outlet temperatures and powers of subassemblies for steady-state normal operation conditions were first determined in a row-by-row basis by a thermal-hydraulic and fuel damage analysis, in which a trial-and-error approach was used to predict the maximum subassembly coolant outlet temperatures and powers that satisfy the design safety criteria for steady-state normal operation conditions. The limiting steady-state temperature and power were then used as the initial subassembly thermal conditions for the off-normal transient analysis to assess the safety performance of the fuel pin for anticipated, unlikely and extremely unlikely events. If the design safety criteria for the off-normal events are not satisfied, then the initial steady-state subassembly temperatures and/or powers are reduced and an iterative procedure is employed until the design safety criteria for off-normal conditions are satisfied, and the initial subassembly outlet coolant temperature and power are the technical specification limits for reactor operation. (author)

A conceptual design of assembly strategy and dedicated tools for assembly of 40o sector

International Nuclear Information System (INIS)

Park, H.K.; Nam, K.O.; Kim, D.J.; Ahn, H.J.; Lee, J.H.; Im, K.; Shaw, R.

2010-01-01

The International Thermanuclear Experimental Reactor (ITER) tokamak device is composed of 9 vacuum vessel (VV)/toroidal field coils (TFCs)/vacuum vessel thermal shields (VVTS) 40 o sectors. Each VV/TFCs/VVTS 40 o sector is made up of one 40 o VV, two 20 o TFCs and associated VVTS segments. The 40 o sectors are sub-assembled at assembly hall respectively and then nine 40 o sectors sub-assembled at assembly hall are finally assembled at tokamak in-pit hall. The assembly strategy and tools for the 40 o sector sub-assembly and final assembly should be developed to satisfy the basic assembly requirements of the ITER tokamak device. Accordingly, the purpose-built assembly tools should be designed and manufactured considering assembly plan, available space, cost, safety, easy operation, efficient maintenance, and so on. The 40 o sector assembly tools are classified into 2 groups. One group is the sub-assembly tools including upending tool, lifting tool, sub-assembly tool, VV supports and bracing tools used at assembly hall and the other group is the in-pit assembly tools including lifting tool, central column, radial beams and their supports. This paper describes the current status of the assembly strategy and major tools for the VV/TFCs/VVTS 40 o sector assembly at in-pit hall and assembly hall. The conceptual design of the major assembly tools and assembly process at assembly hall and tokamak in-pit hall are presented also.

Research Performance Progress Report: Diverging Supernova Explosion Experiments on NIF

Energy Technology Data Exchange (ETDEWEB)

Plewa, Tomasz [Florida State Univ., Tallahassee, FL (United States)

2016-10-25

The aim of this project was to design a series of blast-wave driven Rayleigh-Taylor (RT) experiments on the National Ignition Facility (NIF). The experiments of this kind are relevant to mixing in core-collapse supernovae (ccSNe) and have the potential to address previously unanswered questions in high-energy density physics (HEDP) and astrophysics. The unmatched laser power of the NIF laser offers a unique chance to observe and study “new physics” like the mass extensions observed in HEDP RT experiments performed on the Omega laser [1], which might be linked to self-generated magnetic fields [2] and so far could not be reproduced by numerical simulations. Moreover, NIF is currently the only facility that offers the possibility to execute a diverging RT experiment, which would allow to observe processes such as inter-shell penetration via turbulent mixing and shock-proximity effects (distortion of the shock by RT spikes).

Inhibition effect of phosphorus-based chemicals on corrosion of carbon steel in secondary-treated municipal wastewater.

Science.gov (United States)

Shen, Zhanhui; Ren, Hongqiang; Xu, Ke; Geng, Jinju; Ding, Lili

2013-01-01

Secondary-treated municipal wastewater (MWW) could supply a viable alternative water resource for cooling water systems. Inorganic salts in the concentrated cooling water pose a great challenge to corrosion control chemicals. In this study, the inhibition effect of 1-hydroxy ethylidene-1,1-diphosphonic acid (HEDP), trimethylene phosphonic acid (ATMP) and 2-phosphonobutane-1,2,4-tricarboxylic acid (PBTCA) on corrosion of carbon steel in secondary-treated MWW was investigated by the means of potentiodynamic polarization and electrochemical impedance spectroscopy. The inhibition effect increased with increasing concentration of inhibitors. The corrosion rates of carbon steel were 1.5, 0.8, 0.2 and 0.5 mm a(-1) for blank, HEDP, ATMP and PBTCA samples at 50 mg L(-1), respectively. The phosphorus-based chemicals could adsorb onto the surface of the carbon steel electrode, form a coat of protective film and then protect the carbon steel from corrosion in the test solution.

HEDP and new directions for fusion energy

Science.gov (United States)

Kirkpatrick, Ronald C.

2010-06-01

Magnetic-confinement fusion energy and inertia-confinement fusion energy (IFE) represent two extreme approaches to the quest for the application of thermonuclear fusion to electrical energy generation. Blind pursuit of these extreme approaches has long delayed the achievement of their common goal. We point out the possibility of an intermediate approach that promises cheaper, and consequently more rapid development of fusion energy. For example, magneto-inertial fusion appears to be possible over a broad range of parameter space. It is further argued that imposition of artificial constraints impedes the discovery of physics solutions for the fusion energy problem.

HEDP and new directions for fusion energy

International Nuclear Information System (INIS)

Kirkpatrick, Ronald C.

2009-01-01

The Quest for fusion energy has a long history and the demonstration of thermonuclear energy release in 1951 represented a record achievement for high energy density. While this first demonstration was in response to the extreme fears of mankind, it also marked the beginning of a great hope that it would usher in an era of boundless cheap energy. In fact, fusion still promises to be an enabling technology that can be compared to the prehistoric utilization of fire. Why has the quest for fusion energy been so long on promises and so short in fulfillment? This paper briefly reviews past approaches to fusion energy and suggests new directions. By putting aside the old thinking and vigorously applying our experimental, computational and theoretical tools developed over the past decades we should be able to make rapid progress toward satisfying an urgent need. Fusion not only holds the key to abundant green energy, but also promises to enable deep space missions and the creation of rare elements and isotopes for wide-ranging industrial applications and medical diagnostics.

Post-irradiation examination of fifteen UO2/PuO2-fuel pins from the experiment DFR-350

International Nuclear Information System (INIS)

Geithoff, D.

1975-06-01

Within the framework of the fuel pin development for a sodium-cooled fast reactor a subassembly containing 77 fuel pins has been irradiated up to 5.65% fima in the Dounreay fast reactor. The pins were prototypes in terms of fuel and cladding material. The fuel consisted of mechanically mixed UO 2 (80%) and PuO 2 (20%) pressed into pellets whereas austenitic steels (W.-No. 1,4961 and 1,4988) were used as cladding material. Furthermore a blanket column of UO 2 pellets and a gas plenum were incorporated in the pin. For irradiation the conditions in a fast breeder were simulated by a linear rod power of 450 W/cm and a maximum cladding temperature of 630 0 C. After the successful completion of the irradiation, the subassembly was dismantled and fifteen pins were selected for a nondestructive and destructive examination. The tests included visual control, measurement of external dimensions, γ-spectroscopy, X-ray radiography, fission gas measurement, ceramography, radiochemical burn-up measurement. The results are presented. The most important results of the examinations seem to be the migration of fission product cesium and the fact that no signs of impending pin failure have been found. Thus the pin specification tested in this experiment is capable of achieving higher burnups under the irradiation conditions described above. (orig./AK) [de

Design of the ITER Tokamak Assembly Tools

International Nuclear Information System (INIS)

Park, Hyunki; Her, Namil; Kim, Byungchul; Im, Kihak; Jung, Kijung; Lee, Jaehyuk; Im, Kisuk

2006-01-01

ITER (International Thermonuclear Experimental Reactor) Procurement allocation among the seven Parties, EU, JA, CN, IN , KO, RF and US had been decided in Dec. 2005. ITER Tokamak assembly tools is one of the nine components allocated to Korea for the construction of the ITER. Assembly tools except measurement and common tools are supplied to assemble the ITER Tokamak and classified into 9 groups according to components to be assembled. Among the 9 groups of assembly tools, large-sized Sector Sub-assembly Tools and Sector Assembly Tools are used at the first stage of ITER Tokamak construction and need to be designed faster than seven other assembly tools. ITER IT (International Team) proposed Korea to accomplish ITA (ITER Transitional Arrangements) Task on detailed design, manufacturing feasibility and contract specification of specific, large sized tools such as Upending Tool, Lifting Tool, Sector Sub-assembly Tool and Sector Assembly Tool in Oct. 2004. Based on the concept design by ITER IT, Korea carried out ITA Task on detailed design of large-sized and specific Sector Sub-assembly and Sector Assembly Tools until Mar. 2006. The Sector Sub-assembly Tools mainly consist of the Upending, Lifting, Vacuum Vessel Support and Bracing, and Sector Sub-assembly Tool, among which the design of three tools are herein. The Sector Assembly Tools mainly consist of the Toroidal Field (TF) Gravity Support Assembly, Sector In-pit Assembly, TF Coil Assembly, Vacuum Vessel (VV) Welding and Vacuum Vessel Thermal Shield (TS) Assembly Tool, among which the design of Sector In-pit Assembly Tool is described herein

Modeling of the acoustic boiling noise of sodium during an assembly blockage in sodium-cooled reactors

International Nuclear Information System (INIS)

Vanderhaegen, M.

2013-01-01

In the framework of the fourth generation of nuclear reactors safety requirements, the acoustic boiling detection is studied to detect subassembly blockages. Boiling, that might occur during subassembly blockages and that can lead to clad failure, generates hydrodynamic noise that can be related to the two-phase flow. A bubble dynamics study shows that the sound source during subassembly boiling is condensation. This particular phenomenon generates most noise as a high subcooling is present in the subassembly and because of the high thermal diffusivity of sodium. This result leads to an estimate of the form of the acoustic spectrum that will be filtered and amplified during propagation inside the liquid. And even though it is unlikely that bubbles will be present inside the subassembly, due to the very gradual temperature profile at the wall and due to the geometry that leads to a strong confinement of the vapor, the historical bubble dynamics approach gives some insight in previous measurements. Additionally, some hypotheses can be disproved. These theoretical ideas are validated with a small water experiment, yet it also shows that a simple experience in sodium doesn't lead to a better knowledge of the acoustic source. A theoretical analysis also revealed that a realistic experiment with a simulant fluid, such as water or mercury, isn't representative. A similar conclusion is obtained when studying cavitation as a simulant acoustic source. As such, the acoustic detection of boiling, in comparison with other detection systems, isn't sufficiently developed yet to be applied as a reactor protective system. (author) [fr

High speed data transmission at the Superconducting Super Collider

International Nuclear Information System (INIS)

Leskovar, B.

1990-04-01

High speed data transmission using fiber optics in the data acquisition system of the Superconducting Super Collider has been investigated. Emphasis is placed on the high speed data transmission system overview, the local data network and on subassemblies, such as optical transmitters and receivers. Also, the performance of candidate subassemblies having a low power dissipation for the data acquisition system is discussed. 14 refs., 5 figs

MIF-SCD computer code for thermal hydraulic calculation of supercritical water cooled reactor core

International Nuclear Information System (INIS)

Galina P Bogoslovskaia; Alexander A Karpenko; Pavel L Kirillov; Alexander P Sorokin

2005-01-01

Full text of publication follows: Supercritical pressure power plants constitute the basis of heat power engineering in many countries to day. Starting from a long-standing experience of their operation, it is proposed to develop a new type of fast breeder reactor cooled by supercritical water, which enables the economical indices of NPP to be substantially improved. In the Thermophysical Department of SSC RF-IPPE, an attempt is made to provide thermal-hydraulic validation of the reactor under discussion. The paper presents the results of analysis of the thermal-hydraulic characteristics of fuel subassemblies cooled by supercritical water based on subchannel analysis. Modification of subchannel code MIF - MIF-SCD Code - developed in the SSC RF IPPE is designed as block code and permits one to calculate the coolant temperature and velocity distributions in fuel subassembly channels, the temperature of fuel pin claddings and fuel subassembly wrapper under conditions of irregular geometry and non-uniform axial and radial power generation. The thermal hydraulics under supercritical pressure of water exhibits such peculiarities as abrupt variation of the thermal physical properties in the range of pseudo-critical temperature, the absence of such phenomenon as the critical heat flux which can lead to fuel element burnout in WWERs. As compared with subchannel code for light water, in order to take account of the variation of the coolant properties versus temperature in more detail, a block for evaluating the thermal physical properties of supercritical water versus the local coolant temperature in the fuel subassembly channels was added. The peculiarities of the geometry and power generation in the fuel subassembly of the supercritical reactor are considered as well in special blocks. The results of calculations have shown that considerable preheating of supercritical coolant (several hundreds degrees) can occur in the fuel subassembly. The test calculations according to

A commutator

Energy Technology Data Exchange (ETDEWEB)

Siokhara, T.; Khibino, Y.; Tiaki, K.; Vatanabe, K.

1981-06-22

Switching of a microswitch, which switches the output electrical circuits of a unit, which supports control of industrial processes in time, is accomplished by a two arm rocking lever. The lever is driven by a cam mechanism, linked with a complex, multistage mechanical drive, which includes cylindrical gear drives and cam couplings for linkage, rotated by a synchronous electric micromotor. The unit is equipped with two subassemblies for establishment of process control: in functions of conditional time and functions of the times of day. The dials of the subassemblies for time establishment, located on the facial panel of the instrument, are equipped with arrows for the assigners with illumination and signal lamps. The lights control the mode of electric power for the instrument and the modes of the output circuits of the commutating subassembly.

Experimental research into operating strength and fatigue life of bodywork of buses and trolleybuses

International Nuclear Information System (INIS)

Dolhof, V.; Kepka, M.; Rehor, P.; Horak, V.; Sima, J.

1992-01-01

Operational strength and fatigue life reliability of trolleybus and bus bodies are usually assessed by computational methods in combination with selected tests. The latter include test runs of vehicles on real routes or on specially designed tracks, tests on complete vehicles under model test conditions and laboratory tests on selected materials, parts and subassemblies. This paper describes a method of experimental investigation of operational strength and reliability developed and applied at the Central Research Institute Skoda for public-transport road vehicles made in Czechoslovakia. (orig.)

Hydraulic experiments on the failed fuel location module of prototype fast breeder reactor

International Nuclear Information System (INIS)

Rajesh, K.; Kumar, S.; Padmakumar, G.; Prakash, V.; Vijayashree, R.; Rajan Babu, V.; Govinda Rajan, S.; Vaidyanathan, G.; Prabhaker, R.

2003-01-01

The design of Prototype Fast Breeder Reactor (PFBR) is based on sound design concepts with emphasis on intrinsic safety. The uncertainties involved in the design of various components, which are difficult to assess theoretically, are experimentally verified before design is validated. In PFBR core, the coolant (liquid sodium) enters the bottom of the fuel subassembly, passes over the fuel pins picking up the fission heat and issues in to a hot pool. If there is any breach in the fuel pins, the fission products come in direct contact with the coolant. This is undesirable and it is necessary to locate the subassembly with the failed fuel pin and to isolate it. A component called Failed Fuel Location Module (FFLM) is employed for locating the failed SA by monitoring the coolant samples coming out of each Subassembly. The coolant sample from each Subassembly is drawn by FFLM using an EM pump through sampling tube and selector valve and is monitored for the presence of delayed neutrons which is an indication of failure of the Subassembly. The pressure drop across the selector valve determines the rating of the EM Pump. The dilution of the coolant sample across the selector valve determines the effectiveness of monitoring for contamination. It is not possible to predict pressure drop across the selector valve and dilution of the coolant sample theoretically. These two parameters are determined using a hydraulic experiment on the FFLM. The experiment was carried out in conditions that simulate the reactor conditions following appropriate similarity laws. The paper discusses the details of the model, techniques of experiments and the results from the studies

Investigation of the three-dimensional thermoelastic deformation of the core structure of a fast breeder reactor under stationary working conditions

International Nuclear Information System (INIS)

Yong-Su, Hoang.

1976-12-01

In this study a method is described which has been developed in order to calculate three-dimensional deformation of the reactor core, taking into account thermal expansion. Two problem areas are of particular importance: 1) The spatial deflection of subassemblies in specified flexible supports and with specified clearances; 2) The investigation of the equilibrium configurations of the subassemblies in the planes of clamping (problem of clamping plane). - The elementary theory of beam deflection has been used to calculate the deformation of subassemblies. However, particular problems have been encountered as a result of flexibly designed support configurations having some spatial clearances. The problem has essentially been solved in two steps: a) Uniqueness analysis of the beam-support configuration; b) Calculation of the support loads and bending line for the unique beam-support configuration. - Basic difficulties currently prevent the problem of clamping plane being solved in a satisfactory manner. Therefore, a simplified clamping model was used for supports without spatial clearance and a parametric study was performed for supports having spatial clearance. The computation method developed is applied to the MARK I core of SNR 300. Core deformations are calculated under different support conditions for the subassemblies in the grid plate and in the upper clamping plane. (orig./HR) [de

A novel method for in-situ estimation of time constant for core temperature monitoring thermocouples of operating reactors

International Nuclear Information System (INIS)

Sylvia, J.I.; Chandar, S. Clement Ravi; Velusamy, K.

2014-01-01

Highlights: • Core temperature sensor was mathematically modeled. • Ramp signal generated during reactor operating condition is used. • Procedure and methodology has been demonstrated by applying it to FBTR. • Same technique will be implemented for all fast reactors. - Abstract: Core temperature monitoring system is an important component of reactor protection system in the current generation fast reactors. In this system, multiple thermocouples are housed inside a thermowell of fuel subassemblies. Response time of the thermocouple assembly forms an important input for safety analysis of fast reactor and hence frequent calibration/time constant estimation is essential. In fast reactors the central fuel subassembly is provided with bare fast response thermocouples to detect under cooling events in reactor and take proper safety action. On the other hand, thermocouples in thermowell are mainly used for blockage detection in individual fuel subassemblies. The time constant of thermocouples in thermowell can drift due to creep, vibration and thermal fatigue of the thermowell assembly. A novel method for in-situ estimation of time constant is proposed. This method uses the Safety Control Rod Accelerated Mechanism (SCRAM) or lowering of control Rod (LOR) signals of the reactor along with response of the central subassembly thermocouples as reference data. Validation of the procedure has been demonstrated by applying it to FBTR

Renal bone disease and extraskeletal calcification during dialysis and after transplantation

International Nuclear Information System (INIS)

Graaf, P. de.

1980-01-01

The author reports 10 studies concerning the diagnosis of renal osteodystrophy and extraskeletal calcification in patients on maintenance hemodialysis as well as some aspects of persistent hyperparathyroidism after renal transplantation. The majority of the studies focus on the value of bone scintigraphy with Tc-99m HEDP in the diagnosis of these disorders. (Auth.)

A method for starting up the gas isolating section of a benzine pyrolysis installation

Energy Technology Data Exchange (ETDEWEB)

Bruskin, Yu.A.; Gorokhov, V.V.; Kotler, L.D.; Shevchenko, K.N.; Zeldin, V.Ye.

1982-01-01

In the method for starting up a gas isolation section of a benzine pyrolysis unit, which includes starting the demethanizer in a steaming mode, starting the ethane and ethylene and propane and propylene towers, filling the ethylene cooling system with a hydrocarbon fraction, starting the ethylene cooling system, switching to a mode of the demethanization unit, starting the subassemblies for hydration and isolation of the ethylene and propylene fractions, in order to reduce the length of the start up period and to reduce the ejection of gas to the burner, after starting the demethanizer in the steaming mode, starting the ethane and ethylene and propane and propylene towers, the ethane and ethylene and propane and propylene fractions are mixed before the hydration subassemblies with an H2 bearing gas for catalytic reforming and then the other units and subassemblies are started.

EIS study on corrosion and scale processes and their inhibition in cooling system media

International Nuclear Information System (INIS)

Marin-Cruz, J.; Cabrera-Sierra, R.; Pech-Canul, M.A.; Gonzalez, I.

2006-01-01

A study of the carbon steel/cooling water interface was carried out using electrochemical impedance spectroscopy (EIS). EIS spectra reveal that a layer of corrosion and scale products forms naturally and evolves with the immersion time modifying the carbon steel/cooling water interface and giving rise to corrosion and scale processes. In addition, the nature of the layer formed on the metal was found to depend on the inhibitor used. It was established that the corrosion inhibitor (hydroxyphosphonoacetic acid (HPA)) chelates with Ca(II) ion generating a layer with resistive properties that provides good protection against corrosion. In contrast, the scale inhibitor (1-hydroxy-ethane-1,1-diphosphonic acid (HEDP)) is incorporated into the calcium carbonate crystals at the surface, modifying the structure and diminishing scale formation in the surface; this additive additionally inhibited corrosion. These observations were supported by scanning electronic microscopy (SEM) and corroborate previous studies performed by other techniques on HPA and HEDP. Finally, a synergistic effect was observed between these inhibitors that provides good protection to steel against corrosion and scaling in cooling media

Corrosion control for open cooling water system

International Nuclear Information System (INIS)

Karweer, S.B.; Ramchandran, R.

2000-01-01

Frequent stoppage of water circulation due to shut down of the Detritiation Plant in Heavy Water Division, Trombay resulted in considerable algae growth. As polyphosphate is a nutrient to algae growth, studies were directed in the evaluation of a nonpolyphosphate formulation for controlling corrosion and scale formation of carbon-steel, copper and aluminium. A blend of HEDP, polyacrylate, zinc, and benzotriazole was used and the optimum condition was determined. In presence of 25 ppm kw-1002 [proprietary formulation, containing HEDP and polyacrylate], 10 ppm kw-201 [active ingredient benzotriazole] and 2 ppm zinc (as zinc sulphate), the corrosion rate of carbon-steel in Mumbai Municipal Corporation (MMC) water at pH 7.5 ± 0.1 for a period of 31 days was 10.4 x 10 -3 μm/h. When MMC water concentrated to half its original volume was used, the corrosion rate was still 9.74 x 10 -3 μm/h close to the original value without concentration. Hence, this formulation was used for controlling scale and corrosion. The results were satisfactory. (author)

EIS study on corrosion and scale processes and their inhibition in cooling system media

Energy Technology Data Exchange (ETDEWEB)

Marin-Cruz, J. [Universidad Autonoma Metropolitana, Departamento de Quimica, Apdo. Postal 55-534, 09340 Mexico, DF (Mexico) and Instituto Mexicano del Petroleo, Coordinacion de Ingenieria Molecular, Competencia de Quimica Aplicada, Eje Central Lazaro Cardenas No. 152, CP 07730, DF (Mexico)]. E-mail: jmarin@imp.mx; Cabrera-Sierra, R. [Universidad Autonoma Metropolitana, Departamento de Quimica, Apdo. Postal 55-534, 09340 Mexico, DF (Mexico); Escuela Superior de Ingenieria Quimica e Industrias Extractivas (ESIQIE-IPN), Departamento de Metalurgia, UPALM Zacatenco AP 75-874, CP 07338, DF (Mexico); Pech-Canul, M.A. [Departamento de Fisica Aplicada, Centro de Investigacion y de Estudios, Avanzados del IPN, AP 73 Cordemex, CP 97310, Merida, Yucatan (Mexico); Gonzalez, I. [Universidad Autonoma Metropolitana, Departamento de Quimica, Apdo. Postal 55-534, 09340 Mexico, DF (Mexico)]. E-mail: igm@xanum.uam.mx

2006-01-20

A study of the carbon steel/cooling water interface was carried out using electrochemical impedance spectroscopy (EIS). EIS spectra reveal that a layer of corrosion and scale products forms naturally and evolves with the immersion time modifying the carbon steel/cooling water interface and giving rise to corrosion and scale processes. In addition, the nature of the layer formed on the metal was found to depend on the inhibitor used. It was established that the corrosion inhibitor (hydroxyphosphonoacetic acid (HPA)) chelates with Ca(II) ion generating a layer with resistive properties that provides good protection against corrosion. In contrast, the scale inhibitor (1-hydroxy-ethane-1,1-diphosphonic acid (HEDP)) is incorporated into the calcium carbonate crystals at the surface, modifying the structure and diminishing scale formation in the surface; this additive additionally inhibited corrosion. These observations were supported by scanning electronic microscopy (SEM) and corroborate previous studies performed by other techniques on HPA and HEDP. Finally, a synergistic effect was observed between these inhibitors that provides good protection to steel against corrosion and scaling in cooling media.

