Analysis of the uncertainties in the physical calculations of water-moderated power reactors of the VVER type by the parameters of models of preparing few-group constants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bryukhin, V. V., E-mail: bryuhin@yandex.ru; Kurakin, K. Yu.; Uvakin, M. A.

The article covers the uncertainty analysis of the physical calculations of the VVER reactor core for different meshes of the reference values of the feedback parameters (FBP). Various numbers of nodes of the parametric axes of FBPs and different ranges between them are investigated. The uncertainties of the dynamic calculations are analyzed using RTS RCCA ejection as an example within the framework of the model with the boundary conditions at the core inlet and outlet.

3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langenbuch, S.; Velkov, K.; Lizorkin, M.

1997-07-01

This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

Safety system augmentation at Russian nuclear power plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scerbo, J.A.; Satpute, S.N.; Donkin, J.Y.

1996-12-31

This paper describes the design and procurement of a Class IE DC power supply system to upgrade plant safety at the Kola Nuclear Power Plant (NPP). Kola NPP is located above the Arctic circle at Polyarnie Zorie, Murmansk, Russia. Kola NPP consists of four units. Units 1 and 2 have VVER-440/230 type reactors: Units 3 and 4 have VVER-440/213 type reactors. The VVER-440 reactor design is similar to the pressurized water reactor design used in the US. This project provided redundant, Class 1E DC station batteries and DC switchboards for Kola NPP, Units 1 and 2. The new DC powermore » supply system was designed and procured in compliance with current nuclear design practices and requirements. Technical issues that needed to be addressed included reconciling the requirements in both US and Russian codes and satisfying the requirements of the Russian nuclear regulatory authority. Close interface with ATOMENERGOPROEKT (AEP), the Russian design organization, KOLA NPP plant personnel, and GOSATOMNADZOR (GAN), the Russian version of US Nuclear Regulatory Commission, was necessary to develop a design that would assure compliance with current Russian design requirements. Hence, this project was expected to serve as an example for plant upgrades at other similar VVER-440 nuclear plants. In addition to technical issues, the project needed to address language barriers and the logistics of shipping equipment to a remote section of the Former Soviet Union (FSU). This project was executed by Burns and Roe under the sponsorship of the US DOE as part of the International Safety Program (INSP). The INSP is a comprehensive effort, in cooperation with partners in other countries, to improve nuclear safety worldwide. A major element within the INSP is the improvement of the safety of Soviet-designed nuclear reactors.« less

Calculation with MCNP of capture photon flux in VVER-1000 experimental reactor.

PubMed

Töre, Candan; Ortego, Pedro

2005-01-01

The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface and 1/4 thickness but a good fitting for deeper points in vessel.

Temperature measuring analysis of the nuclear reactor fuel assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Urban, F., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Kučák, L., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Bereznai, J., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk

2014-08-06

Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuelmore » assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.« less

Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

NASA Astrophysics Data System (ADS)

Shchelik, S. V.; Pavlov, A. S.

2013-07-01

Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

Engineering Margin Factors Used in the Design of the VVER Fuel Cycles

NASA Astrophysics Data System (ADS)

Lizorkin, M. P.; Shishkov, L. K.

2017-12-01

The article describes methods for determination of the engineering margin factors currently used to estimate the uncertainties of the VVER reactor design parameters calculated via the KASKAD software package developed at the National Research Center Kurchatov Institute. These margin factors ensure the meeting of the operating (design) limits and a number of other restrictions under normal operating conditions.

Test case for VVER-1000 complex modeling using MCU and ATHLET

NASA Astrophysics Data System (ADS)

Bahdanovich, R. B.; Bogdanova, E. V.; Gamtsemlidze, I. D.; Nikonov, S. P.; Tikhomirov, G. V.

2017-01-01

The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.

Extension of the Bgl Broad Group Cross Section Library

NASA Astrophysics Data System (ADS)

Kirilova, Desislava; Belousov, Sergey; Ilieva, Krassimira

2009-08-01

The broad group cross-section libraries BUGLE and BGL are applied for reactor shielding calculation using the DOORS package based on discrete ordinates method and multigroup approximation of the neutron cross-sections. BUGLE and BGL libraries are problem oriented for PWR or VVER type of reactors respectively. They had been generated by collapsing the problem independent fine group library VITAMIN-B6 applying PWR and VVER one-dimensional radial model of the reactor middle plane using the SCALE software package. The surveillance assemblies (SA) of VVER-1000/320 are located on the baffle above the reactor core upper edge in a region where geometry and materials differ from those of the middle plane and the neutron field gradient is very high which would result in a different neutron spectrum. That is why the application of the fore-mentioned libraries for the neutron fluence calculation in the region of SA could lead to an additional inaccuracy. This was the main reason to study the necessity for an extension of the BGL library with cross-sections appropriate for the SA region. Comparative analysis of the neutron spectra of the SA region calculated by the VITAMIN-B6 and BGL libraries using the two-dimensional code DORT have been done with purpose to evaluate the BGL applicability for SA calculation.

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

NASA Astrophysics Data System (ADS)

Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

2017-09-01

Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Precise calculation of neutron-capture reactions contribution in energy release for different types of VVER-1000 fuel assemblies

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Bahdanovich, Rynat; Pham, Phu

2017-09-01

Precise calculation of energy release in a nuclear reactor is necessary to obtain the correct spatial power distribution and predict characteristics of burned nuclear fuel. In this work, previously developed method for calculation neutron-capture reactions - capture component - contribution in effective energy release in a fuel core of nuclear reactor is discussed. The method was improved and implemented to the different models of VVER-1000 reactor developed for MCU 5 and MCNP 4 computer codes. Different models of equivalent cell and fuel assembly in the beginning of fuel cycle were calculated. These models differ by the geometry, fuel enrichment and presence of burnable absorbers. It is shown, that capture component depends on fuel enrichment and presence of burnable absorbers. Its value varies for different types of hot fuel assemblies from 3.35% to 3.85% of effective energy release. Average capture component contribution in effective energy release for typical serial fresh fuel of VVER-1000 is 3.5%, which is 7 MeV/fission. The method will be used in future to estimate the dependency of capture energy on fuel density, burn-up, etc.

Development of data base with mechanical properties of un- and pre-irradiated VVER cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Asmolov, V.; Yegorova, L.; Kaplar, E.

1998-03-01

Analysis of recent RIA test with PWR and VVER high burnup fuel, performed at CABRI, NSRR, IGR reactors has shown that the data base with mechanical properties of the preirradiated cladding is necessary to interpret the obtained results. During 1997 the corresponding cycle of investigations for VVER clad material was performed by specialists of NSI RRC KI and RIAR in cooperation with NRC (USA), IPSN (France) in two directions: measurements of mechanical properties of Zr-1%Nb preirradiated cladding versus temperature and strain rate; measurements of failure parameters for gas pressurized cladding tubes. Preliminary results of these investigations are presented in thismore » paper.« less

On the interpretation of the inverted kinetics equation and space-time calculations of the effectiveness of the VVER-1000 reactor scram system

NASA Astrophysics Data System (ADS)

Zizin, M. N.; Ivanov, L. D.

2013-12-01

In the present paper, an attempt is made to analyze the accuracy of calculating the effectiveness of the VVER-1000 reactor scram system by means of the inverted solution of the kinetics equation (ISKE). In the numerical studies in the intellectual ShIPR software system, the actuation of the reactor scram system with the possible jamming of one of the two most effective rods is simulated. First, the connection of functionals calculated in the space-time computation in different approximations with the kinetics equation is considered on the theoretical level. The formulas are presented in a manner facilitating their coding. Then, the results of processing of several such functions by the ISKE are presented. For estimating the effectiveness of the VVER-1000 reactor scram system, it is proposed to use the measured currents of ionization chambers (IC) jointly with calculated readings of IC imitators. In addition, the integral of the delayed neutron (DN) generation rate multiplied by the adjoint DN source over the volume of the reactor, calculated for the instant of time when insertion of safety rods ends, is used. This integral is necessary for taking into account the spatial reactivity effects. Reasonable agreement was attained for the considered example between the effectiveness of the scram system evaluated by this method and the values obtained by steady-state calculations as the difference of the reciprocal effective multiplication factors with withdrawn and inserted control rods. This agreement was attained with the use of eight-group DN parameters.

Study of PRIMAVERA steel samples by a positron annihilation spectroscopy technique

NASA Astrophysics Data System (ADS)

Grafutin, V.; Ilyukhina, O.; Krsjak, V.; Burcl, R.; Hähner, P.; Erak, D.; Zeman, A.

2010-11-01

In the present article, a positron annihilation spectroscopy investigation of VVER-440/230 weld materials is discussed. Important characteristics of metals such as Fermi energy, concentration of electrons in the conduction band, size and concentration of defects were experimentally determined for three model materials with higher level of copper (0.16 wt.%) and phosphorus (0.027-0.038 wt.%). The impact of neutron irradiation and subsequent annealing on crystal lattice parameters was investigated. The experiments with the angular correlation of positron annihilation radiation (ACAR) complement the published positron annihilation spectroscopy (PAS) studies of the radiation treated VVER materials as well as previous experiments on PRIMAVERA materials. The availability of the experimental reactor to prepare strong 64Cu positron sources provided for unique experimental conditions, such as good resolution of spectra (0.4 mrad) and reasonable short time of measurement (36 h). The present paper aims to contribute to further understanding of RPV (reactor pressure vessel) steels behaviour under irradiation conditions as well as annealing recovery procedures, which have already been applied at several VVER NPP units in Europe.

Methodological studies on the VVER-440 control assembly calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hordosy, G.; Kereszturi, A.; Maraczy, C.

1995-12-31

The control assembly regions of VVER-440 reactors are represented by 2-group albedo matrices in the global calculations of the KARATE code system. Some methodological aspects of calculating albedo matrices with the COLA transport code are presented. Illustrations are given how these matrices depend on the relevant parameters describing the boron steel and steel regions of the control assemblies. The calculation of the response matrix for a node consisting of two parts filled with different materials is discussed.

Embrittlement of low copper VVER 440 surveillance samples neutron-irradiated to high fluences

NASA Astrophysics Data System (ADS)

Miller, M. K.; Russell, K. F.; Kocik, J.; Keilova, E.

2000-11-01

An atom probe tomography microstructural characterization of low copper (0.06 at.% Cu) surveillance samples from a VVER 440 reactor has revealed manganese and silicon segregation to dislocations and other ultrafine features in neutron-irradiated base and weld materials (fluences 1×10 25 m-2 and 5×10 24 m-2, E>0.5 MeV, respectively). The results indicate that there is an additional mechanism of embrittlement during neutron irradiation that manifests itself at high fluences.

Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

2010-10-01

The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

Almost twenty years' search of transuranium isotopes in effluents discharged to air from nuclear power plants with VVER reactors.

PubMed

Hölgye, Z; Filgas, R

2006-04-01

Airborne effluents of 5 stacks (stacks 1-5) of three nuclear power plants, with 9 pressurized water reactors VVER of 4,520 MWe total power, were searched for transuranium isotopes in different time periods. The search started in 1985. The subject of this work is a presentation of discharge data for the period of 1998-2003 and a final evaluation. It was found that 238Pu, 239,240Pu, 241Am, 242Cm, and 244Cm can be present in airborne effluents. Transuranium isotope contents in most of the quarterly effluent samples from stacks 2, 4 and 5 were not measurable. Transuranium isotopes were present in the effluents from stack l during all 9 years of the study and from stack 3 since the 3rd quarter of 1996 as a result of a defect in the fuel cladding. A relatively high increase of transuranium isotopes in effluents from stack 3 occurred in the 3rd quarter of 1999, and a smaller increase occurred in the 3rd quarter of 2003. In each instance 242Cm prevailed in the transuranium isotope mixtures. 238Pu/239,240Pu, 241Am/239,240Pu, 242Cm/239,240Pu, and 244Cm/239,240Pu ratios in fuel for different burn-up were calculated, and comparison of these ratios in fuel and effluents was performed.

Post-Test Analysis of 11% Break at PSB-VVER Experimental Facility using Cathare 2 Code

NASA Astrophysics Data System (ADS)

Sabotinov, Luben; Chevrier, Patrick

The best estimate French thermal-hydraulic computer code CATHARE 2 Version 2.5_1 was used for post-test analysis of the experiment “11% upper plenum break”, conducted at the large-scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. A computer model has been developed for CATHARE 2 V2.5_1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separated loops, pressurizer, horizontal multitube steam generators, break section. The secondary side is represented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses. The comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as pressures, temperatures, void fractions, loop seal clearance, etc. The experimental and calculation results are very sensitive regarding the fuel cladding temperature, which show a periodical nature. With the applied CATHARE 1D modeling, the global thermal-hydraulic parameters and the core heat up have been reasonably predicted.

On the equilibrium isotopic composition of the thorium-uranium-plutonium fuel cycle

NASA Astrophysics Data System (ADS)

Marshalkin, V. Ye.; Povyshev, V. M.

2016-12-01

The equilibrium isotopic compositions and the times to equilibrium in the process of thorium-uranium-plutonium oxide fuel recycling in VVER-type reactors using heavy water mixed with light water are estimated. It is demonstrated thEhfat such reactors have a capacity to operate with self-reproduction of active isotopes in the equilibrium mode.

On the equilibrium isotopic composition of the thorium–uranium–plutonium fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshalkin, V. Ye., E-mail: marshalkin@vniief.ru; Povyshev, V. M.

2016-12-15

The equilibrium isotopic compositions and the times to equilibrium in the process of thorium–uranium–plutonium oxide fuel recycling in VVER-type reactors using heavy water mixed with light water are estimated. It is demonstrated thEhfat such reactors have a capacity to operate with self-reproduction of active isotopes in the equilibrium mode.

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

NASA Astrophysics Data System (ADS)

Nevinitsa, V. A.; Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N.; Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu.

2015-12-01

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing 233U from 232Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

The basic features of a closed fuel cycle without fast reactors

NASA Astrophysics Data System (ADS)

Bobrov, E. A.; Alekseev, P. N.; Teplov, P. S.

2017-01-01

In this paper the basic features of a closed fuel cycle with thermal reactors are considered. The three variants of multiple Pu and U recycling in VVER reactors was investigated. The comparison of MOX and REMIX fuel approaches for closed fuel cycle with thermal reactors is presented. All variants make possible to recycle several times the total amount of Pu and U obtained from spent fuel. The reported study was funded by RFBR according to the research project № 16-38-00021

Accumulation of radioactive corrosion products on steel surfaces of VVER-type nuclear reactors. II. 60Co

NASA Astrophysics Data System (ADS)

Varga, Kálmán; Hirschberg, Gábor; Németh, Zoltán; Myburg, Gerrit; Schunk, János; Tilky, Péter

2001-10-01

In the case of intact fuel claddings, the predominant source of radioactivity in the primary circuits of water-cooled nuclear reactors is the activation of corrosion products in the core. The most important corrosion product radionuclides in the primary coolant of pressurized water reactors (PWRs) are 60Co, 58Co, 51Cr, 54Mn, 59Fe (as well as 110mAg in some Soviet-made VVER-type reactor). The second part of this series is focused on the complex studies of the formation and build-up of 60Co-containing species on an austenitic stainless steel type 08X18H10T (GOST 5632-61) and magnetite-covered carbon steel often to be used in Soviet-planned VVERs. The kinetics and mechanism of the cobalt accumulation were studied by a combination (coupling) of an in situ radiotracer method and voltammetry in a model solution of the primary circuit coolant. In addition, independent techniques such as X-ray photoelectron spectroscopic (XPS) and ICP-OES are also used to analyze the chemical state of Co species in the passive layer formed on stainless steel as well as the chemical composition of model solution. The experimental results have revealed that: (i) The passive behavior of the austenitic stainless steel at open-circuit conditions, the slightly alkaline pH and the reducing water chemistry can be considered to be optimal to minimize the 60Co contamination. (ii) The highly potential dependent deposition of various Co-oxides at E>1.10 V (vs. RHE) offers a unique possibility to elaborate a novel electrochemical method for the decrease or removal of cobalt traces from borate-containing coolants contaminated with 60Co and/or 58Co radionuclides.

Reactor Dosimetry State of the Art 2008

NASA Astrophysics Data System (ADS)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G. Williams, A. P. Ribaric and T. Schnauber. Agile high-fidelity MCNP model development techniques for rapid mechanical design iteration / J. A. Kulesza.Extension of Raptor-M3G to r-8-z geometry for use in reactor dosimetry applications / M. A. Hunter, G. Longoni and S. L. Anderson. In vessel exposure distributions evaluated with MCNP5 for Atucha II / J. M. Longhino, H. Blaumann and G. Zamonsky. Atucha I nuclear power plant azimutal ex-vessel flux profile evaluation / J. M. Longhino ... [et al.]. UFTR thermal column characterization and redesign for maximized thermal flux / C. Polit and A. Haghighat. Activation counter using liquid light-guide for dosimetry of neutron burst / M. Hayashi ... [et al.]. Control rod reactivity curves for the annular core research reactor / K. R. DePriest ... [et al.]. Specification of irradiation conditions in VVER-440 surveillance positions / V. Kochkin ... [et al.]. Simulations of Mg-Ar ionisation and TE-TE ionisation chambers with MCNPX in a straightforward gamma and beta irradiation field / S. Nievaart ... [et al.]. The change of austenitic stainless steel elements content in the inner parts of VVER-440 reactor during operation / V. Smutný, J. Hep and P. Novosad. Fast neutron environmental spectrometry using disk activation / G. Lövestam ... [et al.]. Optimization of the neutron activation detector location scheme for VVER-lOOO ex-vessel dosimetry / V. N. Bukanov ... [et al.]. Irradiation conditions for surveillance specimens located into plane containers installed in the WWER-lOOO reactor of unit 2 of the South-Ukrainian NPP / O. V. Grytsenko. V. N. Bukanov and S. M. Pugach. Conformity between LRO mock-ups and VVERS NPP RPV neutron flux attenuation / S. Belousov. Kr. Ilieva and D. Kirilova. FLUOLE: a new relevant experiment for PWR pressure vessel surveillance / D. Beretz ... [et al.]. Transport of neutrons and photons through the iron and water layers / M. J. Kost'ál ... [et al.]. Condition evaluation of spent nuclear fuel assemblies from the first-generation nuclear-powered submarines by gamma scanning / A. F. Usatyi. L. A. Serdyukova and B. S. Stepennov -- Oral session 3: Power plant surveillance. Upgraded neutron dosimetry procedure for VVER-440 surveillance specimens / V. Kochkin ... [et al.]. Neutron dosimetry on the full-core first generation VVER-440 aimed to reactor support structure load evaluation / P. Borodkin ... [et al.]. Ex-vessel neutron dosimetry programs for PWRs in Korea / C. S. Yoo. B. C. Kim and C. C. Kim. Comparison of irradiation conditions of VVER-1000 reactor pressure vessel and surveillance specimens for various core loadings / V. N. Bukanov ... [et al.]. Re-evaluation of dosimetry in the new surveillance program for the Loviisa 1 VVER-440 reactor / T. Serén -- Oral session 4: Benchmarks, intercomparisons and adjustment methods. Determination of the neutron parameter's uncertainties using the stochastic methods of uncertainty propagation and analysis / G. Grégoire ... [et al.].Covariance matrices for calculated neutron spectra and measured dosimeter responses / J. G. Williams ... [et al.]. The role of dosimetry at the high flux reactor / S. C. van der Marek ... [et al.]. Calibration of a manganese bath relative to Cf-252 nu-bar / D. M. Gilliam, A. T. Yue and M. Scott Dewey. Major upgrade of the reactor dosimetry interpretation methodology used at the CEA: general principle / C. Destouches ... [et al.] -- Oral session 5: power plant surveillance. The role of ex-vessel neutron dosimetry in reactor vessel surveillance in South Korea / B.-C. Kim ... [et al.]. Spanish RPV surveillance programmes: lessons learned and current activities / A. Ballesteros and X. Jardí. Atucha I nuclear power plant extended dosimetry and assessment / H. Blaumann ... [et al.]. Monitoring of radiation load of pressure vessels of Russian VVER in compliance with license amendments / G. Borodkin ... [et al.] -- Poster session 2: Test reactors, accelerators and advanced systems; cross sections, nuclear data, damage correlations. Two-dimensional mapping of the calculated fission power for the full-size fuel plate experiment irradiated in the advanced test reactor / G. S. Chang and M. A. Lillo. The radiation safety information computational center: a resource for reactor dosimetry software and nuclear data / B. L. Kirk. Irradiated xenon isotopic ratio measurement for failed fuel detection and location in fast reactor / C. Ito, T. Iguchi and H. Harano. Characterization of dosimetry of the BMRR horizontal thimble tubes and broad beam facility / J.-P. Hu, R. N. Reciniello and N. E. Holden. 2007 nuclear data review / N. E. Holden. Further dosimetry studies at the Rhode Island nuclear science / R. N. Reciniello ... [et al.]. Characterization of neutron fields in the experimental fast reactor Joyo MK-III core / S. Maeda ... [et al.]. Measuring [symbol]Li(n, t) and [symbol]B(n, [symbol]) cross sections using the NIST alpha-gamma apparatus / M. S. Dewey ... [et al.]. Improvement of neutron/gamma field evaluation for restart of JMTR / Y. Nagao ... [et al.]. Monitoring of the irradiated neutron fluence in the neutron transmutation doping process of HANARO / M.-S. Kim and S.-J. Park.Training reactor VR-l neutron spectrum determination / M. Vins, A. Kolros and K. Katovsky. Differential cross sections for gamma-ray production by 14 MeV neutrons on iron and bismuth / V. M. Bondar ... [et al.]. The measurements of the differential elastic neutron cross-sections of carbon for energies from 2 to 133 ke V / O. Gritzay ... [et al.]. Determination of neutron spectrum by the dosimetry foil method up to 35 Me V / S. P. Simakov ... [et al.]. Extension of the BGL broad group cross section library / D. Kirilova, S. Belousov and Kr. Ilieva. Measurements of neutron capture cross-section for tantalum at the neutron filtered beams / O. Gritzayand V. Libman. Measurements of microscopic data at GELINA in support of dosimetry / S. Kopecky ... [et al.]. Nuclide guide and international chart of nuclides - 2008 / T. Golashvili -- Oral session 6: Test reactors, accelerators and advanced systems. Neutronic analyses in support of the HFIR beamline modifications and lifetime extension / I. Remec and E. D. Blakeman. Characterization of neutron test facilities at Sandia National Laboratories / D. W. Vehar ... [et al.]. LYRA irradiation experiments: neutron metrology and dosimetry / B. Acosta and L. Debarberis. Calculated neutron and gamma-ray spectra across the prismatic very high temperature reactor core / J. W. Sterbentz. Enhancement of irradiation capability of the experimental fast reactor joyo / S. Maeda ... [et al.]. Neutron spectrum analyses by foil activation method for high-energy proton beams / C. H. Pyeon ... [et al.] -- Oral session 7: Cross sections, nuclear data, damage correlations. Investigation of new reaction cross-section evaluations in order to update and extend the IRDF-2002 reactor dosimetry library / É. M. Zsolnay, H. J. Nolthenius and A. L. Nichols. A novel approach towards DPA calculations / A. Hogenbirk and D. F. Da Cruz. A new ENDFIB-VII.O based multigroup cross-section library for reactor dosimetry / F. A. Alpan and S. L. Anderson. Activities at the NEA for dosimetry applications / H. Henriksson and I. Kodeli. Validation and verification of covariance data from dosimetry reaction cross-section evaluations / S. Badikov. Status of the neutron cross section standards / A. D. Carlson -- Oral session 8: transport calculations. A dosimetry assessment for the core restraint of an advanced gas cooled reactor / D. A. Thornton ... [et al.]. Neutron dosimetry study in the region of the support structure of a VVER-1000 type reactor / G. Borodkin ... [et al.]. SNS moderator poison design and experiment validation of the moderator performance / W. Lu ... [et al.]. Analysis of OSIRIS in-core surveillance dosimetry for GONDOLE steel irradiation program by using TRIPOLI-4 Monte Carlo code / Y. K. Lee and F. Malouch.Reactor dosimetry applications using RAPTOR-M3G: a new parallel 3-D radiation transport code / G. Longoni and S. L. Anderson.