A New Approach for Heating the Plastics Injection Units

Directory of Open Access Journals (Sweden)

Virgilius Vasilache

2010-06-01

Full Text Available The plastics injection molding machines are one of the most eager consumers of energy. The plasticizing unit itself is the most important energetic consumer among the subassemblies of these machines; that is why this subassembly is the target of most actions of consumption decreasing on such machines. Our concerns on this direction got the shape of developing a new heating system for the plasticizing unit, which system was already patented [1].

Contribution of materials investigations and operating experience of reactor vessel internals to PWRs' safety, performance and reliability

International Nuclear Information System (INIS)

Lemaire, E.; Monteil, N.; Jardin, N.; Doll, M.

2015-01-01

The Reactor Pressure Vessel Internals (RVI) include all the components inside the pressure vessel, except the nuclear fuel, the rod cluster assemblies and the instrumentation. The RVI consist of bolted and welded structures that are divided into two sub-assemblies: the upper internals which are removed at every refueling outage and the lower internals which are systematically removed for inspection at every 10-year outage. The main functions of the RVI are to position the core, to support it, and to provide a coolant flow by channeling the fluid. Moreover, the lower internals contribute to a neutron protection of the reactor pressure vessel by absorbing most of the neutron flux from the core. Depending on their location and material composition, the RVI components can face different ageing phenomena, that are actual or potential (such as wear, fatigue, stress corrosion cracking, irradiation assisted stress corrosion cracking, hardening and loss of ductility due to neutron irradiation, irradiation creep and irradiation swelling). EDF has developed a strategy for managing ageing and demonstrating the capacity of the RVI to perform their design functions over 40 years of operation. This overall approach is periodically revisited to take into account the most recent knowledge obtained from the following main topics: Safety Analyses, Research-Development programs, In-Service Inspection (ISI) results, Maintenance programs and Metallurgical Examinations. Based on continuous improvements in those fields, the goal of this paper is to present the way that materials investigations and operating experience obtained on RVI are managed by EDF to improve RVI safety, performance and reliability. It is shown that a perspective of 60 years of operation of RVI components is supported by large Research-Development efforts combined with field experience. (authors)

Destructive Examination of Shipping Package 9975-02019

Energy Technology Data Exchange (ETDEWEB)

Daugherty, W. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-06-13

Destructive and non-destructive examinations have been performed on the components of shipping package 9975-02019 as part of a comprehensive SRS surveillance program for plutonium material stored in the K-Area Complex (KAC). During the field surveillance inspection of this package in KAC, two non-conforming conditions were noted: the axial gap of 1.577 inch exceeded the 1 inch maximum criterion, and two areas of dried glue residue were noted on the upper fiberboard subassembly. This package was subsequently transferred to SRNL for more detailed inspection and destructive examination. In addition to the conditions noted in KAC, the following conditions were noted: - Numerous small spots of corrosion were observed along the bottom edge of the drum. - In addition to the smeared glue residue on the upper fiberboard subassembly, there was also a small dark stain. - Mold was present on the side and bottom of the lower fiberboard subassembly. Dark stains from elevated moisture content were also present in these areas. - A dark spot with possible light corrosion was observed on the primary containment vessel flange, and corresponding rub marks were observed on the secondary containment vessel ID. - The fiberboard thermal conductivity in the radial orientation was above the specified range. When the test was repeated with slightly lower moisture content, the result was acceptable. The moisture content for both tests was within a range typical of other packages in storage. The observed conditions must be fully evaluated by KAC to ensure the safety function of the package is being maintained. Several factors can contribute to the concentration of moisture in the fiberboard, including higher than average initial moisture content, higher internal temperature (due to internal heat load and placement within the array of packages), and the creation of additional moisture as the fiberboard begins to degrade.

Determination of mechanical properties of carbon/epoxy plates by tensile stress test

Science.gov (United States)

Bere, Paul; Krolczyk, Jolanta B.

2017-10-01

The polymeric composite materials used in aerospace, military, medical or racing cars manufacturing end up being used in our daily life Whether we refer to the performing vehicles, subassemblies or parts for aircrafts, wind, telegraph poles, or medical prostheses they all are present in our lives and they are made of composite materials (CM). This paper presents research regarding three different composite materials, plates by carbon fiber, in epoxy matrix. Starting with materials presentation, manufacturing methodology and determination of mechanical properties at carbon fiber/epoxy were done. Vacuum bag technology to obtain the composite structure offer opportunity to get a very compact and homogeny composite structure. For the moment this technology are adequate for high performances pieces. The mechanical characteristics of plates made of composite materials reinforced presented indicates closed value like metal materials. Based on the results, a comparative study between the reinforced materials typically used to manufacture the plates of CM is carried out.

Study of burned optimization for minor actinides in European Sodium Fast Reactor (ESFR) by use of moderator materials

International Nuclear Information System (INIS)

Ramos, R L; Villanueva, A J; Buiront, L

2012-01-01

The minor actinides (MA) burn up optimization in the European Sodium Fast Reactor (ESFR) core was studied by adding different moderating materials in the Minor Actinides Bearing Blanket subassemblies (MABB SA) using the ERANOS neutron code package. These SA are of hexagonal shape and are composed of pellets inside of pins. These pellets contain a mixture of uranium dioxide (UO 2 ) and americium dioxide (AmO 2 ). If some of these pins are replaced by other identical ones containing moderating material instead of minor actinides, a shift in the spectrum towards lower energies is expected, which might enhance the burn up performance. The results of this work demonstrated that the use of compounds of hydrogen and magnesium as moderators produces a shift in the neutron spectrum, improving the porcentual minor actinides consumption. ZrH 2 moderator material was found to exhibit the best performances for this propose, followed by MgO and MgAl 2 O 4 , in that order. The use of SiC, BeO, TiC, LiO 2 and ZrC material produced no effect on the shift of the neutron spectrum. For safety reasons, it seems hardly realistic to use hydrogenous compounds in sodium fast reactors. So, compounds with magnesium are selected to be placed into the pins to improve the porcentual minor actinides consumption. The ESFR core is composed by 817 SA, 453 of them are fuel SA, 247 are reflectors SA, 84 are MABB (Minor Actinides Bearing Blankets) SA and 33 are control and shutdown rods. When about half of the total pins in each MABB were substituted by moderator pins with MgO pellets (135 of 271 pins), the porcentual consumption of minor actinides was of 30.85 %, i.e., 227.22 kg of minor actinides were consumed out of 736.65 kg in the initial configuration. In the case where all the pins of the MABB contained pellets of minor actinides, the porcentual consumption of minor actinides was of 21.26 %, i.e., 312.13 kg of minor actinides were consumed of 1467.87 kg in the initial configuration (author)

Selenide isotope generators for the Galileo Mission: SIG hermetic bimetal weld transition joint

International Nuclear Information System (INIS)

Barnett, W.J.

1979-08-01

The successful development of the commercial 6061-T651/Silver/304L explosive clad plate material as a bimetal weld transition joint material, as described herein, satisfies all SIG Galileo design requirements for hermetic weld attachment of stainless steel subassemblies to aluminum alloy generator housing or end cover structures. The application of this type weld transition joint to the hermetic attachment of stainless steel shell connectors is well-developed and tested. Based on on-going life tests of stainless steel receptacle/bimetal ring attachment assemblies and metallurgical characterization studies of this transition joint material, it appears evident that this transition joint material has more than adequate capability to meet the 250 to 300 0 F and 50,000 hr. design life of the SIG/Galileo mission. Its extended life temperture capability may well approach 350 to 400 0 F

Hardware concepts for a large low-energetics LMFBR core. Final report

International Nuclear Information System (INIS)

Hutter, E.; Batch, R.V.

1980-12-01

A design study was made to identify a practical set of hardware configurations that would embody the requirements developed in the numerical study of a low-energetics core and blanket for a prototype large breeder reactor. Dimensioned drawings are presented for fuel, blanket, reflector/shield, and control rod subassemblies. A horizontal cross section drawing shows how these subassemblies are arranged in the total core/blanket assembly. A core support is illustrated showing a dual plenums arrangement

EVALUATION OF THE MACHINE MODERNITY IN THE MOTOR INDUSTRY

OpenAIRE

Manuela Krystyna Ingaldi

2014-01-01

Most manufacturing companies realize its technologies, implemented through concrete machinery parts. They differ in terms of importance, the relevance of their selection and the level of their modernity. The purpose of this article is to analyse the chosen production machine in terms of its modernity. The ABC technology method was chosen do this research. All parts of the machine were divided into three groups: parts of main subassembly A, parts of supportive subassembly B, parts of collatera...

Study on natural convection in core barrel. Experimental and numerical results for band type spacer pads

International Nuclear Information System (INIS)

Hayashi, Kenji; Kawamata, Nobuhiro; Kamide, Hideki

2003-03-01

In a fast reactor an Inter-Wrapper Flow (IWF) is one of significant phenomena for decay heat removal under natural circulation condition, when a direct reactor auxiliary cooling system (DRACS) is adopted for decay heat removal system. Cold coolant provided by dipped heat exchangers (DHX) of DRACS can penetrate into the core barrel (region between the subassemblies) and it makes natural convection int he core barrel. Such IWF will depend on a spacer pad geometry of subassemblies. Water experiment, TRIF (Test Rig for Inter-wrapper Flow), was carried out for IWF in a reactor core. The test section modeled a 1/12th sector of the core and upper plenum of reactor vessel. Experimental parameters were the spacer pad geometry and flow path geometries connecting the upper plenum and core barrel. Numerical simulation using AQUA code was also performed to confirm applicability of a simulation method. An experimental series using a button type spacer pad had been carried out. Here a band type spacer pad was examined. Temperatures at subassembly wall were measured with parameter of the flow path geometries; one was a connection pipe between the upper plenum and core barrel and the other was flow hole in core former plates between the outermost subassemblies and the core barrel. It was found that these flow paths were effective to remove heat in the core in case of the band type spacer pad. A general purpose three dimensional analysis code, AQUA, was applied to the experimental analysis. Each subassembly and inter wrapper gap region were modeled by slab mesh geometry. Pressure loss coefficient at the pacer pad was set based on the geometry. The numerical simulation results were in good agreement with measured temperature profiles in the core. (author)

A Conceptual Design and Structural Analysis for ITER Mid-plane Brace Tools

Energy Technology Data Exchange (ETDEWEB)

Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Jae Hyuk; Kim, Kyung kyu [SFA Engineering Corp., Changwon (Korea, Republic of); Im, Kihak; Robert, Shaw [ITER Organization, St Paul lez Durance Cedex (France)

2010-10-15

The ITER, International Thermonuclear Experimental Reactor, Tokamak machine is mainly composed of 9 vacuum vessel (VV)/toroidal field coils (TFCs)/vacuum vessel thermal shields (VVTS) 40 .deg. sectors. Each VV/TFCs/VVTS 40 .deg. sector is made up of one 40 .deg. VV, two 20 .deg. TFCs and associated VVTS segments. The ITER Tokamak assembly tools are purpose-built tools to assemble the ITER Tokamak machine which includes the cryostat and the components contained therein. Based on the design description document prepared by the IO (ITER international organization), Korea has carried out the conceptual design of assembly tools with IO cooperation. The 40 .deg. sector assemblies attached mid-plane brace tools sub-assembled at assembly hall are transferred to Tokamak hall using the lifting tool operated by Tokamak main cranes. The sector sub-assembly tools are composed of the upending tool, the sector sub-assembly tool, the sector lifting tool and the vacuum vessel support and mid-plane brace tools. The mid-plane brace tool is assembled to inner surface of VV and TFCs in phase of sector sub-assembly after completion of all sector components. VV, TFC and VVTS are separated fully before completion of 9 sectors at Tokamak in-pit. In this paper the mid-plane brace tools is introduced about function, structure and status of research and development are also described

JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Ohkawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi; Aoyama, Takafumi

2003-04-01

The 'JOYO' MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. Total control rod worth, reactor kinetic parameters and the MK-II core performance test results were included per user's requests. The core characteristics obtained from the 32 nd to 35 th operational cycles, which were conducted in the MK-III transition core, were newly added in this revised version. The MK-II core management data and core characteristics data were recorded to CD-ROM for user convenience. The Configuration Data' include the core arrangement and refueling record for each operational cycle. The 'Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The 'Output Data' contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The 'Core Characteristics Data' include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. (author)

Scintillation crystal mounting apparatus

International Nuclear Information System (INIS)

Engdahl, L.W.; Deans, A.J.

1982-01-01

An improved detector head for a gamma camera is disclosed. The detector head includes a housing and a detector assembly mounted within the housing. Components of the detector assembly include a crystal sub-assembly, a phototube array, and a light pipe between the phototube array and crystal sub-assembly. The invention provides a unique structure for maintaining the phototubes in optical relationship with the light pipe and preventing the application of forces that would cause the camera's crystal to crack

Diffraction scattering and disintegration of 3He nuclei by atomic nuclei

International Nuclear Information System (INIS)

Koval'chuk, V.I.

2006-01-01

Within diffraction model framework a method of cross sections calculation for scattering and disintegration of weakly-bounded two-clustered nuclei by nuclei when both of its clusters are changed has been proposed. The experimental elastic scattering cross sections of 3 He by 40 Ca, 90 Zr and coincidence spectra of disintegration products from 28 Si( 3 He,dp) have been described

A study on fission product retention capability in a sodium coolant system

International Nuclear Information System (INIS)

Satoh, K.; Kubo, S.; Hashiguchi, Y.; Itooka, S.; Akatsu, Y.; Miyagi, K.; Wakamatsu, M.; Endo, H.; Tachino, T.

1992-01-01

Three kinds of separate model tests have been performed using water and air, focusing on the transport behavior of FP gas bubbles from subassembly outlets into a cover gas region, to study the dominant processes regarding the retention for volatiles ejected with inert gas into sodium after fuel failures. In the case that whole fuel pin failures occurring coherently in a subassembly were assumed, a periodic formation of globules was observed at the subassembly outlet. The globules rapidly broke up into small bubbles of less than 10 mm in mean diameter. The small bubbles at the top region had a tendency to be coalesced during rising through the upper plenum. As the coolant flow rate increased, bubble deformation and breakup were accelerated, but the bubble transport time did not vary remarkably. It is expected that bubbles in sodium would play in a similar way as in the water test, and the importance of the bubble behavior for the retention capability of volatiles has been confirmed. (author)

Analysis of steady state temperature distribution in rod arrays exposed to stagnant gaseous environment

International Nuclear Information System (INIS)

Pal, G.; MaarKandeya, S.G.; Raj, V.V.

1991-01-01

This paper deals with the calculation of radiative heat exchange in a rod array exposed to stagnant gaseous environment. a computer code has been developed for this purpose and has been used for predicting the steady state temperature distribution in a nuclear fuel sub-assembly. Nuclear fuels continue to generate heat even after their removal from the core. During the transfer of the nuclear fuel sub-assemblies from the core to the storage bay, they pass through stagnant gaseous environment and may remain there for extended periods under abnormal conditions. Radiative heat exchange will be the dominant mode within the sub-assembly involved, since axial heat conduction through the fuel pins needs to be accounted for. a computer code RHEINA-3D (Radiative Heat Exchange In Nuclear Assemblies -3D) has been developed based on a simplified numerical model which considers both the above-mentioned phenomena. The analytical model and the results obtained are briefly discussed in this paper

A time relay

Energy Technology Data Exchange (ETDEWEB)

Yosimura, K.; Sudzuki, Y.

1981-06-18

The synchronous micromotor of the time relay by means of a two staged cylindrical gear drive drives the gear wheel and the shaft of an actuating mechanism. The shaped drum of a cam mechanism, equipped with a vertical groove, which interacts in its upper part with a lever for driving the first commutating subassembly and in the lower, with a bent sector of a spring and plate movable contact of the second commutating subassembly, is attached to the lower end of the mechanism's shaft (V). The L-shaped lever of the second commutating subassembly's drive rests on a vertical rocking axle, located parallel to the shaft. Both pairs of spring and plate contacts are bracketed in two dielectric brackets which provide for a plane parallel disposition of the cited contacts. The operational time setting for the unit is a function of the initial angular position of the shaft, which is provided for by the attachment of a handle on its upper end.

Evaluation of wrapper tube temperatures of fast neutron reactors using the TRANSCOEUR-2 code

Energy Technology Data Exchange (ETDEWEB)

Valentin, B.; Brun P. [CEA/DRN/DEC/SECA/LHC CEN, St Paul Lez Durance (France); Chaigne, G. [FRAMATOME/NOVATOME, Lyon (France)

1995-09-01

This paper deals with the thermal loading estimation of wrapper tubes using the TRANSCOEUR-2 code. This estimation requires a knowledge of two temperature fields: the first involves the peripheral sub-channel temperatures of each sub-assembly calculated by the design code CADET, and the second, outside the sub-assemblies, is the inter-wrapper flow temperature field calculated by the thermal-hydraulic code TRIO-VF with boundary conditions taken from CADET. Theoretical models of the three codes are presented as well as the first TRANSCOEUR-2 wrapper tube temperature calculation performed on the European Fast Reactor (EFR) Core Design 6/91 (CD 6/91) under nominal power conditions. The results show a temperature variation of 115{degrees}C between the bottom of the lower blanket and the top of the upper blanket fuel sub-assemblies in the center of the core and 95{degrees}C at the core periphery. The wrapper tube temperatures are higher in the center than in the external core.

JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Tabuchi, Shiro; Aoyama, Takafumi; Nagasaki, Hideaki; Kato, Yuichi

1998-12-01

The experimental fast reactor JOYO served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, extensive data were accumulated from the core characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database. The code system MAGI has been developed and used for core management of JOYO MK-II, and the core characteristics and the irradiation test conditions were calculated using MAGI on the basis of three dimensional diffusion theory with seven neutron energy groups. The core management data include extensive data, which were recorded on CD-ROM for user convenience. The data are specifications and configurations of the core, and for about 300 driver fuel subassemblies and about 60 uninstrumented irradiation subassemblies are core composition before and after irradiation, neutron flux, neutron fluences, fuel and control rod burn-up, and temperature and power distributions. MK-II core characteristics and test conditions were stored in the database for post analysis. Core characteristics data include excess reactivities, control rod worths, and reactivity coefficients, e.g., temperature, power and burn-up. Test conditions include both measured and calculated data for irradiation conditions. (author)

Steady natural convection heat transfer experiments in a horizontal annulus for the United States Spent Fuel Shipping Cask Technology Program

International Nuclear Information System (INIS)

Boyd, R.D.

1981-04-01

This experimental study deals with the measurement of the heat transfer across a horizontal annulus which is formed by an inner hexagonal cylinder and an outer concentric circular cylinder. The geometry simulates, in two dimensions, a liquid metal fast breeder reactor radioactive fuel subassembly inside a shipping container. This geometry is also similar to a radioactive fuel pin inside a horizontal reactor subassembly. The objective of the experiments is to measure the local and mean heat transfer at the surface of the inner hexagonal cylinder

Study of the neutronic performances of cores with mixed nitride fuel [(U,Pu)N] for fast neutron reactors

International Nuclear Information System (INIS)

Merzouk, Hamid

1992-01-01

This paper proposes a core design of fast reactor using mixed nitride fuel [(U,Pu)N], having small loss of reactivity and reaching a maximum thermal burn-up rate from 150 GWd/t, while being managed in single batch (renewal of the fuel in only one time for all the subassemblies of the core). This work was completed with aid of the studies of sensibilities of the fast reactors cores to principal parameters: general design of the core, volumetric percentages of the various mixture of materials composing the core, initial enrichments of the fuel. A detailed optimization study on the selected core was conducted complying with safety criteria taking into consideration of consequences of nitride material presence on fuel assembly design rules. (author) [fr

Laser welding of tailored blanks

International Nuclear Information System (INIS)

Pecas, P.; Gouveia, H.; Quintino, L.; Olsen, F.O.

1998-01-01

Laser welding has an increasing role in the automotive industry, namely on the sub-assemblies manufacturing. Several sheet-shape parts are laser welded, on a dissimilar combination of thicknesses and materials, and are afterwards formed (stamped) being transformed in a vehicle body component. In this paper low carbon CO 2 laser welding, on the thicknesses of 1,25 and 0.75 mm, formability investigation is described. There will be a description of how the laser welded blanks behave in different forming tests, and the influence of misalignment and undercut on the formability. The quality is evaluated by measuring the limit strain and limit effective strain for the laser welded sheets and the base material, which will be presented in a forming limit diagram. (Author) 14 refs

Complexes of technetium with pyrophosphate, etidronate, and medronate

International Nuclear Information System (INIS)

Russell, C.D.; Cash, A.G.

1979-01-01

The reduction of [ 99 Tc]pertechnetate was studied as a function of pH in complexing media of pyrophosphate, methylene diphosphonate (MDP), and ethane-1, hydroxy-1, and 1-diphosphonate (HEDP). Test (sampled d-c) and normal-pulse polarography were used to study the reduction of pertechnetate, and normal-pulse polarography (sweeping in the anodic direction) to study the reoxidation of the products. Below pH 6 TcO 4 - was reduced to Tc(III), which could be reoxidized to Tc(IV). Above pH 10, TcO 4 - was reduced in two steps to Tc(V) and Tc(IV), each of which could be reoxidized to TcO 4 - . Between pH 6 and 10 the results differed according to the ligand present. In pyrophosphate and MDP, TcO 4 - was reduced in two steps to Tc(IV) and Tc(III); Tc(III) could be reoxidized in two steps to Tc(IV) and TcO 4 - . In HEDP, on the other hand, TcO 4 - was reduced in two steps to Tc(V) and Tc(III), and could be reoxidized to Tc(IV) and TcO 4 - . Additional waves were observed; they apparently led to unstable products

Nuclear medicine program progress report for quarter ending December 31, 1996

Energy Technology Data Exchange (ETDEWEB)

Knapp, F.F. Jr.; Beets, A.L.; Boll, R.; Luo, H.; McPherson, D.W.; Mirzadeh, S.