Breeding of 233U in the thorium-uranium fuel cycle in VVER reactors using heavy water

NASA Astrophysics Data System (ADS)

Marshalkin, V. E.; Povyshev, V. M.

2015-12-01

A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the 233U-232Th oxide fuel of water-moderated reactors with variable water composition (D2O, H2O) that ensures breeding of the 233U and 235U isotopes. The method is comparatively simple to implement.

Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the 233U isotope in the VVER reactors using thorium and heavy water

NASA Astrophysics Data System (ADS)

Marshalkin, V. E.; Povyshev, V. M.

2015-12-01

A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium-uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D2O, H2O) is proposed. The method is characterized by efficient breeding of the 233U isotope and safe reactor operation and is comparatively simple to implement.

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Dudnikov, A. A.; Blandinskiy, V. Yu.

2015-12-15

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

Application of CFX-10 to the Investigation of RPV Coolant Mixing in VVER Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moretti, Fabio; Melideo, Daniele; Terzuoli, Fulvio

2006-07-01

Coolant mixing phenomena occurring in the pressure vessel of a nuclear reactor constitute one of the main objectives of investigation by researchers concerned with nuclear reactor safety. For instance, mixing plays a relevant role in reactivity-induced accidents initiated by de-boration or boron dilution events, followed by transport of a de-borated slug into the vessel of a pressurized water reactor. Another example is constituted by temperature mixing, which may sensitively affect the consequences of a pressurized thermal shock scenario. Predictive analysis of mixing phenomena is strongly improved by the availability of computational tools able to cope with the inherent three-dimensionality ofmore » such problem, like system codes with three-dimensional capabilities, and Computational Fluid Dynamics (CFD) codes. The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed by the ANSYS CFX-10 CFD code. In particular, the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor has been addressed, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. Results have been compared against experimental data from V1000CT-2 Benchmark. Moreover, a boron mixing problem has been investigated, in the hypothesis that a de-borated slug, transported by natural circulation, enters the vessel. Sensitivity analyses have been conducted on some geometrical features, model parameters and boundary conditions. (authors)« less

Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshalkin, V. E., E-mail: marshalkin@vniief.ru; Povyshev, V. M.

A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D{sub 2}O, H{sub 2}O) is proposed. The method is characterized by efficient breeding of the {sup 233}U isotope and safe reactor operation and is comparatively simple to implement.

The current state, main problems and directions in improving water chemistry at NPSs

NASA Astrophysics Data System (ADS)

Tyapkov, V. F.; Sharafutdinov, R. B.

2007-05-01

An analysis of the current state of managing water-chemistry (WC) at Russian nuclear power plants with type-VVER and-RBMK reactors presently in operation is presented. The main directions for improvement of WC are shown.

Breeding of {sup 233}U in the thorium–uranium fuel cycle in VVER reactors using heavy water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshalkin, V. E., E-mail: marshalkin@vniief.ru; Povyshev, V. M.

A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the {sup 233}U–{sup 232}Th oxide fuel of water-moderated reactors with variable water composition (D{sub 2}O, H{sub 2}O) that ensures breeding of the {sup 233}U and {sup 235}U isotopes. The method is comparatively simple to implement.

The virtual digital nuclear power plant: A modern tool for supporting the lifecycle of VVER-based nuclear power units

NASA Astrophysics Data System (ADS)

Arkadov, G. V.; Zhukavin, A. P.; Kroshilin, A. E.; Parshikov, I. A.; Solov'ev, S. L.; Shishov, A. V.

2014-10-01

The article describes the "Virtual Digital VVER-Based Nuclear Power Plant" computerized system comprising a totality of verified initial data (sets of input data for a model intended for describing the behavior of nuclear power plant (NPP) systems in design and emergency modes of their operation) and a unified system of new-generation computation codes intended for carrying out coordinated computation of the variety of physical processes in the reactor core and NPP equipment. Experiments with the demonstration version of the "Virtual Digital VVER-Based NPP" computerized system has shown that it is in principle possible to set up a unified system of computation codes in a common software environment for carrying out interconnected calculations of various physical phenomena at NPPs constructed according to the standard AES-2006 project. With the full-scale version of the "Virtual Digital VVER-Based NPP" computerized system put in operation, the concerned engineering, design, construction, and operating organizations will have access to all necessary information relating to the NPP power unit project throughout its entire lifecycle. The domestically developed commercial-grade software product set to operate as an independently operating application to the project will bring about additional competitive advantages in the modern market of nuclear power technologies.

Absolute determination of power density in the VVER-1000 mock-up on the LR-0 research reactor.

PubMed

Košt'ál, Michal; Švadlenková, Marie; Milčák, Ján

2013-08-01

The work presents a detailed comparison of calculated and experimentally determined net peak areas of selected fission products gamma lines. The fission products were induced during a 2.5 h irradiation on the power level of 9.5 W in selected fuel pins of the VVER-1000 Mock-Up. The calculations were done with deterministic and stochastic (Monte Carlo) methods. The effects of different nuclear data libraries used for calculations are discussed as well. The Net Peak Area (NPA) may be used for the determination of fission density across the mock-up. This fission density is practically identical to power density. Copyright © 2013 Elsevier Ltd. All rights reserved.

Transportation and storage of MOX and LEU assemblies at the Balakovo Nuclear Power Plant

DOT National Transportation Integrated Search

2001-01-01

The VVER-1000-type Balakovo Nuclear Power Plant has been chosen to dispose of the : plutonium created as part of Russian weapons program. The plutonium will be converted to mixed-oxide : (MOX), fabricated into assemblies and loaded into the reactor. ...

Comparison of ENDF/B-VII.1 and JEFF-3.2 in VVER-1000 operational data calculation

NASA Astrophysics Data System (ADS)

Frybort, Jan

2017-09-01

Safe operation of a nuclear reactor requires an extensive calculational support. Operational data are determined by full-core calculations during the design phase of a fuel loading. Loading pattern and design of fuel assemblies are adjusted to meet safety requirements and optimize reactor operation. Nodal diffusion code ANDREA is used for this task in case of Czech VVER-1000 reactors. Nuclear data for this diffusion code are prepared regularly by lattice code HELIOS. These calculations are conducted in 2D on fuel assembly level. There is also possibility to calculate these macroscopic data by Monte-Carlo Serpent code. It can make use of alternative evaluated libraries. All calculations are affected by inherent uncertainties in nuclear data. It is useful to see results of full-core calculations based on two sets of diffusion data obtained by Serpent code calculations with ENDF/B-VII.1 and JEFF-3.2 nuclear data including also decay data library and fission yields data. The comparison is based directly on fuel assembly level macroscopic data and resulting operational data. This study illustrates effect of evaluated nuclear data library on full-core calculations of a large PWR reactor core. The level of difference which results exclusively from nuclear data selection can help to understand the level of inherent uncertainties of such full-core calculations.

Heuristic rules embedded genetic algorithm for in-core fuel management optimization

NASA Astrophysics Data System (ADS)

Alim, Fatih

The objective of this study was to develop a unique methodology and a practical tool for designing loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and BP placement in order to achieve maximum cycle length while satisfying the safety constraints. Genetic Algorithms (GA) have been already used to solve this problem for LP optimization for both PWR and Boiling Water Reactor (BWR). The GA, which is a stochastic method works with a group of solutions and uses random variables to make decisions. Based on the theories of evaluation, the GA involves natural selection and reproduction of the individuals in the population for the next generation. The GA works by creating an initial population, evaluating it, and then improving the population by using the evaluation operators. To solve this optimization problem, a LP optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed in the framework of this thesis. This code is applicable for all types of PWR cores having different geometries and structures with an unlimited number of FA types in the inventory. To reach this goal, an innovative GA is developed by modifying the classical representation of the genotype. To obtain the best result in a shorter time, not only the representation is changed but also the algorithm is changed to use in-core fuel management heuristics rules. The improved GA code was tested to demonstrate and verify the advantages of the new enhancements. The developed methodology is explained in this thesis and preliminary results are shown for the VVER-1000 reactor hexagonal geometry core and the TMI-1 PWR. The improved GA code was tested to verify the advantages of new enhancements. The core physics code used for VVER in this research is Moby-Dick, which was developed to analyze the VVER by SKODA Inc. The SIMULATE-3 code, which is an advanced two-group nodal code, is used to analyze the TMI-1.

Simulation of a small cold-leg-break experiment at the PMK-2 test facility using the RELAP5 and ATHLET codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ezsoel, G.; Guba, A.; Perneczky, L.

Results of a small-break loss-of-coolant accident experiment, conducted on the PMK-2 integral-type test facility are presented. The experiment simulated a 1% break in the cold leg of a VVER-440-type reactor. The main phenomena of the experiment are discussed, and in the case of selected events, a more detailed interpretation with the help of measured void fraction, obtained by a special measurement device, is given. Two thermohydraulic computer codes, RELAP5 and ATHLET, are used for posttest calculations. The aim of these calculations is to investigate the code capability for modeling natural circulation phenomena in VVER-440-type reactors. Therefore, the results of themore » experiment and both calculations are compared. Both codes predict most of the transient events well, with the exception that RELAP5 fails to predict the dryout period in the core. In the experiment, the hot- and cold-leg loop-seal clearing is accompanied by natural circulation instabilities, which can be explained by means of the ATHLET calculation.« less

Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zizin, M. N.; Zimin, V. G.; Zizina, S. N., E-mail: zizin@adis.vver.kiae.ru

2010-12-15

The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit ofmore » the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.« less

Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

NASA Astrophysics Data System (ADS)

Zizin, M. N.; Zimin, V. G.; Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A.

2010-12-01

The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit of the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.

Procedure of recovery of pin-by-pin fields of energy release in the core of VVER-type reactor for the BIPR-8 code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gordienko, P. V., E-mail: gorpavel@vver.kiae.ru; Kotsarev, A. V.; Lizorkin, M. P.

2014-12-15

The procedure of recovery of pin-by-pin energy-release fields for the BIPR-8 code and the algorithm of the BIPR-8 code which is used in nodal computation of the reactor core and on which the recovery of pin-by-pin fields of energy release is based are briefly described. The description and results of the verification using the module of recovery of pin-by-pin energy-release fields and the TVS-M program are given.

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

NASA Astrophysics Data System (ADS)

Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

2014-08-01

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

Experimental study of burnout in channels with twisted fuel rods

NASA Astrophysics Data System (ADS)

Bol'Shakov, V. V.; Bashkirtsev, S. M.; Kobzar', L. L.; Morozov, A. G.

2007-05-01

The results of experimental studies of pressure drop and critical heat flux in the models of fuel assemblies (FAs) with fuel rod simulators twisted relative to the longitudinal axis and a three-ray cross section are considered. The experimental data are compared to the results obtained with the use of techniques adopted for design calculations with fuel rod bundles of type-VVER reactors.

The role of PRA in the safety assessment of VVER Nuclear Power Plants in Ukraine.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kot, C.

1999-05-10

Ukraine operates thirteen (13) Soviet-designed pressurized water reactors, VVERS. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs), in accordance with new SAR content requirements issued in September 1995, by the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine. The requirements are in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The last two requirements, on PRA and BDBA, are new, and the DBA requirements are an expanded version of the older SAR requirements. The US Departmentmore » of Energy (USDOE), as part of its Soviet-Designed Reactor Safety activities, is providing assistance and technology transfer to Ukraine to support their nuclear power plants (NPPs) in developing a Western-type technical basis for the new SARs. USDOE sponsored In-Depth Safety Assessments (ISAs) are in progress at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1, and a follow-on study has been initiated at Khmenytskyy Unit 1. The ISA projects encompass most areas of plant safety evaluation, but the initial emphasis is on performing a detailed, plant-specific Level 1 Internal Events PRA. This allows the early definition of the plant risk profile, the identification of risk significant accident sequences and plant vulnerabilities and provides guidance for the remainder of the safety assessments.« less

Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gonzalez-Pintor, S.; Verdu, G.; Ginestar, D.

Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmarkmore » and with the results provided by PARCS code. (authors)« less

Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50-400)°C

NASA Astrophysics Data System (ADS)

Kuleshova, E. A.; Gurovich, B. A.; Bukina, Z. V.; Frolov, A. S.; Maltsev, D. A.; Krikun, E. V.; Zhurko, D. A.; Zhuchkov, G. M.

2017-07-01

This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50-400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔTK) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects - dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔTK shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔTK shift in the studied range of irradiation temperature and fluence.

The fast neutron fluence and the activation detector activity calculations using the effective source method and the adjoint function

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hep, J.; Konecna, A.; Krysl, V.

2011-07-01

This paper describes the application of effective source in forward calculations and the adjoint method to the solution of fast neutron fluence and activation detector activities in the reactor pressure vessel (RPV) and RPV cavity of a VVER-440 reactor. Its objective is the demonstration of both methods on a practical task. The effective source method applies the Boltzmann transport operator to time integrated source data in order to obtain neutron fluence and detector activities. By weighting the source data by time dependent decay of the detector activity, the result of the calculation is the detector activity. Alternatively, if the weightingmore » is uniform with respect to time, the result is the fluence. The approach works because of the inherent linearity of radiation transport in non-multiplying time-invariant media. Integrated in this way, the source data are referred to as the effective source. The effective source in the forward calculations method thereby enables the analyst to replace numerous intensive transport calculations with a single transport calculation in which the time dependence and magnitude of the source are correctly represented. In this work, the effective source method has been expanded slightly in the following way: neutron source data were performed with few group method calculation using the active core calculation code MOBY-DICK. The follow-up neutron transport calculation was performed using the neutron transport code TORT to perform multigroup calculations. For comparison, an alternative method of calculation has been used based upon adjoint functions of the Boltzmann transport equation. Calculation of the three-dimensional (3-D) adjoint function for each required computational outcome has been obtained using the deterministic code TORT and the cross section library BGL440. Adjoint functions appropriate to the required fast neutron flux density and neutron reaction rates have been calculated for several significant points within the RPV and RPV cavity of the VVER-440 reacto rand located axially at the position of maximum power and at the position of the weld. Both of these methods (the effective source and the adjoint function) are briefly described in the present paper. The paper also describes their application to the solution of fast neutron fluence and detectors activities for the VVER-440 reactor. (authors)« less

PACER -- A fast running computer code for the calculation of short-term containment/confinement loads following coolant boundary failure. Volume 2: User information

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, J.J.

A fast running and simple computer code has been developed to calculate pressure loadings inside light water reactor containments/confinements under loss-of-coolant accident conditions. PACER was originally developed to calculate containment/confinement pressure and temperature time histories for loss-of-coolant accidents in Soviet-designed VVER reactors and is relevant to the activities of the US International Nuclear Safety Center. The code employs a multicompartment representation of the containment volume and is focused upon application to early time containment phenomena during and immediately following blowdown. PACER has been developed for FORTRAN 77 and earlier versions of FORTRAN. The code has been successfully compiled and executedmore » on SUN SPARC and Hewlett-Packard HP-735 workstations provided that appropriate compiler options are specified. The code incorporates both capabilities built around a hardwired default generic VVER-440 Model V230 design as well as fairly general user-defined input. However, array dimensions are hardwired and must be changed by modifying the source code if the number of compartments/cells differs from the default number of nine. Detailed input instructions are provided as well as a description of outputs. Input files and selected output are presented for two sample problems run on both HP-735 and SUN SPARC workstations.« less

Investigation of high temperature annealing effectiveness for recovery of radiation-induced structural changes and properties of 18Cr-10Ni-Ti austenitic stainless steels

NASA Astrophysics Data System (ADS)

Gurovich, B. A.; Kuleshova, E. A.; Frolov, A. S.; Maltsev, D. A.; Prikhodko, K. E.; Fedotova, S. V.; Margolin, B. Z.; Sorokin, A. A.

2015-10-01

A complex study of structural state and properties of 18Cr-10Ni-Ti austenitic stainless steel after irradiation in BOR-60 fast research reactor (in the temperature range 330-400 °С up to damaging doses of 145 dpa) and in VVER-1000 light water reactor (at temperature ∼320 °С and damaging doses ∼12-14 dpa) was performed. The possibility of recovery of structural-phase state and mechanical properties to the level almost corresponding to the initial state by the recovery annealing was studied. The principal possibility of the recovery annealing of pressurized water reactor internals that ensures almost complete recovery of its mechanical properties and microstructure was shown. The optimal mode of recovery annealing was established: 1000 °C during 120 h.

MEASUREMENTS OF THE CONFINEMENT LEAKTIGHTNESS AT THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

DOE Office of Scientific and Technical Information (OSTI.GOV)

GREENE,G.A.; GUPPY,J.G.

1998-08-01

This is the final report on the INSP project entitled, ``Kola Confinement Leaktightness'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.1. This project was initiated in February 1993 to assist the Russians to reduce risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Units 1 and 2, through upgrades in the confinement performance to reduce the uncontrolled leakage rate. The major technical objective of this-project was to improve the leaktightness of the Kola NPP VVER confinement boundaries, through the application of a variety of sealants to penetrations, doors andmore » hatches, seams and surfaces, to the extent that current technology permitted. A related objective was the transfer, through training of Russian staff, of the materials application procedures to the staff of the Kola NPP. This project was part of an overall approach to minimizing uncontrolled releases from the Kola NPP VVER440/230s in the event of a serious accident, and to thereby significantly mitigate the consequences of such an accident. The US provided materials, application technology, and applications equipment for application of sealant materials, surface coatings, potting materials and gaskets, to improve the confinement leaktightness of the Kola VVER-440/23Os. The US provided for training of Russian personnel in the applications technology.« less

INSTALLATION OF A POST-ACCIDENT CONFINEMENT HIGH-LEVEL RADIATION MONITORING SYSTEM IN THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

DOE Office of Scientific and Technical Information (OSTI.GOV)

GREENE,G.A.; GUPPY,J.G.

1998-09-01

This is the final report on the INSP project entitled, ``Post-Accident Confinement High-Level Radiation Monitoring System'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.6 (Attachment 1). This project was initiated in February 1993 to assist the Russians in reducing risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Unit 2, through improved accident detection capability, specifically by the installation of a dual train high-level radiation detection system in the confinement of Unit 2 of the Kola NPP. The major technical objective of this project was to provide, install andmore » make operational the necessary hardware inside the confinement of the Kola NPP Unit 2 to provide early and reliable warning of the release of radionuclides from the reactor into the confinement air space as an indication of the occurrence of a severe accident at the plant. In addition, it was intended to provide hands-on experience and training to the Russian plant workers in the installation, operation, calibration and maintenance of the equipment in order that they may use the equipment without continued US assistance as an effective measure to improve reactor safety at the plant.« less

Water chemistry of the secondary circuit at a nuclear power station with a VVER power reactor

NASA Astrophysics Data System (ADS)

Tyapkov, V. F.; Erpyleva, S. F.

2017-05-01

Results of implementation of the secondary circuit organic amine water chemistry at Russian nuclear power plant (NPP) with VVER-1000 reactors are presented. The requirements for improving the reliability, safety, and efficiency of NPPs and for prolonging the service life of main equipment items necessitate the implementation of new technologies, such as new water chemistries. Data are analyzed on the chemical control of power unit coolant for quality after the changeover to operation with the feed of higher amines, such as morpholine and ethanolamine. Power units having equipment containing copper alloy components were converted from the all-volatile water chemistry to the ethanolamine or morpholine water chemistry with no increase in pH of the steam generator feedwater. This enables the iron content in the steam generator feedwater to be decreased from 6-12 to 2.0-2.5 μg/dm3. It is demonstrated that pH of high-temperature water is among the basic factors controlling erosion and corrosion wear of the piping and the ingress of corrosion products into NPP steam generators. For NPP power units having equipment whose construction material does not include copper alloys, the water chemistries with elevated pH of the secondary coolant are adopted. Stable dosing of correction chemicals at these power units maintains pH25 of 9.5 to 9.7 in the steam generator feedwater with a maximum iron content of 2 μg/dm3 in the steam generator feedwater.

The in-depth safety assessment (ISA) pilot projects in Ukraine.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kot, C. A.

1998-02-10

Ukraine operates pressurized water reactors of the Soviet-designed type, VVER. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs). After approval of the SARS by the Ukrainian Nuclear Regulatory Authority, the plants will be granted longer-term operating licenses. In September 1995, the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine issued a new contents requirement for the safety analysis reports of VVERs in Ukraine. It contains requirements in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. Themore » DBA requirements are an expanded version of the older SAR requirements. The last two requirements, on PRA and BDBA, are new. The US Department of Energy (USDOE), through the International Nuclear Safety Program (INSP), has initiated an assistance and technology transfer program to Ukraine to assist their nuclear power stations in developing a Western-type technical basis for the new SARS. USDOE sponsored In-Depth Safety Assessments (ISAs) have been initiated at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1. USDOE/INSP have structured the ISA program in such a way as to provide maximum assistance and technology transfer to Ukraine while encouraging and supporting the Ukrainian plants to take the responsibility and initiative and to perform the required assessments.« less

The prospect of nuclear energy in Türkiye especially after Fukushima accident

NASA Astrophysics Data System (ADS)

Şahin, Sümer

2014-09-01

Türkiye considers since mid-50's to use nuclear electricity, but Government and bureaucracy have continuously postponed reactor construction. However, since 2010 the case has gained a real shape. Official agreement has been signed for the construction of 4 units of Russian VVER type reactors with installed power of 4×1200 MWel. It is expected that they will begin to deliver electricity early 20's. Further negotiations are being conducted with Japanese Mitsubashi and French AREVA. The target is to have nuclear electricity by 2023 at the 100th anniversary of Turkish Republic. Turkish Nuclear Energy Strategy aims; • Decrease country's dependency on foreign suppliers of energy sources • Provide fuel supply mix diversification • Utilization of environmentally friendly energy production technologies Possess advanced and prestigious power generation technologies.