1997-03-20

In this report the authors describe the use of an effective method for concentration of the rhenium-188 bolus and the results of the first Phase 1 clinical studies for bone pain palliation with rhenium-188 obtained from the tungsten-188/rhenium-188 generator. Initial studies with therapeutic levels of Re-188-HEDP at the Clinic for Nuclear Medicine at the University of Bonn, Germany, have demonstrated the expected good metastatic uptake of Re-188-HEDP in four patients who presented with skeletal metastases from disseminated prostatic cancer with good pain palliation and minimal marrow suppression. In addition, skeletal metastatic targeting of tracer doses of Re-188(V)-DMSA has been evaluated in several patients with metastases from prostatic cancer at the Department of Nuclear Medicine at the Canterbury and Kent Hospital in Canterbury, England. In this report the authors also describe further studies with the E-(R,R)-IQNP ligand developed in the ORNL Nuclear Medicine Program as a potential imaging agent for detection of changes which may occur in the cerebral muscarinic-cholinergic receptors (mAChR) in Alzheimer`s and other diseases.

Ship construction and welding

CERN Document Server

Mandal, Nisith R

2017-01-01

This book addresses various aspects of ship construction, from ship types and construction materials, to welding technologies and accuracy control. The contents of the book are logically organized and divided into twenty-one chapters. The book covers structural arrangement with longitudinal and transverse framing systems based on the service load, and explains basic structural elements like hatch side girders, hatch end beams, stringers, etc. along with structural subassemblies like floors, bulkheads, inner bottom, decks and shells. It presents in detail double bottom construction, wing tanks & duct keels, fore & aft end structures, etc., together with necessary illustrations. The midship sections of various ship types are introduced, together with structural continuity and alignment in ship structures. With regard to construction materials, the book discusses steel, aluminum alloys and fiber reinforced composites. Various methods of steel material preparation are discussed, and plate cutting and form...

Producing charcoal from wastes

Energy Technology Data Exchange (ETDEWEB)

Pogorelov, V.A.

1983-01-01

Experimental works to use wood wastes for producing charcoal are examined, which are being conducted in the Sverdlovsk assembly and adjustment administration of Soyuzorglestekhmontazh. A wasteless prototype installation for producing fine charcoal is described, along with its subsequent briqueting, which is made on the basis of units which are series produced by the factories of the country. The installation includes subassemblies for preparing and drying the raw material and for producing the charcoal briquets. In the opinion of specialists, the charcoal produced from the wastes may be effectively used in ferrous and nonferrous metallurgy and in the production of pipes.

Methodologies for Verification and Validation of Space Launch System (SLS) Structural Dynamic Models: Appendices

Science.gov (United States)

Coppolino, Robert N.

2018-01-01

Verification and validation (V&V) is a highly challenging undertaking for SLS structural dynamics models due to the magnitude and complexity of SLS subassemblies and subassemblies. Responses to challenges associated with V&V of Space Launch System (SLS) structural dynamics models are presented in Volume I of this paper. Four methodologies addressing specific requirements for V&V are discussed. (1) Residual Mode Augmentation (RMA). (2) Modified Guyan Reduction (MGR) and Harmonic Reduction (HR, introduced in 1976). (3) Mode Consolidation (MC). Finally, (4) Experimental Mode Verification (EMV). This document contains the appendices to Volume I.

Pressurized solid oxide fuel cell integral air accumular containment

Science.gov (United States)

Gillett, James E.; Zafred, Paolo R.; Basel, Richard A.

2004-02-10

A fuel cell generator apparatus contains at least one fuel cell subassembly module in a module housing, where the housing is surrounded by a pressure vessel such that there is an air accumulator space, where the apparatus is associated with an air compressor of a turbine/generator/air compressor system, where pressurized air from the compressor passes into the space and occupies the space and then flows to the fuel cells in the subassembly module, where the air accumulation space provides an accumulator to control any unreacted fuel gas that might flow from the module.

SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Basehore, K.L.; Todreas, N.E.

1980-08-01

Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

International Nuclear Information System (INIS)

Basehore, K.L.; Todreas, N.E.

1980-08-01

Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries

Mechanical behavior of a fast reactor core: Application of the 3D codes to SUPER PHENIX 1

International Nuclear Information System (INIS)

Bernard, A.; Masoni, P.; Dorsselaere, J.P. van

1983-01-01

The series of the 3-dimensional mechanical codes of a fast reactor core was used for the first time within the framework of a design study of an industrial reactor: SUPER-PHENIX 1. These codes are the following ones: - ARGOH which calculates the behavior of an isolated subassembly. - HARMONIE which calculates the core mechanical equilibrium - TRACAR which yields a graphic visualization of HARMONIE results, and calculates the handling forces and support reactions - HARMOREA which calculates the reactivity variations between given equilibrium states (for instance: pads effect and diagrid effect); now at the end of its development. The calculations were performed on 1/3 of the SPX1 core. Their purpose is double: - on the one hand, to check design criteria, and provide the loadings for the subassembly mechanical design studies; on the other hand, to evaluate the reactivity effects, related to the horizontal core deformations, and useful for operation and safety studies. The results of these calculations showed that the design criteria were verified for the contractual lifetime of the subassemblies. (orig.)

Measurement of burnup in FBR MOX fuel irradiated to high burnup

International Nuclear Information System (INIS)

Koyama, Shin-ichi; Osaka, Masahiko; Sekine, Takashi; Morozumi, Katsufumi; Namekawa, Takashi; Itoh, Masahiko

2003-01-01

The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28% FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector. (author)

Development of whole core thermal-hydraulic analysis program ACT. 3. Coupling core module with primary heat transport system module

International Nuclear Information System (INIS)

Ohtaka, Masahiko; Ohshima, Hiroyuki

1998-10-01

A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including inter-wrapper flow under various reactor operation conditions. In this work, the core module as a main part of the ACT developed last year, which simulates thermal-hydraulics in the subassemblies and the inter-subassembly gaps, was coupled with an one dimensional plant system thermal-hydraulic analysis code LEDHER to simulate transients in the primary heat transport system and to give appropriate boundary conditions to the core model. The effective algorithm to couple these two calculation modules was developed, which required minimum modification of them. In order to couple these two calculation modules on the computing system, parallel computing technique using PVM (Parallel Virtual Machine) programming environment was applied. The code system was applied to analyze an out-of-pile sodium experiment simulating core with 7 subassemblies under transient condition for code verification. It was confirmed that the analytical results show a similar tendency of experimental results. (author)

Constitutive correlations for wire-wrapped subchannel analysis under forced and mixed convection conditions. Part 1

International Nuclear Information System (INIS)

Cheng, S.K.; Todreas, N.E.

1984-08-01

A simple subchannel analysis method based on the ENERGY series of codes, ENERGY-IV, has been established for predicting the temperature field in a single isolated wire-wrapped Liquid Metal Fast Breeder Reactor (LMFBR) subassembly under steady state forced and mixed convection conditions. The ENERGY-IV is a totally empirical code employed for fast running purposes and requires well calibrated lead length averaged input parameters to achieve satisfactory predictions. These input parameters were identified to be the inlet flow split parameters, the subchannel friction factors, the interchannel mixing parameters, the conduction shape factor, and the transverse velocity at the edge gap. Experiments were performed in a 37-pin wire-wrapped rod bundle with a geometry between that of a typical LMFBR fuel subassembly and blanket subassembly for filling the gap in the available data base for the input parameters. The isokinetic extraction method for measuring subchannel velocity, the pitot-static probe for measuring pressure drop, and the salt tracer injection method for estimating the interchannel mixing, were used in these experiments

Methods for batch fabrication of cold cathode vacuum switch tubes

Science.gov (United States)

Walker, Charles A [Albuquerque, NM; Trowbridge, Frank R [Albuquerque, NM

2011-05-10

Methods are disclosed for batch fabrication of vacuum switch tubes that reduce manufacturing costs and improve tube to tube uniformity. The disclosed methods comprise creating a stacked assembly of layers containing a plurality of adjacently spaced switch tube sub-assemblies aligned and registered through common layers. The layers include trigger electrode layer, cathode layer including a metallic support/contact with graphite cathode inserts, trigger probe sub-assembly layer, ceramic (e.g. tube body) insulator layer, and metallic anode sub-assembly layer. Braze alloy layers are incorporated into the stacked assembly of layers, and can include active metal braze alloys or direct braze alloys, to eliminate costs associated with traditional metallization of the ceramic insulator layers. The entire stacked assembly is then heated to braze/join/bond the stack-up into a cohesive body, after which individual switch tubes are singulated by methods such as sawing. The inventive methods provide for simultaneously fabricating a plurality of devices as opposed to traditional methods that rely on skilled craftsman to essentially hand build individual devices.

Analysis of excess reactivity of JOYO MK-III performance test core

International Nuclear Information System (INIS)

Maeda, Shigetaka; Yokoyama, Kenji

2003-10-01

JOYO is currently being upgraded to the high performance irradiation bed JOYO MK-III core'. The MK-III core is divided into two fuel regions with different plutonium contents. To obtain a higher neutron flux, the active core height was reduced from 55 cm to 50 cm. The reflector subassemblies were replaced by shielding subassemblies in the outer two rows. Twenty of the MK-III outer core fuel subassemblies in the performance test core were partially burned in the transition core. Four irradiation test rigs, which do not contain any fuel material, were loaded in the center of the performance test core. In order to evaluate the excess reactivity of MK-III performance test core accurately, we evaluated it by applying not only the JOYO MK-II core management code system MAGI, but also the MK-III core management code system HESTIA, the JUPITER standard analysis method and the Monte Carlo method with JFS-3-J3.2R content set. The excess reactivity evaluations obtained by the JUPITER standard analysis method were corrected to results based on transport theory with zero mesh-size in space and angle. A bias factor based on the MK-II 35th core, which sensitivity was similar to MK-III performance test core's, was also applied, except in the case where an adjusted nuclear cross-section library was used. Exact three-dimensional, pin-by-pin geometry and continuous-energy cross sections were used in the Monte Carlo calculation. The estimated error components associated with cross-sections, methods correction factors and the bias factor were combined based on Takeda's theory. Those independently calculated values agree well and range from 2.8 to 3.4%Δk/kk'. The calculation result of the MK-III core management code system HESTLA was 3.13% Δk/kk'. The estimated errors for bias method range from 0.1 to 0.2%Δk/kk'. The error in the case using adjusted cross-section was 0.3%Δk/kk'. (author)

Acoustic emissions of a boiling liquid - an experimental survey in water and extrapolation to SFRs

International Nuclear Information System (INIS)

Vanderhaegen, M.; Paumel, K.; Tourin, A.

2013-06-01

The acoustic detection of sodium boiling is seen as a promising and innovative surveillance technique for sodium-cooled fast reactors (SFRs). It could be especially useful to detect in-core boiling that are the consequence of initiating accidents or whilst the mean subassembly temperature is very close to the nominal value. This latter is a consequence of the fuel assembly design of SFRs. Furthermore, it is a technique that has been proven to be successful in the past to follow the boiling behavior during SFR experiments that were aimed at simulating accidental conditions. However its effectiveness as in-core instrumentation still has to be demonstrated. In that aim, the acoustic emissions of sodium boiling in subassemblies are studied. Experimental studies are however limited to the boiling of common coolants due to the complications that arise when boiling liquid metals. As such, simple water experiments are performed. And although the results of these experiments are not completely representative for sodium boiling due to the incomplete thermo-hydraulic similarities between sodium and water, they can provide an interesting knowledge of the many influences that control the acoustic pressure field. In this article we'll specifically show how the condensation of vapor in subcooled liquid, the principal contribution to the acoustic emissions during boiling and hence the acoustic spectrum, is influenced by a pin-bundle geometry. We study this influence by comparing pool boiling experimental acoustic recordings with those of a simple pin-bundle geometry boiling experiment. The qualitative link, between this relatively simple pin-bundle experiment and the condensation phenomena that take place during sodium boiling inside SFR subassemblies, is used as a basis for this analysis. This simple experimental evidence, together with other theoretical arguments based on a thorough analysis of the sodium material properties, enables us to deduce that simple sodium

Structural evaluation of fast reactor core restraint with irradiation creep-swelling opposition effects

International Nuclear Information System (INIS)

Kalinowski, J.E.

1979-01-01

Irradiation creep and swelling correlations are derived from primary loading in-reactor experiments in which irradiation creep and swelling act in the same direction. When correlation uncertainty bands are applied in core restraint evaluations, significant variability in sub-assembly behavior is predicted. For example, sub-assemblies in the outer core region where neutron flux and duct temperature gradients are significant exhibit bowing responses ranging from a creep dominated outward bow to a swelling dominated inward bow. Furthermore, solutions based on upper bound and lower bound correlation uncertainty combinations are observed to cross-over indicating that such combinations are physically unrealistic in the assessment of creep-swelling opposition effects. In order to obtain realistic upper and lower bound sub-assembly responses, judgement must be applied in the selection of creep-swelling equation uncertainty combinations. Experimental programs have been defined which will provide the needed basic as well as prototypic creep-swelling opposition data for reference and advanced sub-assembly duct alloys. The first of these is an irradiation of cylindrical capsules subjected to a through-wall temperature gradient. This test which is presently underway in the EBR-II reactor will provide the data needed to refine irradiation creep and swelling correlations and their associated uncertainties when applied to core restraint evaluations. Restrained pin and duct bowing experiments in FFTF have also been defined. These will provide the prototypic data necessary to verify irradiated duct bowing methodology. The results of this experimental program are expected to reduce creep and swelling uncertainties and permit better definition of the design window for load plane gaps. (orig.)

Fast reactor safety and computational thermo-fluid dynamics approaches

International Nuclear Information System (INIS)

Ninokata, Hisashi; Shimizu, Takeshi

1993-01-01

This article provides a brief description of the safety principle on which liquid metal cooled fast breeder reactors (LMFBRs) is based and the roles of computations in the safety practices. A number of thermohydraulics models have been developed to date that successfully describe several of the important types of fluids and materials motion encountered in the analysis of postulated accidents in LMFBRs. Most of these models use a mixture of implicit and explicit numerical solution techniques in solving a set of conservation equations formulated in Eulerian coordinates, with special techniques included to specific situations. Typical computational thermo-fluid dynamics approaches are discussed in particular areas of analyses of the physical phenomena relevant to the fuel subassembly thermohydraulics design and that involve describing the motion of molten materials in the core over a large scale. (orig.)

Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

International Nuclear Information System (INIS)

Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

1979-11-01

A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

International Nuclear Information System (INIS)

Okuda, Eiji; Ito, Hiromichi; Yoshihara, Shizuya

2014-01-01

An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

Device and process for cleaning a steam generator plate by a sludge lance

International Nuclear Information System (INIS)

Michel, D.; Berard, P.; Denuit, J.

1995-01-01

The cleaning system comprises a carriage which is able to move along a rail and is equipped with a serie of oscillating nozzles. The carriage is made up of a first subassembly with a reversible lance module carrying the nozzles and is equipped at each end with a male hydraulic coupling which is able to connect either with a pressurized water circuit. Its second subassembly comprises the module which connects with and rotates the nozzles, a reversible observation and detection unit and an independent drive mechanism with at least two toothed wheels able to engage with racks. 11 figs

Improvements in remote equipment torquing and fastening

International Nuclear Information System (INIS)

Garin, J.

1978-01-01

Remote torquing and fastening is a requirement of generic interest for application in an environment not readily accessible to man. The developments over the last 30 years in torque-controlled equipment above 200 nm (150 ft/lb) have not been emphasized. The development of specialized subassemblies to torque and fasten equipment in a remotely controlled environment is an integral part of the Advanced Fuel Recycle Program at Oak Ridge National Laboratory. Commercially available subassemblies have been adapted into a system that would provide remote torquing and fastening in the range of 200 to 750 nm (150 to 550 ft/lb). 9 figures

NCEL: two dimensional finite element code for steady-state temperature distribution in seven rod-bundle

International Nuclear Information System (INIS)

Hrehor, M.

1979-01-01

The paper deals with an application of the finite element method to the heat transfer study in seven-pin models of LMFBR fuel subassembly. The developed code NCEL solves two-dimensional steady state heat conduction equation in the whole subassembly model cross-section and enebles to perform the analysis of thermal behaviour in both normal and accidental operational conditions as eccentricity of the central rod or full or partial (porous) blockage of some part of the cross-flow area. The heat removal is simulated by heat sinks in coolant under conditions of subchannels slug flow approximation

Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Design and performance of reference cores. Research project 620-25

International Nuclear Information System (INIS)

Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

1979-11-01

A parameter study was conducted to determine the interrelated effects of: loosely of tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

The miniature optical transmitter and transceiver for the High-Luminosity LHC (HL-LHC) experiments

International Nuclear Information System (INIS)

Liu, C; Zhao, X; Deng, B; Gong, D; Guo, D; Li, X; Liang, F; Liu, G; Liu, T; Xiang, A C; Ye, J; Chen, J; Huang, D; Hou, S; Teng, P-K

2013-01-01

We present the design and test results of the Miniature optical Transmitter (MTx) and Transceiver (MTRx) for the high luminosity LHC (HL-LHC) experiments. MTx and MTRx are Transmitter Optical Subassembly (TOSA) and Receiver Optical Subassembly (ROSA) based. There are two major developments: the Vertical Cavity Surface Emitting Laser (VCSEL) driver ASIC LOCld and the mechanical latch that provides the connection to fibers. In this paper, we concentrate on the justification of this work, the design of the latch and the test results of these two modules with a Commercial Off-The-Shelf (COTS) VCSEL driver

Assessment calculation of MARS-LMR using EBR-II SHRT-45R

Energy Technology Data Exchange (ETDEWEB)

Choi, C.; Ha, K.S.

2016-10-15

Highlights: • Neutronic and thermal-hydraulic behavior predicted by MARS-LMR is validated with EBR-II SHRT-45R test data. • Decay heat model of ANS-94 give better prediction of the fission power. • The core power is well predicted by reactivity feedback during initial transient, however, the predicted power after approximately 200 s is over-estimated. The study of the reactivity feedback model of the EBR-II is necessary for the better calculation of the power. • Heat transfer between inter-subassemblies is the most important parameter, especially, a low flow and power subassembly, like non-fueled subassembly. - Abstract: KAERI has designed a prototype Gen-IV SFR (PGSFR) with metallic fuel. And the safety analysis code for the PGSFR, MARS-LMR, is based on the MARS code, and supplemented with various liquid metal related features including sodium properties, heat transfer, pressure drop, and reactivity feedback models. In order to validate the newly developed MARS-LMR, KAERI has joined the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on “Benchmark Analysis of an EBR-II Shutdown Heat Removal Test (SHRT)”. Argonne National Laboratory (ANL) has technically supported and participated in this program. One of benchmark analysis tests is SHRT-45R, which is an unprotected loss of flow test in an EBR-II. So, sodium natural circulation and reactivity feedbacks are major phenomena of interest. A benchmark analysis was conducted using MARS-LMR with original input data provided by ANL. MARS-LMR well predicts the core flow and power change by reactivity feedbacks in the core. Except the results of the XX10, the temperature and flow in the XX09 agreed well with the experiments. Moreover, sensitivity tests were carried out for a decay heat model, reactivity feedback model, inter-subassembly heat transfer, internal heat structures and so on, to evaluate their sensitivity and get a better prediction. The decay heat model of ANS-94 shows

Assessment calculation of MARS-LMR using EBR-II SHRT-45R

International Nuclear Information System (INIS)

Choi, C.; Ha, K.S.

2016-01-01

Highlights: • Neutronic and thermal-hydraulic behavior predicted by MARS-LMR is validated with EBR-II SHRT-45R test data. • Decay heat model of ANS-94 give better prediction of the fission power. • The core power is well predicted by reactivity feedback during initial transient, however, the predicted power after approximately 200 s is over-estimated. The study of the reactivity feedback model of the EBR-II is necessary for the better calculation of the power. • Heat transfer between inter-subassemblies is the most important parameter, especially, a low flow and power subassembly, like non-fueled subassembly. - Abstract: KAERI has designed a prototype Gen-IV SFR (PGSFR) with metallic fuel. And the safety analysis code for the PGSFR, MARS-LMR, is based on the MARS code, and supplemented with various liquid metal related features including sodium properties, heat transfer, pressure drop, and reactivity feedback models. In order to validate the newly developed MARS-LMR, KAERI has joined the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on “Benchmark Analysis of an EBR-II Shutdown Heat Removal Test (SHRT)”. Argonne National Laboratory (ANL) has technically supported and participated in this program. One of benchmark analysis tests is SHRT-45R, which is an unprotected loss of flow test in an EBR-II. So, sodium natural circulation and reactivity feedbacks are major phenomena of interest. A benchmark analysis was conducted using MARS-LMR with original input data provided by ANL. MARS-LMR well predicts the core flow and power change by reactivity feedbacks in the core. Except the results of the XX10, the temperature and flow in the XX09 agreed well with the experiments. Moreover, sensitivity tests were carried out for a decay heat model, reactivity feedback model, inter-subassembly heat transfer, internal heat structures and so on, to evaluate their sensitivity and get a better prediction. The decay heat model of ANS-94 shows

Classical and ablative Richtmyer-Meshkov instability and other ICF-relevant plasma flows diagnosed with monochromatic x-ray imaging

International Nuclear Information System (INIS)

Aglitskiy, Y; Metzler, N; Karasik, M; Velikovich, A L; Zalesak, S T; Schmitt, A J; Serlin, V; Weaver, J; Obenschain, S P; Gardner, J H

2008-01-01

In inertial confinement fusion (ICF) and high-energy density physics (HEDP), the most important manifestations of the hydrodynamic instabilities and other mixing processes involve lateral motion of the accelerated plasmas. In order to understand the experimental observations and to advance the numerical simulation codes to the point of predictive capability, it is critically important to accurately diagnose the motion of the dense plasma mass. The most advanced diagnostic technique recently developed for this purpose is the monochromatic x-ray imaging that combines large field of view with high contrast, high spatial resolution and large throughput, ensuring high temporal resolution at large magnification. Its application made it possible for the experimentalists to observe for the first time important hydrodynamic effects that trigger compressible turbulent mixing in laser targets, such as ablative Richtmyer-Meshkov (RM) instability, feedout, interaction of an RM-unstable interface with shock and rarefaction waves. It also helped to substantially improve the accuracy of diagnosing many other important plasma flows, ranging from laser-produced jets to electromagnetically driven wires in a Z-pinch, and to test various methods suggested for mitigation of the Rayleigh-Taylor instability. We will review the results obtained with the aid of this technique in ICF-HEDP studies at the Naval Research Laboratory and the prospects of its future applications.

Shield requirement estimation for pin storage room in fuel fabrication plant

International Nuclear Information System (INIS)

Shanthi, M.M.; Keshavamurthy, R.S.; Sivashankaran, G.

2012-01-01

Fast Reactor Fuel Cycle Facility (FRFCF) is an upcoming project in Kalpakkam. It has the facility to recycle the fuel from PFBR. It is an integrated facility, consists of fuel reprocessing plant, fuel fabrication plant (FFP), core subassembly plant, Reprocessed Uranium plant (RUP) and waste management plant. The spent fuel from PFBR would be reprocessed in fuel reprocessing plant. The reprocessed fuel material would be sent to fuel fabrication plant. The main activity of fuel fabrication plant is the production of MOX fuel pins. The fuel fabrication plant has a fuel pin storage room. The shield requirement for the pin storage room has been estimated by Monte Carlo method. (author)

Fast reactor fuel failures and steam generator leaks: Transient and accident analysis approaches

International Nuclear Information System (INIS)

1996-10-01

This report consists of a survey of activities on transient and accident analysis for the LMFR. It is focused on the following subjects: Fuel transient tests and analyses in hypothetical incident/accident situations; sodium-water interaction in steam generators, and sodium fires: test and analyses. There are also sections dealing with the experimental and analytical studies of: fuel subassembly failures; sodium boiling, molten fuel-coolant interaction; molten material movement and relocation in fuel bundles; heat removal after an accident or incident; sodium-water reaction in steam generator; steam generator protection systems; sodium-water contact in steam generator building; fire-fighting methods and systems to deal with sodium fires. Refs, figs, tabs

Evaluation of integrally finned cladding for LMFBR fuel pins

International Nuclear Information System (INIS)

Cantley, D.A.; Sutherland, W.H.

1975-01-01

An integral fin design effectively reduces the coolant temperature gradients within an LMFBR subassembly by redistributing coolant flow so as to reduce the maximum cladding temperature and increase the duct wall temperature. The reduced cladding temperatures are offset by strain concentrations resulting from the fin geometry, so there is little net effect on predicted fuel pin performance. The increased duct wall temperatures, however, significantly reduce the duct design lifetime so that the final conclusion is that the integral fin design is inferior to the standard wire wrap design. This result, however, is dependent upon the material correlations used. Advanced alloys with improved irradiation properties could alter this conclusion

The FUBR-1B experiment and BEATRIX-I

International Nuclear Information System (INIS)

Hollenberg, G.W.; Knight, R.C.; Pember, L.A.; Johnson, C.E.; Poeppel, R.B.; Yang, L.

1987-01-01

The objective of this work is to irradiate lithium ceramics in a fast neutron environment to high burnup levels and with large temperature gradients. The first insertion of two subassemblies completed its irradiation in December 1986. This irradiation exposed Li 2 O and LiAlO and LiA102 to not only high temperatures but also large temperature gradients which are expected in fusion blankets. In addition, it included other materials such as Li 2 ZrO 3 , Li 8 ZrO 6 , Li 4 SiO 4 , and LiAlO 2 (spheres and large grain size) some of which will go to high burnups

Status of fast reactor technology in China

International Nuclear Information System (INIS)

Xu Mi

1992-01-01

The paper has introduced briefly the recent news about the Chinese nuclear programme on PWR and FBR. Concerning the FFR design, some issues under consideration have been presented, including the matches between thermo-parameters of primary sodium and of steam, the arrangement of control and safety rods which correspond to first and second shut-down systems, the structure of inner vessel and the axial length of subassembly. With regard to the R and D of FBR technology, some results on sodium technology and on the cladding materials have been given in the paper. Finally, some progress and troubles on site selection for this reactor have also been outlined. (author)

Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

Energy Technology Data Exchange (ETDEWEB)

Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

2016-05-15

Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

Flow visualization on a natural circulation inter-wrapper flow. Experimental and numerical results under a geometric condition of button type spacer pads

Energy Technology Data Exchange (ETDEWEB)

Yasuda, A.; Miyakoshi, H.; Hayashi, K.; Nishimura, M.; Kamide, H.; Hishida, K. [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

1999-04-01

Investigations on the inter-wrapper flow (IWF) in a liquid metal cooled fast breeder reactor core have been carried out. The IWF is a natural circulation flow between wrapper tubes in the core barrel where cold fluid is coming from a direct heat exchanger (DHX) in the upper plenum. It was shown by the sodium experiment using 7-subassembly core model that the IWF can cool the subassemblies. To clarify thermal-hydraulic characteristics of the IWF in the core, the water experiment was performed using the flow visualization technique. The test rig for IWF (TRIF) has the core simulating the fuel subassemblies and radial reflectors. The subassemblies are constructed featuring transparent heater to enable both Joule heating and flow visualization. The transparent heater was made of glass with thin conductor film coating of tin oxide, and the glass heater was embedded on the wall of modeled wrapper tube made of acrylic plexiglass. In the present experiment, influences of peripheral geometric parameters such as flow holes of core formers on the thermal-hydraulic field were investigated with the button type spacer pads of the wrapper tube. Through the water tests, flow patterns of the IWF were revealed and velocity fields were quantitatively measured with a particle image velocimetry (PIV). Also, no substantial influence of peripheral geometry was found on the temperature field of the IWF, as far as the button type spacer pad was applied. Numerical simulation was applied to the experimental analysis of IWF by using multidimensional code with porous body model. The numerical results reproduced the flow patterns within TRIF and agreed well to experimental temperature distributions, showing capability of predicting IWF with porous body model. (author)

JOYO MK-II core characteristics database. Update to JFS-3-J3.2R

International Nuclear Information System (INIS)

Ohkawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi

2003-04-01

The 'JOYO' MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition in 2001. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. However, after the database was published, it was recently found that there were errors in the process of making the group constant set JFS-3-J3.2, and it was revised at JFS-3-J3.2R. Then, the group constant set was updated at JFS-3-J3.2R in this database. The MK-II core management data nad core characteristics data were recorded on CD-ROM for user convenience. The structure of the database is the same as in the first edition. The 'Configuration Data' include the core arrangement and refueling record for each operational cycle. The 'Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The 'Output Data' contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The 'Core Characteristics Data' include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. The effect of updating the group constant set, the calculation results of excess reactivity decreased by about 0.15Δk/kk', and the effects to other core characteristics were negligible. (author)

Flow induced vibration studies for LMFBR in Japan: Past and recent studies of FIV for JOYO and MONJU

Energy Technology Data Exchange (ETDEWEB)

Sato, K [Sodium Engineering Division, O-arai Engineering Centre, Power Reactor and Nuclear Fuel, Development Corporation, Narita-cho, O-arai Machi, Ibaraki-ken (Japan)

1977-12-01

This paper presents the past and recent studies of flow induced vibration of the reactor components for the experimental fast breeder reactor JOYO and the prototype fast breeder reactor MONJU, in which many suggestive results for the higher flow velocity systems in a future reactor are contained. The fuel subassembly is the most important from the view point of the vibration. Thus, the studies were carried out with a mock-up subassembly for JOYO. In this experiment, statistical analysis results of the vibration characteristics of single core subassembly and the effects of external forced vibration, flow disturbance and fuel pin bundle vibration were reported. The further more detailed investigations are now being performed for MONJU. In addition to the above studies, the vibration failure of a sodium valve is reported. The valve is a 8-inch stop valve in SODIUM FLOW AND HEAT TRANSFER TEST LOOP at O-arai Engineering Center. The failure occurred in 1969 during the performance test of the mechanical pump, and this resulted in a small sodium leak. The cause of the failure was found to be the vibration fatigue of the metal bellows. (author)

Variation in computer time with geometry prescription in monte carlo code KENO-IV

International Nuclear Information System (INIS)

Gopalakrishnan, C.R.

1988-01-01

In most studies, the Monte Carlo criticality code KENO-IV has been compared with other Monte Carlo codes, but evaluation of its performance with different box descriptions has not been done so far. In Monte Carlo computations, any fractional savings of computing time is highly desirable. Variation in computation time with box description in KENO for two different fast reactor fuel subassemblies of FBTR and PFBR is studied. The K eff of an infinite array of fuel subassemblies is calculated by modelling the subassemblies in two different ways (i) multi-region, (ii) multi-box. In addition to these two cases, excess reactivity calculations of FBTR are also performed in two ways to study this effect in a complex geometry. It is observed that the K eff values calculated by multi-region and multi-box models agree very well. However the increase in computation time from the multi-box to the multi-region is considerable, while the difference in computer storage requirements for the two models is negligible. This variation in computing time arises from the way the neutron is tracked in the two cases. (author)

Experience with the instrumentation tests in large sodium test facilities

International Nuclear Information System (INIS)

Lauhoff, Th.; Ruppert, E.; Stehle, H.; Vinzens, K.

1976-01-01

A facility is described for fast breeder core components (AKB) to test specially instrumented fuel dummies and blanket elements, and also absorber elements under simulated normal and extreme reactor conditions. In addition to endurance testing of a special sodium and high temperature sub-assembly, instrumentation is provided to investigate thermohydraulic and vibrational behaviour of core elements. During tests of > 3000 h at temperatures above 820 K the main sub-assembly characteristics, e.g. pressure drop, leakage flow, vibration and noise spectra can be reproduced. The use of eddy current flow meters, strain gauges, magnetostrictive noise sensors, pressure transducers, thermocouples, and acoustic surveillance devices, are described. (U.K.)

Rod drop in the LR-0 reactor core comprising 55 fuel assemblies

International Nuclear Information System (INIS)

Hadek, J.; Grundmann, U.

1989-09-01

Data from the third stage of kinetic measurements on the LR-0 reactor, performed in 1988, were employed for additional calculations using the 3-dimensional neutron kinetics code HEXDYN3D. The reactor consists of subassemblies similar to those in the WWER-1000 (PWR) reactor. The theoretical and experimental results are compared for the time behavior of the neutron flux caused by drop of the control rod cluster in various subassemblies of the reactor. The results demonstrate that the HEXDYN3D code is well suited to the treatment of the space-time behavior of the neutron flux. (author). 21 figs., 2 tabs., 16 refs

Gamma heated subassembly for sodium boiling experiments

International Nuclear Information System (INIS)

Artus, S.C.

1975-01-01

The design of a system to boil sodium in an LMFBR is examined. This design should be regarded as a first step in a series of boiling experiments. The reactor chosen for the design of the boiling apparatus is the Experimental Breeder Reactor-II (EBR-II), located at the National Reactor Testing Station in Idaho. Criteria broadly classified as design objectives and design requirements are discussed

Gamma heated subassembly for sodium boiling experiments

Energy Technology Data Exchange (ETDEWEB)

Artus, S.C.

1975-01-01

The design of a system to boil sodium in an LMFBR is examined. This design should be regarded as a first step in a series of boiling experiments. The reactor chosen for the design of the boiling apparatus is the Experimental Breeder Reactor-II (EBR-II), located at the National Reactor Testing Station in Idaho. Criteria broadly classified as design objectives and design requirements are discussed.

High-Temperature Corrosion Study for the RPP Low Activity Waste Melter

International Nuclear Information System (INIS)

Marshall, K.M.

2003-01-01

The River Protection Program (RPP) low activity waste (LAW) melter design incorporates a series of bubblers used to increase convection in the molten glass. Through runs of a pilot melter at Duratek, Inc. in Columbia, Maryland, the bubblers have been identified as the major component limiting LAW melter availability, requiring frequent replacement due to corrosive degradation, primarily at the melt line. Laboratory experiments were performed to evaluate the performance of several alloys and coatings in simulated RPP low activity waste melter vapor space and molten glass environments. The performance of the alloys and coatings was studied in order to advance our understanding of how these materials react at the melt/air interface inside the melter. The ultimate goal was to identify a material with superior performance compared to that of Inconel 693, and to deliver a bubbler sub-assembly made of that material to the RPP LAW melter pilot facility for further testing

Evolution of Flat Roofs

Directory of Open Access Journals (Sweden)

Şt. Vasiliu

2009-01-01

Full Text Available Roofs are constructive subassembles that are located at the top of buildings, which toghether with perimetral walls and some elements of the infrastructure belongs to the subsystem elements that close the building. Roofs must meet resistance requirements to mechanical action, thermal insulating, waterproofing and acoustic, fire resistance, durability, economy and aesthetics. The man saw the need to build roofs from the oldest ancient times. Even if the design of buildings has an empirical character, are known and are preserved until today constructions that are made in antiquity, by the Egyptians, Greeks and Romans with architectural achievements, worthy of admiration and in present time. General composition of civil construction has been influenced throughout the evolution of construction history by the level of production forces and properties of building materials available in every historical epoch. For over five millennia, building materials were stone, wood and ceramic products (concrete was used by theRomans only as filling material.

Corrosion Performance of Friction Stir Linear Lap Welded AM60B Joints

Science.gov (United States)

Kish, J. R.; Birbilis, N.; McNally, E. M.; Glover, C. F.; Zhang, X.; McDermid, J. R.; Williams, G.

2017-11-01

A corrosion investigation of friction stir linear lap welded AM60B joints used to fabricate an Mg alloy-intensive automotive front end sub-assembly was performed. The stir zone exhibited a slightly refined grain size and significant break-up and re-distribution of the divorced Mg17Al12 (β-phase) relative to the base material. Exposures in NaCl (aq) environments revealed that the stir zone was more susceptible to localized corrosion than the base material. Scanning vibrating electrode technique measurements revealed differential galvanic activity across the joint. Anodic activity was confined to the stir zone surface and involved initiation and lateral propagation of localized filaments. Cathodic activity was initially confined to the base material surface, but was rapidly modified to include the cathodically-activated corrosion products in the filament wake. Site-specific surface analyses revealed that the corrosion observed across the welded joint was likely linked to variations in Al distribution across the surface film/metal interface.

Extended MHD Effects in High Energy Density Experiments

Science.gov (United States)

Seyler, Charles

2016-10-01

The MHD model is the workhorse for computational modeling of HEDP experiments. Plasma models are inheritably limited in scope, but MHD is expected to be a very good model for studying plasmas at the high densities attained in HEDP experiments. There are, however, important ways in which MHD fails to adequately describe the results, most notably due to the omission of the Hall term in the Ohm's law (a form of extended MHD or XMHD). This talk will discuss these failings by directly comparing simulations of MHD and XMHD for particularly relevant cases. The methodology is to simulate HEDP experiments using a Hall-MHD (HMHD) code based on a highly accurate and robust Discontinuous Galerkin method, and by comparison of HMHD to MHD draw conclusions about the impact of the Hall term. We focus on simulating two experimental pulsed power machines under various scenarios. We examine the MagLIF experiment on the Z-machine at Sandia National Laboratories and liner experiments on the COBRA machine at Cornell. For the MagLIF experiment we find that power flow in the feed leads to low density plasma ablation into the region surrounding the liner. The inflow of this plasma compresses axial magnetic flux onto the liner. In MHD this axial flux tends to resistively decay, whereas in HMHD a force-free current layer sustains the axial flux on the liner leading to a larger ratio of axial to azimuthal flux. During the liner compression the magneto-Rayleigh-Taylor instability leads to helical perturbations due to minimization of field line bending. Simulations of a cylindrical liner using the COBRA machine parameters can under certain conditions exhibit amplification of an axial field due to a force-free low-density current layer separated by some distance from the liner. This results in a configuration in which there is predominately axial field on the liner inside the current layer and azimuthal field outside the layer. We are currently attempting to experimentally verify the simulation

Biodistribution and dosimetric evaluation of 186Re hydroxy ethylen diphosphate

International Nuclear Information System (INIS)

Noto, M.G.; Manzini, Alberto

1987-01-01

The pharmacokinetics and the dose of radiation absorbed in different body tissues by the administration of 186 Re HEDP (hydroxyethylendiphosphate). The radiation dose in the standard man was established between 5,5 and 25 rad/mCi for red marrow, and red marrow and bone respectively; the radiation dose in metastases would be of 125 rad/mCi. It is concluded that this radiopharmaceutical is suitable for palliative treatment for pain with the mentioned patollogy. (M.E.L.) [es

Report of the Interagency Task Force on High Energy Density Physics

Energy Technology Data Exchange (ETDEWEB)

None

2007-08-01

Identifies the needs for improving Federal stewardship of specific aspects of high energy density physics, particularly the study of high energy density plasmas in the laboratory, and strengthening university activities in this latter discipline. The report articulates how HEDP fits into the portfolio of federally funded missions and includes agency actions to be taken that are necessary to further this area of study consistent with Federal priorities and plans, while being responsive to the needs of the scientific community.

Report of the Interagency Task Force on High Energy Density Physics

International Nuclear Information System (INIS)

2007-01-01

Identifies the needs for improving Federal stewardship of specific aspects of high energy density physics, particularly the study of high energy density plasmas in the laboratory, and strengthening university activities in this latter discipline. The report articulates how HEDP fits into the portfolio of federally funded missions and includes agency actions to be taken that are necessary to further this area of study consistent with Federal priorities and plans, while being responsive to the needs of the scientific community

Past, present and future of safeguards implementation for the on-load RMBK-1500 reactors in Ignalina

International Nuclear Information System (INIS)

Zendel, M.; Yim, S.; Monticone, C.; Kurselis, S.

1999-01-01

The on-load refueled RBMKs ('Reactor Bolshoy Moschnosti Kanalniy - Large Power Channel Type Reactor') are very different from all other power reactors which the Agency has been safeguarding over the past decades. Distinct differences in fuel properties and handling necessitated the formulation of separate, facility specific approaches. The spent fuel management at the RBMKs in Ignalina uses hot cells to cut each spent fuel assembly into two subassemblies. A large number of subassemblies are subsequently stored in large capacity, compact storage baskets at the spent fuel storage ponds adjacent to the reactor hall. The development of the safeguards approach is presented considering limitation in core access, technological feasibility, operation mode and financial as well as human resources of the Agency. The safeguards approach is based on a quarterly inspection scheme using Containment and Surveillance (C/S) measures, verification of fresh and spent fuel by Non Destructive Assay (NDA), establishing of flow balances to complement the material accountancy and the application of neutron/gamma monitors in a continuous, unattended mode. The implementation of these safeguards measures is discussed and actual inspection experience with an emphasis on the application of the neutron/gamma monitors is given. The neutron/gamma monitors serve multiple safeguards functions, such as monitoring shipments of waste from cutting operations for irradiated fuel in the hot cells, confirming the unloading history for the on-load reactors, complementing C/S by detecting movements of irradiated fuel materials in the reactor halls and verifying the operational status and the power output of the reactors. Actual measurement results are presented to demonstrate their effectiveness. Power Considerations are given for future safeguards implementation matters at Ignalina Nuclear plant (INPP) including measures for the Strengthened Safeguards System (SSS). (author)

Status of radiation shield design for liquid metal fast breeder reactor spent fuel shipping cask application

International Nuclear Information System (INIS)

Dupree, S.A.; Rack, H.J.

1976-09-01

Neutron and gamma-ray transport calculations in one-dimensional cylindrical geometry have been performed on a trial reference LMFBR spent-fuel shipping cask that could transport one CRBR subassembly. In the study it was assumed that a layer of depleted U and a layer of neutron shielding materials were sandwiched between 5.08-cm-thick (2-in.) layers of stainless steel. The thicknesses of the internal layers were adjusted until a balanced dose rate (50 percent neuton and 50 percent gamma-ray) of 5 mrem/hr was achieved at a point 1.83 m (6 ft) from the cask surface. Neutron-shield materials considered were LiH, Be, B 4 C, DiH 2 . 5 , and C (graphite). Of these materials, LiH provided the smallest, lightest, and least expensive cask; however, its use would be contigent on expansion of production facilities for LiH and development of a canning or cladding procedure. The B 4 C shielded cask would offer the best alternative if the designs were limited to those using currently available materials

Analytical evaluation of local fault in sodium cooled small fast reactor (4S). Preliminary evaluation of partial blockage in coolant channel

International Nuclear Information System (INIS)

Nishimura, Satoshi; Ueda, Nobuyuki

2007-01-01

Local faults are fuel failures that result from heat removal imbalance within a single subassembly especially in FBRs. Although the occurrence frequency of local faults is quite low, the licensing body required local faults evaluations in previous FBR plants to confirm the potential for the occurrence of severe fuel subassembly failure and its propagation. A conceptual design of 4S (Super-Safe, Small and Simple) is a sodium cooled fast reactor, which aims at an application to dispersed energy source and long core lifetime. It has a dense arrangement of fuel pins to achieve a long lifetime. Therefore, from the viewpoint of thermal hydraulics, the 4S reactor is considered to have more potential for coolant boiling and fuel pin failure caused by formation of local blockage, comparing these potential in the conventional FBRs. The objective of the present study is to evaluate the effect of local blockage on the coolant flow pattern and temperature rise in the 4S-type fuel subassembly under the normal operation condition. A series of three-dimensional thermal-hydraulic analysis in a single subassembly with local blockage was conducted by the commercialized CFD code 'PHOENICS'. Analytical results show that the peak coolant temperature behind the blockage rises with increasing the blockage area, however, the coolant boiling does not occur under the present analytical conditions. On the other hand, it is found that the liquid phase formation caused by eutectic reactions will occur between the metallic fuel and the cladding under the local blockage condition. However, the penetration rate of liquid phase at fuel-cladding interface is quit low. Therefore, it is expected that rapid fuel pin failure and its propagation to surrounding pins due to liquid phase formation will not occur. (author)

Response of hexagonal fuel assembly coupled with internal hydrodynamics

International Nuclear Information System (INIS)

Marchertas, A.H.; Julke, R.T.

1975-01-01

For safety considerations of sodium cooled fast breeder reactors the mechanistic accident-initiating conditions must be studied. In previous investigations of such initiating accidents the models assumed axisymmetric configurations and in general neglected the coupling effects with the subassembly boundary. This paper presents a more precise treatment of the subassembly boundary and also provides feedback of the boundary response to the pressure source. This is accomplished by marking use of two computer codes: REXCO-HT and SADCAT. The internal hydrodynamics of the fuel subassembly is simulated by the REXCO-HT code which possesses certain models of fuel-coolant interactions (MFCI) to be used as a pressure source. The hexagonal boundary of the fuel subassembly is modeled by the SADCAT code. Since both codes involve explicit time integration, coupling between the two is effected at each time step. The pressure at the outside boundary of the REXCO-HT model provides the loading on the SADCAT model. Given the load, the SADCAT model yields the three-dimensional deformation of the hexagonal boundary. With the deformation known, the outside REXCO-HT model boundary is adjusted and the computation cycle of the coupling is completed. In effect, the coupling of the two codes substitutes a cylindrical vessel of the REXCO-HT code by a hexagonal duct. It is shown by the use of this procedure that the assumption of a cylindrical vessel of the same thickness as that of the hexcan is quite erroneous. The maximum deformation of the flat of the hexcan in the illustrative examples is larger by as much as one order of magnitude. The maximum strains at the inside CORNER of the hexcan are also underestimated by a similar amount

A conceptual redesign of an inter-building fuel transfer cask

International Nuclear Information System (INIS)

Klann, R.T.; Picker, B.A. Jr.

1993-01-01

The Inter-Building Fuel Transfer Cask, referred to as the IBC, is a lead shielded cask for transporting subassemblies between buildings on the Argonne National Laboratory-West site near Idaho Falls, Idaho. The cask transports both newly fabricated and spent reactor subassemblies between the Experimental Breeder Reactor-2 (EBR-2), the Fuel Cycle Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The IBC will play a key role in the Integral Fast Reactor (IFR) fuel recycling demonstration project. The existing IBC technology, designed and fabricated in the late fifties, is outdated and is a source of personnel exposure at ANL-W. The current IBC system requires forced argon cooling and has extremely limited passive cooling capabilities due to existing design features. A conceptual redesign of the IBC has been performed. The objective of the conceptual design was to increase the passive heat removal capabilities, reduce the personnel radiation exposure and incorporate enhanced safety features into the design. The heat transfer, radiation and thermal-hydraulic properties of the IBC were analytically modeled to determine the principal factors controlling the design. The scoping studies that were performed determined the vital physical characteristics (i.e., size, shielding, pumps, etc.) of the IBC conceptual design. The conceptual design for the IBC allows subassemblies with up to 800 Watts of decay heat to be passively cooled, a significant increase over the existing system. The new design which incorporates better passive cooling mechanisms will prevent inadvertent damage to the subassembly during postulated loss-of-power and loss-of-flow accident scenarios. The new design also decreases the radiation hazard to personnel by having fewer external systems, a better shield plug design, and surfaces that are easier to decontaminate. The control and monitoring system will also be state-of-the-art technology

Multi-Axial Simulation Table (MAST)

Data.gov (United States)

Federal Laboratory Consortium — The MAST delivers an extensive array of testing applications providing rapid, flexible and reliable analysis for ground vehicle components and subassemblies. Using...