Accumulation of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg

NASA Astrophysics Data System (ADS)

Hirschberg, Gábor; Baradlai, Pál; Varga, Kálmán; Myburg, Gerrit; Schunk, János; Tilky, Péter; Stoddart, Paul

Formation, presence and deposition of corrosion product radionuclides (such as 60Co, 51Cr, 54Mn, 59Fe and/or 110mAg) in the primary circuits of water-cooled nuclear reactors (PWRs) throw many obstacles in the way of normal operation. During the course of the work presented in this series, accumulations of such radionuclides have been studied at austenitic stainless steel type 08X18H10T (GOST 5632-61) surfaces (this austenitic stainless steel corresponds to AISI 321). Comparative experiments have been performed on magnetite-covered carbon steel (both materials are frequently used in some Soviet VVER type PWRs). For these laboratory-scale investigations a combination of the in situ radiotracer `thin gap' method and voltammetry is considered to be a powerful tool due to its high sensitivity towards the detection of the submonolayer coverages of corrosion product radionuclides. An independent technique (XPS) is also used to characterize the depth distribution and chemical state of various contaminants in the passive layer formed on austenitic stainless steel. In the first part of the series the accumulation of 110mAg has been investigated. Potential dependent sorption of Ag + ions (cementation) is found to be the predominant process on austenitic steel, while in the case of magnetite-covered carbon steel the silver species are mainly depleted in the form of Ag 2O. The XPS depth profile of Ag gives an evidence about the embedding of metallic silver into the entire passive layer of the austenitic stainless steel studied.

Dependency of the Reynolds number on the water flow through the perforated tube

DOE Office of Scientific and Technical Information (OSTI.GOV)

Závodný, Zdenko, E-mail: zdenko.zavodny@stuba.sk; Bereznai, Jozef, E-mail: jozef.bereznai@stuba.sk; Urban, František

Safe and effective loading of nuclear reactor fuel assemblies demands qualitative and quantitative analysis of the relationship between the coolant temperature in the fuel assembly outlet, measured by the thermocouple, and the mean coolant temperature profile in the thermocouple plane position. It is not possible to perform the analysis directly in the reactor, so it is carried out using measurements on the physical model, and the CFD fuel assembly coolant flow models. The CFD models have to be verified and validated in line with the temperature and velocity profile obtained from the measurements of the cooling water flowing in themore » physical model of the fuel assembly. Simplified physical model with perforated central tube and its validated CFD model serve to design of the second physical model of the fuel assembly of the nuclear reactor VVER 440. Physical model will be manufactured and installed in the laboratory of the Institute of Energy Machines, Faculty of Mechanical Engineering of the Slovak University of Technology in Bratislava.« less

In-vessel melt retention as a severe accident management strategy for the Loviisa Nuclear Power Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kymaelaeinen, O.; Tuomisto, H.; Theofanous, T.G.

1997-02-01

The concept of lower head coolability and in-vessel retention of corium has been approved as a basic element of the severe accident management strategy for IVO`s Loviisa Plant (VVER-440) in Finland. The selected approach takes advantage of the unique features of the plant such as low power density, reactor pressure vessel without penetrations at the bottom and ice-condenser containment which ensures flooded cavity in all risk significant sequences. The thermal analyses, which are supported by experimental program, demonstrate that in Loviisa the molten corium on the lower head of the reactor vessel is coolable externally with wide margins. This papermore » summarizes the approach and the plant modifications being implemented. During the approval process some technical concerns were raised, particularly with regard to thermal loadings caused by contact of cool cavity water and hot corium with the reactor vessel. Resolution of these concerns is also discussed.« less

The prospect of nuclear energy in Türkiye especially after Fukushima accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Şahin, Sümer, E-mail: ssahin@atilim.edu.tr

2014-09-30

Türkiye considers since mid-50's to use nuclear electricity, but Government and bureaucracy have continuously postponed reactor construction. However, since 2010 the case has gained a real shape. Official agreement has been signed for the construction of 4 units of Russian VVER type reactors with installed power of 4×1200 MW{sub el}. It is expected that they will begin to deliver electricity early 20's. Further negotiations are being conducted with Japanese Mitsubashi and French AREVA. The target is to have nuclear electricity by 2023 at the 100{sup th} anniversary of Turkish Republic. Turkish Nuclear Energy Strategy aims; • Decrease country's dependency onmore » foreign suppliers of energy sources • Provide fuel supply mix diversification • Utilization of environmentally friendly energy production technologies Possess advanced and prestigious power generation technologies.« less

Influence of structural parameters on the tendency of VVER-1000 reactor pressure vessel steel to temper embrittlement

NASA Astrophysics Data System (ADS)

Gurovich, B.; Kuleshova, E.; Zabusov, O.; Fedotova, S.; Frolov, A.; Saltykov, M.; Maltsev, D.

2013-04-01

In this paper the influence of structural parameters on the tendency of steels to reversible temper embrittlement was studied for assessment of performance properties of reactor pressure vessel steels with extended service life. It is shown that the growth of prior austenite grain size leads to an increase of the critical embrittlement temperature in the initial state. An embrittlement heat treatment at the temperature of maximum manifestation of temper embrittlement (480 °C) shifts critical embrittlement temperature to higher values due to the increase of the phosphorus concentration on grain boundaries. There is a correlation between phosphorus concentration on boundaries of primary austenite grains and the share of brittle intergranular fracture (that, in turn, depends on impact test temperature) in the fracture surfaces of the tested Charpy specimens.

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, Upendra S.

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary ofmore » appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/« less

Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

NASA Astrophysics Data System (ADS)

Muratov, V. G.; Lopatkin, A. V.

An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

Erosion-corrosion entrainment of iron-containing compounds as a source of deposits in steam generators used at nuclear power plants equipped with VVER reactors

NASA Astrophysics Data System (ADS)

Tomarov, G. V.; Shipkov, A. A.

2011-03-01

The main stages and processes through which deposits are generated, migrate, and precipitate in the metal-secondary coolant system of power units at nuclear power plants are analyzed and determined. It is shown that substances produced by the mechanism of general erosion-corrosion are the main source of the ionic-colloid form of iron, which is the main component of deposits in a steam generator. Ways for controlling the formation of deposits in a nuclear power plant's steam generator are proposed together with methods for estimating their efficiency.

Design issues concerning Iran`s Bushehr nuclear power plant VVER-1000 conversion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carson, C.F.

On January 8, 1995, the Atomic Energy Organization of Iran (AEOI) signed a contract for $800 million with the Russian Federation Ministry for Atomic Energy (Minatom) to complete Bushehr nuclear power plant (BNPP) unit 1. The agreement called for a Russian VVER-1000/320 pressurized water reactor (PWR) to be successfully installed into the existing German-built BNPP facilities in 5 yr. System design differences, bomb damage, and environmental exposure are key issues with which Minatom must contend in order to fulfill the contract. The AEOI under the Shah of Iran envisioned Bushehr as the first of many nuclear power plants, with Iranmore » achieving 24 GW(electric) by 1993 and 34 GW(electric) by 2000. Kraftwerk Union AG (KWU) began construction of the two-unit plant near the Persian Gulf town of Halileh in 1975. Unit 1 was {approx}80% complete and unit 2 was {approx}50% complete when construction was interrupted by the 1979 Iranian Islamic revolution. Despite repeated AEOI attempts to lure KWU and other companies back to Iran to complete the plant, Western concerns about nuclear proliferation in Iran and repeated bombings of the plant during the 1980-1988 Iran-Iraq war dissuaded Germany from resuming construction.« less

Assessment of environmental public exposure from a hypothetical nuclear accident for Unit-1 Bushehr nuclear power plant.

PubMed

Sohrabi, M; Ghasemi, M; Amrollahi, R; Khamooshi, C; Parsouzi, Z

2013-05-01

Unit-1 of the Bushehr nuclear power plant (BNPP-1) is a VVER-type reactor with 1,000-MWe power constructed near Bushehr city at the coast of the Persian Gulf, Iran. The reactor has been recently operational to near its full power. The radiological impact of nuclear power plant (NPP) accidents is of public concern, and the assessment of radiological consequences of any hypothetical nuclear accident on public exposure is vital. The hypothetical accident scenario considered in this paper is a design-basis accident, that is, a primary coolant leakage to the secondary circuit. This scenario was selected in order to compare and verify the results obtained in the present paper with those reported in the Final Safety Analysis Report (FSAR 2007) of the BNPP-1 and to develop a well-proven methodology that can be used to study other and more severe hypothetical accident scenarios for this reactor. In the present study, the version 2.01 of the PC COSYMA code was applied. In the early phase of the accidental releases, effective doses (from external and internal exposures) as well as individual and collective doses (due to the late phase of accidental releases) were evaluated. The surrounding area of the BNPP-1 within a radius of 80 km was subdivided into seven concentric rings and 16 sectors, and distribution of population and agricultural products was calculated for this grid. The results show that during the first year following the modeled hypothetical accident, the effective doses do not exceed the limit of 5 mSv, for the considered distances from the BNPP-1. The results obtained in this study are in good agreement with those in the FSAR-2007 report. The agreement obtained is in light of many inherent uncertainties and variables existing in the two modeling procedures applied and proves that the methodology applied here can also be used to model other severe hypothetical accident scenarios of the BNPP-1 such as a small and large break in the reactor coolant system as well as beyond design-basis accidents. Such scenarios are planned to be studied in the near future, for this reactor.

Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru; Pinegin, A. A.

The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

Radioactive waste management and plutonium recovery within the context of the development of nuclear energy in Russia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kushnikov, V.

1996-05-01

The Russian strategy for radioactive waste and plutonium management is based on the concept of the closed fuel cycle that has been adopted in Russia, and, to a great degree, falls under the jurisdiction of the existing Russian nuclear energy structures. From its very beginning, Russian atomic energy policy was based on finding the most effective method of developing the new fuel direction with the maximum possible utilization of the energy potential from the fission of heavy atoms and the achievement of fuel self-sufficiency through the recycling of secondary fuel. Although there can be no doubt about the importance ofmore » economic considerations (for the future), concerns for the safety of the environment are currently of the utmost importance. In this context, spent NPP fuel can be viewed as a waste to be buried only if there is persuasive evidence that such an approach is both economically and environmentally sound. The production of I GW of energy per year is accompanied by the accumulation of up to 800-1000 kg of highly radioactive fission products and approximately 250 kg of plutonium. Currently, spent fuel from the VVER 100 and the RBNK reactors contains approximately 25 tons of plutonium. There is an additional 30 tons of fuel-grade plutonium in the form of purified oxide, separated from spent fuels used in VVER440 reactors and other power production facilities, as well as approximately 100 tons of weapons-grade plutonium from dismantled warheads. The spent fuel accumulates significant amounts of small actinoids - neptunium americium, and curium. Science and technology have not yet found technical solutions for safe and secure burial of non-reprocessed spent fuel with such a broad range of products, which are typically highly radioactive and will continue to pose a threat for hundreds of thousands of years.« less

Comparison of various hours living fission products for absolute power density determination in VVER-1000 mock up in LR-0 reactor.

PubMed

Košťál, Michal; Švadlenková, Marie; Koleška, Michal; Rypar, Vojtěch; Milčák, Ján

2015-11-01

Measuring power level of zero power reactor is a quite difficult task. Due to the absence of measurable cooling media heating, it is necessary to employ a different method. The gamma-ray spectroscopy of fission products induced within reactor operation is one of possible ways of power determination. The method is based on the proportionality between fission product buildup and released power. The (92)Sr fission product was previously preferred as nuclide for LR-0 power determination for short-time irradiation experiments. This work aims to find more appropriate candidates, because the (92)Sr, however suitable, has a short half-life, which limits the maximal measurable amount of fuel pins within a single irradiation batch. The comparison of various isotopes is realized for (92)Sr, (97)Zr, (135)I, (91)Sr, and (88)Kr. The comparison between calculated and experimentally determined (C/E-1 values) net peak areas is assessed for these fission products. Experimental results show that studied fission products, except (88)Kr, are in comparable agreement with (92)Sr results. Since (91)Sr has notably higher half-life than (92)Sr, (91)Sr seems to be more appropriate marker in experiments with a large number of measured fuel pins. Copyright © 2015 Elsevier Ltd. All rights reserved.

Prediction and modeling of the two-dimensional separation characteristic of a steam generator at a nuclear power station with VVER-1000 reactors

NASA Astrophysics Data System (ADS)

Parchevsky, V. M.; Guryanova, V. V.

2017-01-01

A computational and experimental procedure for construction of the two-dimensional separation curve (TDSC) for a horizontal steam generator (SG) at a nuclear power station (NPS) with VVER-reactors. In contrast to the conventional one-dimensional curve describing the wetness of saturated steam generated in SG as a function of the boiler water level at one, usually rated, load, TDSC is a function of two variables, which are the level and the load of SGB that enables TDSC to be used for wetness control in a wide load range. The procedure is based on two types of experimental data obtained during rated load operation: the nonuniformity factor of the steam load at the outlet from the submerged perforated sheet (SPS) and the dependence of the mass water level in the vicinity of the "hot" header on the water level the "cold" end of SG. The TDSC prediction procedure is presented in the form of an algorithm using SG characteristics, such as steam load and water level as the input and giving the calculated steam wetness at the output. The zoneby-zone calculation method is used. The result is presented in an analytical form (as an empirical correlation) suitable for uploading into controllers or other controls. The predicted TDSC can be used during real-time operation for implementation of different wetness control scenarios (for example, if the effectiveness is a priority, then the minimum water level, minimum wetness, and maximum turbine efficiency should be maintained; if safety is a priority, then the maximum level at the allowable wetness and the maximum water inventory should be kept), for operation of NPS in controlling the frequency and power in a power system, at the design phase (as a part of the simulation complex for verification of design solutions), during construction and erection (in developing software for personnel training simulators), during commissioning tests (to reduce the duration and labor-intensity of experimental activities), and for training.

Three-dimensional pin-to-pin analyses of VVER-440 cores by the MOBY-DICK code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lehmann, M.; Mikolas, P.

1994-12-31

Nuclear design for the Dukovany (EDU) VVER-440s nuclear power plant is routinely performed by the MOBY-DICK system. After its implementation on Hewlett Packard series 700 workstations, it is able to perform routinely three-dimensional pin-to-pin core analyses. For purposes of code validation, the benchmark prepared from EDU operational data was solved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Samohyl, P.

The application of the LBB requires also fatigue flaw growth assessment. This analysis was performed for PWR nuclear power plants types VVER 440/230, VVER 440/213c, VVER 1000/320. Respecting that these NPP`s were designed according to Russian codes that differ from US codes it was needed to compare these approaches. Comparison with our experimental data was accomplished, too. Margins of applicability of the US methods and their modifications for the materials used for construction of Czech and Slovak NPP`s are shown. Computer code accomplishing the analysis according to described method is presented. Some measurement and calculations show that thermal stratifications inmore » horizontal pipelines can lead to additive loads that are not negligible and can be dangerous. An attempt to include these loads induced by steady-state stratification was made.« less

Material distribution in light water reactor-type bundles tested under severe accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Noack, V.; Hagen, S.J.L.; Hofmann, P.

1997-02-01

Severe fuel damage experiments simulating small-break loss-of-coolant accidents have been carried out in the CORA out-of-pile test facility at Forschungszentrum Karlsruhe. Rod bundles with electrically heated fuel rod simulators containing annular UO{sub 2} pellets, UO{sub 2} full pellet rods, and absorber rods of two kinds (Ag/In/Cd to represent pressurized water reactor conditions and B{sub 4}C to represent boiling water reactor and VVER-1000 fuel elements) were subjected to temperature transients up to 2,300 K. A special method was applied to determine the axial mass distribution of bundle materials. The low-temperature melt formation by various interactions between zirconium and components of absorbermore » and spacer grids strongly influences the bundle degradation and material relocation. Absorber materials can separate from the fuel by a noncoherent relocation of the materials at different temperatures. The distributions of solidified materials in the different test bundles show a clear dependence on the axial temperature profile. Coolant channel blockages are observed mainly at the lower end of the bundle, i.e., near the lowest elevation at which an oxidation excursion resulting from the highly exothermic zirconium-steam reaction had been experienced. This elevation corresponds with a steep axial temperature gradient in the maximum temperature attained. Oxide layers on Zircaloy result in reduced melt formation.« less

Adaptive Neural Network Algorithm for Power Control in Nuclear Power Plants

NASA Astrophysics Data System (ADS)

Masri Husam Fayiz, Al

2017-01-01

The aim of this paper is to design, test and evaluate a prototype of an adaptive neural network algorithm for the power controlling system of a nuclear power plant. The task of power control in nuclear reactors is one of the fundamental tasks in this field. Therefore, researches are constantly conducted to ameliorate the power reactor control process. Currently, in the Department of Automation in the National Research Nuclear University (NRNU) MEPhI, numerous studies are utilizing various methodologies of artificial intelligence (expert systems, neural networks, fuzzy systems and genetic algorithms) to enhance the performance, safety, efficiency and reliability of nuclear power plants. In particular, a study of an adaptive artificial intelligent power regulator in the control systems of nuclear power reactors is being undertaken to enhance performance and to minimize the output error of the Automatic Power Controller (APC) on the grounds of a multifunctional computer analyzer (simulator) of the Water-Water Energetic Reactor known as Vodo-Vodyanoi Energetichesky Reaktor (VVER) in Russian. In this paper, a block diagram of an adaptive reactor power controller was built on the basis of an intelligent control algorithm. When implementing intelligent neural network principles, it is possible to improve the quality and dynamic of any control system in accordance with the principles of adaptive control. It is common knowledge that an adaptive control system permits adjusting the controller’s parameters according to the transitions in the characteristics of the control object or external disturbances. In this project, it is demonstrated that the propitious options for an automatic power controller in nuclear power plants is a control system constructed on intelligent neural network algorithms.

Investigation of the possibility of using residual heat reactor energy

NASA Astrophysics Data System (ADS)

Aminov, R. Z.; Yurin, V. E.; Bessonov, V. N.

2017-11-01

The largest contribution to the probable frequency of core damage is blackout events. The main component of the heat capacity at each reactor within a few minutes following a blackout is the heat resulting from the braking of beta-particles and the transfer of gamma-ray energy by the fission fragments and their decay products, which is known as the residual heat. The power of the residual heat changes gradually over a long period of time and for a VVER-1000 reactor is about 15-20 MW of thermal power over 72 hours. Current cooldown systems increase the cost of the basic nuclear power plants (NPP) funds without changing the amount of electricity generated. Such systems remain on standby, accelerating the aging of the equipment and accordingly reducing its reliability. The probability of system failure increases with the duration of idle time. Furthermore, the reactor residual heat energy is not used. A proposed system for cooling nuclear power plants involves the use of residual thermal power to supply the station’s own needs in emergency situations accompanied by a complete blackout. The thermal power of residual heat can be converted to electrical energy through an additional low power steam turbine. In normal mode, the additional steam turbine generates electricity, which makes it possible to ensure spare NPP and a return on the investment in the reservation system. In this work, experimental data obtained from a Balakovo NPP was analyzed to determine the admissibility of cooldown of the reactors through the 2nd circuit over a long time period, while maintaining high-level parameters for the steam generated by the steam generators.

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skutnik, Steven E.

The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared tomore » a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.« less

Development and application of an information-analytic system on the problem of flow accelerated corrosion of pipeline elements in the secondary coolant circuit of VVER-440-based power units at the Novovoronezh nuclear power plant

NASA Astrophysics Data System (ADS)

Tomarov, G. V.; Povarov, V. P.; Shipkov, A. A.; Gromov, A. F.; Kiselev, A. N.; Shepelev, S. V.; Galanin, A. V.

2015-02-01

Specific features relating to development of the information-analytical system on the problem of flow-accelerated corrosion of pipeline elements in the secondary coolant circuit of the VVER-440-based power units at the Novovoronezh nuclear power plant are considered. The results from a statistical analysis of data on the quantity, location, and operating conditions of the elements and preinserted segments of pipelines used in the condensate-feedwater and wet steam paths are presented. The principles of preparing and using the information-analytical system for determining the lifetime to reaching inadmissible wall thinning in elements of pipelines used in the secondary coolant circuit of the VVER-440-based power units at the Novovoronezh NPP are considered.

RELAP5 posttest calculation of IAEA-SPE-4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petelin, S.; Mavko, B.; Parzer, I.

The International Atomic Energy Agency`s Fourth Standard Problem Exercise (IAEA-SPE-4) was performed at the PMK-2 facility. The PMK-2 facility is designed to study processes following small- and medium-size breaks in the primary system and natural circulation in VVER-440 plants. The IAEA-SPE-4 experiment represents a cold-leg side small break, similar to the IAEA-SPE-2, with the exception of the high-pressure safety injection being unavailable, and the secondary side bleed and feed initiation. The break valve was located at the dead end of a vertical downcomer, which in fact simulates a break in the reactor vessel itself, and should be unlikely to happenmore » in a real nuclear power plant (NPP). Three different RELAP5 code versions were used for the transient simulation in order to assess the calculations with test results.« less

Horizontal steam generator thermal-hydraulics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ubra, O.; Doubek, M.

1995-09-01

Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. Themore » 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.« less

Pretest analysis of natural circulation on the PWR model PACTEL with horizontal steam generators

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kervinen, T.; Riikonen, V.; Ritonummi, T.

A new tests facility - parallel channel tests loop (PACTEL)- has been designed and built to simulate the major components and system behavior of pressurized water reactors (PWRs) during postulated small- and medium-break loss-of-coolant accidents. Pretest calculations have been performed for the first test series, and the results of these calculations are being used for planning experiments, for adjusting the data acquisition system, and for choosing the optimal position and type of instrumentation. PACTEL is a volumetrically scaled (1:305) model of the VVER-440 PWR. In all the calculated cases, the natural circulation was found to be effective in removing themore » heat from the core to the steam generator. The loop mass flow rate peaked at 60% mass inventory. The straightening of the loop seals increased the mass flow rate significantly.« less

Determination of power distribution in the VVER-440 core on the basis of data from in-core monitors by means of a metric analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kryanev, A. V.; Udumyan, D. K.; Kurchenkov, A. Yu., E-mail: s327@vver.kiae.ru

2014-12-15

Problems associated with determining the power distribution in the VVER-440 core on the basis of a neutron-physics calculation and data from in-core monitors are considered. A new mathematical scheme is proposed for this on the basis of a metric analysis. In relation to the existing mathematical schemes, the scheme in question improves the accuracy and reliability of the resulting power distribution.