Feasibility study of the design of homogeneously mixed thorium-uranium oxide and all-uranium fueled reactor cores for civil nuclear marine propulsion - 15082

International Nuclear Information System (INIS)

Alam, S.B.; Lindley, B.A.; Parks, G.T.

2015-01-01

In this reactor physics study, we attempt to design a civil marine reactor core that can operate over a 10 effective-full-power-years life at 333 MWth using ThUO 2 and all-UO 2 fuel. We use WIMS to develop subassembly designs and PANTHER to examine whole-core arrangements, optimizing: subassembly and core geometry; fuel enrichment; burnable and moveable poison design; and whole-core loading patterns. We compare designs with a 14% fissile loading for ThUO 2 and all-UO 2 fuel in 13*13 assemblies with ZrB 2 integral fuel burnable absorber pins for reactivity control. Taking advantage of self-shielding effects, the ThUO 2 option shows greater promise in the final burnable poison design while maintaining low, stable reactivity with minimal burnup penalty. For the final poisoning design with ZrB 2 , ThUO 2 contributes 2.5% more initial reactivity suppression, although the all-UO 2 design exhibits lower reactivity swing. All the candidate materials show greater rod worth for the ThUO 2 design. For both fuels, B 4 C has the highest reactivity worth, providing 10% higher control rod worth for ThUO 2 fuel than all-UO 2 . Finally, optimized assemblies were loaded into a 3D reactor model in PANTHER. The PANTHER results show that after 10 years, the core is on the border of criticality, confirming the fissile loading is well-designed. (authors)

Review of fast reactor activities

Energy Technology Data Exchange (ETDEWEB)

Balz, W [Commission of the European Communities, Brussels (Belgium)

1978-07-01

The Commission of the European Communities continued its activities on the following lines: activities aimed at preparing for commercialization of fast breeder reactors which are essentially performed in the frame of Fast Reactor Coordinating Committee (FRCC); the execution of its own research program in the Joint Research Center. The report covers activities of the FRCC, of the Safety Working Group (SWG), the Whole Core Accident Code (WAC) subgroup, Containment (CONT) subgroup, Codes and Standards Working Group (CSWG). Research and development activities are concerned with LMFBR safety, subassembly thermal hydraulics, fuel-coolant interactions, post-accident heat removal, dynamic load response, safety related material properties, utilization limits of fast breeder fuels, plutonium and actinide aspects of nuclear fuel cycle.

Fast-mixed spectrum reactor interim report initial feasibility study

International Nuclear Information System (INIS)

Fischer, G.J.; Cerbone, R.J.

1979-01-01

The report summarizes the results of an initial four-month feasibility study of the Fast-Mixed Spectrum Reactor (FMSR). Reactor physics, fuel cycle, and thermal-hydraulic analyses were performed on a reference design. These results when coupled to a fuel and materials evaluation performed in cooperation with the Argonne National Laboratory indicate that the FMSR is feasible provided the fuels, cladding, and subassembly ducts can survive a peak fuel burnup of 15 to 20 atom percent heavy metal and peak fluences of 8 x 10 23 (nvt > 0.1 MeV). The results of this short study have also provided a basis for exploring alternative designs requiring significantly lower peak burnup and fluences for their operation

Review of fast reactor activities

International Nuclear Information System (INIS)

Balz, W.

1978-01-01

The Commission of the European Communities continued its activities on the following lines: activities aimed at preparing for commercialization of fast breeder reactors which are essentially performed in the frame of Fast Reactor Coordinating Committee (FRCC); the execution of its own research program in the Joint Research Center. The report covers activities of the FRCC, of the Safety Working Group (SWG), the Whole Core Accident Code (WAC) subgroup, Containment (CONT) subgroup, Codes and Standards Working Group (CSWG). Research and development activities are concerned with LMFBR safety, subassembly thermal hydraulics, fuel-coolant interactions, post-accident heat removal, dynamic load response, safety related material properties, utilization limits of fast breeder fuels, plutonium and actinide aspects of nuclear fuel cycle

Whole-core damage analysis of EBR-II driver fuel elements following SHRT program

International Nuclear Information System (INIS)

Chang, L.K.; Koenig, J.F.; Porter, D.L.

1987-01-01

In the Shutdown Heat Removal Testing (SHRT) program in EBR-II, fuel element cladding temperatures of some driver subassemblies were predicted to exceed temperatures at which cladding breach may occur. A whole-core thermal analysis of driver subassemblies was performed to determine the cladding temperatures of fuel elemnts, and these temperatures were used for fuel element damage calculation. The accumulated cladding damage of fuel element was found to be very small and fuel element failure resulting from SHRT transients is unlikely. No element breach was noted during the SHRT transients. The reactor was immediately restarted after the most severe SHRT transient had been completed and no driver fuel breach has been noted to date. (orig.)

Model for determining stresses in the structure of a fast reactor fuel assembly

International Nuclear Information System (INIS)

Kervevan, J.-J.

1974-01-01

Deformations in a reactor core are due to two metallurgical phenomena, swelling of the steel under irradiation and irradiation creep when the structure is under stress. The first step is to determine the deformation of each sub-assembly supposed free, subjected to a neutron flux or temperature gradient, and the second is to study the interactions amongst most of the sub-assemblies. Under the influence of the deformations the interaction value will change with time, and this development must be determined. Calculation methods were developed for the purpose. A number of computing codes already exist and it is necessary to complete them, modify them if necessary, create new ones as the case arises and form a coherent whole [fr

Mid-frequency Band Dynamics of Large Space Structures

Science.gov (United States)

Coppolino, Robert N.; Adams, Douglas S.

2004-01-01

High and low intensity dynamic environments experienced by a spacecraft during launch and on-orbit operations, respectively, induce structural loads and motions, which are difficult to reliably predict. Structural dynamics in low- and mid-frequency bands are sensitive to component interface uncertainty and non-linearity as evidenced in laboratory testing and flight operations. Analytical tools for prediction of linear system response are not necessarily adequate for reliable prediction of mid-frequency band dynamics and analysis of measured laboratory and flight data. A new MATLAB toolbox, designed to address the key challenges of mid-frequency band dynamics, is introduced in this paper. Finite-element models of major subassemblies are defined following rational frequency-wavelength guidelines. For computational efficiency, these subassemblies are described as linear, component mode models. The complete structural system model is composed of component mode subassemblies and linear or non-linear joint descriptions. Computation and display of structural dynamic responses are accomplished employing well-established, stable numerical methods, modern signal processing procedures and descriptive graphical tools. Parametric sensitivity and Monte-Carlo based system identification tools are used to reconcile models with experimental data and investigate the effects of uncertainties. Models and dynamic responses are exported for employment in applications, such as detailed structural integrity and mechanical-optical-control performance analyses.

Experimental study of core thermohydraulics in fast reactors during transition from forced to natural circulation. Influence of inter-wrapper flow

International Nuclear Information System (INIS)

Kamide, H.; Hayashi, K.; Momoi, K.

1997-01-01

The evaluation of core thermohydraulics under natural circulation conditions is important to utilize inherent safety features of Fast Reactors. When heat exchangers of a decay heat removal system are operated in an upper plenum of reactor vessel, cold sodium is provided by the heat exchangers. Core-plenum interactions will occur during a natural circulation condition due to this cold sodium in the upper plenum, e.g., it can penetrate into gap regions between fuel subassemblies (inter-wrapper flow, IWF) and the flow may reverse in low power core channels. These interactions will significantly modify the flow and temperature distributions in the core. Sodium experiments were carried out to study these phenomena. In a test section, seven subassemblies are housed and connected to an upper plenum. The influences of core-plenum interactions on the core thermohydraulics were investigated under steady state conditions and also in transitions from forced to natural circulation. Cooling effects of IWF on the fuel subassemblies were found in spite of natural circulation flow reduction in the primary loop due to temperature decreases in the upper non-heated section in the core. The inter-wrapper flow can effectively cool the core under extreme conditions of low flow rates through the core. (author)

Accuracy of fuel motion measurements using in-core detectors

International Nuclear Information System (INIS)

Dupree, S.A.

1975-01-01

An initial assessment has been made as to how accurately fuel motion can be measured with in-core detectors. A portion of this assessment has involved the calculation of the response of various detectors to fuel motion and the development of a formalism for correlating uncertainties in a neutron flux measurement to uncertainties in the fuel motion. Initially, four idealized configurations were studied in one dimension. These configurations consisted of (1) a single fuel-pin test using ACPR, (2) a seven fuel-pin test using ACPR, (3) a full subassembly (271 pin) test using a Class I ANL-type SAREF, and (4) a full subassembly plus six partial subassemblies (approximately 1000 pin) test using a Class III GE-type SAREF. It was assumed that melt would occur symmetrically at the center of the test fuel and that fuel would therefore disappear from the center of the geometry. For each case of series of calculations was performed in which detector responses were determined at several radial locations for the unperturbed core and for the core with various fractions of the fuel replaced with Na. This fuel loss was assumed to occur essentially instantaneously such that the power level in the remaining portion of the test fuel remained unchanged from that of the initial unperturbed condition

Coupled MCNP - SAS-SFR calculations for sodium fast reactor core at steady-state - 15460

International Nuclear Information System (INIS)

Ponomarev, A.; Travleev, A.; Pfrang, W.; Sanchez, V.

2015-01-01

The prediction of core parameters at steady state is the first step when studying core accident transient behaviour. At this step thermal hydraulics (TH) and core geometry parameters are calculated corresponding to initial operating conditions. In this study we present the coupling of the SAS-SFR code to the Monte-Carlo neutron transport code MCNP at steady state together with application to the European Sodium Fast Reactor (ESFR). The SAS-SFR code employs a multi-channel core representation where each channel represents subassemblies with similar power, thermal-hydraulics and pin mechanics conditions. For every axial node of every channel the individual geometry and material compositions parameters are calculated in accord with power and cooling conditions. This requires supplying the SAS-SFR-code with nodal power values which should be calculated by neutron physics code with given realistic core parameters. In the conventional approach the neutron physics model employs some core averaged TH and geometry data (fuel temperature, coolant density, core axial and radial expansion). In this study we organize a new approach coupling the MCNP neutron physics models and the SAS-SFR models, so that calculations of power can be improved by using distributed core parameters (TH and geometry) taken from SAS-SFR. The MCNP code is capable to describe cores with distributed TH parameters and even to model non-uniform axial expansion of fuel subassemblies. In this way, core TH and geometrical data calculated by SAS-SFR are taken into account accurately in the neutronics model. The coupling implementation is done by data exchange between two codes with help of processing routines managed by driver routine. Currently it is model-specific and realized for the ESFR 'Reference Oxide' core. The Beginning-Of-Life core state is considered with 10 channel representation for fuel subassemblies. For this model several sets of coupled calculations are performed, in which different

GFR demonstrator ALLEGRO design status

International Nuclear Information System (INIS)

Poette, C.; Malo, J.Y.; Brun-Magaud, V.; Morin, F.; Dor, I.; Mathieu, B.; Duhamel, H.; Stainsby, R.; Mikityuk, K.

2009-01-01

The ALLEGRO project has the ambitious goal of building and operating the first Gas Cooled Fast Reactor (GFR). It will be a low power experimental reactor with the main objective to validate on a pilot scale the specific GFR technologies (fuel element and sub-assembly, safety systems). It is a loop type, non electricity generating reactor. Its power is about 80 MW. The approach for the core includes first MOX cores loaded with some ceramic mixed carbide or nitride sub-assemblies with SiC/SiCf cladding and wrappers. When such unit test will be considered convincing enough, the diagrid and circuits are designed to accept full high temperature ceramic cores. The core neutrons can also be used to irradiate structural materials with fast neutron spectrum and in a large temperature range. The core can also include innovative irradiation fuel devices (samples or full bundles) for other reactor systems. Finally, the primary circuit can be connected to a test loop to validate the reactor coupled operation of a high temperature process or component. The paper deals with the current ALLEGRO design studies on a mid term roadmap aiming at ending the viability phase in 2012 in order to make a decision in 2013 for further detailed design and construction. Since 2005, the ALLEGRO design studies are shared in the GCFR 6th Framework Program which gathers 10 partners from 6 European countries. The paper will give an overview of recent progresses in various areas such as: - Last 3D core physics analysis of the MOX cores and their irradiation performances in terms of fast flux, dose/burnup, irradiation locations. - The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the MOX core. - Fuel handling principles and solutions. - System design and global reactor architecture which is largely influenced by the Decay Heat Removal strategy (DHR) for depressurized accidents. - An overview of the system transient analysis performed by the partners

75 FR 38988 - Notice of Petitions by Firms for Determination of Eligibility To Apply for Trade Adjustment...

Science.gov (United States)

2010-07-07

..., Romeoville, IL mechanical and mechanical subassemblies 60446. for the automotive, medical, appliance, and consumer electronics industries. DJ Acquisition Management Corp. 6364 Dean Parkway, 6/18/2010 The Company...

Automated cleaning of electronic components

International Nuclear Information System (INIS)

Drotning, W.; Meirans, L.; Wapman, W.; Hwang, Y.; Koenig, L.; Petterson, B.

1994-01-01

Environmental and operator safety concerns are leading to the elimination of trichloroethylene and chlorofluorocarbon solvents in cleaning processes that remove rosin flux, organic and inorganic contamination, and particulates from electronic components. Present processes depend heavily on these solvents for manual spray cleaning of small components and subassemblies. Use of alternative solvent systems can lead to longer processing times and reduced quality. Automated spray cleaning can improve the quality of the cleaning process, thus enabling the productive use of environmentally conscious materials, while minimizing personnel exposure to hazardous materials. We describe the development of a prototype robotic system for cleaning electronic components in a spray cleaning workcell. An important feature of the prototype system is the capability to generate the robot paths and motions automatically from the CAD models of the part to be cleaned, and to embed cleaning process knowledge into the automatically programmed operations

Dictionary materials engineering, materials testing

International Nuclear Information System (INIS)

1994-01-01

This dictionary contains about 9,500 entries in each part of the following fields: 1) Materials using and selection; 2) Mechanical engineering materials -Metallic materials - Non-metallic inorganic materials - Plastics - Composites -Materials damage and protection; 3) Electrical and electronics materials -Conductor materials - Semiconductors - magnetic materials - Dielectric materials - non-conducting materials; 4) Materials testing - Mechanical methods - Analytical methods - Structure investigation - Complex methods - Measurement of physical properties - Non-destructive testing. (orig.) [de

Geopolymer resin materials, geopolymer materials, and materials produced thereby

Energy Technology Data Exchange (ETDEWEB)

Seo, Dong-Kyun; Medpelli, Dinesh; Ladd, Danielle; Mesgar, Milad

2018-01-09

A product formed from a first material including a geopolymer resin material, a geopolymer material, or a combination thereof by contacting the first material with a fluid and removing at least some of the fluid to yield a product. The first material may be formed by heating and/or aging an initial geopolymer resin material to yield the first material before contacting the first material with the fluid. In some cases, contacting the first material with the fluid breaks up or disintegrates the first material (e.g., in response to contact with the fluid and in the absence of external mechanical stress), thereby forming particles having an external dimension in a range between 1 nm and 2 cm.

Fiber-Based 589 nm Laser for Sodium Guide Star

National Research Council Canada - National Science Library

Nilsson, Lars J; Jeong, Yoonchan; Dupriez, Pascal

2006-01-01

...: The work will entail the construction of the required sub-assemblies, their integration into the laser source, and the exploration and adjustment of parameters in order to obtain adequate performance...

Device for facilitating the insertion and withdrawal of fuel assemblies from a nuclear reactor

International Nuclear Information System (INIS)

Andrea, C.; Siegel, E.A.

1976-01-01

A device is provided which is installed in a reactor prior to carrying out refueling operations and which accurately locates and isolates a selected core location to permit rapid withdrawal and insertion of fuel subassemblies at that location. A shielded plug designed to cooperate with the refueling apparatus is inserted into an access port in the reactor head. A structural shroud extends down from the plug and carries at its lower end a radially floating, hexagonal spreader tube with mechanisms to rotate it for angular alignment purposes and a linear drive for inserting it into the core. The upper end of the spreader tube serves as a guide for leading the fuel handling apparatus into alignment with the chosen subassembly

The status of studies on fast reactor core thermal hydraulics at PNC

International Nuclear Information System (INIS)

Nishimura, M.; Ohshima, H.; Kamide, H.; Yamaguchi, K.; Yamaguchi, A.

2000-01-01

An outlook was addressed on investigative activities of the fast reactor core thermal-hydraulics at Power Reactor and Nuclear Fuel Development Corporation. Firstly, a computational modeling to predict flow field under natural circulation decay heat removal condition using multi-dimensional codes and its validation were presented. The validation was carried out through calculations of sodium experiments on an inter-subassembly heat transfer, a transient from forced to natural circulation and an inter-wrapper flow. Secondly, experimental and computational studies were expressed on local blockage with porous media in a fuel subassembly. Lastly, information was presented on an advanced computational code based on a subchannel analysis code. The code is under the development and extended to perform whole core simulation. (author)

Evaluation of neutron streaming in fast breeder reactor fuel assembly by double heterogeneous modelling

International Nuclear Information System (INIS)

Unesaki, Hironobu; Takeda, Toshikazu

1988-01-01

Neutron streaming in a fast breeder reactor fuel assembly caused by the double heterogeneity structure is estimated by double heterogeneous modelling. The conventional pin cell model, a two-region subassembly model and the exact pin cluster model are used to take into account the streaming effect caused by the pin cell structure and the surrounding wrapper tube structure. The heterogeneity of wrapper tube and its surrounding sodium is explicitly considered. The streaming effect is evaluated based on Benoist's diffusion coefficient. The total streaming effect caused by the double heterogeneity structure of a fuel subassembly is found to be -0.2 % dk/kk' for k eff , which is almost twice that obtained from the conventional pin cell model of -0.1 % dk/kk'. (author)

Temperature fluctuations: an assessment of their use in the detection of fast reactor coolant blockages

International Nuclear Information System (INIS)

Greef, C.P.

1979-01-01

The temperature noise technique for the detection of local blockages in fast reactor subassemblies is discussed. The main factors involved in an assessment of the technique are outlined and the experimental and theoretical work that has been carried out at BNL on the various aspects of the problem is described. It is concluded that blockings appreciably smaller than those predicted to produce boiling should be detectable against a background noise level due to subassembly power tilts, on a time scale giving protection against rapidly developing incidents. Further work required to increase confidence in the application of the technique to the reactor is outlined, including measurements in fully representative geometries, data from sodium rigs and further information on reactor background noise levels. (Auth.)

Transmutation of Tc-99 in fission reactors

International Nuclear Information System (INIS)

Kloosterman, J.L.; Li, J.M.

1994-12-01

Transmutation of Tc-99 in three different types of fission reactors is considered: A heavy water reactor, a fast reactor and a light water reactor. For the first type a CANDU reactor was chosen, for the second one the Superphenix reactor, and for the third one a PWR. The three most promising Tc-99 transmuters are the fast reactor with a moderated subassembly in the inner core, a fast reactor with a non-moderated subassembly in the inner core, and a heavy water reactor with Tc-99 target pins in the moderator between the fuel bundles. Transmutation half lives of 15 to 25 years can be achieved, with yearly transmuted Tc-99 masses of about 100 kg at a thermal reactor power of about 3000 MW. (orig.)

Geopolymer resin materials, geopolymer materials, and materials produced thereby

Science.gov (United States)

Seo, Dong-Kyun; Medpelli, Dinesh; Ladd, Danielle; Mesgar, Milad

2016-03-29

A product formed from a first material including a geopolymer resin material, a geopolymer resin, or a combination thereof by contacting the first material with a fluid and removing at least some of the fluid to yield a product. The first material may be formed by heating and/or aging an initial geopolymer resin material to yield the first material before contacting the first material with the fluid. In some cases, contacting the first material with the fluid breaks up or disintegrates the first material (e.g., in response to contact with the fluid and in the absence of external mechanical stress), thereby forming particles having an external dimension in a range between 1 nm and 2 cm.

An interim report on the materials and selection criteria analysis for the Compact Ignition Tokamak Toroidal Field Coil Turn-to-Turn Insulation System

International Nuclear Information System (INIS)

Campbell, V.W.; Dooley, J.B.; Hubrig, J.G.; Janke, C.J.; McManamy, T.J.; Welch, D.E.

1990-01-01

Design criteria for the Compact Ignition Tokamak, Toroidal-Field (TF) Coil, Turn-to-Turn Insulation System require an insulation sheet and bonding system that will survive cryogenic cycling in a radiation environment and maintain structural integrity during exposure to the significant compressive and shear loads associated with each operating cycle. For thermosetting resin systems, a complex interactive dependency exists between optimum peak value, in-service property performance capabilities of candidate generic materials; key handling and processing parameters required to achieve their optimum in-service property performance as an insulation system; and suitability of their handling and processing parameters as a function of design configuration and assembly methodology. This dependency is assessed in a weighted study matrix in which two principal programmatic approaches for the development of the TF Coil Subassembly Insulation System have been identified. From this matrix study, two viable approaches to the fabrication of the insulation sheet were identified: use of a press-formed sheet bonded in place with epoxy for mechanical bonding and tolerance take-up and formation of the insulation sheet by placement of dry cloth and subsequent vacuum pressure impregnation. Laboratory testing was conducted to screen a number of combinations of resins and hardeners on a generic basis. These combinations were chosen for their performance in similar applications. Specimens were tested to screen viscosity, thermal-shock tolerance, and cryogenic tolerance. Cryogenic shock and cryogenic temperature proved to be extremely lethal to many combinations of resin, hardener, and cure. Two combinations survived: a heavily flexibilized bisphenol A resin with a flexibilized amine hardener and a bisphenol A resin with cycloaliphatic amine hardener. 7 refs., 12 figs., 6 tabs

Polymeric dispersants for control of steam generator fouling

International Nuclear Information System (INIS)

Balakrishnan, P.V.; Klimas, S.J.; Lepine, L.; Turner, C.W.

1999-05-01

Fouling of steam generators by corrosion products from the feedtrain leads to loss of heat-transfer efficiency, disturbances in thermalhydraulics, and potential corrosion problems resulting from the development of sites for localized accumulation of aggressive chemicals. This report summarizes studies of the use of polymeric dispersants for the control of fouling, which were conducted at the Chalk River Laboratories. High-temperature settling studies on magnetite suspensions were performed to screen available generic dispersants, and the dispersants were ranked in terms of their dispersion efficiency; polyacrylic acid (PAA) and the phosphonate, HEDP, were ranked as the most efficient. Polyacrylic acid was considered more suitable than HEDP for nuclear steam generators, and more emphasis was given to the former in these studies. The dispersants had no effect on the particle deposition rates under single-phase forced-convective flow, but did reduce the deposition rates under flow-boiling conditions. The extent to which the deposition rates were reduced increased in proportion to the dispersant concentration. Preliminary corrosion tests indicated that pitting or general corrosion of steam generator tube materials in the presence of PAA was negligible. Corrosion of carbon steel, although higher in a magnetite-packed crevice under heat flux than in bulk water, was lower in the presence of PAA than in its absence. Some impurities (e.g., sulphate, sodium) were observed in commercially available PAA products at small, though significant concentrations, making these products unacceptable for use in nuclear plants. However, the PAA could be purified by ion exchange. Preliminary experiments, to assess the thermal stability of PAA at steam generator operating temperature, showed the polymer to break down in deaerated solutions and under argon cover to give hydrogen and carbon dioxide as the two major products in the gas phase and variable concentrations of acetate and formate

Development of a lumped parametric model for scoping investigations of uncertainties in fast reactor probabilistic safety analysis. Progress report, October 10, 1974--October 10, 1975

International Nuclear Information System (INIS)

Ott, K.O.; Luck, L.B.

1975-01-01

The objective of the researh reported is to explore the possibility of the development of a novel reactor safety analysis methodology suitable for a parametric investigation of uncertainties in the progression of severe fast reactor accidents. The essential feature of this approach is a description of the reactor state by means of volumetric distributions (the distribution of volume of reactor materials, such as coolant, clad, and fuel, with temperature and in the case of fuel material, also with power). Stationary volumetric distributions are computed from detailed spatial temperature and power distributions of materials in the steady state reactor. Stationary volumetric distributions and other reactor physics quantities form the input for the reactor transient calculations in which the accident progression is described by the behavior of transient volumetric distributions. The report discusses the representation of spatial temperature distributions, the theory and calculation of stationary volumetric distributions, and includes examples of single subassembly and reactor distributions. The status of reactor neutronic code development and application is discussed and results are displayed

CONSIDERATIONS UPON DESIGNING MODULAR CONSTRUCTIONS FOR IMPROVING THE PRODUCTS ASSEMBLING, MAINTENANCE AND RECYCLING PROCESSES

Directory of Open Access Journals (Sweden)

BÂRSAN Lucian

2015-11-01

Full Text Available Modular constructions are frequently used in industry because of their multiple advantages. Used from the antiquity as a measuring system that ensured good proportions for the objects or buildings, the module is used in present industry as a tool for improving the product maintenance, repair, upgrading, and/or recycling. Modular constructions can be assembled and disassembled easily, facilitating the postuse actions like subassemblies reuse, or materials recovering for the recycling process. An important aspect of this paper is that designers should create the modular solution even from the conceptual design stage and build a structure of functions based on well motivated arguments and which can easily be brake out according to technological possibilities, product functioning and assembly solutions.

UNION-K 2.0 User's Manual

Energy Technology Data Exchange (ETDEWEB)

Yoo, Jae Woon; Jang, Jin Wook; Kim, Yeong Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-12-15

The unified Input/Output processor UNION-K(UNified Input/Output Processor for Neutronics Analysis by K-CORE) has been developed to provide the automatic input file generation and output processing for the fast reactor core analysis. UNION-K includes the material database for reactor analysis and produces the input files for effective cross section generation, core analysis, and reactivity coefficient calculations. UNION-K can execute corresponding core analysis module and edit core performance parameters from the outputs. UNION-K also provides the graphical interface to check the geometrical arrangement of core and fuel sub-assembly and produces the picture files that can be used for reporting. This report has been prepared for instruction of using UNION-K Ver. 2.0 with examples.

UNION-K 2.0 User's Manual

International Nuclear Information System (INIS)

Yoo, Jae Woon; Jang, Jin Wook; Kim, Yeong Il

2008-12-01

The unified Input/Output processor UNION-K(UNified Input/Output Processor for Neutronics Analysis by K-CORE) has been developed to provide the automatic input file generation and output processing for the fast reactor core analysis. UNION-K includes the material database for reactor analysis and produces the input files for effective cross section generation, core analysis, and reactivity coefficient calculations. UNION-K can execute corresponding core analysis module and edit core performance parameters from the outputs. UNION-K also provides the graphical interface to check the geometrical arrangement of core and fuel sub-assembly and produces the picture files that can be used for reporting. This report has been prepared for instruction of using UNION-K Ver. 2.0 with examples

Transmutation in ASTRID

International Nuclear Information System (INIS)

Grouiller, Jean-Paul; Buiron, Laurent; Mignot, Gérard; Palhier, Raphael

2013-01-01

Summary and future prospects for incorporating Am in ASTRID: → Potential to demonstrate the minor actinide transmutation on an industrial scale in the CFV V1 core of ASTRID: • Homogeneous concept: 2% of Am in a standard fuel; • Heterogeneous concept: 10% on UO 2 in the radial blanket. • The objective of ensuring a balance in the Am (and total minor actinides) flow in the ASTRID fuel cycle may be obtained without any impact on the design of the core and handling systems for the management of the new and spent fuel subassemblies. • Several experimental phases in ASTRID to implement different transmutation scenarios using homogeneous and heterogeneous concepts. ⇒ the availability of facilities involved in the ASTRID material cycles

The Jefferson Lab Quality Assurance Program for the SNS Superconducting Linac Construction Project

International Nuclear Information System (INIS)

Joseph Ozelis

2003-01-01

As part of a multi-laboratory collaboration, Jefferson Lab is currently engaged in the fabrication, assembly, and testing of 23 cryomodules for the superconducting linac portion of the Spallation Neutron Source (SNS) being built at Oak Ridge National Laboratory. As with any large accelerator construction project, it is vitally important that these components be built in a cost effective and timely manner, and that they meet the stringent performance requirements dictated by the project specifications. A comprehensive Quality Assurance (QA) program designed to help accomplish these goals has been implemented as an inherent component of JLab's SNS construction effort. This QA program encompasses the traditional spectrum of component performance, from incoming parts inspection, raw materials testing, through to sub-assembly and finished article performance evaluation

Intersubassembly incoherencies and grouping techniques in LMFBR hypothetical overpower accident

International Nuclear Information System (INIS)

Wilburn, N.P.

1977-10-01

A detailed analysis was made of the FTR core using the 100-channel MELT-IIIA code. Results were studied for the transient overpower accident (where 0.5$/sec and 1$/sec ramps) and in which the Damage Parameter and the Failure Potential criteria were used. Using the information obtained from these series of runs, a new method of grouping the subassemblies into channels has been developed. Also, it was demonstrated that a 7-channel representation of the FTR core using this method does an adequate job of representing the behavior during a hypothetical disruptive transient overpower core accident. It has been shown that this new 7-channel grouping method does a better job than an earlier 20-channel grouping. It has also been demonstrated that the incoherency effects between subassemblies as shown during the 76-channel representation of the reactor can be adequately modeled by 7-channels, provided the 7-channels are selected according to the criteria stated in the report. The overall results of power and net reactivity were shown to be only slightly different in the two cases of the 7-channel and the 76-channel runs. Therefore, it can be concluded that any intersubassembly incoherencies can be modeled adequately by a small number of channels, provided the subassemblies making up these channels are selected according to the criteria stated

Augmented cooling vest system subassembly: Design and analysis

International Nuclear Information System (INIS)

D’Angelo, Maurissa; D’Angelo, Joseph; Almajali, Mohammad; Lafdi, Khalid; Delort, Antoine; Elmansori, Mohamed

2014-01-01

Highlights: • Thermoelectric cooler (TEC) was employed to provide cooling air to cooling vest. • Aluminum cooling fins were used to exchange heat for hot and cold sides of TEC. • Performance of the system was determined and the experimental technique was described. • Heat sink is capable to remove additional heat and heat exchanger provides cooling air. • Future work is proposed to optimize the efficiency of the system. - Abstract: A prototype cooling engine consisting of thermoelectric coolers (TECs) was developed and designed. In this prototype, aluminum cooling fins were employed as the heat exchange method for both the hot and cold sides of the TEC. Aluminum fins were used to cool the ambient air through a heat exchanger and dissipate heat build up from the heat sink. This system was modeled and performance capabilities were determined. The experimental technique used to monitor parameters affecting the efficiency of the designed system was described. These parameters include the temperatures of the inlets and outlets of both heat exchanger and heat sink and the flow rate of the cooled air. The experiment was run under three input DC powers; 15 V, 18 V, and 21 V. As the power increased, both the flow rate and the temperature difference between the hot and cold side of thermoelectric cooler increased, demonstrating the heat sink capability to remove the additional heat. However, the temperature difference between the inlet and outlet of the heat exchanger decreases as the power increase. The findings demonstrated the effectiveness of this cooling system and future work is proposed to optimize the heat

Applications of ultrasonic phased array technique during fabrication of nuclear tubing and other components for the Indian nuclear power program

International Nuclear Information System (INIS)

Kapoor, K.

2015-01-01

Ultrasonic phased array technique has been applied in fabrication of nuclear fuel and structural at NFC. The integrity of the nuclear fuel and structural components is most crucial as they are exposed to severe environment during operation leading to rapid degradation of its properties during its lifecycle. Nuclear Fuel Complex has mandate for the fabrication of the nuclear fuel and core structurals for Indian PHWRs/BWR, sub-assemblies for the PFBR and steam generator tubing for PFBR and PHWRs which are the most critical materials for the Indian Nuclear Power program. NDE during fabrication of these materials is thus most crucial as it provides the confidence to the designer for safe operation during its lifetime. Many of these techniques have to be developed in-house to meet unique requirements of high sensitivity, resolution and shape of the components. Some of the advancements in the NDE during the fabrication include use of ultrasonic phased array which is detailed in this paper

Validation of the REBUS-3/RCT methodologies for EBR-II core-follow analysis

International Nuclear Information System (INIS)

McKnight, R.D.

1992-01-01

One of the many tasks to be completed at EBR-2/FCF (Fuel Cycle Facility) regarding fuel cycle closure for the Integral Fast Reactor (IFR) is to develop and install the systems to be used for fissile material accountancy and control. The IFR fuel cycle and pyrometallurgical process scheme determine the degree of actinide of actinide buildup in the reload fuel assemblies. Inventories of curium, americium and neptunium in the fuel will affect the radiation and thermal environmental conditions at the fuel fabrication stations, the chemistry of reprocessing, and the neutronic performance of the core. Thus, it is important that validated calculational tools be put in place for accurately determining isotopic mass and neutronic inputs to FCF for both operational and material control and accountancy purposes. The primary goal of this work is to validate the REBUS-2/RCT codes as tools which can adequately compute the burnup and isotopic distribution in binary- and ternary-fueled Mark-3, Mark-4, and Mark-5 subassemblies. 6 refs

Comparison of effect of insulating blockages on metal and oxide fuel elements

International Nuclear Information System (INIS)

Tilbrook, R.W.; Dever, D.J.

1988-01-01

The safety philosophy of the new liquid metal reactor (LMR) plant designs is oriented towards inherent protection against loss of coolable geometry and other entries to core disruption. On potential entry is via propagation of local faults. Within this category is a wide range of initiators which each require assessment of their probability and consequences in order to determine their contribution to plant risk. Local faults include those initiators which cause local power/flow disturbances restricted either to a single subassembly or to a local region of the bundle. The concern is that these localized initiators may start a sequence of events in which fuel failure may propagate first within a subassembly envelope and finally cause loss of coolable geometry in adjacent. This document discusses these scenarios. 3 refs., 1 fig

Method for cooling a breeder reactor and breeder reactor for applying the method

International Nuclear Information System (INIS)

Gast, K.

1977-01-01

The fuel assemblies of the fission zone and the breeder subassemblies in the radial breeding blanket are supported on a double bottom. The coolant gets into the interspace of the double bottom below the blanket. This part of the space is separated from the interspace of the double bottom below the fission zone. Each breeder subassembly consists of a twin tube. The coolant enters the inner tube, flows through it upwards in axial direction, is then deflected on the upper end, and afterwards, in the annulus of the twin tube, flows down again in axial direction into the inlet region, below the double bottom. From there on it flows upwards through the fuel assemblies of the fission zone from below. Thereby a uniformly high coolant outlet temperature is obtained. (DG) [de

Status of Preliminary Design on the Assembly Tools for ITER Tokamak Machine

International Nuclear Information System (INIS)

Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin; Moon, Jae Hwan; Kim, Byung Seok; Lee, Jae Hyuk; Shaw, Robert

2012-01-01

The ITER Tokamak device is principally composed of nine 40 .deg. sectors. Each 40 .deg. sector is made up of one 40 .deg. vacuum vessel (VV), two 20 .deg. toroidal filed coils (TFC) and associated vacuum vessel thermal shield (VVTS) segments which consist of one inboard and two outboard vacuum vessel thermal shields. Based on the design description document and final report prepared by the ITER organization (IO) and conceptual design, Korea has carried out the preliminary design of these assembly tools. The assembly strategy and relevant tools for the 40 .deg. sector sub-assembly and sector assembly at in-pit should be developed to satisfy the basic assembly requirements of the ITER Tokamak machine. Assembly strategy, preliminary design of the sector sub-assembly and assembly tools are described in this paper

The internal core catcher in Super Phenix 1

International Nuclear Information System (INIS)

Le Rigoleur, C.; Kayser, G.; Maurin, G.; Magnon, B.

1982-07-01

The internal core catcher in SUPER PHENIX 1 is described here in some detail. The fuel retention capabilities are presented for situations of increasing severity. The first situation corresponds to the core catcher design. It relates to a hypothetical subassembly accident that would cause a limited quantity of fuel, corresponding to the mass of seven subassemblies, to be deposited on the core catcher. For this situation and at all levels of the analysis, the most conservative assumptions are made in order to prove the integrity of the core catcher. The second situation corresponds to a hypothetical larger core melt accident. In this case, for some of the parameters, assumptions are made that correspond to the most likely situations based on engineering considerations. Then the maximum retention capabilities are presented

Extraction Compression and Acceleration of High Line Charge Density Ion Beams

CERN Document Server

Henestroza, Enrique; Grote, D P; Peters, Craig; Yu, Simon

2005-01-01

HEDP applications require high line charge density ion beams. An efficient method to obtain this type of beams is to extract a long pulse, high current beam from a gun at high energy, and let the beam pass through a decelerating field to compress it. The low energy beam bunch is loaded into a solenoid and matched to a Brillouin flow. The Brillouin equilibrium is independent of the energy if the relationship between the beam size (a), solenoid magnetic field strength (B) and line charge density is such that (Ba)2

The First Experiments on the National Ignition Facility

International Nuclear Information System (INIS)

Landen, O L; Glenzer, S; Froula, D; Dewald, E; Suter, L J; Schneider, M; Hinkel, D; Fernandez, J; Kline, J; Goldman, S; Braun, D; Celliers, P; Moon, S; Robey, H; Lanier, N; Glendinning, G; Blue, B; Wilde, B; Jones, O; Schein, J; Divol, L; Kalantar, D; Campbell, K; Holder, J; MacDonald, J; Niemann, C; Mackinnon, A; Collins, R; Bradley, D; Eggert, J; Hicks, D; Gregori, G; Kirkwood, R; Young, B; Foster, J; Hansen, F; Perry, T; Munro, D; Baldis, H; Grim, G; Heeter, R; Hegelich, B; Montgomery, D; Rochau, G; Olson, R; Turner, R; Workman, J; Berger, R; Cohen, B; Kruer, W; Langdon, B; Langer, S; Meezan, N; Rose, H; Still, B; Williams, E; Dodd, E; Edwards, J; Monteil, M; Stevenson, M; Thomas, B; Coker, R; Magelssen, G; Rosen, P; Stry, P; Woods, D; Weber, S; Alvarez, S; Armstrong, G; Bahr, R; Bourgade, J; Bower, D; Celeste, J; Chrisp, M; Compton, S; Cox, J; Constantin, C; Costa, R; Duncan, J; Ellis, A; Emig, J; Gautier, C; Greenwood, A; Griffith, R; Holdner, F; Holtmeier, G; Hargrove, D; James, T; Kamperschroer, J; Kimbrough, J; Landon, M; Lee, D; Malone, R; May, M; Montelongo, S; Moody, J; Ng, E; Nikitin, A; Pellinen, D; Piston, K; Poole, M; Rekow, V; Rhodes, M; Shepherd, R; Shiromizu, S; Voloshin, D; Warrick, A; Watts, P; Weber, F; Young, P; Arnold, P; Atherton, L J; Bardsley, G; Bonanno, R; Borger, T; Bowers, M; Bryant, R; Buckman, S; Burkhart, S; Cooper, F; Dixit, S; Erbert, G; Eder, D; Ehrlich, B; Felker, B; Fornes, J; Frieders, G; Gardner, S; Gates, C; Gonzalez, M; Grace, S; Hall, T; Haynam, C; Heestand, G; Henesian, M; Hermann, M; Hermes, G; Huber, S; Jancaitis, K; Johnson, S; Kauffman, B; Kelleher, T; Kohut, T; Koniges, A E; Labiak, T; Latray, D; Lee, A; Lund, D; Mahavandi, S; Manes, K R; Marshall, C; McBride, J; McCarville, T; McGrew, L; Menapace, J.

2005-01-01

A first set of laser-plasma interaction, hohlraum energetics and hydrodynamic experiments have been performed using the first 4 beams of the National Ignition Facility (NIF), in support of indirect drive Inertial Confinement Fusion (ICF) and High Energy Density Physics (HEDP). In parallel, a robust set of optical and x-ray spectrometers, interferometer, calorimeters and imagers have been activated. The experiments have been undertaken with laser powers and energies of up to 8 TW and 17 kJ in flattop and shaped 1-9 ns pulses focused with various beam smoothing options

The complexes of Ho with methylenediphosphonate and 1 hydroxyethylidenephosphonate

International Nuclear Information System (INIS)

Vanura, P.; Jedinakova-Krizova, V.; Hakenova, L.; Munesawa, Y.

1999-01-01

The composition and stability of holmium methylenediphosphonate (MDP) and 1-hydroxyethylidenephosphonate (HEDP) complexes in the aqueous solution of 0.1 M NaCl at the temperature of 25 grad C were studied by potentiometric titration methods. The complexes of the composition HoH n L have been found in the aqueous solution if the concentration of the ligand is higher than the concentration of holmium. The protonation constants of both acids and stability constants of all complexes were determined and the comparison with literature data of analogical complexes of other lanthanides was performed. (authors)

e-Commerce and supply chains: Modelling of dynamics through ...

Indian Academy of Sciences (India)

R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

The dynamics associated with two production planning and control policies are modelled, viz. .... Hence, there is a strong need to design a dynamic knowledge inference system .... sell a variety of components to the subassembly manufacturer.

Overproduction and Improved Strategies for Isolation and ...

African Journals Online (AJOL)

Yajing Zhou

2012-04-24

Apr 24, 2012 ... purification of human DNA polymerase delta and its subassemblies ... that can result from loss of genomic stability. Therefore, a ..... different enzyme assemblies as shown in panel A. PCNA, proliferating cell nuclear antigen.

Aerospace materials and material technologies

CERN Document Server

Wanhill, R

2017-01-01

This book is a comprehensive compilation of chapters on materials (both established and evolving) and material technologies that are important for aerospace systems. It considers aerospace materials in three Parts. Part I covers Metallic Materials (Mg, Al, Al-Li, Ti, aero steels, Ni, intermetallics, bronzes and Nb alloys); Part II deals with Composites (GLARE, PMCs, CMCs and Carbon based CMCs); and Part III considers Special Materials. This compilation has ensured that no important aerospace material system is ignored. Emphasis is laid in each chapter on the underlying scientific principles as well as basic and fundamental mechanisms leading to processing, characterization, property evaluation and applications. A considerable amount of materials data is compiled and presented in appendices at the end of the book. This book will be useful to students, researchers and professionals working in the domain of aerospace materials.

Transient effects in unstable ablation fronts and mixing layers in HEDP

International Nuclear Information System (INIS)

Clarisse, J-M; Gauthier, S; Dastugue, L; Vallet, A; Schneider, N

2016-01-01

We report results obtained for two elementary unstable flow configurations relevant to high energy density physics: the ablation front instability and the Rayleigh–Taylor -instability induced mixing layer. These two flows are characterized by a transience of their perturbation dynamics. In the ablative flow case, this perturbation dynamics transience takes the form of finite-durations of successive linear-perturbation evolution phases until reaching regimes of decaying oscillations. This behaviour is observed in various regimes: weakly or strongly accelerated ablation fronts, irradiation asymmetries or initial external-surface defects, and is a result of the mean-flow unsteadiness and stretching. In the case of the Rayleigh–Taylor-instability induced mixing layer, perturbation dynamics transience manifests itself through the extinction of turbulence and mixing as the flow reaches a stable state made of two stably stratified layers of pure fluids separated by an unstratified mixing layer. A second feature, also due to compressibility, takes the form of an intense acoustic wave production, mainly localized in the heavy fluid. Finally, we point out that a systematic short-term linear-perturbation dynamics analysis should be undertaken within the framework of non-normal stability theory. (paper)

Formation of Imploding Plasma Liners for HEDP and MIF Applications - Diagnostics

Energy Technology Data Exchange (ETDEWEB)

Gilmore, Mark [Univ. of New Mexico, Albuquerque, NM (United States). Dept. of Electrical and Computer Engineering. Dept. of Physics and Astronomy; Hsu, Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Witherspoon, F. Douglas [HyperV Technologies Corp., Chantilly, VA (United States); Cassibry, Jason [Univ. of Alabama, Huntsville, AL (United States); Bauer, Bruno S. [Univ. of Nevada, Reno, NV (United States)

2015-04-27

The goal of the plasma liner experiment (PLX) was to explore and demonstrate the feasibility of forming imploding spherical plasma liners that can reach High Energy Density (HED)-relevant (~ 0.1 Mbar) pressures upon stagnation. The plasma liners were to be formed by a spherical array of 30 – 36 railgun-driven hypervelocity plasma jets (Mach 10 – 50). Due to funding and project scope reductions in year two of the project, this initial goal was revised to focus on studies of individual jet propagation, and on two jet merging physics. PLX was a collaboration between a number of partners including Los Alamos National Laboratory, HyperV Technologies, University of New Mexico (UNM), University of Alabama, Huntsville, and University of Nevada, Reno. UNM’s part in the collaboration was primary responsibility for plasma diagnostics. Though full plasma liner experiments could not be performed, the results of single and two jet experiments nevertheless laid important groundwork for future plasma liner investigations. Though challenges were encountered, the results obtained with one and two jets were overwhelmingly positive from a liner formation point of view, and were largely in agreement with predictions of hydrodynamic models.

78 FR 60248 - Foreign-Trade Zone (FTZ) 183-Austin, Texas; Notification of Proposed Production Activity...

Science.gov (United States)

2013-10-01

... shroud assemblies; mechanism bases; storage; busbars; button dim links; electromagnetic interference fans...; connector brackets; frames; holders; insulators; link torsion; manifold exhausts; stiffeners; subassemblies; thermal pads; insert mold torsion bars; torsion springs; vapor chambers; power supplies; housing magnets...