Test facility for investigation of heat transfer of promising coolants for the nuclear power industry

NASA Astrophysics Data System (ADS)

Belyaev, I. A.; Sviridov, V. G.; Batenin, V. M.; Biryukov, D. A.; Nikitina, I. S.; Manchkha, S. P.; Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, E. V.

2017-11-01

The results are presented of experimental investigations into liquid metal heat transfer performed by the joint research group consisting of specialist in heat transfer and hydrodynamics from NIU MPEI and JIHT RAS. The program of experiments has been prepared considering the concept of development of the nuclear power industry in Russia. This concept calls for, in addition to extensive application of water-cooled, water-moderated (VVER-type) power reactors and BN-type sodium cooled fast reactors, development of the new generation of BREST-type reactors, fusion power reactors, and thermonuclear neutron sources. The basic coolants for these nuclear power installations will be heavy liquid metals, such as lead and lithium-lead alloy. The team of specialists from NRU MPEI and JIHT RAS commissioned a new RK-3 mercury MHD-test facility. The major components of this test facility are a unique electrical magnet constructed at Budker Nuclear Physics Institute and a pressurized liquid metal circuit. The test facility is designed for investigating upward and downward liquid metal flows in channels of various cross-sections in a transverse magnetic field. A probe procedure will be used for experimental investigation into heat transfer and hydrodynamics as well as for measuring temperature, velocity, and flow parameter fluctuations. It is generally adopted that liquid metals are the best coolants for the Tokamak reactors. However, alternative coolants should be sought for. As an alternative to liquid metal coolants, molten salts, such as fluorides of lithium and beryllium (so-called FLiBes) or fluorides of alkali metals (so-called FLiNaK) doped with uranium fluoride, can be used. That is why the team of specialists from NRU MPEI and JIHT RAS, in parallel with development of a mercury MHD test facility, is designing a test facility for simulating molten salt heat transfer and hydrodynamics. Since development of this test facility requires numerical predictions and verification of numerical codes, all examined configurations of the MHD flow are also investigated numerically.

Modified ring stretch tensile testing of Zr-1Nb cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cohen, A.B.; Majumdar, S.; Ruther, W.E.

1998-03-01

In a round robin effort between the US Nuclear Regulatory Commission, Institut de Protection et de Surete Nucleaire in France, and the Russian Research Centre-Kurchatov Institute, Argonne National Laboratory conducted 16 modified ring stretch tensile tests on unirradiated samples of zr-1Nb cladding, which is used in Russian VVER reactors. Test were conducted at two temperatures (25 and 400 C) and two strain rates (0.001 and 1 s{sup {minus}1}). At 25 C and 0.001 s{sup {minus}1}, the yield strength (YS), ultimate tensile strength (UTS), uniform elongation (UE), and total elongation (TE) were 201 MPa, 331 MPa, 18.2%, and 57.6%, respectively. Atmore » 400 C and 0.001 s{sup {minus}1}, the YS, UTS, UE, and TE were 109 MPa, 185 MPa, 15.4%, and 67.7%, respectively. Finally, at 400 C and 1 s{sup {minus}1}, the YS, UTS, UE, and TE were 134 MPa, 189 MPa, 18.9%, and 53.4%, respectively. The high strain rate tests at room temperature were not successful. Test results proved to be very sensitive to the amount of lubrication used on the inserts; because of the large contact area between the inserts and specimen, too little lubrication leads to significantly higher strengths and lower elongations being reported. It is also important to note that only 70 to 80% of the elongation takes place in the gauge section, depending on specimen geometry. The appropriate percentage can be estimated from a simple model or can be calculated from finite-element analysis.« less

On formation of the asymptotic spectrum of delayed neutron emitters in measuring the VVER-1000 scram system effectiveness

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shishkov, L. K., E-mail: slk@vver.kiae.ru; Zizin, M. N., E-mail: zizin_m@mail.ru

The process of formation of an asymptotic distribution of the neutron flux density in the reactor systems after introducing different negative reactivities is considered. The impact of two factors after the reactivity introduction is evaluated: (1) nonuniformity of perturbation of core properties, on one hand, and (2) a sharp reduction in the density of prompt neutrons, which prevents the appearance of new delayed neutron emitters distributed in accordance with the “new” prompt neutron distribution, on the other hand. The results of calculations show that the errors of measuring the scram system effectiveness using the method of inverse solution of themore » kinetics equation are caused by the fact that, after the negative reactivity insertion, the sources of prompt and delayed neutrons have different spatial distributions. In the case of high negative reactivities, this difference remains while the system still has neutrons, which can be measured.« less

Optimized planning of in-service inspections of local flow-accelerated corrosion of pipeline elements used in the secondary coolant circuit of the VVER-440-based units at the Novovoronezh NPP

NASA Astrophysics Data System (ADS)

Tomarov, G. V.; Povarov, V. P.; Shipkov, A. A.; Gromov, A. F.; Budanov, V. A.; Golubeva, T. N.

2015-03-01

Matters concerned with making efficient use of the information-analytical system on the flow-accelerated corrosion problem in setting up in-service examination of the metal of pipeline elements operating in the secondary coolant circuit of the VVER-440-based power units at the Novovoronezh NPP are considered. The principles used to select samples of pipeline elements in planning ultrasonic thickness measurements for timely revealing metal thinning due to flow-accelerated corrosion along with reducing the total amount of measurements in the condensate-feedwater path are discussed.

Investigations on optimization of accident management measures following a station blackout accident in a VVER-1000 pressurized water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tusheva, P.; Schaefer, F.; Kliem, S.

2012-07-01

The reactor safety issues are of primary importance for preserving the health of the population and ensuring no release of radioactivity and fission products into the environment. A part of the nuclear research focuses on improvement of the safety of existing nuclear power plants. Studies, research and efforts are a continuing process at improving the safety and reliability of existing and newly developed nuclear power plants at prevention of a core melt accident. Station blackout (loss of AC power supply) is one of the dominant accidents taken into consideration at performing accident analysis. In case of multiple failures of safetymore » systems it leads to a severe accident. To prevent an accident to turn into a severe one or to mitigate the consequences, accident management measures must be performed. The present paper outlines possibilities for application and optimization of accident management measures following a station blackout accident. Assessed is the behaviour of the nuclear power plant during a station blackout accident without accident management measures and with application of primary/secondary side oriented accident management measures. Discussed are the possibilities for operators ' intervention and the influence of the performed accident management measures on the course of the accident. Special attention has been paid to the effectiveness of the passive feeding and physical phenomena having an influence on the system behaviour. The performed simulations show that the effectiveness of the secondary side feeding procedure can be limited due to an early evaporation or flashing effects in the feed water system. The analyzed cases show that the effectiveness of the accident management measures strongly depends on the initiation criteria applied for depressurization of the reactor coolant system. (authors)« less

77 FR 60039 - Non-Power Reactor License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-02

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [NRC-2011-0087] RIN 3150-AI96 Non-Power Reactor... the final regulatory basis for rulemaking to streamline non-power reactor license renewal. This final... Reactor (RTR) License Renewal Process. This contemplated rulemaking also recommends conforming changes to...

The Design of PSB-VVER Experiments Relevant to Accident Management

NASA Astrophysics Data System (ADS)

Nevo, Alessandro Del; D'Auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes(1), which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility (2), operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed (3). The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

Results of an Analysis of Field Studies of the Intrinsic Dynamic Characteristics Important for the Safety of Nuclear Power Plant Equipment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kaznovsky, A. P., E-mail: kaznovskyap@atech.ru; Kasiyanov, K. G.; Ryasnyj, S. I.

2015-01-15

A classification of the equipment important for the safety of nuclear power plants is proposed in terms of its dynamic behavior under seismic loading. An extended bank of data from dynamic tests over the entire range of thermal and mechanical equipment in generating units with VVER-1000 and RBMK-1000 reactors is analyzed. Results are presented from a study of the statistical behavior of the distribution of vibrational frequencies and damping decrements with the “small perturbation” factor that affects the measured damping decrements taken into account. A need to adjust the regulatory specifications for choosing the values of the damping decrements withmore » specified inertial loads on equipment owing to seismic effects during design calculations is identified. Minimum values of the decrements are determined and proposed for all types of equipment as functions of the directions and natural vibration frequencies of the dynamic interactions to be adopted as conservative standard values in the absence of actual experimental data in the course of design studies of seismic resistance.« less

Uranium oxide fuel cycle analysis in VVER-1000 with VISTA simulation code

NASA Astrophysics Data System (ADS)

Mirekhtiary, Seyedeh Fatemeh; Abbasi, Akbar

2018-02-01

The VVER-1000 Nuclear power plant generates about 20-25 tons of spent fuel per year. In this research, the fuel transmutation of Uranium Oxide (UOX) fuel was calculated by using of nuclear fuel cycle simulation system (VISTA) code. In this simulation, we evaluated the back end components fuel cycle. The back end component calculations are Spent Fuel (SF), Actinide Inventory (AI) and Fission Product (FP) radioisotopes. The SF, AI and FP values were obtained 23.792178 ton/y, 22.811139 ton/y, 0.981039 ton/y, respectively. The obtained value of spent fuel, major actinide, and minor actinide and fission products were 23.8 ton/year, 22.795 ton/year, 0.024 ton/year and 0.981 ton/year, respectively.

Methodology and measures for preventing unacceptable flow-accelerated corrosion thinning of pipelines and equipment of NPP power generating units

NASA Astrophysics Data System (ADS)

Tomarov, G. V.; Shipkov, A. A.; Lovchev, V. N.; Gutsev, D. F.

2016-10-01

Problems of metal flow-accelerated corrosion (FAC) in the pipelines and equipment of the condensate- feeding and wet-steam paths of NPP power-generating units (PGU) are examined. Goals, objectives, and main principles of the methodology for the implementation of an integrated program of AO Concern Rosenergoatom for the prevention of unacceptable FAC thinning and for increasing operational flow-accelerated corrosion resistance of NPP EaP are worded (further the Program). A role is determined and potentialities are shown for the use of Russian software packages in the evaluation and prediction of FAC rate upon solving practical problems for the timely detection of unacceptable FAC thinning in the elements of pipelines and equipment (EaP) of the secondary circuit of NPP PGU. Information is given concerning the structure, properties, and functions of the software systems for plant personnel support in the monitoring and planning of the inservice inspection of FAC thinning elements of pipelines and equipment of the secondary circuit of NPP PGUs, which are created and implemented at some Russian NPPs equipped with VVER-1000, VVER-440, and BN-600 reactors. It is noted that one of the most important practical results of software packages for supporting NPP personnel concerning the issue of flow-accelerated corrosion consists in revealing elements under a hazard of intense local FAC thinning. Examples are given for successful practice at some Russian NPP concerning the use of software systems for supporting the personnel in early detection of secondary-circuit pipeline elements with FAC thinning close to an unacceptable level. Intermediate results of working on the Program are presented and new tasks set in 2012 as a part of the updated program are denoted. The prospects of the developed methods and tools in the scope of the Program measures at the stages of design and construction of NPP PGU are discussed. The main directions of the work on solving the problems of flow-accelerated corrosion of pipelines and equipment in Russian NPP PGU are defined.

Modeling the transport of nitrogen in an NPP-2006 reactor circuit

NASA Astrophysics Data System (ADS)

Stepanov, O. E.; Galkin, I. Yu.; Sledkov, R. M.; Melekh, S. S.; Strebnev, N. A.

2016-07-01

Efficient radiation protection of the public and personnel requires detecting an accident-initiating event quickly. Specifically, if a heat-exchange tube in a steam generator is ruptured, the 16N radioactive nitrogen isotope, which contributes to a sharp increase in the steam activity before the turbine, may serve as the signaling component. This isotope is produced in the core coolant and is transported along the circulation circuit. The aim of the present study was to model the transport of 16N in the primary and the secondary circuits of a VVER-1000 reactor facility (RF) under nominal operation conditions. KORSAR/GP and RELAP5/Mod.3.2 codes were used to perform the calculations. Computational models incorporating the major components of the primary and the secondary circuits of an NPP-2006 RF were constructed. These computational models were subjected to cross-verification, and the calculation results were compared to the experimental data on the distribution of the void fraction over the steam generator height. The models were proven to be valid. It was found that the time of nitrogen transport from the core to the heat-exchange tube leak was no longer than 1 s under RF operation at a power level of 100% N nom with all primary circuit pumps activated. The time of nitrogen transport from the leak to the γ-radiation detection unit under the same operating conditions was no longer than 9 s, and the nitrogen concentration in steam was no less than 1.4% (by mass) of its concentration at the reactor outlet. These values were obtained using conservative approaches to estimating the leak flow and the transport time, but the radioactive decay of nitrogen was not taken into account. Further research concerned with the calculation of thermohydraulic processes should be focused on modeling the transport of nitrogen under RF operation with some primary circuit pumps deactivated.

Results of operation and current safety performance of nuclear facilities located in the Russian Federation

NASA Astrophysics Data System (ADS)

Kuznetsov, V. M.; Khvostova, M. S.

2016-12-01

After the NPP radiation accidents in Russia and Japan, a safety statu of Russian nuclear power plants causes concern. A repeated life time extension of power unit reactor plants, designed at the dawn of the nuclear power engineering in the Soviet Union, power augmentation of the plants to 104-109%, operation of power units in a daily power mode in the range of 100-70-100%, the use of untypical for NPP remixed nuclear fuel without a careful study of the results of its application (at least after two operating periods of the research nuclear installations), the aging of operating personnel, and many other management actions of the State Corporation "Rosatom", should attract the attention of the Federal Service for Ecological, Technical and Atomic Supervision (RosTekhNadzor), but this doesn't happen. The paper considers safety issues of nuclear power plants operating in the Russian Federation. The authors collected statistical information on violations in NPP operation over the past 25 years, which shows that even after repeated relaxation over this period of time of safety regulation requirements in nuclear industry and highly expensive NPP modernization, the latter have not become more safe, and the statistics confirms this. At a lower utilization factor high-power pressure-tube reactors RBMK-1000, compared to light water reactors VVER-440 and 1000, have a greater number of violations and that after annual overhauls. A number of direct and root causes of NPP mulfunctions is still high and remains stable for decades. The paper reveals bottlenecks in ensuring nuclear and radiation safety of nuclear facilities. Main outstanding issues on the storage of spent nuclear fuel are defined. Information on emissions and discharges of radioactive substances, as well as fullness of storages of solid and liquid radioactive waste, located at the NPP sites are presented. Russian NPPs stress test results are submitted, as well as data on the coming removal from operation of NPP units is analyzed.

Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trambauer, K.

1997-07-01

The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonablemore » accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.« less

A custom-tailored FAMOS burn-up meter for VVER 440 fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simon, G.G.; Golochtchapov, S.; Glazov, A.G.

1995-12-31

The FAMOS fuel assembly monitoring system had been originally developed for monitoring irradiated fuel assemblies of the Karlsruhe Nuclear Research Center concentrating on neutron detection systems for special applications.The measurements in the past had demonstrated that FAMOS can perform precise measurements to control or measure with accuracy the main physical parameters of spent fuel. The FAMOS 3 system is specialized for burn-up determination of fuel assemblies. Thus it is possible to take into account the burn-up for the purposes of storage and transportation. The Kola NPP VVER 440 requirements necessitated developing an especially adopted FAMOS 3 system. In addition tomore » the passive neutron measurement, a gross gamma detection and a boron concentration monitoring system are implemented. The new system was constructed as well as tested in laboratory experiments. The monitoring system has been delivered to the customer and is ready for use.« less

Assessment of segregation kinetics in water-moderated reactors pressure vessel steels under long-term operation

NASA Astrophysics Data System (ADS)

Kuleshova, E. A.; Gurovich, B. A.; Lavrukhina, Z. V.; Saltykov, M. A.; Fedotova, S. V.; Khodan, A. N.

2016-08-01

In reactor pressure vessel (RPV) bcc-lattice steels temper embrittlement is developed under the influence of both operating temperature of ∼300 °C and neutron irradiation. Segregation processes in the grain boundaries (GB) begin to play a special role in the assessment of the safe operation of the RPV in case of its lifetime extension up to 60 years or more. The most reliable information on the RPV material condition can be obtained by investigating the surveillance specimens (SS) that are exposed to operational factors simultaneously with the RPV itself. In this paper the GB composition in the specimens with different thermal exposure time at the RPV operating temperature as well as irradiated by fast neutrons (E ≥ 0.5 MeV) to different fluences (20-71)·1022 m-2 was studied by means of Auger electron spectroscopy (AES) including both impurity and main alloying elements content. The data obtained allowed to trace the trend of the operating temperature and radiation-stimulated diffusion influence on the overall segregants level in GB. The revealed differences in the concentration levels of GB segregants in different steels, are due to the different chemical composition of the steels and also due to different grain boundary segregation levels in initial (unexposed) state. The data were used to estimate the RPV steels working capacity for 60 years. The estimation was carried out using both the well-known Langmuir-McLean model and the one specially developed for RPV steels, which takes into account the structure and phase composition of VVER-1000 RPV steels, as well as the long-term influence of operational factors.

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

A Course in Chemical Reactor Design.

ERIC Educational Resources Information Center

Takoudis, Christos G.

1983-01-01

Presents course outline, topics covered, and final project (doubling as a take home final exam) for a one-semester, interdisciplinary course on the design and behavior of chemical reactors. Interplay of chemical and physical rate processes is stressed in the course. (JM)

Scientific basis for modelling and calculation of acoustic vibrations in the nuclear power plant coolant

NASA Astrophysics Data System (ADS)

Proskuryakov, K. N.

2017-11-01

Created new scientific direction: “Diagnosis, prognosis and prevention of vibration - acoustic resonances in the nuclear power plant (NPP) equipment. The possibility of using methods for calculating and analyzing electric oscillation systems in the study of the properties of acoustic systems with a two-phase medium is proved, based on the similarity of the differential equations describing the state of these systems. Is shown that the developed methods can be used to predict and prevent the occurrence of vibration - acoustic resonances in the NPP equipment. Is shown that the volume of pressurizer at NPPs with VVER and PWR as well as boiling water reactor that exploded at Japan’s NPP Fukushima Daiichi is a Helmholtz resonator, which contain water and steam volumes and able many times increases the impact on them of outside periodic oscillations. Paper presents most important results published long before the severe accidents at NPPs Three Mile Island (TMI), Chernobyl and Fukushima Daiichi that could be used for the prediction of a severe accident scenario, identification of measuring data and process control in order to minimize the damage. Worked out results also could be useful in another industrial technologies based on applications of single and two-phase flows.

Treatment of smuggled cigarette tobacco by composting process in facultative reactors.

PubMed

Zittel, Rosimara; Pinto da Silva, Cleber; Domingues, Cinthia Eloise; de Oliveira Stremel, Tatiana Roselena; de Almeida, Thiago Eduardo; Vieira Damiani, Gislaine; Xavier de Campos, Sandro

2018-01-01

This paper presents a study on the degradation of smuggled cigarette tobacco combined with domestic organic waste and sawdust or wood chips, using facultative reactor. Four reactors with different amounts of residue were assembled. For the study of the quality of the compost obtained, physicochemical, phytotoxicity and microbiological analyses were carried out. The mixture with wood chips presented the best temperature conditions and pH variation optimizing the degradation. The final germination index (GI) values of all treatments were above the recommended GI value (50%) and the final C/N ratio between 8 and 13 indicated a mature compost. The concentration of metals under study was below the limit allowed for the commercialization. The composting carried out in all facultative reactors provided ideal conditions for the total sterilization of the final compost. Therefore, the treatment of smuggled cigarettes through facultative reactors was efficient to produce stable and mature compost. Copyright © 2017 Elsevier Ltd. All rights reserved.

75 FR 57302 - Advisory Committee on Reactor Safeguards; Public Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-20

... Monday, October 14, 2009, (74 FR 52829-52830). Thursday, October 7, 2010, Conference Room T2-B1, Two....: Final Safety Evaluation Report Associated with the Economic Simplified Boiling Water Reactor (ESBWR..., Inc. regarding the final Safety Evaluation Report associated with the ESBWR design certification...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

76 FR 52715 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-23

... review Draft Final Regulatory Guide (RG) 1.93, ``Availability of Electric Power Sources,'' Revision 1 and new Draft Final RG 1.218, ``Condition Monitoring Techniques for Electric Cables Used in Nuclear Power... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS...

76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-16

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2014 CFR

2014-01-01

...), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation... 10 Energy 1 2014-01-01 2014-01-01 false Finality of partial decision on site suitability issues in a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES...

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2013 CFR

2013-01-01

...), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation... 10 Energy 1 2013-01-01 2013-01-01 false Finality of partial decision on site suitability issues in a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES...

NORTHROP REACTOR. REVISION NO. 1 TO THE FINAL SAFEGUARDS REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duncan, J.M.; Shimizu, B.; Romine, R.A.

1962-10-01

Additions and changes related to the original application concerning construction and operation of the Northrop Reactor are given. Answers to 13 questions relative to the Final Safeguards Report are included. Answers are also included concerning 12 questions regarding receipt, possession, and storage of fuel elements. Other information is included concerning personnel changes and changes from Al-clad fuel elements to improved stainless steelclad hydride elements. It is concluded that the improved Northrop Reactor does not present any undue hazard to the health and safety of the operating personnel or the public. (J.R.D.)

Risk-informed regulation and safety management of nuclear power plants--on the prevention of severe accidents.

PubMed

Himanen, Risto; Julin, Ari; Jänkälä, Kalle; Holmberg, Jan-Erik; Virolainen, Reino

2012-11-01

There are four operating nuclear power plant (NPP) units in Finland. The Teollisuuden Voima (TVO) power company has two 840 MWe BWR units supplied by Asea-Atom at the Olkiluoto site. The Fortum corporation (formerly IVO) has two 500 MWe VVER 440/213 units at the Loviisa site. In addition, a 1600 MWe European Pressurized Water Reactor supplied by AREVA NP (formerly the Framatome ANP--Siemens AG Consortium) is under construction at the Olkiluoto site. Recently, the Finnish Parliament ratified the government Decision in Principle that the utilities' applications to build two new NPP units are in line with the total good of the society. The Finnish utilities, Fenno power company, and TVO company are in progress of qualifying the type of the new nuclear builds. In Finland, risk-informed applications are formally integrated in the regulatory process of NPPs that are already in the early design phase and these are to run through the construction and operation phases all through the entire plant service time. A plant-specific full-scope probabilistic risk assessment (PRA) is required for each NPP. PRAs shall cover internal events, area events (fires, floods), and external events such as harsh weather conditions and seismic events in all operating modes. Special attention is devoted to the use of various risk-informed PRA applications in the licensing of Olkiluoto 3 NPP. © 2012 Society for Risk Analysis.

The slightly-enriched spectral shift control reactor. Final report, September 30, 1988--September 30, 1991

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martin, W.R.; Lee, J.C.; Larsen, E.W.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

A low-cost municipal sewage treatment system with a combination of UASB and the "fourth-generation" downflow hanging sponge reactors.

PubMed

Tandukar, M; Uemura, S; Machdar, I; Ohashi, A; Harada, H

2005-01-01

This paper presents an evaluation of the process performance of a pilot-scale "fourth generation" downflow hanging sponge (DHS) post-treatment system combined with a UASB pretreatment unit treating municipal wastewater. After the successful operation of the second- and third-generation DHS reactors, the fourth-generation DHS reactor was developed to overcome a few shortcomings of its predecessors. This reactor was designed to further enhance the treatment efficiency and simplify the construction process in real scale, especially for the application in developing countries. Configuration of the reactor was modified to enhance the dissolution of air into the wastewater and to avert the possible clogging of the reactor especially during sudden washout from the UASB reactor. The whole system was operated at a total hydraulic retention time (HRT) of 8 h (UASB: 6 h and DHS: 2 h) for a period of over 600 days. The combined system was able to remove 96% of unfiltered BOD with only 9 mg/L remaining in the final effluent. Likewise, F. coli were removed by 3.45 log with the final count of 10(3) to 10(4) MPN/100 ml. Nutrient removal by the system was also satisfactory.

Applying U.S. EOP Analytical Justification Experience for VVER Plants in the Ukraine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linn, Paul A.; Julian, Harold V.; Chapman, James R.

2002-07-01

The foundation for new Emergency Operating Instructions (EOIs) being developed at several plants in the Ukraine is the Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERGs) developed in the U.S. The ERGs were chosen as a base for the new EOIs for several reasons. First the overall structure and format was adaptable to VVER Pressurized Water Reactor (PWR) designs. Second, the ERGs have served as a base for many plant EOIs in both the U.S. and internationally. Third, key information supporting the ERGs was available. This paper describes the method used at one of the Ukrainian plants to provide anmore » analytical justification for their EOIs. The method being employed by a second plant is very similar, differing only slightly in how it is implemented. The WOG ERG development program, which started shortly after the accident at Three Mile Island Unit 2, used many sources of technical information on plant and system transient response, which were available in support of the plant design and licensing efforts. In addition, operating experience from many operating PWR plants in the U.S. and around the world was used. For example, design basis accident (DBA) analyses, documented in a plant's Safety Analysis Report (SAR) and other design documents, had been performed by Nuclear Steam Supply System (NSSS) vendors, utilities, or the Architect/Engineer. All relevant sources were considered in the development of the ERGs. Limited Probabilistic Risk Assessment (PRA) analyses were available during that time period. When a technical basis for a recovery strategy and associated operator actions was not available, an analysis was defined and performed. In general, these analyses were performed on a generic basis, and addressed the different categories of design (e.g., number of reactor coolant loops and/or low/high pressure safety injection system design). U.S. Nuclear Power plants that were in the WOG program were responsible for implementing the generic ERGs. This required the utilities to review the generic analyses to ensure that they were applicable and to justify any deviations from the ERG methodology. Modern PRA analyses are similar to the analyses supporting the ERGs since they address multiple failures and assume better estimate or expected assumptions for equipment availability and operator performance. The process being employed by Ukrainian plants is similar to the WOG. That is, available analyses and operating experience are being reviewed and pertinent information extracted to assist in the analytical justification of the EOIs. This includes the use of recently updated PRA and DBA analyses, other 'original' design information and operating experience. A systematic review of the EOIs is being conducted to identify items requiring analytical justification. For each analysis identified, the specific purpose of the analysis is being documented. The analysis needs are then compared to the available analyses and operating experience. From this review, new analyses needed to justify the EOIs are developed, and a basis for using existing analyses is established. The work is being conducted in two phases. The first phase performs all of the reviews and assessments necessary to determine the new analyses required to justify the EOIs. In the second phase, these new analyses will be conducted and documented, and the EOI Analytical Justification (AJ) report will be written. (authors)« less

Glasses for immobilization of low- and intermediate-level radioactive waste

NASA Astrophysics Data System (ADS)

Laverov, N. P.; Omel'yanenko, B. I.; Yudintsev, S. V.; Stefanovsky, S. V.; Nikonov, B. S.

2013-03-01

Reprocessing of spent nuclear fuel (SNF) for recovery of fissionable elements is a precondition of long-term development of nuclear energetics. Solution of this problem is hindered by the production of a great amount of liquid waste; 99% of its volume is low- and intermediate-level radioactive waste (LILW). The volume of high-level radioactive waste (HLW), which is characterized by high heat release, does not exceed a fraction of a percent. Solubility of glasses at an elevated temperature makes them unfit for immobilization of HLW, the insulation of which is ensured only by mineral-like matrices. At the same time, glasses are a perfect matrix for LILW, which are distinguished by low heat release. The solubility of borosilicate glass at a low temperature is so low that even a glass with relatively low resistance enables them to retain safety of under-ground LILW depositories without additional engineering barriers. The optimal technology of liquid confinement is their concentration and immobilization in borosilicate glasses, which are disposed in shallow-seated geological repositories. The vitrification of 1 m3 liquid LILW with a salt concentration of ˜300 kg/m3 leaves behind only 0.2 m3 waste, that is, 4-6 times less than by bitumen impregnation and 10 times less than by cementation. Environmental and economic advantages of LILW vitrification result from (1) low solubility of the vitrified LILW in natural water; (2) significant reduction of LILW volume; (3) possibility to dispose the vitrified waste without additional engineering barriers under shallow conditions and in diverse geological media; (4) the strength of glass makes its transportation and storage possible; and finally (5) reliable longterm safety of repositories. When the composition of the glass matrix for LILW is being chosen, attention should be paid to the factors that ensure high technological and economic efficiency of vitrification. The study of vitrified LILW from the Kursk nuclear power plant with high-power channel reactors (HPCR; equivalent Russian acronym, RBMK) and the Kalinin nuclear power plant with pressurized water reactors (PWR; equivalent Russian acronym VVER) after their 14-yr storage in the shallow-seated repository at the MosNPO Radon testing ground has confirmed the safety of repositories ensured by confinement properties of borosilicate matrix. The most efficient vitrification technology is based on cold crucible induction melting. If the content of a chemical element in waste exceeds its solubility in glass, a crystalline phase is formed in the course of vitrification, so that the glass ceramics become a matrix for such waste. Vitrified waste with high Fe; Na and Al; Na, Fe, and Al; Na and B is characterized. The composition of frit and its proportion to waste depends on waste composition. This procedure requires careful laboratory testing.

The Munich accelerator for fission fragments MAFF

NASA Astrophysics Data System (ADS)

Habs, D.; Groß, M.; Assmann, W.; Ames, F.; Bongers, H.; Emhofer, S.; Heinz, S.; Henry, S.; Kester, O.; Neumayr, J.; Ospald, F.; Reiter, P.; Sieber, T.; Szerypo, J.; Thirolf, P. G.; Varentsov, V.; Wilfart, T.; Faestermann, T.; Krücken, R.; Maier-Komor, P.

2003-05-01

The Munich Accelerator for Fission Fragments MAFF has been designed for the new Munich research reactor FRM-II. It will deliver several intense beams (˜3×10 11 s -1) of very neutron-rich fission fragments with a final energy of 30 keV (low-energy beam) or energies between 3.7 and 5.9 MeV· A (high-energy beam). Such beams are of interest for the creation of super-heavy elements by fusion reactions, nuclear spectroscopy of exotic nuclei, but they also have a potential for applications, e.g. in medicine. Presently the Munich research reactor FRM-II is ready for operation, but authorities delay the final permission to turn the reactor critical probably till the end of 2002. Only after this final permission the financing of the major parts of MAFF can start. On the other hand all major components have been designed and special components have been tested in separate setups.

SNAP (Space Nuclear Auxiliary Power) reactor overview. Final report, June 1982-December 1983

DOE Office of Scientific and Technical Information (OSTI.GOV)

Voss, S.S.

1984-08-01

The SNAP reactor programs are outlined in this report. A summary of the program is included along with a technical outline of the SER, S2DR, SNAP 10A/SNAPSHOT, S8ER, and S8DR reactor systems. Specifications of the designs, the design logic and a conclusion outlining some of the program weaknesses are given.

Modelling of the anti-neutrino production and spectra from a Magnox reactor

NASA Astrophysics Data System (ADS)

Mills, Robert W.; Mountford, David J.; Coleman, Jonathon P.; Metelko, Carl; Murdoch, Matthew; Schnellbach, Yan-Jie

2018-01-01

The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor operation was calculated. The results of the preliminary calculations are shown and limitations in the methods and data discussed.

The slightly-enriched spectral shift control reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martin, W.R.; Lee, J.C.; Larsen, E.W.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

Advanced shield development for a fission surface power system for the lunar surface

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. E. Craft; I. J. Silver; C. M. Clark

A nuclear reactor power system such as the affordable fission surface power system enables a potential outpostonthemoon.Aradiation shieldmustbe included in the reactor system to reduce the otherwise excessive dose to the astronauts and other vital system components. The radiation shield is typically the most massive component of a space reactor system, and thus must be optimized to reduce mass asmuchas possible while still providing the required protection.Various shield options for an on-lander reactor system are examined for outpost distances of 400m and 1 kmfromthe reactor. Also investigated is the resulting mass savings from the use of a high performance cermetmore » fuel. A thermal analysis is performed to determine the thermal behaviours of radiation shields using borated water. For an outpost located 1000m from the core, a tetramethylammonium borohydride shield is the lightest (5148.4 kg), followed by a trilayer shield (boron carbide–tungsten–borated water; 5832.3 kg), and finally a borated water shield (6020.7 kg). In all of the final design cases, the temperature of the borated water remains below 400 K.« less

The effect of mixing ratio variation of sludge and organic solid waste on biodrying process

NASA Astrophysics Data System (ADS)

Nasution, A. C.; Kristanto, G. A.

2018-01-01

In this study, organic waste was co-biodried with sludge cake to determine which mixing ratio gave the best result. The organic waste was consisted of dried leaves and green leaves, while the sludge cake was obtained from a waste water treatment plant in Bekasi. The experiment was performed on 3 lab-scale reactors with same specifications. After 21 days of experiment, it was found that the reactor with the lowest mixing fraction of sludge (5:1) has the best temperature profile and highest moisture content depletion compared with others. Initial moisture content and initial volatile solid content of this reactor’s feedstock was 52.25% and 82.4% respectively. The airflow rate was 10 lpm. After biodrying was done, the final moisture content of the feedstock from Reactor C was 22.0% and the final volatile solid content was 75.9%.The final calorific value after biodrying process was 3179,28kcal/kg.

Reactor-Scale Cultivation of the Hyperthermophilic Methanarchaeon Methanococcus jannaschii to High Cell Densities

PubMed Central

Mukhopadhyay, Biswarup; Johnson, Eric F.; Wolfe, Ralph S.

1999-01-01

For the hyperthermophilic and barophilic methanarchaeon Methanococcus jannaschii, we have developed a medium and protocols for reactor-scale cultivation that improved the final cell yield per liter from ∼0.5 to ∼7.5 g of packed wet cells (∼1.8 g dry cell mass) under autotrophic growth conditions and to ∼8.5 g of packed wet cells (∼2 g dry cell mass) with yeast extract (2 g liter−1) and tryptone (2 g liter−1) as medium supplements. For growth in a sealed bottle it was necessary to add Se to the medium, and a level of 2 μM for added Se gave the highest final cell yield. In a reactor M. jannaschii grew without added Se in the medium; it is plausible that the cells received Se as a contaminant from the reactor vessel and the H2S supply. But, for the optimal performance of a reactor culture, an addition of Se to a final concentration of 50 to 100 μM was needed. Also, cell growth in a reactor culture was inhibited at much higher Se concentrations. These observations and the data from previous work with methanogen cell extracts (B. C. McBride and R. S. Wolfe, Biochemistry 10:4312–4317, 1971) suggested that from a continuously sparged reactor culture Se was lost in the exhaust gas as volatile selenides, and this loss raised the apparent required level of and tolerance for Se. In spite of having a proteinaceous cell wall, M. jannaschii withstood an impeller tip speed of 235.5 cms−1, which was optimal for achieving high cell density and also was the higher limit for the tolerated shear rate. The organism secreted one or more acidic compounds, which lowered pH in cultures without pH control; this secretion continued even after cessation of growth. PMID:10543823

Graphite distortion ``C`` Reactor. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, N.H.

1962-02-08

This report covers the efforts of the Laboratory in an investigation of the graphite distortion in the ``C`` reactor at Hanford. The particular aspects of the problem to be covered by the Laboratory were possible ``fixes`` to the control rod sticking problem caused by VSR channel distortion.

Decommissioning of the Northrop TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cozens, George B.; Woo, Harry; Benveniste, Jack

1986-07-01

An overview of the administrative and operational aspects of decommissioning and dismantling the Northrop Mark F TRIGA Reactor, including: planning and preparation, personnel requirements, government interfacing, costs, contractor negotiations, fuel shipments, demolition, disposal of low level waste, final survey and disposition of the concrete biological shielding. (author)

Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fromm, Bradley; Hauch, Benjamin; Sridharan, Kumar

2016-12-01

A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulationmore » tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.« less

Synthesis of carbohydrates in a continuous flow reactor by immobilized phosphatase and aldolase.

PubMed

Babich, Lara; Hartog, Aloysius F; van Hemert, Lieke J C; Rutjes, Floris P J T; Wever, Ron

2012-12-01

Herein, we report a new flow process with immobilized enzymes to synthesize complex chiral carbohydrate analogues from achiral inexpensive building blocks in a three-step cascade reaction. The first reactor contained immobilized acid phosphatase, which phosphorylated dihydroxyacetone to dihydroxyacetone phosphate using pyrophosphate as the phosphate donor. The second flow reactor contained fructose-1,6-diphosphate aldolase (RAMA, rabbit muscle aldolase) or rhamnulose-1-phosphate aldolase (RhuA from Thermotoga maritima) and acid phosphatase. The immobilized aldolases coupled the formed dihydroxyacetone phosphate to aldehydes, resulting in phosphorylated carbohydrates. A final reactor containing acid phosphatase that dephosphorylated the phosphorylated product yielded the final product. Different aldehydes were used to synthesize carbohydrates on a gram scale. To demonstrate the feasibility of the flow systems, we synthesized 0.6 g of the D-fagomine precursor. By using immobilized aldolase RhuA we were also able to obtain other stereoisomers of the D-fagomine precursor. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

PubMed

Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

2011-01-01

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. Copyright © 2010 Elsevier Ltd. All rights reserved.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ryan, B.C.

1997-05-01

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

10 CFR 52.171 - Finality of manufacturing licenses; information requests.

Code of Federal Regulations, 2010 CFR

2010-01-01

... required for issuance of a construction permit, operating license, combined license, in any hearing under... manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases for not incorporating severe accident mitigation design alternatives into the design of the reactor to...

Safety approach to the selection of design criteria for the CRBRP reactor refueling system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meisl, C J; Berg, G E; Sharkey, N F

1979-01-01

The selection of safety design criteria for Liquid Metal Fast Breeder Reactor (LMFBR) refueling systems required the extrapolation of regulations and guidelines intended for Light Water Reactor refueling systems and was encumbered by the lack of benefit from a commercially licensed predecessor other than Fermi. The overall approach and underlying logic are described for developing safety design criteria for the reactor refueling system (RRS) of the Clinch River Breeder Reactor Plant (CRBRP). The complete selection process used to establish the criteria is presented, from the definition of safety functions to the finalization of safety design criteria in the appropriate documents.more » The process steps are illustrated by examples.« less

Reactor-Produced Medical Radionuclides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mirzadeh, Saed; Mausner, Leonard; Garland, Marc A

2011-01-01

The therapeutic use of radionuclides in nuclear medicine, oncology and cardiology is the most rapidly growing use of medical radionuclides. Since most therapeutic radionuclides are neutron rich and decay by beta emission, they are reactor-produced. This chapter deals mainly with production approaches with neutrons. Neutron interactions with matter, neutron transmission and activation rates, and neutron spectra of nuclear reactors are discussed in some detail. Further, a short discussion of the neutron-energy dependence of cross sections, reaction rates in thermal reactors, cross section measurements and flux monitoring, and general equations governing the reactor production of radionuclides are presented. Finally, the chaptermore » is concluded by providing a number of examples encompassing the various possible reaction routes for production of a number of medical radionuclides in a reactor.« less

SP-100 Program: space reactor system and subsystem investigations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harty, R.B.

1983-09-30

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

SP-100 program: Space reactor system and subsystem investigations

NASA Astrophysics Data System (ADS)

Harty, R. B.

1983-09-01

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. The nuclear safety review/approval process that is required for a space reactor system is summarized. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that is expected and to provide information that could be usable in future programs.

Predicting the conditions under which vibroacoustic resonances with external periodic loads occur in the primary coolant circuits of VVER-based NPPs

NASA Astrophysics Data System (ADS)

Proskuryakov, K. N.; Fedorov, A. I.; Zaporozhets, M. V.

2015-08-01

The accident at the Japanese Fukushima Daiichi nuclear power plant (NPP) caused by an earthquake showed the need of taking further efforts aimed at improving the design and engineering solutions for ensuring seismic resistance of NPPs with due regard to mutual influence of the dynamic processes occurring in the NPP building structures and process systems. Resonance interaction between the vibrations of NPP equipment and coolant pressure pulsations leads to an abnormal growth of dynamic stresses in structural materials, accelerated exhaustion of equipment service life, and increased number of sudden equipment failures. The article presents the results from a combined calculation-theoretical and experimental substantiation of mutual amplification of two kinds of external periodic loads caused by rotation of the reactor coolant pump (RCP) rotor and an earthquake. The data of vibration measurements at an NPP are presented, which confirm the predicted multiple amplification of vibrations in the steam generator and RCP at a certain combination of coolant thermal-hydraulic parameters. It is shown that the vibration frequencies of the main equipment may fall in the frequency band corresponding to the maximal values in the envelope response spectra constructed on the basis of floor accelerograms. The article presents the results from prediction of conditions under which vibroacoustic resonances with external periodic loads take place, which confirm the occurrence of additional earthquake-induced multiple growth of pressure pulsation intensity in the steam generator at the 8.3 Hz frequency and additional multiple growth of vibrations of the RCP and the steam generator cold header at the 16.6 Hz frequency. It is shown that at the elastic wave frequency equal to 8.3 Hz in the coolant, resonance occurs with the frequency of forced vibrations caused by the rotation of the RCP rotor. A conclusion is drawn about the possibility of exceeding the design level of equipment vibrations under the effect of external periodic loads caused by an earthquake when the vibration frequency of the reactor plant main equipment and the frequency of elastic waves fall in the frequency band corresponding to the maximal values of envelope response spectra.

10 CFR 50.82 - Termination of license.

Code of Federal Regulations, 2011 CFR

2011-01-01

... to the NRC, consistent with the requirements of § 50.4(b)(8); (ii) Once fuel has been permanently... fuel from the reactor vessel, or when a final legally effective order to permanently cease operations... emplacement or retention of fuel into the reactor vessel. (3) Decommissioning will be completed within 60...

Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 2, Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

1992-03-01

This report contains a review of design studies for Inertial Confinement reactor. This second of three volumes discussions is some detail the following: Objectives, requirements, and assumptions; rationale for design option selection; key technical issues and R&D requirements; and conceptual design selection and description.

Bioaugmentation of activated sludge towards 3-chloroaniline removal with a mixed bacterial population carrying a degradative plasmid.

PubMed

Bathe, Stephan; Schwarzenbeck, Norbert; Hausner, Martina

2009-06-01

A bioaugmentation approach combining several strategies was applied to achieve degradation of 3-chloroaniline (3CA) in semicontinuous activated sludge reactors. In a first step, a 3CA-degrading Comamonas testosteroni strain carrying the degradative plasmid pNB2 was added to a biofilm reactor, and complete 3CA degradation together with spread of the plasmid within the indigenous biofilm population was achieved. A second set of reactors was then bioaugmented with either a suspension of biofilm cells removed from the carrier material or with biofilm-containing carrier material. 3CA degradation was established rapidly in all bioaugmented reactors, followed by a slow adaptation of the non-bioaugmented control reactors. In response to variations in 3CA concentration, all reactors exhibited temporary performance breakdowns. Whereas duplicates of the control reactors deviated in their behaviour, the bioaugmented reactors appeared more reproducible in their performance and population dynamics. Finally, the carrier-bioaugmented reactors showed an improved performance in the presence of high 3CA influent concentrations over the suspension-bioaugmented reactors. In contrast, degradation in one control reactor failed completely, but was rapidly established in the remaining control reactor.

Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

DTIC Science & Technology

1987-06-01

computer modeling remains at best semiempirical (C-i), this large variation in scaling factor makes extrapolation of data impossible. The DIDO Water...in a full scale PWR are not practical. The reactor plant is not controlled to tolerances necessary for research, and utilities are reluctant to vary...MIT Reactor Safeguards Committee, in revision 1 to the PCCL Safety Evaluation Report (SER), for final approval to begin in-pile testing and

DE-NE0008277_PROTEUS final technical report 2018

DOE Office of Scientific and Technical Information (OSTI.GOV)

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

Calculation to experiment comparison of SPND signals in various nuclear reactor environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien

2015-07-01

In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first partmore » of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)« less

Final environmental statement, Liquid Metal Fast Breeder Reactor Program. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1975-12-01

Information is presented under the following section headings: LMFBR program options and their compatibility with the major issues affecting commercial development, Proposed Final Environmental Statement for the LMFBR program, December 1974, WASH-1535, supplemental material, and material relating to Proposed Final Environmental Statement review. (DG)

Taming The Next Set of Strategic Weapons Threats

DTIC Science & Technology

2006-06-01

Reactors Victor Gilinsky . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 6. Coping with Biological Threats after...Regime (MTCR) is not yet optimized to cope with these challenges. Finally, nuclear technologies have become much more difficult to control. New...resistance of the most popular type of power reactor concludes that the current international nuclear safeguards system needs to be modified to cope

Simulation of hydrostatic water level measuring system for pressure vessels with the ATHLET-code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hampel, R.; Vandreier, B.; Kaestner, W.