Optimization of FBR fuel element for high burnup

International Nuclear Information System (INIS)

Marbach, G.; Millet, P.

1985-03-01

After a brief historical background showing evolution of the French fast reactor fuel element from RAPSODIE to PHENIX and SUPER PHENIX we have examined the main points which have permitted to increase irradiation performance of the subassembly

Microstructure examination of the interface of the glass-ceramic insulator of the molybdenum frame of a vacuum tube

International Nuclear Information System (INIS)

Spears, R.K.

1980-01-01

A common technique used in examining the structural integrity of a glass-ceramic insulator-molybdenum cylinder bond in a vacuum tube subassembly is to slit the outer molybdenum cylinder and separate it from the glass-ceramic insulator. Typically, a black glassy layer (0.001 to 0.002 in. thick) remains on the cylinder. This layer has been interpreted as a requirement for an adequate seal. A subassembly was found that did not exhibit this feature. Further investigation of approximately 100 subassemblies revealed four more parts lacking a black glassy layer. These parts were found to be from two production runs and from three glass-ceramic lots. A microstructural analysis showed that on those parts having a black glassy layer, the crystalline phase in the glass-ceramic grew to within one to two microns of the metal interface and then terminated. A dark region existed in the insulator between the interface and the termination of the crystalline phase. This was attributed to molybdenum oxide dissolved in the glass. On those parts where the glass-ceramic broke clean from the cylinder, the crystalline phase extended up to the metal. Also observed on these parts was the appearance of a dark region adjacent to the metal that extended approximately one to two microns into the glass-ceramic. This was assumed to be an oxide of molybdenum. This report presents information concerning the microstructure of the interface

An overview of IPPE research on liquid metal fast reactor thermohydraulics

International Nuclear Information System (INIS)

Sorokin, A. P.; Efanov, A. D.; Zhukov, A. V.; Bogoslovskaia, G. P.

2003-01-01

The paper presents brief information on the most significant researches in the fields of liquid metal hydrodynamics and heat transfer performed in the State Scientific Center of Russian Federation 'Institute for Physics and Power Engineering' named after A.I.Leypunski applied to sodium-cooled fast reactors. Experimental methods for studying liquid metal thermohydraulics and applied measurement techniques are overviewed briefly in the paper. Some results of fundamental thermohydraulic investigations, such as quasi-universal character of velocity and temperature profile in liquid metals, if considered normally to the channel wall etc. are presented. Specific features of heat transfer in liquid metal cooled fuel subassembly are mentioned, among them there are: high level of coolant temperature; significant influence of an interchannel exchange on velocity and temperature distribution; an availability of contact thermal resistance; large azimuthal non-uniformity of velocity and temperature; 'conjugate' problem of heat transfer in combined geometry of fuel pin; an absence of stabilization of heat transfer in non-standard channels; an influence of non-uniform heat generation. Special attention is given to the temperature fields in fuel subassembly subjected to deformation because of radioactive swelling and creeping, as well as in case of blockage of a part of subassembly cross section. Some results of thermohydraulic investigation are demonstrated for intermediate heat exchangers, pressurized head collectors. Also the developed methods and codes of thermohydraulic calculations applied to fast reactor core are considered: subchannel approach, porous body model

Two-dimensional calculation by finite element method of velocity field and temperature field development in fast reactor fuel assembly. II

International Nuclear Information System (INIS)

Schmid, J.

1985-11-01

A package of updated computer codes for velocity and temperature field calculations for a fast reactor fuel subassembly (or its part) by the finite element method is described. Isoparametric triangular elements of the second degree are used. (author)

STEP flight experiments Large Deployable Reflector (LDR) telescope

Science.gov (United States)

Runge, F. C.

1984-01-01

Flight testing plans for a large deployable infrared reflector telescope to be tested on a space platform are discussed. Subsystem parts, subassemblies, and whole assemblies are discussed. Assurance of operational deployability, rigidization, alignment, and serviceability will be sought.

Temperature field downstream of an heated bundle mock-up results for different power distribution

International Nuclear Information System (INIS)

Girard, J.P.; Buravand, Y.

1982-10-01

The aim of these peculiar experiments performed on the ML4 loop in ISPRA is to evaluate the characteristics of the temperature field over a length of 20 to 30 dias downstream of a rod bundle for different temperatures profiles at the bundle outlet. The final purpose of this work will be to establish either directly or through models whether it is possible or not to detect subassembly failures using suitable of the subassembly outlet temperature signal. 15 hours of digital and analog recording were taped for five different power distributions in the bundle. The total power dissipation remained constant during the whole run. Two flow rates and seven axial location were investigated. It is shown that the different temperature profiles produce slight differences in the variance and skewness of the temperature signal measured along the axis of the pipe over 20 dias

Gamma scanning of mixed carbide and oxide fuel pins irradiated in FBTR

International Nuclear Information System (INIS)

Jayaraj, V.V.; Padalakshmi, M.; Ulaganathan, T.; Venkiteswaran, C.N.; Divakar, R.; Joseph, Jojo; Bhaduri, A.K.

2016-01-01

Fission in nuclear fuels results in a number of fission products that are gamma emitters in the energy range of 100 keV to 3 MeV. The gamma emitting fission products are therefore amenable for detection by gamma detectors. Assessment of the fission product distribution and their migration behavior through gamma scanning is important for characterizing the in reactor behavior of the fuel. Gamma scanning is an important non destructive technique used to evaluate the behavior of irradiated fuels. As a part of Post Irradiation Examinations (PIE), axial gamma scanning has been carried out on selected fuel pins of the FBTR Mark I mixed carbide fuel sub-assemblies and PFBR MOX test fuel sub-assembly irradiated in FBTR. This paper covers the results of gamma scanning and correlation of gamma scanning results with other PIE techniques

In-core flow rate distribution measurement test of the JOYO irradiation core

International Nuclear Information System (INIS)

Suzuki, Toshihiro; Isozaki, Kazunori; Suzuki, Soju

1996-01-01

A flow rate distribution measurement test was carried out for the JOYO irradiation core (the MK-II core) after the 29th duty cycle operation. The main object of the test is to confirm the proper flow rate distribution at the final phase of the MK-II core. The each flow rate at the outlet of subassemblies was measured by the permanent magnetic flowmeter inserted avail of fuel exchange hole in the rotating plug. This is third test in the MK-II core, after 10 years absence from the final test (1985). Total of 550 subassemblies were exchanged and accumulated reactor operation time reached up to 38,000 hours from the previous test. As a conclusion, it confirmed that the flow rate distribution has been kept suitable in the final phase of the MK-II core. (author)

Material efficiency: providing material services with less material production.

Science.gov (United States)

Allwood, Julian M; Ashby, Michael F; Gutowski, Timothy G; Worrell, Ernst

2013-03-13

Material efficiency, as discussed in this Meeting Issue, entails the pursuit of the technical strategies, business models, consumer preferences and policy instruments that would lead to a substantial reduction in the production of high-volume energy-intensive materials required to deliver human well-being. This paper, which introduces a Discussion Meeting Issue on the topic of material efficiency, aims to give an overview of current thinking on the topic, spanning environmental, engineering, economics, sociology and policy issues. The motivations for material efficiency include reducing energy demand, reducing the emissions and other environmental impacts of industry, and increasing national resource security. There are many technical strategies that might bring it about, and these could mainly be implemented today if preferred by customers or producers. However, current economic structures favour the substitution of material for labour, and consumer preferences for material consumption appear to continue even beyond the point at which increased consumption provides any increase in well-being. Therefore, policy will be required to stimulate material efficiency. A theoretically ideal policy measure, such as a carbon price, would internalize the externality of emissions associated with material production, and thus motivate change directly. However, implementation of such a measure has proved elusive, and instead the adjustment of existing government purchasing policies or existing regulations-- for instance to do with building design, planning or vehicle standards--is likely to have a more immediate effect.

Studies and research concerning BNFP: shearing tests conducted at Allied-General Nuclear Services for the Consolidated Fuel Reprocessing Program

International Nuclear Information System (INIS)

Weil, B.; Townes, G.

1979-09-01

An experiment conducted to shear two dummy PWR subassemblies is described. Results pertain to the removal of end hardware by shearing, spacer grid fragmentation, the character of sheared product, product leachability, shearing force requirements, and the effects of compaction

Improved control system power unit for large parachutes

Science.gov (United States)

Chandler, J. A.; Grubbs, T. M.

1968-01-01

Improved control system power unit drives the control surfaces of very large controllable parachutes. The design features subassemblies for determining control surface position and cable loading, and protection of the load sensor against the possibility of damage during manipulation.

Brazing of sensors for high-temperature steam instrumentation systems

International Nuclear Information System (INIS)

Moorhead, A.J.; Morgan, C.S.; Woodhouse, J.J.; Reed, R.W.

1981-01-01

Procedures are developed for brazing a ceramic-to-metal seal and for laser welding of sensor subassemblies into tube walls, induction brazing thermocouples through a tube wall, and furnace brazing triaxial cables, thermocouples, and a vent tube to a guide tube

Sodium boiling detection in LMFBRs (Phase I). 5th quarterly technical progress report, July 1, 1975--October 31, 1975

International Nuclear Information System (INIS)

Albrecht, R.W.; McCormick, N.J.

1975-01-01

Progress summarized includes the design of a gamma heated subassembly for sodium boiling experiments and an experiment showing that neutronic noise and acoustic noise caused by sodium boiling are highly correlated in a wide frequency band about the bubble repetition frequency

Theoretical and experimental investigations of the thermo-hydraulics of deformed wire-wrapped bundles in nominal flow conditions

International Nuclear Information System (INIS)

Leteinturier, D.; Cartier, L.

1979-01-01

Theoretical and experimental studies undertaken in CEN Cadarache on deformed subassemblies are presented. After the mainlines description of this program first temperature distribution results are given on an in-pile experiment in RAPSODIE (61 pins). Comparison with calculation is made

Replacement of fluid-filter elements without interruption of flow

Science.gov (United States)

Kotler, R. A.; Ward, J. B.

1969-01-01

Gatling-type filter assembly, preloaded with several filter elements enables filter replacement without breaking into the operative fluid system. When the filter element becomes contaminated, a unit inner subassembly is rotated 60 degrees to position a clean filter in the line.

Commentary: The Materials Project: A materials genome approach to accelerating materials innovation

Directory of Open Access Journals (Sweden)

Anubhav Jain

2013-07-01

Full Text Available Accelerating the discovery of advanced materials is essential for human welfare and sustainable, clean energy. In this paper, we introduce the Materials Project (www.materialsproject.org, a core program of the Materials Genome Initiative that uses high-throughput computing to uncover the properties of all known inorganic materials. This open dataset can be accessed through multiple channels for both interactive exploration and data mining. The Materials Project also seeks to create open-source platforms for developing robust, sophisticated materials analyses. Future efforts will enable users to perform ‘‘rapid-prototyping’’ of new materials in silico, and provide researchers with new avenues for cost-effective, data-driven materials design.

Computational Materials Science | Materials Science | NREL

Science.gov (United States)

Computational Materials Science Computational Materials Science An image of interconnecting, sphere science capabilities span many research fields and interests. Electronic, Optical, and Transport Properties of Photovoltaic Materials Material properties and defect physics of Si, CdTe, III-V, CIGS, CZTS

Materials Discovery | Materials Science | NREL

Science.gov (United States)

Discovery Materials Discovery Images of red and yellow particles NREL's research in materials characterization of sample by incoming beam and measuring outgoing particles, with data being stored and analyzed Staff Scientist Dr. Zakutayev specializes in design of novel semiconductor materials for energy

Materials 2014: a great success for materials sector

International Nuclear Information System (INIS)

Isnard, Olivier; Crepin, Jerome

2014-01-01

In this work are presented the summaries of the 19 symposiums presented at the conference: 'Materials 2014' and whose topics were: eco-materials, materials for energy storage and conversion, strategic materials, rare elements and recycling, surfaces functionalization and physico-chemical characterization, interfaces and coatings, corrosion, aging, durability, damage mechanical behaviours, disordered materials, glasses and their functionalization, materials and health, functional materials, porous, granular and with a high surface area materials, nano-materials, nano-structured systems, assembling processes, carbonaceous materials, great instruments and studies of materials, materials in severe conditions, powder forming processes, metallic materials and structures lightening. (O.M.)

Concept of operations for channel characterization and simulation of coaxial transmission channels at the National Ignition Facility (NIF)

Energy Technology Data Exchange (ETDEWEB)

Brown, Jr., Charles G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-03-23

The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) executes experiments for inertial con nement fusion (ICF), world-class high energy density physics (HEDP), and critical national security missions. While the laser systems, target positioners, alignment systems, control systems, etc. enable the execution of such experiments, NIF’s utility would be greatly reduced without its suite of diagnostics. It would be e ectively “blind” to the incredible physics unleashed in its target chamber. Since NIF diagnostics are such an important part of its mission, the quality and reliability of the diagnostics, and of the data recorded from them, is crucial.

The VULKIN code used for evaluation of the cladding tube's performance

International Nuclear Information System (INIS)

Marbach, G.

1979-01-01

Full text: 1 - Introduction. The French approach for fast subassembly project is to analyse each component part of the subassembly and each basic phenomenon to estimate the total behaviour. The VULKIN code describes the mechanical behaviour of a clad alone. A cladding damage parameter is calculated from the observed deformations. When it is greater than a fixed value we consider that the rupture probability is not negligible. But this function is not the only limit for the irradiation project. Other limits are bound to other problems: no fuel melting bundle, interaction behaviour. 2 - VULKIN code - Presentation. The VULKIN code gives the evolution of stresses and strains distribution in the thickness of the clad with the hypothesis of revolution symmetry. This program takes into account temperature dilatation and radial thermal gradient, fission gas pressure and steel swelling due to neutron flux. The fuel clad mechanical interaction is not described by this model. Experimental results show that its influence is negligible for the most unusual subassemblies but, if it is necessary, a special calculation is obtained using a specific code like TUREN, described in another paper. This model does not consider the stresses and strains resulting from interaction between bundle and wrapper. Another model describes the bundle behaviour and determines diametral deformation limit from the subassembly geometrical characteristics. The clad is considered as an elasto-plastic element. Plastic flows instantaneous, thermal creep or irradiation creep are determined at each time. The data of this code are the geometry, the irradiation parameters (temperature, dose), the fission gas pressure evolution, the swelling law and the experimental relations for thermal and irradiation creep. The mechanical resolution is classical: the clad is divided into concentric rings. At each time the equations resulting from the equilibrium of strengths and compatibility of displacements are resolved

Towards Materials Sustainability through Materials Stewardship

Directory of Open Access Journals (Sweden)

Christopher D. Taylor

2016-10-01

Full Text Available Materials sustainability requires a concerted change in philosophy across the entire materials lifecycle, orienting around the theme of materials stewardship. In this paper, we address the opportunities for improved materials conservation through dematerialization, durability, design for second life, and diversion of waste streams through industrial symbiosis.

75 FR 32806 - Notice of Issuance of Final Determination Concerning Certain Upright and Recumbent Exercise Bikes

Science.gov (United States)

2010-06-09

... resistor/bracket/cable assembly; the PCB/ battery assembly; the reed switch/bracket subassembly; the shroud...; (alternator-pulley assembly) 9. Assembling resistor, resistor brackets, resistor rod and covering the assembly with cardboard insulator; (rear resistor/ bracket/cable assembly) 10. Installing wire harness to the...

Brazing of special metallic materials and material combinations using a special material

International Nuclear Information System (INIS)

Lison, R.

1981-01-01

The special materials include metals of groups IVa, Va and VIa of the periodic tables and their alloys. Their particular properties have won them applications in many highly specialized industries. For these materials to be used, mastery of thermal joining methods appropriate to their characteristics is necessary. High-temperature brazing is one such method for joining special materials. This paper presents variants of this technique suitable for each individual special material. Compatibility tests between various brazing metals and various special materials have been carried out by simulating the temperature/time cycle involved in brazing procedures. Special materials are relatively expensive, and their special properties are not required at every point in a structure: elsewhere they can be replaced by a different special material or by other metals or alloys. This means that joints must be made between two special materials or between a special material and a conventional material. When certain conditions are fulfilled, such joins can be made by high-temperature brazing. This paper also shows the extent to which the geometry of the join determines the choice of process. Example of applications are also given. (orig.)

CAT reconstruction and potting comparison of a LMFBR fuel bundle

International Nuclear Information System (INIS)

Betten, P.R.; Tow, D.M.

1984-04-01

A standard Liquid Metal Fast Breeder Reactor (LMFBR) subassembly used in the Experimental Breeder Reactor II (EBR-II) was investigated, by remote techniques, for fuel bundle distortion by both nondestructive and destructive methods, and the results from both methods were compared. The non-destructive method employed neutron tomography to reconstruct the locations of fuel elements through the use of a maximum entropy reconstruction algorithm known as MENT. The destructive method consisted of ''potting'' (a technique that embeds and permanently fixes the fuel elements in a solid matrix) the subassembly, and then cutting and polishing the individual sections. The comparison indicated that the tomography reconstruction provided good results in describing the bundle geometry and spacer-wire locations, with the overall resolution being on the order of a spacer-wire diameter. A dimensional consistency check indicated that the element and spacer-wire dimensions were accurately reproduced in the reconstruction

Implications of manufacturing deviations on fuel performance

International Nuclear Information System (INIS)

Chellapandi, P.; Clement Ravi Chandar, S.; Chetal, S.C.; Baldev Raj

2009-01-01

Prototype Fast Breeder Reactor (PFBR) core consists of 181 Fuel subassembly (FSA), 120 blanket SA and shielding SA of steel and B 4 C besides 9 Control Safety Rod SA and 3 Diverse Safety Rod SA. All the subassemblies (SA) stand vertically on the grid plate. PFBR FSA consists of 217 fuel pins of 2540 mm length arranged in a triangular pitch standing vertically on the rails inside a hexagonal duct. The bottom of the hexagon is screwed and welded to a cylindrical foot of length 600 mm. The SA foot has radial slots that provide an entry to the coolant sodium and also houses flow control devices. While the foot of the A handling head is welded to the top of hexagon, provides an aid for fuel handling machine to insert and withdraw the FSA from the grid plate. The length of the FSA is 4500 mm.

EBRPOCO - a program to calculate detailed contributions of power reactivity components of EBR-II

International Nuclear Information System (INIS)

Meneghetti, D.; Kucera, D.A.

1981-01-01

The EBRPOCO program has been developed to facilitate the calculations of the power coefficients of reactivity of EBR-II loadings. The program enables contributions of various components of the power coefficient to be delineated axially for every subassembly. The program computes the reactivity contributions of the power coefficients resulting from: density reduction of sodium coolant due to temperature; displacement of sodium coolant by thermal expansions of cladding, structural rods, subassembly cans, and lower and upper axial reflectors; density reductions of these steel components due to temperature; displacement of bond-sodium (if present) in gaps by differential thermal expansions of fuel and cladding; density reduction of bond-sodium (if present) in gaps due to temperature; free axial expansion of fuel if unrestricted by cladding or restricted axial expansion of fuel determined by axial expansion of cladding. Isotopic spatial contributions to the Doppler component my also be obtained. (orig.) [de

Method for assembling dynamoelectric machine end shield parts

International Nuclear Information System (INIS)

Thomson, J.M.

1984-01-01

Methods, apparatus, and systems are provided for automatically assembling end shield assemblies of subassemblies for electric motors. In a preferred form, a system and methods are provided that utilize a non-palletized, non-synchronous concept to convey end shields through a number of assembly stations. At process stations situated along a conveyor, operations are performed on components. One method includes controlling traffic of sub-assemblies by toggle type escapements. A stop or latch of unique design stops end shield components in midstream, and ''lifts'' of unique design disengage parts from the conveyor and also support such parts during various operations. Photo-optic devices and proximity and reed switch mechanisms are utilized for control purposes. The work stations involved in one system include a unique assembly and pressing station involving oil well covers; a unique feed wick seating system; a unique lubricant adding operation; and unique ''building block'' mechanisms and methods

Bowing behavior of subassemblies in experimental fast reactor ''JOYO''

International Nuclear Information System (INIS)

Ikegami, T.; Mizoo, N.; Matsuno, Y.; Watari, Y.

1984-01-01

In JOYO, the measured power coefficients in the beginning of the operation cycle of MK-I and MK-II cores showed power dependence, while the calculation without taking account of bowing predicted little power dependence. The bowing analysis was performed in order to investigate the power dependence observed in the measured power coefficients and the following conclusions were obtained. (1) The evaluated power coefficients taking account of bowing effect agree better with measured ones than the calculated ones without taking account of bowing effect in MK-I core. (2) In MK-II core, although the analytical results show not so good agreement quantitatively with the measured power coefficients, it is suggested that they agree better depending on the uncertain parameters such as the heat generation in the reflector region, the threshold moment for leaning and the stiffness of the inner reflector. (3) It becomes clear from these results that the power dependence observed in the measured power coefficients in JOYO is due to the bowing effect. (author)

Helium leak testing of large pressure vessels or subassemblies

International Nuclear Information System (INIS)

Hopkins, J.S.; Valania, J.J.

1977-01-01

Specifications for pressure-vessel components [such as the intermediate heat exchangers (IHX)] for service in the liquid metal fast breeder reactor facilities require helium leak testing of pressure boundaries to very exacting standards. The experience of Foster Wheeler Energy Corporation (FWEC) in successfully leak-testing the IHX shells and bundle assemblies now installed in the Fast Flux Test Facility at Richland, WA is described. Vessels of a somewhat smaller size for the closed loop heat exchanger system in the Fast Flux Test Facility have also been fabricated and helium leak tested for integrity of the pressure boundary by FWEC. Specifications on future components call for helium leak testing of the tube to tubesheet welds of the intermediate heat exchangers

Materials Informatics: Statistical Modeling in Material Science.

Science.gov (United States)

Yosipof, Abraham; Shimanovich, Klimentiy; Senderowitz, Hanoch

2016-12-01

Material informatics is engaged with the application of informatic principles to materials science in order to assist in the discovery and development of new materials. Central to the field is the application of data mining techniques and in particular machine learning approaches, often referred to as Quantitative Structure Activity Relationship (QSAR) modeling, to derive predictive models for a variety of materials-related "activities". Such models can accelerate the development of new materials with favorable properties and provide insight into the factors governing these properties. Here we provide a comparison between medicinal chemistry/drug design and materials-related QSAR modeling and highlight the importance of developing new, materials-specific descriptors. We survey some of the most recent QSAR models developed in materials science with focus on energetic materials and on solar cells. Finally we present new examples of material-informatic analyses of solar cells libraries produced from metal oxides using combinatorial material synthesis. Different analyses lead to interesting physical insights as well as to the design of new cells with potentially improved photovoltaic parameters. © 2016 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Engineering of automated assembly of beam-shaping optics

Science.gov (United States)

Haag, Sebastian; Sinhoff, Volker; Müller, Tobias; Brecher, Christian

2014-03-01

Beam-shaping is essential for any kind of laser application. Assembly technologies for beam-shaping subassemblies are subject to intense research and development activities and their technical feasibility has been proven in recent years while economic viability requires more efficient engineering tools for process planning and production ramp up of complex assembly tasks for micro-optical systems. The work presented in this paper aims for significant reduction of process development and production ramp up times for the automated assembly of micro-optical subassemblies for beam-collimation and beam-tilting. The approach proposed bridges the gap between the product development phase and the realization of automation control through integration of established software tools such as optics simulation and CAD modeling as well as through introduction of novel software tools and methods to efficiently describe active alignment strategies. The focus of the paper is put on the methodological approach regarding the engineering of assembly processes for beam-shaping micro-optics and the formal representation of assembly objectives similar to representation in mechanical assemblies. Main topic of the paper is the engineering methodology for active alignment processes based on the classification of optical functions for beam-shaping optics and corresponding standardized measurement setups including adaptable alignment algorithms. The concepts are applied to industrial use-cases: (1) integrated collimation module for fast- and slow-axis and (2) beam-tilting subassembly consisting of a fast-axis collimator and micro-lens array. The paper concludes with an overview of current limitations as well as an outlook on the next development steps considering adhesive bonding processes.