1996-11-01

The static and dynamic behavior of measuring systems determine the value indicated by the measuring systems in relation to the true operating conditions. This paper demonstrates the necessity to involve the behavior of measuring systems in accident analysis with the thermohydraulic code ATHLET (developed by GRS Germany) by the example of hydrostatic water level measurement for horizontal steam generators on NPP (VVER). The modelling of a comparison vessel for the level measuring system with high sensitivity and a limited range of measurement (narrow-range level measuring system) by using ATHLET components and the checking of the function of the module weremore » realized. A good correspondence (maximal deviation 3%) between the measured and calculated narrow-range water level by the module was obtained for a realized post calculation of a measured operational transient in a NPP (VVER). The research carried out was sponsored by the Federal Ministry for Research and Technology within the projects {open_quotes}Basic research of process and system behaviour of NPP, control technique for accident management{close_quotes} (Project number 150 0855/7) and the project RS 978. The research work appertains to the theoretic and experimental work of institute {open_quotes}Institut fuer ProzeBtechnik, ProzeBautomatisierung und MeBtechnik (IPM){close_quotes} for accident analysis and accident management.« less

Chemical vapor deposition of epitaxial silicon

DOEpatents

Berkman, Samuel

1984-01-01

A single chamber continuous chemical vapor deposition (CVD) reactor is described for depositing continuously on flat substrates, for example, epitaxial layers of semiconductor materials. The single chamber reactor is formed into three separate zones by baffles or tubes carrying chemical source material and a carrier gas in one gas stream and hydrogen gas in the other stream without interaction while the wafers are heated to deposition temperature. Diffusion of the two gas streams on heated wafers effects the epitaxial deposition in the intermediate zone and the wafers are cooled in the final zone by coolant gases. A CVD reactor for batch processing is also described embodying the deposition principles of the continuous reactor.

Changing concepts of geologic structure and the problem of siting nuclear reactors: Examples from Washington State

NASA Astrophysics Data System (ADS)

Tabor, R. W.

1986-09-01

The conflict between regulation and healthy evolution of geological science has contributed to the difficulties of siting nuclear reactors. On the Columbia Plateau in Washington, but for conservative design of the Hanford reactor facility, the recognition of the little-understood Olympic-Wallowa lineament as a major, possibly still active structural alinement might have jeopardized the acceptability of the site for nuclear reactors. On the Olympic Peninsula, evolving concepts of compressive structures and their possible recent activity and the current recognition of a subducting Juan de Fuca plate and its potential for generating great earthquakes—both concepts little-considered during initial site selection—may delay final acceptance of the Satsop site. Conflicts of this sort are inevitable but can be accommodated if they are anticipated in the reactor-licensing process. More important, society should be increasing its store of geologic knowledge now, during the current recess in nuclear reactor siting.

Development of a Model and Computer Code to Describe Solar Grade Silicon Production Processes

NASA Technical Reports Server (NTRS)

Srivastava, R.; Gould, R. K.

1979-01-01

Mathematical models and computer codes based on these models, which allow prediction of the product distribution in chemical reactors for converting gaseous silicon compounds to condensed-phase silicon were developed. The following tasks were accomplished: (1) formulation of a model for silicon vapor separation/collection from the developing turbulent flow stream within reactors of the Westinghouse (2) modification of an available general parabolic code to achieve solutions to the governing partial differential equations (boundary layer type) which describe migration of the vapor to the reactor walls, (3) a parametric study using the boundary layer code to optimize the performance characteristics of the Westinghouse reactor, (4) calculations relating to the collection efficiency of the new AeroChem reactor, and (5) final testing of the modified LAPP code for use as a method of predicting Si(1) droplet sizes in these reactors.

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

Development of a reactor with carbon catalysts for modular-scale, low-cost electrochemical generation of H 2O 2

DOE PAGES

Chen, Zhihua; Chen, Shucheng; Siahrostami, Samira; ...

2017-03-01

The development of small-scale, decentralized reactors for H 2O 2 production that can couple to renewable energy sources would be of great benefit, particularly for water purification in the developing world. Herein, we describe our efforts to develop electrochemical reactors for H 2O 2 generation with high Faradaic efficiencies of >90%, requiring cell voltages of only ~1.6 V. The reactor employs a carbon-based catalyst that demonstrates excellent performance for H 2O 2 production under alkaline conditions, as demonstrated by fundamental studies involving rotating-ring disk electrode methods. Finally, the low-cost, membrane-free reactor design represents a step towards a continuous, modular-scale, de-centralizedmore » production of H 2O 2.« less

Long lifetime fast spectrum reactor for lunar surface power system

NASA Astrophysics Data System (ADS)

Kambe, Mitsuru

1993-01-01

In the framework of innovative reactor research activities, a conceptual design study of fast spectrum reactor and primary system for 800 kWe lunar surface power system to be combined with potassium Rankine cycle power conversion has been conducted to meet the power requirements of the lunar base activities in the next century. The reactor subsystem is characterized by RAPID (Refueling by All Pins Integrated Design) concept to enhance inherent safety and to enable quick and simplifed refueling in every 10 years. RAPID concept affords power plant design lifetime of up to 30 years. Integrity of the reactor structure and replacement of failed primary circuits are also discussed. Substantial reduction in per-kWh cost on considering launch, emplacement, and final disposition can be expected by a long system lifetime.

Microprocessor tester for the treat upgrade reactor trip system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

78 FR 3470 - DTE Electric Company (Formerly the Detroit Edison Company), Notice of Availability of Final...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-16

...--specific Web page at http://www.nrc.gov/reactors/new-reactors/col/fermi.html . The Ellis Library and... possesses and are publicly-available, using any of the following methods: Federal Rulemaking Web site: Go to... Documents Access and Management System (ADAMS): You may access publicly-available documents online in the...

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2012 CFR

2012-01-01

... complete and acceptable for docketing under § 2.101(a)(3), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, shall determine whether the... a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION RULES OF...

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2011 CFR

2011-01-01

... complete and acceptable for docketing under § 2.101(a)(3), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, shall determine whether the... a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION RULES OF...

Optimization of the water chemistry of the primary coolant at nuclear power plants with VVER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barmin, L. F.; Kruglova, T. K.; Sinitsyn, V. P.

2005-01-15

Results of the use of automatic hydrogen-content meter for controlling the parameter of 'hydrogen' in the primary coolant circuit of the Kola nuclear power plant are presented. It is shown that the correlation between the 'hydrogen' parameter in the coolant and the 'hydrazine' parameter in the makeup water can be used for controlling the water chemistry of the primary coolant system, which should make it possible to optimize the water chemistry at different power levels.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalugin, A. V., E-mail: Kalugin-AV@nrcki.ru; Tebin, V. V.

The specific features of calculation of the effective multiplication factor using the Monte Carlo method for weakly coupled and non-asymptotic multiplying systems are discussed. Particular examples are considered and practical recommendations on detection and Monte Carlo calculation of systems typical in numerical substantiation of nuclear safety for VVER fuel management problems are given. In particular, the problems of the choice of parameters for the batch mode and the method for normalization of the neutron batch, as well as finding and interpretation of the eigenvalue spectrum for the integral fission matrix, are discussed.

Control console replacement at the WPI Reactor. [Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

A simple, space constrained NIRIM type reactor for chemical vapour deposition of diamond

NASA Astrophysics Data System (ADS)

Thomas, Evan L. H.; Ginés, Laia; Mandal, Soumen; Klemencic, Georgina M.; Williams, Oliver A.

2018-03-01

In this paper the design of a simple, space constrained chemical vapour deposition reactor for diamond growth is detailed. Based on the design by NIRIM, the reactor is composed of a quartz discharge tube placed within a 2.45 GHz waveguide to create the conditions required for metastable growth of diamond. Utilising largely off-the-shelf components and a modular design, the reactor allows for easy modification, repair, and cleaning between growth runs. The elements of the reactor design are laid out with the CAD files, parts list, and control files made easily available to enable replication. Finally, the quality of nanocrystalline diamond films produced are studied with SEM and Raman spectroscopy, with the observation of clear faceting and a large diamond fraction suggesting the design offers deposition of diamond with minimal complexity.

Analysis of the Browns Ferry Unit 3 irradiation experiments. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simmons, G.L.

1984-11-01

The results of the analysis of two experiments performed at the Browns Ferry-3 reactor are presented. These calculations utilize state-of-the-art neutron transport techniques and a new neutron cross-section library that has been developed for LWR applications. The calculations agree well with the experimental data obtained in irradiations inside the reactor vessel. For the measurements performed in the reactor cavity, the calculations agree well at the reactor midplane. Accurate determination of the axial distribution of the neutron fluence in the reactor cavity depends on having a concise representation of the axial-void distribution in the core. Detailed data are presented describing themore » procedures used in the generation of the new cross-section library that has been named SAILOR. This library is available from the Radiation-Shielding Information Center.« less

Damper mechanism for nuclear reactor control elements

DOEpatents

Taft, William Elwood

1976-01-01

A damper mechanism which provides a nuclear reactor control element decelerating function at the end of the scram stroke. The total damping function is produced by the combination of two assemblies, which operate in sequence. First, a tapered dashram assembly decelerates the control element to a lower velocity, after which a spring hydraulic damper assembly takes over to complete the final damping.

Glycerol Production and Transformation: A Critical Review with Particular Emphasis on Glycerol Reforming Reaction for Producing Hydrogen in Conventional and Membrane Reactors.

PubMed

Bagnato, Giuseppe; Iulianelli, Adolfo; Sanna, Aimaro; Basile, Angelo

2017-03-23

Glycerol represents an emerging renewable bio-derived feedstock, which could be used as a source for producing hydrogen through steam reforming reaction. In this review, the state-of-the-art about glycerol production processes is reviewed, with particular focus on glycerol reforming reactions and on the main catalysts under development. Furthermore, the use of membrane catalytic reactors instead of conventional reactors for steam reforming is discussed. Finally, the review describes the utilization of the Pd-based membrane reactor technology, pointing out the ability of these alternative fuel processors to simultaneously extract high purity hydrogen and enhance the whole performances of the reaction system in terms of glycerol conversion and hydrogen yield.

Glycerol Production and Transformation: A Critical Review with Particular Emphasis on Glycerol Reforming Reaction for Producing Hydrogen in Conventional and Membrane Reactors

PubMed Central

Bagnato, Giuseppe; Iulianelli, Adolfo; Sanna, Aimaro; Basile, Angelo

2017-01-01

Glycerol represents an emerging renewable bio-derived feedstock, which could be used as a source for producing hydrogen through steam reforming reaction. In this review, the state-of-the-art about glycerol production processes is reviewed, with particular focus on glycerol reforming reactions and on the main catalysts under development. Furthermore, the use of membrane catalytic reactors instead of conventional reactors for steam reforming is discussed. Finally, the review describes the utilization of the Pd-based membrane reactor technology, pointing out the ability of these alternative fuel processors to simultaneously extract high purity hydrogen and enhance the whole performances of the reaction system in terms of glycerol conversion and hydrogen yield. PMID:28333121

Enzymatic membrane reactors for biodegradation of recalcitrant compounds. Application to dye decolourisation.

PubMed

López, C; Mielgo, I; Moreira, M T; Feijoo, G; Lema, J M

2002-11-13

Membrane bioreactors are being increasingly used in enzymatic catalysed transformations. However, the application of enzymatic-based treatment systems in the environmental field is rather unusual. The aim of this paper is to overview the application of enzymatic membrane reactors to wastewater treatment, more specifically to dye decolourisation. Firstly, the basic aspects such as different configurations of enzymatic reactors, advantages and disadvantages associated to their utilisation are revised as well as the application of this technology to wastewater treatment. Secondly, dye decolourisation by white-rot fungi and their oxidative enzymes are discussed, presenting an overall view from for in vivo and in vitro systems. Finally, dye decolourisation by manganese peroxidase in an enzymatic membrane reactor in continuous operation is presented.

Corrosion and Corrosion Control in Light Water Reactors

NASA Astrophysics Data System (ADS)

Gordon, Barry M.

2013-08-01

Serious corrosion problems have plagued the light water reactor (LWR) industry for decades. The complex corrosion mechanisms involved and the development of practical engineering solutions for their mitigation will be discussed in this article. After a brief overview of the basic designs of the boiling water reactor (BWR) and pressurized water reactor (PWR), emphasis will be placed on the general corrosion of LWR containments, flow-accelerated corrosion of carbon steel components, intergranular stress corrosion cracking (IGSCC) in BWRs, primary water stress corrosion cracking (PWSCC) in PWRs, and irradiation-assisted stress corrosion cracking (IASCC) in both systems. Finally, the corrosion future of both plants will be discussed as plants extend their period of operation for an additional 20 to 40 years.

75 FR 32972 - Office of New Reactors: Notice of Availability of the Final Staff Guidance; Section 14.3.12 on...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-10

... the Final Staff Guidance; Section 14.3.12 on Physical Security Hardware Inspections, Tests, Analyses...: The NRC is issuing its Final Revision 1 to NUREG-0800, ``Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants,'' Section 14.3.12 on ``Physical Security Hardware...

75 FR 68009 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Standard Review Plan...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-04

... the Final Staff Guidance Standard Review Plan Section 13.6.3, Revision 1 on Physical Security--Early... NRC is issuing its Final Revision 1 to NUREG-0800, ``Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants,'' Section 13.6.3, Revision 1 on Physical Security...

75 FR 68009 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Standard Review Plan...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-04

... the Final Staff Guidance Standard Review Plan Section 13.6.2, Revision 1 on Physical Security--Design... issuing its Final Revision 1 to NUREG-0800, ``Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants,'' Section 13.6.2, Revision 1 on Physical Security--Design Certification...

75 FR 68009 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Standard Review Plan...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-04

... the Final Staff Guidance Standard Review Plan, Section 13.6.1, Revision 1 on Physical Security... Availability. SUMMARY: The NRC is issuing its Final Revision 1 to NUREG-0800, ``Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants,'' Section 13.6.1, Revision 1 on...

Entropy Production in Chemical Reactors

NASA Astrophysics Data System (ADS)

Kingston, Diego; Razzitte, Adrián C.

2017-06-01

We have analyzed entropy production in chemically reacting systems and extended previous results to the two limiting cases of ideal reactors, namely continuous stirred tank reactor (CSTR) and plug flow reactor (PFR). We have found upper and lower bounds for the entropy production in isothermal systems and given expressions for non-isothermal operation and analyzed the influence of pressure and temperature in entropy generation minimization in reactors with a fixed volume and production. We also give a graphical picture of entropy production in chemical reactions subject to constant volume, which allows us to easily assess different options. We show that by dividing a reactor into two smaller ones, operating at different temperatures, the entropy production is lowered, going as near as 48 % less in the case of a CSTR and PFR in series, and reaching 58 % with two CSTR. Finally, we study the optimal pressure and temperature for a single isothermal PFR, taking into account the irreversibility introduced by a compressor and a heat exchanger, decreasing the entropy generation by as much as 30 %.

Interim Safe Storage of Plutonium Production Reactors at the US DOE Hanford Site - 13438

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schilperoort, Daryl L.; Faulk, Darrin

2013-07-01

Nine plutonium production reactors located on DOE's Hanford Site are being placed into an Interim Safe Storage (ISS) period that extends to 2068. The Environmental Impact Statement (EIS) for ISS [1] was completed in 1993 and proposed a 75-year storage period that began when the EIS was finalized. Remote electronic monitoring of the temperature and water level alarms inside the safe storage enclosure (SSE) with visual inspection inside the SSE every 5 years are the only planned operational activities during this ISS period. At the end of the ISS period, the reactor cores will be removed intact and buried inmore » a landfill on the Hanford Site. The ISS period allows for radioactive decay of isotopes, primarily Co-60 and Cs-137, to reduce the dose exposure during disposal of the reactor cores. Six of the nine reactors have been placed into ISS by having an SSE constructed around the reactor core. (authors)« less

Bioelectrochemical enhancement of methane production from highly concentrated food waste in a combined anaerobic digester and microbial electrolysis cell.

PubMed

Park, Jungyu; Lee, Beom; Tian, Donjie; Jun, Hangbae

2018-01-01

A microbial electrolysis cell (MEC) is a promising technology for enhancing biogas production from an anaerobic digestion (AD) reactor. In this study, the effects of the MEC on the rate of methane production from food waste were examined by comparing an AD reactor with an AD reactor combined with a MEC (AD+MEC). The use of the MEC accelerated methane production and stabilization via rapid organic oxidation and rapid methanogenesis. Over the total experimental period, the methane production rate and stabilization time of the AD+MEC reactor were approximately 1.7 and 4.0 times faster than those of the AD reactor. Interestingly however, at the final steady state, the methane yields of both the reactors were similar to the theoretical maximum methane yield. Based on these results, the MEC did not increase the methane yield over the theoretical value, but accelerated methane production and stabilization by bioelectrochemical reactions. Copyright © 2017 Elsevier Ltd. All rights reserved.

NACA Zero Power Reactor Facility Hazards Summary

NASA Technical Reports Server (NTRS)

1957-01-01

The Lewis Flight Propulsion Laboratory of the National Advisory Committee for Aeronautics proposes to build a zero power research reactor facility which will be located in the laboratory grounds near Clevelaurd, Ohio. The purpose of this report is to inform the Advisory Commit tee on Reactor Safeguards of the U. S. Atomic Energy Commission in re gard to the design of the reactor facility, the cha,acteristics of th e site, and the hazards of operation at this location, The purpose o f this reactor is to perform critical experiments, to measure reactiv ity effects, to serve as a neutron source, and to serve as a training tool. The reactor facility is described. This is followed by a discu ssion of the nuclear characteristics and the control system. Site cha racteristics are then discussed followed by a discussion of the exper iments which may be conducted in the facility. The potential hazards of the facility are then considered, particularly, the maximum credib le accident. Finally, the administrative procedure is discussed.

Conceptual design of BNCT facility based on the TRR medical room

NASA Astrophysics Data System (ADS)

Golshanian, M.; Rajabi, A. A.; Kasesaz, Y.

2017-10-01

This paper presents a conceptual design of the Boron Neutron Capture Therapy (BNCT) facility based on the medical room of Tehran Research Reactor (TRR). The medical room is located behind the east wall of the reactor pool. The designed beam line is an in-pool Beam Shaping Assembly (BSA) which is considered between the reactor core and the medical room wall. The final designed BSA can provide 2.96× 109 n/cm2ṡs epithermal neutron flux at the irradiation position with acceptable beam contamination to use as a clinical BNCT.

Calculation and Experiment of Adding Top Beryllium Shims for Iran MNSR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ebadati, Javad; Rezvanifard, Mehdi; Shahabi, Iraj

2006-07-01

Miniature Neutron Source Reactor which is called MNSR were put into operation on June 1994 in Esfahan Nuclear Technology Center (ENTC). At that time the excess reactivity at the cold condition was 3.85 mk. After 7 years of operation and fuel consumption the reactivity was reduced to 2.90 mk. To compensate this reduction and upgrade the reactor, Beryllium Shim were used at the top of the core. This paper discusses the steps for this accurate and sensitive task. Finally a layer of 1.5 mm Beryllium were added to restore the reactor life. (authors)

N.S. SAVANNAH, DRAFT OF FINAL SAFEGUARDS REPORT TEST, START-UP AND TRIALS, NEW YORK SHIPBUILDING CORPORATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-04-01

The N. S. Savannah program for testing, start-up, and initial operation of all reactor and propulsion components and systems is discussed. Definitions of test phases are given and various stages of the test program are outlined. A list of tests for the various reactor, propulsion, and other system components is included. (C.J.G.)

10 CFR 52.171 - Finality of manufacturing licenses; information requests.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Finality of manufacturing licenses; information requests. 52.171 Section 52.171 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND... manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases...

10 CFR 52.171 - Finality of manufacturing licenses; information requests.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Finality of manufacturing licenses; information requests. 52.171 Section 52.171 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND... manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases...

10 CFR 52.171 - Finality of manufacturing licenses; information requests.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Finality of manufacturing licenses; information requests. 52.171 Section 52.171 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND... manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases...

10 CFR 52.171 - Finality of manufacturing licenses; information requests.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Finality of manufacturing licenses; information requests. 52.171 Section 52.171 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND... manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases...

The prototype fast reactor at Dounreay, Scotland. Process and engineering development for sodium removal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mann, A.; Herrick, R.; Gunn, J.

2007-07-01

Dounreay was home to commercial fast reactor development in the UK. Following the construction and operation of the Dounreay Fast Reactor, a sodium-cooled Prototype Fast Reactor (PFR), was constructed. PFR started operating in 1974, closed in 1994 and is presently being decommissioned. To date the bulk of the sodium has been removed and treated. Due to the design of the existing extraction system however, a sodium pool will remain in the heel of the reactor. To remove this sodium, a pump/camera system was developed, tested and deployed. The Water Vapour Nitrogen (WVN) process has been selected to allow removal ofmore » the final sodium residues from the reactor. Due to the design of the reactor and potential for structural damage should Normal WVN (which produces hydrated sodium hydroxide) be used, Low Concentration WVN (LC WVN) has been developed. Pilot scale testing has shown that it is possible treat the reactor within 18 months at a WVN concentration of up to 4% v/v and temperature of 120 deg. C. At present the equipment that will be used to apply LC WVN to the reactor is being developed at the detail design stage. and is expected to be deployed within the next few years. (authors)« less

Final report on the oxidation of energetic materials in supercritical water. Final Air Force report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buelow, S.J.; Allen, D.; Anderson, G.K.

1995-04-03

The objective of this project was to determine the suitability of oxidation in supercritical fluids (SCO), particularly water (SCWO), for disposal of propellants, explosives, and pyrotechnics (PEPs). The SCO studies of PEPs addressed the following issues: The efficiency of destruction of the substrate. The products of destruction contained in the effluents. Whether the process can be conducted safely on a large scale. Whether energy recovery from the process is economically practicable. The information essential for process development and equipment design was also investigated, including issues such as practical throughput of explosives through a SCWO reactor, reactor materials and corrosion, andmore » models for process design and optimization.« less

Basic principles of creating a new generation of high- temperature brazing filler alloys

NASA Astrophysics Data System (ADS)

Kalin, B. A.; Suchkov, A. N.