An assessment of methods of calculating Doppler effects in plutonium fuelled sodium cooled fast reactors

International Nuclear Information System (INIS)

Butland, A.T.D.; Reddell, G.

1979-01-01

After a survey of the requirements, an assessment of UK methods and data is made on the basis of the following work. First, the analysis of the SEFOR Doppler experiments, carried out using the UK FGL5 fine group nuclear data library, the MURAL cell code and whole reactor diffusion theory calculations of the neutron flux. Second, the analysis of some Japanese FCA central sample perturbation measurements of structural material Doppler effects. Third, an assessment of the accuracy of Doppler predictions in a sodium voided core using results from Zebra 5 and BIZET, and theoretical studies of additional effects relevant to power reactors and not covered by the above analyses, including the following, the calculation of Doppler effects at high temperature, fuel cycle and burn-up effects, and the heterogeneity effects of large fuelled subassemblies in pin geometry. The importance of crystalline binding effects in the fuel are discussed as is the importance of reactor material boundaries in the calculation of resonance shielding effects. Some suggestions for further Doppler studies are made. (U.K.)

Gas storage materials, including hydrogen storage materials

Science.gov (United States)

Mohtadi, Rana F; Wicks, George G; Heung, Leung K; Nakamura, Kenji

2013-02-19

A material for the storage and release of gases comprises a plurality of hollow elements, each hollow element comprising a porous wall enclosing an interior cavity, the interior cavity including structures of a solid-state storage material. In particular examples, the storage material is a hydrogen storage material such as a solid state hydride. An improved method for forming such materials includes the solution diffusion of a storage material solution through a porous wall of a hollow element into an interior cavity.

Journal of Biosciences | Indian Academy of Sciences

Indian Academy of Sciences (India)

We have calculated the most probable mass distributions for Lumbricus and Riftia assemblies and their globin and linker subassemblies, based on the Lumbricus Er stoichiometry and using accurate subunit masses obtained by electrospray ionization mass spectrometry. The expected masses of Lumbricus and Riftia Ers ...

Philips Electronics synchronizes its supply chain to end the bullwhip effect

NARCIS (Netherlands)

Kok, de A.G.; Janssen, F.B.S.L.P.; van Doremalen, J.B.M.; Wachem, van E.; Clerkx, M.J.R.; Peeters, W.

2005-01-01

Demand variability increases as one moves up a supply chain. The demand for finished products is less variable than for subassemblies, which is less variable than for individual components. This phenomenon is known as the bullwhip or Forrester effect. It increases inventory unnecessarily and makes

Materials, critical materials and clean-energy technologies

Science.gov (United States)

Eggert, R.

2017-07-01

Modern engineered materials, components and systems depend on raw materials whose properties provide essential functionality to these technologies. Some of these raw materials are subject to supply-chain risks, and such materials are known as critical materials. This paper reviews corporate, national and world perspectives on material criticality. It then narrows its focus to studies that assess "what is critical" to clean-energy technologies. The focus on supply-chain risks is not meant to be alarmist but rather to encourage attention to monitoring these risks and pursuing technological innovation to mitigate the risks.

Parametric study for the fire safety design of steel structures

DEFF Research Database (Denmark)

Aiuti, Riccardo; Giuliani, Luisa

2013-01-01

the considered time of fire exposure. A deeper knowledge on the failure mode of steel structure is however important in order to ensure the safety of the people and properties outside the building. Aim of this paper is to analyze the behaviour of single elements, sub-assemblies and frames exposed to fire...... or hindered thermal expansion induced on the element by the rest of the structure. Nevertheless, restrained thermal expansion is known to significantly affect the behaviour of steel structures in fire, and the compliance with a prescribed resistance class doesn’t ensure the integrity of the building after...... and find out the basic collapse mechanisms of structural elements in fire conditions, considering the rest of the construction with appropriate constraints. The analysis is carried out taking into account material and geometrical nonlinearities as well as the degradation of steel properties at high...

Tower Shielding Reactor II design and operation report. Vol. 3. Assembling and testing of the control mechanism assembly

International Nuclear Information System (INIS)

Ward, D.R.; Holland, L.B.

1979-09-01

The mechanisms that are operated to control the reactivity of the Tower Shielding Reactor II(TSR-II) are mounted on a Control Mechanism Housing (CMH) that is centered inside the reactor core. The information required to procure, fabricate, inspect, and assemble a CMH is contained in the ORNL engineering drawings listed in the appropriate sections. The components are fabricated and inspected from these drawings in accordance with a Quality Assurance Plan and a Manufacturing Plan. The material in this report describes the acceptance and performance tests of CMH subassemblies used ty the Tower Shielding Facility (TSF) staff but it can also be used by personnel fabricating the components. This information which was developed and used before the advent of the formalized QA Program and Manufacturing Plans evolved during the fabrication and testing of the first five CMHs

Material Science

Energy Technology Data Exchange (ETDEWEB)

Won, Dong Yeon; Kim, Heung

1987-08-15

This book introduces material science, which includes key of a high-tech industry, new materials of dream like new metal material and semiconductor, classification of materials, microstructure of materials and characteristic. It mentions magic new materials such as shape memory alloy, fine ceramics, engineering fine ceramics, electronic ceramics, engineering plastic, glass, silicone conductor, optical fiber mixed materials and integrated circuit, challenge for new material and development of new materials.

Material Science

International Nuclear Information System (INIS)

Won, Dong Yeon; Kim, Heung

1987-08-01

This book introduces material science, which includes key of a high-tech industry, new materials of dream like new metal material and semiconductor, classification of materials, microstructure of materials and characteristic. It mentions magic new materials such as shape memory alloy, fine ceramics, engineering fine ceramics, electronic ceramics, engineering plastic, glass, silicone conductor, optical fiber mixed materials and integrated circuit, challenge for new material and development of new materials.

Materials, critical materials and clean-energy technologies

Directory of Open Access Journals (Sweden)

Eggert R.

2017-01-01

Full Text Available Modern engineered materials, components and systems depend on raw materials whose properties provide essential functionality to these technologies. Some of these raw materials are subject to supply-chain risks, and such materials are known as critical materials. This paper reviews corporate, national and world perspectives on material criticality. It then narrows its focus to studies that assess “what is critical” to clean-energy technologies. The focus on supply-chain risks is not meant to be alarmist but rather to encourage attention to monitoring these risks and pursuing technological innovation to mitigate the risks.

Development of a 3D cell-centered Lagrangian scheme for the numerical modeling of the gas dynamics and hyper-elasticity systems

International Nuclear Information System (INIS)

Georges, Gabriel

2016-01-01

High Energy Density Physics (HEDP) flows are multi-material flows characterized by strong shock waves and large changes in the domain shape due to rare faction waves. Numerical schemes based on the Lagrangian formalism are good candidates to model this kind of flows since the computational grid follows the fluid motion. This provides accurate results around the shocks as well as a natural tracking of multi-material interfaces and free-surfaces. In particular, cell-centered Finite Volume Lagrangian schemes such as GLACE (Godunov-type Lagrangian scheme Conservative for total Energy) and EUCCLHYD (Explicit Unstructured Cell-Centered Lagrangian Hydrodynamics) provide good results on both the modeling of gas dynamics and elastic-plastic equations. The work produced during this PhD thesis is in continuity with the work of Maire and Nkonga [JCP, 2009] for the hydrodynamic part and the work of Kluth and Despres [JCP, 2010] for the hyper elasticity part. More precisely, the aim of this thesis is to develop robust and accurate methods for the 3D extension of the EUCCLHYD scheme with a second-order extension based on MUSCL (Monotonic Upstream-centered Scheme for Conservation Laws) and GRP (Generalized Riemann Problem) procedures. A particular care is taken on the preservation of symmetries and the monotonicity of the solutions. The scheme robustness and accuracy are assessed on numerous Lagrangian test cases for which the 3D extensions are very challenging. (author) [fr

Fatigue Analysis of an Outer Bearing Bush of a Kaplan Turbine

Directory of Open Access Journals (Sweden)

Doina Frunzaverde

2011-01-01

Full Text Available The paper presents the fatigue analysis of an outer bearing bush of aKaplan turbine. This outer bush, together with an inner one, bear thepin lever - trunion - blade subassembly of the runner blade operatingmechanism. For modeling and simulation, SolidWorks software is used.

The behaviour of Phenix fuel pin bundle under irradiation

International Nuclear Information System (INIS)

Marbach, G.; Millet, P.; Blanchard, P.; Huillery, R.

1979-07-01

An entire Phenix sub-assembly has been mounted and sectioned after irradiation. The examination of cross-sections revealed the effects of mechanical interaction in the bundle (ovalisations and contacts between clads). According to analysis of the sodium channels, cooling of the pin bundle remained uniform. (author)

Variable-speed, portable routing skate

Science.gov (United States)

Pesch, W. A.

1967-01-01

Lightweight, portable, variable-speed routing skate is used on heavy metal subassemblies which are impractical to move to a stationary machine. The assembly, consisting of the housing with rollers, router, and driving mechanism with transmission, weighs about forty pounds. Both speed and depth of cut are adjustable.

26 CFR 1.460-1 - Long-term contracts.

Science.gov (United States)

2010-04-01

... the manufacture of personal property is a manufacturing contract. In contrast, a contract for the... performance of engineering and design services, and the production of components and subassemblies that are..., enters into a single long-term contract to design and manufacture a satellite and to develop computer...

Executive summary: Mod-1 wind turbine generator analysis and design report

Science.gov (United States)

1979-01-01

Activities leading to the detail design of a wind turbine generator having a nominal rating of 1.8 megawatts are reported. Topics covered include (1) system description; (2) structural dynamics; (3) stability analysis; (4) mechanical subassemblies design; (5) power generation subsystem; and (6) control and instrumentation subsystem.

Photorefractive Materials and Their Applications 2 Materials

CERN Document Server

Günter, Peter

2007-01-01

Photorefractive Materials and Their Applications 2: Materials is the second of three volumes within the Springer Series in Optical Sciences. The book gives a comprehensive review of the most important photorefractive materials and discusses the physical properties of organic and inorganic crystals as well as poled polymers. In this volume, photorefractive effects have been investigated at wavelengths covering the UV, visible and near infrared. Researchers in the field and graduate students of solid-state physics and engineering will gain a thorough understanding of the properties of materials in photorefractive applications. The other two volumes are: Photorefractive Materials and Their Applications 1: Basic Effects. Photorefractive Materials and Their Applications 3: Applications.

Monte Carlo criticality source convergence in a loosely coupled fuel storage system

International Nuclear Information System (INIS)

Blomquist, Roger N.; Gelbard, Ely M.

2003-01-01

The fission source convergence of a very loosely coupled array of 36 fuel subassemblies with slightly non-symmetric reflection is studied. The fission source converges very slowly from a uniform guess to the fundamental mode in which about 40% of the fissions occur in one corner subassembly. Eigenvalue and fission source estimates are analyzed using a set of statistical tests similar to those used in MCNP, including the 'drift-in-mean' test and a new drift-in-mean test using a linear fit to the cumulative estimate drift, the Shapiro-Wilk test for normality, the relative error test, and the '1/N' test. The normality test does not detect a drifting eigenvalue or fission source. Applied to eigenvalue estimates, the other tests generally fail to detect an unconverged solution, but they are sometimes effective when evaluating fission source distributions. None of the tests provides completely reliable indication of convergence, although they can detect nonconvergence. (author)

Assessment of void fraction prediction using the RETRAN-3d and CORETRAN-01/VIPRE-02 codes

International Nuclear Information System (INIS)

Aounallah, Y.; Coddington, P.; Gantner, U.

2000-01-01

A review of wide-range void fraction correlations against an extensive database has been undertaken to identify the correlations best suited for nuclear safety applications. Only those based on the drift-flux model have been considered. The survey confirmed the application range of the Chexal-Lellouche correlation, and the database was also used to obtain new parameters for the Inoue drift-flux correlation, which was also found suitable. A void fraction validation study has also been undertaken for the codes RETRAN-3D and CORETRAN-01/VIPRE-02 at the assembly and sub-assembly levels. The study showed the impact of the RETRAN-03 user options on the predicted void fraction, and the RETRAN-3D limitation at very low fluid velocity. At the sub-assembly level, CORETRAN-01/VIPRE-02 substantially underestimates the void in regions with low power-to-flow ratios. Otherwise, a generally good predictive performance was obtained with both RETRAN-3D and CORETRAN-01/VIPRE-02. (authors)

Assessment of void fraction prediction using the RETRAN-3d and CORETRAN-01/VIPRE-02 codes

Energy Technology Data Exchange (ETDEWEB)

Aounallah, Y.; Coddington, P.; Gantner, U

2000-07-01

A review of wide-range void fraction correlations against an extensive database has been undertaken to identify the correlations best suited for nuclear safety applications. Only those based on the drift-flux model have been considered. The survey confirmed the application range of the Chexal-Lellouche correlation, and the database was also used to obtain new parameters for the Inoue drift-flux correlation, which was also found suitable. A void fraction validation study has also been undertaken for the codes RETRAN-3D and CORETRAN-01/VIPRE-02 at the assembly and sub-assembly levels. The study showed the impact of the RETRAN-03 user options on the predicted void fraction, and the RETRAN-3D limitation at very low fluid velocity. At the sub-assembly level, CORETRAN-01/VIPRE-02 substantially underestimates the void in regions with low power-to-flow ratios. Otherwise, a generally good predictive performance was obtained with both RETRAN-3D and CORETRAN-01/VIPRE-02. (authors)

Illustrating the disassembly of 3D models

KAUST Repository

Guo, Jianwei

2013-06-11

We present a framework for the automatic disassembly of 3D man-made models and the illustration of the disassembly process. Given an assembled 3D model, we first analyze the individual parts using sharp edge loops and extract the contact faces between each pair of neighboring parts. The contact faces are then used to compute the possible moving directions of each part. We then present a simple algorithm for clustering the sets of the individual parts into meaningful sub-assemblies, which can be used for a hierarchical decomposition. We take the stability of sub-assemblies into account during the decomposition process by considering the upright orientation of the input models. Our framework also provides a user-friendly interface to enable the superimposition of the constraints for the decomposition. Finally, we visualize the disassembly process by generating an animated sequence. The experiments demonstrate that our framework works well for a variety of complex models. © 2013 Elsevier Ltd.

CDF central detector installation. An efficient merge of digital photogrammetry and laser tracker metrology

International Nuclear Information System (INIS)

Greenwood, John A.; Wojcik, George J.

2003-01-01

Metrology for Run II at the Collider Detector at Fermilab (CDF) required a very complex geodetic survey. The Collision Hall network, surveyed with a Laser Tracker and digital level, provides a constraining network for the positioning of the Central Detector (CD). A part-based Laser Tracker network, which surrounded the 2,000-ton CD, was used as control for assembly. Subassembly surveys of the Detector's major components were measured as independent networks using Laser Tracker, V-STARS/S (Video-Simultaneous Triangulation And Resection System/Single camera) digital photogrammetry system, and BETS (Brunson Electronic Theodolite System) theodolite triangulation system. Each subassembly survey was transformed into and constrained by the part-based network. For roll-in, the CD part-based network was transformed into the Collision Hall network coordinate system. The CD was positioned in the Collision Hall using two Laser Trackers in 'stakeout mode.' This paper discusses the survey, adjustment, transformation, and precision of the various networks. (author)

Validation of detailed thermal hydraulic models used for LMR safety and for improvement of technical specifications

Energy Technology Data Exchange (ETDEWEB)

Dunn, F.E.

1995-12-31

Detailed steady-state and transient coolant temperatures and flow rates from an operating reactor have been used to validate the multiple pin model in the SASSYS-1 liquid metal reactor systems analysis code. This multiple pin capability can be used for explicit calculations of axial and lateral temperature distributions within individual subassemblies. Thermocouples at a number of axial locations and in a number of different coolant sub-channels m the XXO9 instrumented subassembly in the EBR-II reactor provided temperature data from the Shutdown Heat Removal Test (SHRT) series. Flow meter data for XXO9 and for the overall system are also available from these tests. Results of consistent SASSYS-1 multiple pin analyses for both the SHRT-45 loss-of-flow-without-scram-test and the S14RT-17 protected loss-of-flow test agree well with the experimental data, providing validation of the SASSYS-1 code over a wide range of conditions.

Method for the determination of technical specifications limiting temperature in EBR-II operation

International Nuclear Information System (INIS)

Chang, L.K.; Hill, D.J.; Ku, J.Y.

1994-01-01

The methodology and analysis procedure to qualify the Mark-V and Mark-VA fuels for the Experimental Breeder Reactor II are summarized in this paper. Fuel performance data and design safety criteria are essential for thermal-hydraulic analysis and safety evaluations. Normal and off-normal operation duty cycles and transient classifications are required for the safety assessment of the fuels. The temperature limits of subassemblies were first determined by a steady-state thermal-structural and fuel damage analysis, in which a trial-and-error approach was used to predict the maximum allowable fuel pin temperature that satisfies the design criteria for steady-state normal operation. The steady-state temperature limits were used as the basis of the off-normal transient analysis to assess the safety performance of the fuel for anticipated, unlikely and extremely unlikely events. If the design criteria for the off-normal events are not satisfied, then the subassembly temperature limit is reduced and an iterative procedure is employed until all design criteria are met

A conceptual redesign of an Inter-Building Fuel Transfer Cask

International Nuclear Information System (INIS)

Klann, R.T.; Picker, B.A. Jr.

1993-01-01

The Inter-Building Fuel Transfer Cask, referred to as the IBC, is a lead shielded cask for transporting subassemblies between buildings on the Argonne National Laboratory-West site near Idaho Falls, Idaho. The cask transports both newly fabricated and spent reactor subassemblies between the Experimental Breeder Reactor-II (EBR-II), the Fuel Cycle Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The IBC will play a key role in the Integral Fast Reactor (IFR) fuel recycling demonstration project. This report discusses a conceptual redesign of the IBC which has been performed. The objective of the conceptual design was to increase the passive heat removal capabilities, reduce the personnel radiation exposure and incorporate enhanced safety features into the design. The heat transfer, radiation and thermal-hydraulic properties of the IBC were analytically modelled to determine the principal factors controlling the desip. The scoping studies that were performed determined the vital physical characteristics (i.e., size, shielding, pumps, etc.) of the MC conceptual design

Establishment and validation of the model of molten pool in fast reactor

International Nuclear Information System (INIS)

Zhou Shufeng; Luo Rui; Wang Zhou; Shi Xiaobo; Yang Xianyong

2007-01-01

Running under the beyond design base accidental condition, sodium boiling and dry-out will soon be brought about in LMFBR. If not stopped timely, the fuel pins of the subassembly will be melt and broken to form a molten pool at the bottom of the subassembly. to present a reasonable analysis about the molten pool accident, a method of establishing model according to the mechanism is selected, by which an integral model of the molten pool is established. Validated on the three power groups of BF1 experiments which belong to the France SCARABEE series experimenters, the model shows good results. After compared with the models of GEYSER and BF2 experiments which had been validated before, some conclusions about mechanism of molten pool are derived. Moreover, through comparing the relative parameters such as the discharged heat and the increment of temperature etc., a reasonable analysis about the type of heat transfer is present, on the basis of which some conclusions are derived as well. (authors)

Materials Characterization and Microelectronic Implementation of Metal-insulator Transition Materials and Phase Change Materials

Science.gov (United States)

2015-03-26

materials like crystalline semiconductors, graphene , and composites, the materials discussed here could have a significant impact. This thesis investigates...diagnosis [124], crystallinity of pharmaceutical materials [125], materials diagnosis for restoration of paintings [126], and materials research [127...temperature dots and paint were placed on samples on the substrate. Temperature dots are typically used in the transportation of goods such as food in order

Planning for closure and deactivation of the EBR-II complex

International Nuclear Information System (INIS)

Michelbacher, J.A.; Henslee, S.P.; Poland, H.F.; Wells, P.B.

1997-01-01

In January 1994, DOE terminated the Integral Fast Reactor (IFR) Program. Argonne National Laboratory-West (ANL-W) prepared a detailed plan to put Experimental Breeder Reactor-II (EBR-II) in a safe condition, including removal of irradiated fueled subassemblies from the plant, transfer of subassemblies, and removal and stabilization of primary and secondary sodium liquid heat transfer metal. The goal of deactivation is to stabilize the EBR-II complex until decontamination and decommissioning (D ampersand D) is implemented, thereby minimizing maintenance and surveillance. Deactivation of a sodium cooled reactor presents unique concerns. Residual sodium in the primary and secondary systems must be either reacted or inerted to preclude concerns with explosive sodium-air reactions. Also, residual sodium on components will effectively solder these items in place, making removal unfeasible. Several special cases reside in the primary system, including primary cold traps, a cesium trap, a cover gas condenser, and systems containing sodium-potassium alloy. The sodium or sodium-potassium alloy in these components must be reacted in place or the components removed. The Sodium Components Maintenance Shop at ANL-W provides the capability for washing primary components, removing residual quantities of sodium while providing some decontamination capacity. Considerations need to be given to component removal necessary for providing access to primary tank internals for D ampersand D activities, removal of hazardous materials, and removal of stored energy sources. ANL-W's plan for the deactivation of EBR-II addresses these issues, providing for an industrially and radiologically safe complex, requiring minimal surveillance during the interim period between deactivation and D ampersand D. Throughout the deactivation and closure of the EBR-II complex, federal environmental concerns will be addressed, including obtaining the proper permits for facility condition and waste processing

Reference material systems: a sourcebook for material assessment

Energy Technology Data Exchange (ETDEWEB)

Bhagat, N. (ed.)

1976-12-01

A reference set of data related to material systems and a framework for carrying out the material technologies assessment are presented. While the bulk of renewables have been considered in this report, the nonrenewable materials dealt with here include structural materials only, such as steel, aluminum, cement and concrete, and bricks. The complete data set is supposed to include material flows, energy requirements, capital and labor inputs, and environmental effects for each process that a resource must go through to become a useful material for an end use. Although effort has been made to obtain as much information as possible, considerable gaps in data, apparent throughout this report, could not be avoided. A new material technology can be evaluated by substituting that technology for appropriate elements of the reference materials system and calculating the net change in material resource, energy, capital and labor requirements, and environmental impacts. This combination of information thus serves as a means of evaluating the potential benefits to be gained by research in various material technologies.