2016-04-01

The development of new materials is based on the formation of a structural-phase state providing the desired properties by selecting the base and the complex of alloying elements. The development of amorphous filler alloys for a high-temperature brazing has its own features that are due to the limited life cycle and the production method of brazing filler alloys. The work presents a cycle of analytical and experimental materials science investigations including justification of the composition of a new amorphous filler alloy for brazing the products from zirconium alloys at the temperature of no more than 800 °C and at the unbrazing temperature of permanent joints of more than 1200 °C. The experimental alloys have been used for manufacture of amorphous ribbons by rapid quenching, of which the certification has been made by X-ray investigations and a differential-thermal analysis. These ribbons were used to obtain permanent joints from the spacer grid cells (made from the alloy Zr-1% Nb) of fuel assemblies of the thermal nuclear reactor VVER-440. The brazed samples in the form of a pair of cells have been exposed to corrosion tests in autoclaves in superheated water at a temperature of 350 °C, a pressure of 160 MPa and duration of up to 6,000 h. They have been also exposed to destructive tests using a tensile machine. The experimental results obtained have made it possible to propose and patent a brazing filler alloy of the following composition: Zr-5.5Fe-(2.5-3.5)Be-1Nb-(5-8)Cu-2Sn-0.4Cr-(0.5-1.0)Ge. Its melting point is 780 °C and the recommended brazing temperature is 800°C.

75 FR 70304 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-17

..., Rockville, Maryland. Thursday, December 2, 2010, Conference Room T2-B1, Two White Flint North, Rockville... opening remarks regarding the conduct of the meeting. 8:35 a.m.-10 a.m.: Final Safety Evaluation Report... Kewaunee, Inc., regarding the final safety evaluation report associated with the License Renewal...

Final environmental statement, Liquid Metal Fast Breeder Reactor Program. Volume 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1975-12-01

Included are copies of thirty-four comment letters on the Proposed Final Environmental Statement together with the ERDA replies to these letters. The letters were received from Federal, State, and local agencies, environmental and public interest groups, members of the academic and industrial communities, and individual citizens. (DG)

Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality.

PubMed

Rezaeian, M; Kamali, J

2017-01-01

Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B 4 C) were investigated, and the minimum required receptacle pitch of the basket was determined. Copyright © 2016 Elsevier Ltd. All rights reserved.

Accounting for the Effect of Noncondensing Gases on Interphasic Heat and Mass Transfer in the Two-Fluid Model Used in the KORSAR Code

NASA Astrophysics Data System (ADS)

Yudov, Yu. V.

2018-03-01

A model is presented of the interphasic heat and mass transfer in the presence of noncondensable gases for the KORSAR/GP design code. This code was developed by FGUP NITI and the special design bureau OKB Gidropress. It was certified by Rostekhnadzor in 2009 for numerical substantiation of the safety of reactor installations with VVER reactors. The model is based on the assumption that there are three types of interphasic heat and mass transfer of the vapor component: vapor condensation or evaporation on the interphase under any thermodynamic conditions of the phases, pool boiling of the liquid superheated above the saturation temperature at the total pressure, and spontaneous condensation in the volume of gas phase supercooled below the saturation temperature at the vapor partial pressure. Condensation and evaporation on the interphase continuously occur in a two-phase flow and control the time response of the interphase heat and mass transfer. Boiling and spontaneous condensation take place only at the metastable condition of the phases and run at a quite high speed. The procedure used for calculating condensation and evaporation on the interphase accounts for the combined diffusion and thermal resistance of mass transfer in all regimes of the two-phase flow. The proposed approach accounts for, in a natural manner, a decrease in the rate of steam condensation (or generation) in the presence of noncondensing components in the gas phase due to a decrease (or increase) in the interphase temperature relative to the saturation temperature at the vapor partial pressure. The model of the interphase heat transfer also accounts for the processes of dissolution or release of noncondensing components in or from the liquid. The gas concentration at the interphase and on the saturation curve is calculated by the Henry law. The mass transfer coefficient in gas dissolution is based on the heat and mass transfer analogy. Results are presented of the verification of the interphase heat and mass transfer used in the KORSAR/GP code based on the data on film condensation of steam-air flows in vertical pipes. The proposed model was also tested by solving a problem of nitrogen release from a supersaturated water solution.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nygaard, E. T.; Williams, M. M. R.; Angelo, P. L.

Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' A 'Level 1' model representing B and W'smore » proposed Medical Isotope Production System (MIPS) reactor has been developed. The Level 1 model couples a series of differential equations representing neutronics, temperature, and voiding. Neutronics are represented by point reactor kinetics while temperature and voiding terms are axially varying (one-dimensional). While this model was developed specifically for the MIPS reactor, its applicability to the Japanese TRACY reactor was assessed. The results from the Level 1 model were in good agreement with TRACY experimental data and found to be conservative over most of the time domains considered. The Level 1 model was used to study the MIPS reactor. An analysis showed the Level 1 model agreed well with a more complex computational model of the MIPS reactor (a FETCH model). Finally, a significant reactivity insertion was simulated with the Level 1 model to study the MIPS reactor's time-dependent response. (authors)« less

Reactor Operations Monitoring System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hart, M.M.

1989-01-01

The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« less

GUM Analysis for TIMS and SIMS Isotopic Ratios in Graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heasler, Patrick G.; Gerlach, David C.; Cliff, John B.

2007-04-01

This report describes GUM calculations for TIMS and SIMS isotopic ratio measurements of reactor graphite samples. These isotopic ratios are used to estimate reactor burn-up, and currently consist of various ratios of U, Pu, and Boron impurities in the graphite samples. The GUM calculation is a propagation of error methodology that assigns uncertainties (in the form of standard error and confidence bound) to the final estimates.

Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing 235U, 233U, and 232Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

NASA Astrophysics Data System (ADS)

Ioffe, B. L.; Kochurov, B. P.

2012-02-01

A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of 235U. It operates in the open-cycle mode involving 233U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

Integrated process of distillation with side reactors for synthesis of organic acid esters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Panchal, Chandrakant B; Prindle, John C; Kolah, Aspri

An integrated process and system for synthesis of organic-acid esters is provided. The method of synthesizing combines reaction and distillation where an organic acid and alcohol composition are passed through a distillation chamber having a plurality of zones. Side reactors are used for drawing off portions of the composition and then recycling them to the distillation column for further purification. Water is removed from a pre-reactor prior to insertion into the distillation column. An integrated heat integration system is contained within the distillation column for further purification and optimizing efficiency in the obtaining of the final product.

Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

Recovery of elemental sulphur from anaerobic effluents through the biological oxidation of sulphides.

PubMed

de Sousa, José Tavares; Lima, Jéssyca de Freitas; da Silva, Valquíria Cordeiro; Leite, Valderi Duarte; Lopes, Wilton Silva

2017-03-01

The aim of the present study was to evaluate the biological oxidation of sulphide in two different UASB reactors by assessing the occurrence of oxidized forms of sulphur in the effluents and the amount of S 0 that could be recovered in the process. The bioreactors employed were an anaerobic hybrid (AH) reactor employing porous polyurethane foam as support media and a micro-aerated UASB reactor equipped with an aeration device above the digestion zone. The AH reactor produced a final effluent containing low concentrations of S 2- (3.87% of total sulphur load). It was achieved due to a complete oxidation of 56.1% of total sulphur. The partial biological oxidation that occurred in the AH reactor allowed the recovery of 30% of the sulphur load as S 0 . The effluent from the micro-aerated UASB reactor contained 5% of the sulphur load in the form of S 2- , while 20.9% was present as dissolved SO 4 2- and 46% was precipitated as S 0 . It is concluded that the AH reactor or micro-aeration carried out above the digestion zone of the UASB reactor favoured the biological oxidation of S 2- and the release of odourless effluents. Both technologies represent feasible and low-cost alternatives for the anaerobic treatment of domestic sewage.

Biodegradation of Methylene Blue Dye by Sequential Treatment Using Anaerobic Hybrid Reactor and Submerged Aerobic Fixed Film Bioreactor

NASA Astrophysics Data System (ADS)

Farooqi, Izharul H.; Basheer, Farrukh; Tiwari, Pradeepika

2017-12-01

Laboratory scale experiments were carried out to access the feasibility of sequential anaerobic/aerobic biological treatment for the biodegradation of Methylene Blue (MB) dye. Anaerobic studies were performed using anaerobic hybrid reactor (consisting of UASB and Anaerobic filter) whereas submerged aerobic fixed film reactor was used as aerobic reactor. Degradation of MB dye was attempted using neutralized acetic acid (1000 mg/L) as co-substrate. MB dye concentration was stepwise increased from 10 to 70 mg/L after reaching steady state in each dye concentration. Such a gradual increase in the dye concentration helps in the proper acclimatization of the sludge to dyes thereby avoiding the possible inhibitory effects to biological activities at high dye concentrations. The overall treatment efficiency of MB through sequential anaerobic-aerobic reactor operation was 90% at maximum attempted dye concentration of 70 mg/L. The effluent from anaerobic reactor was analysed for intermediate biodegradation products through HPLC. It was observed that catechol, quinone, amino pyrine, 1,4 diamino benzene were present. However they were absent in final effluent.

Waste tyre pyrolysis: modelling of a moving bed reactor.

PubMed

Aylón, E; Fernández-Colino, A; Murillo, R; Grasa, G; Navarro, M V; García, T; Mastral, A M

2010-12-01

This paper describes the development of a new model for waste tyre pyrolysis in a moving bed reactor. This model comprises three different sub-models: a kinetic sub-model that predicts solid conversion in terms of reaction time and temperature, a heat transfer sub-model that calculates the temperature profile inside the particle and the energy flux from the surroundings to the tyre particles and, finally, a hydrodynamic model that predicts the solid flow pattern inside the reactor. These three sub-models have been integrated in order to develop a comprehensive reactor model. Experimental results were obtained in a continuous moving bed reactor and used to validate model predictions, with good approximation achieved between the experimental and simulated results. In addition, a parametric study of the model was carried out, which showed that tyre particle heating is clearly faster than average particle residence time inside the reactor. Therefore, this fast particle heating together with fast reaction kinetics enables total solid conversion to be achieved in this system in accordance with the predictive model. Copyright © 2010 Elsevier Ltd. All rights reserved.

POWER-BURST FACILITY (PBF) CONCEPTUAL DESIGN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wasserman, A.A.; Johnson, S.O.; Heffner, R.E.

1963-06-21

A description is presented of the conceptual design of a high- performance, pulsed reactor called the Power Burst Facility (PBF). This reactor is designed to generate power bursts with initial asymptotic periods as short as 1 msec, producing energy releases large enough to destroy entire fuel subassemblies placed in a capsule or flow loop mounted in the reactor, all without damage to the reactor itself. It will be used primarily to evaluate the consequences and hazards of very rapid destructive accidents in reactors representing the entire range of current nuclear technology as applied to power generation, propulsion, and testing. Itmore » will also be used to carry out detailed studies of nondestructive reactivity feedback mechanisms in the shortperiod domain. The facility was designed to be sufficiently flexible to accommodate future cores of even more advanced design. The design for the first reactor core is based upon proven technology; hence, completion of the final design of this core will involve no significant development delays. Construction of the PBF is proposed to begin in September 1984, and is expected to take approximately 20 months to complete. (auth)« less

The Effect of Stochastic Perturbation of Fuel Distribution on the Criticality of a One Speed Reactor and the Development of Multi-Material Multinomial Line Statistics

NASA Technical Reports Server (NTRS)

Jahshan, S. N.; Singleterry, R. C.

2001-01-01

The effect of random fuel redistribution on the eigenvalue of a one-speed reactor is investigated. An ensemble of such reactors that are identical to a homogeneous reference critical reactor except for the fissile isotope density distribution is constructed such that it meets a set of well-posed redistribution requirements. The average eigenvalue, , is evaluated when the total fissile loading per ensemble element, or realization, is conserved. The perturbation is proven to increase the reactor criticality on average when it is uniformly distributed. The various causes of the change in reactivity, and their relative effects are identified and ranked. From this, a path towards identifying the causes. and relative effects of reactivity fluctuations for the energy dependent problem is pointed to. The perturbation method of using multinomial distributions for representing the perturbed reactor is developed. This method has some advantages that can be of use in other stochastic problems. Finally, some of the features of this perturbation problem are related to other techniques that have been used for addressing similar problems.

GUM Analysis for SIMS Isotopic Ratios in BEP0 Graphite Qualification Samples, Round 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerlach, David C.; Heasler, Patrick G.; Reid, Bruce D.

2009-01-01

This report describes GUM calculations for TIMS and SIMS isotopic ratio measurements of reactor graphite samples. These isotopic ratios are used to estimate reactor burn-up, and currently consist of various ratios of U, Pu, and Boron impurities in the graphite samples. The GUM calculation is a propagation of error methodology that assigns uncertainties (in the form of standard error and confidence bound) to the final estimates.

Diagnosing ion-beam targets, data acquisition, reactor conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mendel, Jr., C. W.

1982-01-01

The final lecture will discuss diagnostics of the target. These are very difficult because of the short times, small spatial extent, and extreme values of temperature and pressure. Diagnostics for temperature, density profile, and neutron production will be discussed. A few minutes will be devoted to data acquisition needs. The lecture will end with a discussion of current areas where improvements are needed and future diagnostics that will be required for reactor conditions.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Venkataraman, M.; Natarajan, R.; Raj, Baldev

The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR)more » spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)« less

Hydrodynamic study of an internal airlift reactor for microalgae culture.

PubMed

Rengel, Ana; Zoughaib, Assaad; Dron, Dominique; Clodic, Denis

2012-01-01

Internal airlift reactors are closed systems considered today for microalgae cultivation. Several works have studied their hydrodynamics but based on important solid concentrations, not with biomass concentrations usually found in microalgae cultures. In this study, an internal airlift reactor has been built and tested in order to clarify the hydrodynamics of this system, based on microalgae typical concentrations. A model is proposed taking into account the variation of air bubble velocity according to volumetric air flow rate injected into the system. A relationship between riser and downcomer gas holdups is established, which varied slightly with solids concentrations. The repartition of solids along the reactor resulted to be homogenous for the range of concentrations and volumetric air flow rate studied here. Liquid velocities increase with volumetric air flow rate, and they vary slightly when solids are added to the system. Finally, liquid circulation time found in each section of the reactor is in concordance with those employed in microalgae culture.

Co-cultivation of Lactobacillus zeae and Veillonella cricetifor the production of propionic acid

PubMed Central

2013-01-01

In this work a defined co-culture of the lactic acid bacterium Lactobacillus zeae and the propionate producer Veillonella criceti has been studied in continuous stirred tank reactor (CSTR) and in a dialysis membrane reactor. It is the first time that this reactor type is used for a defined co-culture fermentation. This reactor allows high mixing rates and working with high cell densities, making it ideal for co-culture investigations. In CSTR experiments the co-culture showed over a broad concentration range an almost linear correlation in consumption and production rates to the supply with complex nutrients. In CSTR and dialysis cultures a strong growth stimulation of L. zeae by V. criceti was shown. In dialysis cultures very high propionate production rates (0.61 g L-1h-1) with final titers up to 28 g L-1 have been realized. This reactor allows an individual, intracellular investigation of the co-culture partners by omic-technologies to provide a better understanding of microbial communities. PMID:23705662

ENGINEERING AND CONSTRUCTING THE HALLAM NUCLEAR POWER FACILITY REACTOR STRUCTURE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mahlmeister, J E; Haberer, W V; Casey, D F

1960-12-15

The Hallam Nuclear Power Facility reactor structure, including the cavity liner, is described, and the design philosophy and special design requirements which were developed during the preliminary and final engineering phases of the project are explained. The structure was designed for 600 deg F inlet and 1000 deg F outlet operating sodium temperatures and fabricated of austenitic and ferritic stainless steels. Support for the reactor core components and adequate containment for biological safeguards were readily provided even though quite conservative design philosophy was used. The calculated operating characteristics, including heat generation, temperature distributions and stress levels for full-power operation, aremore » summarized. Ship fabrication and field installation experiences are also briefly related. Results of this project have established that the sodium graphite reactor permits practical and economical fabrication and field erection procedures; considerably higher operating design temperatures are believed possible without radical design changes. Also, larger reactor structures can be similarly constructed for higher capacity (300 to 1000 Mwe) nuclear power plants. (auth)« less

Successive and large-scale synthesis of InP/ZnS quantum dots in a hybrid reactor and their application to white LEDs

NASA Astrophysics Data System (ADS)

Kim, Kyungnam; Jeong, Sohee; Woo, Ju Yeon; Han, Chang-Soo

2012-02-01

We report successive and large-scale synthesis of InP/ZnS core/shell nanocrystal quantum dots (QDs) using a customized hybrid flow reactor, which is based on serial combination of a batch-type mixer and a flow-type furnace. InP cores and InP/ZnS core/shell QDs were successively synthesized in the hybrid reactor in a simple one-step process. In this reactor, the flow rate of the solutions was typically 1 ml min-1, 100 times larger than that of conventional microfluidic reactors. In order to synthesize high-quality InP/ZnS QDs, we controlled both the flow rate and the crystal growth temperature. Finally, we obtained high-quality InP/ZnS QDs in colors from bluish green to red, and we demonstrated that these core/shell QDs could be incorporated into white-light-emitting diode (LED) devices to improve color rendering performance.

Successive and large-scale synthesis of InP/ZnS quantum dots in a hybrid reactor and their application to white LEDs.

PubMed

Kim, Kyungnam; Jeong, Sohee; Woo, Ju Yeon; Han, Chang-Soo

2012-02-17

We report successive and large-scale synthesis of InP/ZnS core/shell nanocrystal quantum dots (QDs) using a customized hybrid flow reactor, which is based on serial combination of a batch-type mixer and a flow-type furnace. InP cores and InP/ZnS core/shell QDs were successively synthesized in the hybrid reactor in a simple one-step process. In this reactor, the flow rate of the solutions was typically 1 ml min(-1), 100 times larger than that of conventional microfluidic reactors. In order to synthesize high-quality InP/ZnS QDs, we controlled both the flow rate and the crystal growth temperature. Finally, we obtained high-quality InP/ZnS QDs in colors from bluish green to red, and we demonstrated that these core/shell QDs could be incorporated into white-light-emitting diode (LED) devices to improve color rendering performance.

Anaerobic digestion of pressed off leachate from the organic fraction of municipal solid waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nayono, Satoto E.; Institute of Biology for Engineers and Biotechnology of Wastewater, University of Karlsruhe, Am Fasanengarten, 76131 Karlsruhe; Winter, Josef, E-mail: josef.winter@iba.uka.d

2010-10-15

A highly polluted liquid ('press water') was obtained from the pressing facility for the organic fraction of municipal solid waste in a composting plant. Methane productivity of the squeezed-off leachate was investigated in batch assays. To assess the technical feasibility of 'press water' as a substrate for anaerobic digestion, a laboratory-scale glass column reactor was operated semi-continuously at 37 {sup o}C. A high methane productivity of 270 m{sup -3} CH{sub 4} ton{sup -1} COD{sub added} or 490 m{sup -3} CH{sub 4} ton{sup -1} VS{sub added} was achieved in the batch experiment. The semi-continuously run laboratory-scale reactor was initially operated atmore » an organic loading rate of 10.7 kg COD m{sup -3} d{sup -1}. The loading was increased to finally 27.7 kg COD m{sup -3} d{sup -1}, corresponding to a reduction of the hydraulic retention time from initially 20 to finally 7.7 days. During the digestion, a stable elimination of organic material (measured as COD elimination) of approximately 60% was achieved. Linearly with the increment of the OLR, the volumetric methane production of the reactor increased from 2.6 m{sup 3} m{sub reactor}{sup -3} d{sup -1} to 7.1 m{sup 3} m{sub reactor}{sup -3} d{sup -1}. The results indicated that 'press water' from the organic fraction of municipal solid waste was a suitable substrate for anaerobic digestion which gave a high biogas yield even at very high loading rates.« less

RADSOURCE. Volume 1, Part 1, A scaling factor prediction computer program technical manual and code validation: Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vance, J.N.; Holderness, J.H.; James, D.W.

1992-12-01

Waste stream scaling factors based on sampling programs are vulnerable to one or more of the following factors: sample representativeness, analytic accuracy, and measurement sensitivity. As an alternative to sample analyses or as a verification of the sampling results, this project proposes the use of the RADSOURCE code, which accounts for the release of fuel-source radionuclides. Once the release rates of these nuclides from fuel are known, the code develops scaling factors for waste streams based on easily measured Cobalt-60 (Co-60) and Cesium-137 (Cs-137). The project team developed mathematical models to account for the appearance rate of 10CFR61 radionuclides inmore » reactor coolant. They based these models on the chemistry and nuclear physics of the radionuclides involved. Next, they incorporated the models into a computer code that calculates plant waste stream scaling factors based on reactor coolant gamma- isotopic data. Finally, the team performed special sampling at 17 reactors to validate the models in the RADSOURCE code.« less

Optimization of post-column reactor radius in capillary high performance liquid chromatography Effect of chromatographic column diameter and particle diameter

PubMed Central

Xu, Hongjuan; Weber, Stephen G.

2006-01-01

A post-column reactor consisting of a simple open tube (Capillary Taylor Reactor) affects the performance of a capillary LC in two ways: stealing pressure from the column and adding band spreading. The former is a problem for very small radius reactors, while the latter shows itself for large reactor diameters. We derived an equation that defines the observed number of theoretical plates (Nobs) taking into account the two effects stated above. Making some assumptions and asserting certain conditions led to a final equation with a limited number of variables, namely chromatographic column radius, reactor radius and chromatographic particle diameter. The assumptions and conditions are that the van Deemter equation applies, the mass transfer limitation is for intraparticle diffusion in spherical particles, the velocity is at the optimum, the analyte’s retention factor, k′, is zero, the post-column reactor is only long enough to allow complete mixing of reagents and analytes and the maximum operating pressure of the pumping system is used. Optimal ranges of the reactor radius (ar) are obtained by comparing the number of observed theoretical plates (and theoretical plates per time) with and without a reactor. Results show that the acceptable reactor radii depend on column diameter, particle diameter, and maximum available pressure. Optimal ranges of ar become narrower as column diameter increases, particle diameter decreases or the maximum pressure is decreased. When the available pressure is 4000 psi, a Capillary Taylor Reactor with 12 μm radius is suitable for all columns smaller than 150 μm (radius) packed with 2–5 μm particles. For 1 μm packing particles, only columns smaller than 42.5 μm (radius) can be used and the reactor radius needs to be 5 μm. PMID:16494886

Biodegradation of endocrine disruptors in urban wastewater using Pleurotus ostreatus bioreactor.

PubMed

Křesinová, Zdena; Linhartová, Lucie; Filipová, Alena; Ezechiáš, Martin; Mašín, Pavel; Cajthaml, Tomáš

2018-07-25

The white rot fungus Pleurotus ostreatus HK 35, which is also an edible industrial mushroom commonly cultivated in farms, was tested in the degradation of typical representatives of endocrine disrupters (EDCs; bisphenol A, estrone, 17β-estradiol, estriol, 17α-ethinylestradiol, triclosan and 4-n-nonylphenol); its degradation efficiency under model laboratory conditions was greater than 90% within 12 days and better than that of another published strain P. ostreatus 3004. A spent mushroom substrate from a local farm was tested for its applicability in various batch and trickle-bed reactors in degrading EDCs in model fortified and real communal wastewater. The reactors were tested under various regimes including a pilot-scale trickle-bed reactor, which was finally tested at a wastewater treatment plant. The result revealed that the spent substrate is an efficient biodegradation agent, where the fungus was usually able to remove about 95% of EDCs together with suppression of the estrogenic activity of the sample. The results showed the fungus was able to operate in the presence of bacterial microflora in wastewater without any substantial negative effects on the degradation abilities. Finally, a pilot-scale trickle-bed reactor was installed in a wastewater treatment plant and successfully operated for 10days, where the bioreactor was able to remove more than 76% of EDCs present in the wastewater. Copyright © 2017 Elsevier B.V. All rights reserved.

76 FR 77025 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Section 1.0, Revision...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-09

... the Final Staff Guidance Section 1.0, Revision 2 on Introduction and Interfaces AGENCY: Nuclear... Plants: LWR Edition,'' Section 1.0, Revision 2 on ``Introduction and Interfaces'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML112730393). The NRC staff issues revisions to SRP...

Method and apparatus for producing synthesis gas

DOEpatents

Hemmings, John William; Bonnell, Leo; Robinson, Earl T.

2010-03-03

A method and apparatus for reacting a hydrocarbon containing feed stream by steam methane reforming reactions to form a synthesis gas. The hydrocarbon containing feed is reacted within a reactor having stages in which the final stage from which a synthesis gas is discharged incorporates expensive high temperature materials such as oxide dispersed strengthened metals while upstream stages operate at a lower temperature allowing the use of more conventional high temperature alloys. Each of the reactor stages incorporate reactor elements having one or more separation zones to separate oxygen from an oxygen containing feed to support combustion of a fuel within adjacent combustion zones, thereby to generate heat to support the endothermic steam methane reforming reactions.

Research on pressure control of pressurizer in pressurized water reactor nuclear power plant

NASA Astrophysics Data System (ADS)

Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang

2010-07-01

Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor.

PubMed

Essadki, A H; Gourich, B; Vial, Ch; Delmas, H; Bennajah, M

2009-09-15

Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12mA/cm(2), but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H(2) microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

Utilization of 134Cs/137Cs in the environment to identify the reactor units that caused atmospheric releases during the Fukushima Daiichi accident

NASA Astrophysics Data System (ADS)

Chino, Masamichi; Terada, Hiroaki; Nagai, Haruyasu; Katata, Genki; Mikami, Satoshi; Torii, Tatsuo; Saito, Kimiaki; Nishizawa, Yukiyasu

2016-08-01

The Fukushima Daiichi nuclear power reactor units that generated large amounts of airborne discharges during the period of March 12-21, 2011 were identified individually by analyzing the combination of measured 134Cs/137Cs depositions on ground surfaces and atmospheric transport and deposition simulations. Because the values of 134Cs/137Cs are different in reactor units owing to fuel burnup differences, the 134Cs/137Cs ratio measured in the environment was used to determine which reactor unit ultimately contaminated a specific area. Atmospheric dispersion model simulations were used for predicting specific areas contaminated by each dominant release. Finally, by comparing the results from both sources, the specific reactor units that yielded the most dominant atmospheric release quantities could be determined. The major source reactor units were Unit 1 in the afternoon of March 12, 2011, Unit 2 during the period from the late night of March 14 to the morning of March 15, 2011. These results corresponded to those assumed in our previous source term estimation studies. Furthermore, new findings suggested that the major source reactors from the evening of March 15, 2011 were Units 2 and 3 and that the dominant source reactor on March 20, 2011 temporally changed from Unit 3 to Unit 2.

Improving biogas production from continuous co-digestion of oily wastewater and waste-activated sludge by hydrodynamic cavitation pre-treatment.

PubMed

Habashi, Nima; Alighardashi, Abolghasem; Mennerich, Artur; Mehrdadi, Nasser; Torabian, Ali

2018-04-01

Hydrodynamic cavitation (HC) was evaluated as a pretreatment for synthetic oily wastewater (OWW) to be co-digested with waste-activated sludge (WAS). The main objective of the present research was the enhancement of biogas production by the application of HC pretreatment. HC was applied to the OWW, and the OWW and WAS were added to a 50 L continuous digestion reactor. As a control system, an identical digestion reactor was set up for co-digestion of the WAS and the OWW without pretreatment. The reactors were initially filled with inoculum and the hydraulic retention time (HRT) was set to 22 d. The HRT was gradually reduced to 19, 16, and finally 13 d, but the substrate quality was kept constant. The loading rate, accordingly, increased from 0.86 to 1.46 g TVS/(L d). The biogas volume was recorded online and its quality was analyzed regularly. The HC improved biogas production up to 43% at 22 d of HRT. Reducing the HRT decreased biogas production from the main reactor while that of the control reactor was more or less constant. HC also increased the biogas methane content; the methane concentration of the main reactor was about 3% higher than the methane concentration of the control reactor. The main reactor experienced no clogging or accumulation of fatty materials.

Utilization of (134)Cs/(137)Cs in the environment to identify the reactor units that caused atmospheric releases during the Fukushima Daiichi accident.

PubMed

Chino, Masamichi; Terada, Hiroaki; Nagai, Haruyasu; Katata, Genki; Mikami, Satoshi; Torii, Tatsuo; Saito, Kimiaki; Nishizawa, Yukiyasu

2016-08-22

The Fukushima Daiichi nuclear power reactor units that generated large amounts of airborne discharges during the period of March 12-21, 2011 were identified individually by analyzing the combination of measured (134)Cs/(137)Cs depositions on ground surfaces and atmospheric transport and deposition simulations. Because the values of (134)Cs/(137)Cs are different in reactor units owing to fuel burnup differences, the (134)Cs/(137)Cs ratio measured in the environment was used to determine which reactor unit ultimately contaminated a specific area. Atmospheric dispersion model simulations were used for predicting specific areas contaminated by each dominant release. Finally, by comparing the results from both sources, the specific reactor units that yielded the most dominant atmospheric release quantities could be determined. The major source reactor units were Unit 1 in the afternoon of March 12, 2011, Unit 2 during the period from the late night of March 14 to the morning of March 15, 2011. These results corresponded to those assumed in our previous source term estimation studies. Furthermore, new findings suggested that the major source reactors from the evening of March 15, 2011 were Units 2 and 3 and that the dominant source reactor on March 20, 2011 temporally changed from Unit 3 to Unit 2.

Utilization of 134Cs/137Cs in the environment to identify the reactor units that caused atmospheric releases during the Fukushima Daiichi accident

PubMed Central

Chino, Masamichi; Terada, Hiroaki; Nagai, Haruyasu; Katata, Genki; Mikami, Satoshi; Torii, Tatsuo; Saito, Kimiaki; Nishizawa, Yukiyasu

2016-01-01

The Fukushima Daiichi nuclear power reactor units that generated large amounts of airborne discharges during the period of March 12–21, 2011 were identified individually by analyzing the combination of measured 134Cs/137Cs depositions on ground surfaces and atmospheric transport and deposition simulations. Because the values of 134Cs/137Cs are different in reactor units owing to fuel burnup differences, the 134Cs/137Cs ratio measured in the environment was used to determine which reactor unit ultimately contaminated a specific area. Atmospheric dispersion model simulations were used for predicting specific areas contaminated by each dominant release. Finally, by comparing the results from both sources, the specific reactor units that yielded the most dominant atmospheric release quantities could be determined. The major source reactor units were Unit 1 in the afternoon of March 12, 2011, Unit 2 during the period from the late night of March 14 to the morning of March 15, 2011. These results corresponded to those assumed in our previous source term estimation studies. Furthermore, new findings suggested that the major source reactors from the evening of March 15, 2011 were Units 2 and 3 and that the dominant source reactor on March 20, 2011 temporally changed from Unit 3 to Unit 2. PMID:27546490

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Powers, Jeffrey J.; Mueller, Don

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy researchmore » and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.« less

Severe accident skyshine radiation analysis by MCNP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eurajoki, T.

1994-12-31

If a severe accident with a considerable core damage occurs at a nuclear power plant whose containment top is remarkably thin compared with the walls, the radiation transported through the top and scattered in air may cause high dose rates at the power plant area. Noble gases and other fission products released to the containment act as sources. The dose rates caused by skyshine have been calculated by MCNP3A for the Loviisa nuclear power plant (two-unit, 445-MW VVER) for the outside area and inside some buildings, taking the attenuation in the roofs of the buildings into account.

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

DOE Office of Scientific and Technical Information (OSTI.GOV)

John Darrell Bess

2008-06-01

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, suchmore » as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.« less

Features of postfailure fuel behavior in transient overpower and transient undercooled/overpower tests in the transient reactor test facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doerner, R.C.; Bauer, T.H.; Morman, J.A.

Prototypic oxide fuel was subjected to simulated, fast reactor severe accident conditions in a series of in-pile tests in the Transient Reactor Test Facility reactor. Seven experiments were performed on fresh and previously irradiated oxide fuel pins under transient overpower and transient undercooled. overpower accident conditions. For each of the tests, fuel motions were observed by the hodoscope. Hodoscope data are correlated with coolant flow, pressure, and temperature data recorded by the loop instrumentation. Data were analyzed from the onset of initial failure to a final mass distribution at the end of the test. In this paper results of thesemore » analyses are compared to pre- and posttest accident calculations and to posttest metallographic accident calculations and to posttest metallographic examinations and computed tomographic reconstructions from neutron radiographs.« less

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D&D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D&D plans for the turbine building were prepared from 1979 through 1990. D&D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and the absence of loosemore » contamination, the D&D activities were completed with no radiation exposure to the workers. The D&D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D D plans for the turbine building were prepared from 1979 through 1990. D D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and themore » absence of loose contamination, the D D activities were completed with no radiation exposure to the workers. The D D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Production of edible carbohydrates from formaldehyde in a spacecraft. pH variations in the calcium hydroxide catalyzed formose reaction. Final Report, 1 Jul. 1973 - 30 Jun. 1974. M.S. Thesis

NASA Technical Reports Server (NTRS)

Weiss, A. H.; Kohler, J. T.; John, T.

1974-01-01

The study of the calcium hydroxide catalyzed condensation of formaldehyde was extended to a batch reactor system. Decreases in pH were observed, often in the acid regime, when using this basic catalyst. This observation was shown to be similar to results obtained by others using less basic catalysts in the batch mode. The relative rates of these reactions are different in a batch reactor than in a continuous stirred tank reactor. This difference in relative rates is due to the fact that at any degree of advancement in the batch system, the products have a history of previous products, pH, and dissolved catalyst. The relative rate differences can be expected to yield a different nature of product sugars for the two types of reactors.

Nuclear fuels - Present and future

NASA Astrophysics Data System (ADS)

Olander, D.

2009-06-01

The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

TEMPEST code simulations of hydrogen distribution in reactor containment structures. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trent, D.S.; Eyler, L.L.

The mass transport version of the TEMPEST computer code was used to simulate hydrogen distribution in geometric configurations relevant to reactor containment structures. Predicted results of Battelle-Frankfurt hydrogen distribution tests 1 to 6, and 12 are presented. Agreement between predictions and experimental data is good. Best agreement is obtained using the k-epsilon turbulence model in TEMPEST in flow cases where turbulent diffusion and stable stratification are dominant mechanisms affecting transport. The code's general analysis capabilities are summarized.

Computational Modeling as a Design Tool in Microelectronics Manufacturing

NASA Technical Reports Server (NTRS)

Meyyappan, Meyya; Arnold, James O. (Technical Monitor)

1997-01-01

Plans to introduce pilot lines or fabs for 300 mm processing are in progress. The IC technology is simultaneously moving towards 0.25/0.18 micron. The convergence of these two trends places unprecedented stringent demands on processes and equipments. More than ever, computational modeling is called upon to play a complementary role in equipment and process design. The pace in hardware/process development needs a matching pace in software development: an aggressive move towards developing "virtual reactors" is desirable and essential to reduce design cycle and costs. This goal has three elements: reactor scale model, feature level model, and database of physical/chemical properties. With these elements coupled, the complete model should function as a design aid in a CAD environment. This talk would aim at the description of various elements. At the reactor level, continuum, DSMC(or particle) and hybrid models will be discussed and compared using examples of plasma and thermal process simulations. In microtopography evolution, approaches such as level set methods compete with conventional geometric models. Regardless of the approach, the reliance on empricism is to be eliminated through coupling to reactor model and computational surface science. This coupling poses challenging issues of orders of magnitude variation in length and time scales. Finally, database development has fallen behind; current situation is rapidly aggravated by the ever newer chemistries emerging to meet process metrics. The virtual reactor would be a useless concept without an accompanying reliable database that consists of: thermal reaction pathways and rate constants, electron-molecule cross sections, thermochemical properties, transport properties, and finally, surface data on the interaction of radicals, atoms and ions with various surfaces. Large scale computational chemistry efforts are critical as experiments alone cannot meet database needs due to the difficulties associated with such controlled experiments and costs.

Hydrogel coated monoliths for enzymatic hydrolysis of penicillin G

PubMed Central

Smeltink, M. W.; Straathof, A. J. J.; Paasman, M. A.; van de Sandt, E. J. A. X.; Kapteijn, F.; Moulijn, J. A.

2008-01-01

The objective of this work was to develop a hydrogel-coated monolith for the entrapment of penicillin G acylase (E. coli, PGA). After screening of different hydrogels, chitosan was chosen as the carrier material for the preparation of monolithic biocatalysts. This protocol leads to active immobilized biocatalysts for the enzymatic hydrolysis of penicillin G (PenG). The monolithic biocatalyst was tested in a monolith loop reactor (MLR) and compared with conventional reactor systems using free PGA, and a commercially available immobilized PGA. The optimal immobilization protocol was found to be 5 g l−1 PGA, 1% chitosan, 1.1% glutaraldehyde and pH 7. Final PGA loading on glass plates was 29 mg ml−1 gel. For 400 cpsi monoliths, the final PGA loading on functionalized monoliths was 36 mg ml−1 gel. The observed volumetric reaction rate in the MLR was 0.79 mol s−1 m−3monolith. Apart from an initial drop in activity due to wash out of PGA at higher ionic strength, no decrease in activity was observed after five subsequent activity test runs. The storage stability of the biocatalysts is at least a month without loss of activity. Although the monolithic biocatalyst as used in the MLR is still outperformed by the current industrial catalyst (immobilized preparation of PGA, 4.5 mol s−1 m−3catalyst), the rate per gel volume is slightly higher for monolithic catalysts. Good activity and improved mechanical strength make the monolithic bioreactor an interesting alternative that deserves further investigation for this application. Although moderate internal diffusion limitations have been observed inside the gel beads and in the gel layer on the monolith channel, this is not the main reason for the large differences in reactor performance that were observed. The pH drop over the reactor as a result of the chosen method for pH control results in a decreased performance of both the MLR and the packed bed reactor compared to the batch system. A different reactor configuration including an optimal pH profile is required to increase the reactor performance. The monolithic stirrer reactor would be an interesting alternative to improve the performance of the monolith-PGA combination. PMID:18427849

Two-phase anaerobic digestion of source sorted OFMSW (organic fraction of municipal solid waste): performance and kinetic study.

PubMed

Pavan, P; Battistoni, P; Cecchi, F; Mata-Alvarez, J

2000-01-01

The results of a two-phase system operated in different conditions, treating the source-sorted organic fraction of municipal solid waste (SS-OFMSW), coming mainly from fruit and vegetable markets, are presented. Hydraulic retention time (HRT) in the hydrolytic reactor and in the methanogenic reactor and also the temperature in the hydrolytic reactor (mesophilic and thermophilic conditions) are varied in order to evaluate the effect of these factors. The methanogenic reactor is always operated within the thermophilic range. Optimum operating conditions are found to be around 12 days (total system) using the mesophilic range of temperature in the first reactor. Specific gas production (SGP) in these conditions is around 0.6 m3/kg TVS. A kinetic study is also carried out, using the first and the step diffusional models. The latter gives much better results, with fitted constants comparable to other studies. Finally, a comparison with a one-phase system is carried out, showing that a two-phase system is much more appropriate for the digestion of this kind of highly biodegradable substrate in thermophilic conditions.

Production of oxalic acid from sugar beet molasses by formed nitrogen oxides.

PubMed

Gürü, M; Bilgesü, A Y; Pamuk, V

2001-03-01

Production of oxalic acid from sugar beet molasses was developed in a series of three reactors. Nitrogen oxides formed were used to manufacture oxalic acid in the second and third reactor. Parameters affecting the reaction were determined to be, air flow rate, temperature, the amount of V2O5 catalyst and the concentrations of molasses and H2SO4. The maximum yields in the second and third reactors were 78.9% and 74.6% of theoretical yield, respectively. Also, kinetic experiments were performed and the first-order rate constants were determined for the glucose consumption rate. Nitrogen oxides in off-gases from the final reactor were absorbed in water and concentrated sulphuric acid and reused in the following reactors giving slightly lower yields under similar conditions. In this novel way, it was possible to recover NO(x) and to prevent air pollution. Meanwhile, it was possible to reduce the unit cost of reactant for oxalic acid production. A maximum 77.5% and 74.1% of theoretical yield was obtained by using the absorption solutions with NO(x).

Improved measurement of the reactor antineutrino flux and spectrum at Daya Bay

DOE PAGES

An, F. P.; Balantekin, A. B.; Band, H. R.; ...

2017-01-01

Here, a new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay reactor neutrino experiment is reported. The antineutrinos were generated by six 2.9 GW th nuclear reactors and detected by eight antineutrino detectors deployed in two near (560 m and 600 m flux-weighted baselines) and one far (1640 m flux-weighted baseline) underground experimental halls. With 621 days of data, more than 1.2 million inverse beta decay (IBD) candidates were detected. The IBD yield in the eight detectors was measured, and the ratio of measured to predicted flux was found to be 0.946 ± 0.020 (0.992more » ± 0.021) for the Huber+Mueller (ILL+Vogel) model. A 2.9σ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4$-$6 MeV was found in the measured spectrum, with a local significance of 4.4σ. Finally, a reactor antineutrino spectrum weighted by the IBD cross section is extracted for model-independent predictions.« less

Review of heat transfer problems associated with magnetically-confined fusion reactor concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hoffman, M.A.; Werner, R.W.; Carlson, G.A.

1976-04-01

Conceptual design studies of possible fusion reactor configurations have revealed a host of interesting and sometimes extremely difficult heat transfer problems. The general requirements imposed on the coolant system for heat removal of the thermonuclear power from the reactor are discussed. In particular, the constraints imposed by the fusion plasma, neutronics, structure and magnetic field environment are described with emphasis on those aspects which are unusual or unique to fusion reactors. Then the particular heat transfer characteristics of various possible coolants including lithium, flibe, boiling alkali metals, and helium are discussed in the context of these general fusion reactor requirements.more » Some specific areas where further experimental and/or theoretical work is necessary are listed for each coolant along with references to the pertinent research already accomplished. Specialized heat transfer problems of the plasma injection and removal systems are also described. Finally, the challenging heat transfer problems associated with the superconducting magnets are reviewed, and once again some of the key unsolved heat transfer problems are enumerated.« less

Types and analysis of defects in welding junctions of the header to steam generator shells on power-generating units with VVER-1000

NASA Astrophysics Data System (ADS)

Ozhigov, L. S.; Voevodin, V. N.; Mitrofanov, A. S.; Vasilenko, R. L.

2016-10-01

Investigation objects were metal templates, which were cut during the repair of welding junction no. 111 (header to the steam generator shell) on a power-generating unit with VVER-1000 of the South-Ukraine NPP, and substances of mud depositions collected from walls of this junction. Investigations were carried out using metallography, optical microscopy, and scanning electron microscopy with energy dispersion microanalysis by an MMO-1600-AT metallurgical microscope and a JEOL JSM-7001F scanning electron microscope with the Shottky cathode. As a result of investigations in corrosion pits and mud depositions in the area of welding junction no. 111, iron and copper-enriched particles were revealed. It is shown that, when contacting with the steel header surface, these particles can form microgalvanic cells causing reactions of iron dissolution and the pit corrosion of metal. Nearby corrosion pits in metal are microcracks, which can be effect of the stress state of metal under corrosion pits along with revealed effects of twinning. The hypothesis is expressed that pitting corrosion of metal occurred during the first operation period of the power-generating unit in the ammonia water chemistry conditions (WCC). The formation of corrosion pits and nucleating cracks from them was stopped with the further operation under morpholine WCC. The absence of macrocracks in metal of templates verifies that, during operation, welding junction no. 111 operated under load conditions not exceeding the permissible ones by design requirements. The durability of the welding junction of the header to the steam generator shell significantly depends on the technological schedule of chemical cleaning and steam generator shut-down cooling.

78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-19

... 100 Grams,'' dated October 10, 2002, and subsequent similar security orders issued to licensees... than 100 grams from theft, diversion, or radiological sabotage. The requirements in the final rule...

Osiris and SOMBRERO inertial confinement fusion power plant designs. Volume 2, Designs, assessments, and comparisons, Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meier, W.R.; Bieri, R.L.; Monsler, M.J.

1992-03-01

The primary objective of the of the IFE Reactor Design Studies was to provide the Office of Fusion Energy with an evaluation of the potential of inertial fusion for electric power production. The term reactor studies is somewhat of a misnomer since these studies included the conceptual design and analysis of all aspects of the IFE power plants: the chambers, heat transport and power conversion systems, other balance of plant facilities, target systems (including the target production, injection, and tracking systems), and the two drivers. The scope of the IFE Reactor Design Studies was quite ambitious. The majority of ourmore » effort was spent on the conceptual design of two IFE electric power plants, one using an induction linac heavy ion beam (HIB) driver and the other using a Krypton Fluoride (KrF) laser driver. After the two point designs were developed, they were assessed in terms of their (1) environmental and safety aspects; (2) reliability, availability, and maintainability; (3) technical issues and technology development requirements; and (4) economics. Finally, we compared the design features and the results of the assessments for the two designs.« less

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Technical Reports Server (NTRS)

Martin, James; Salvail, Pat

2003-01-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final "wet in". A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/- 1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to less than10(exp -10) std cc/sec helium and vacuum conditioned at 250 C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Astrophysics Data System (ADS)

Martin, James; Salvail, Pat

2004-02-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final ``wet in''. A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/-1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to <10-10 std cc/sec helium and vacuum conditioned at 250 °C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 °C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

NPR design basis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Locke, G.L.

1958-09-08

The design basis is composed of requirements and conditions for the design of the reactor plant (composed of the reactor and heat dissipation system). Its intent is to insure that the final product meets the economic, safety, and technical objectives of the project. The design basis is dependent on the ground rules, objectives, technical criteria, and practical design considerations. This document is being issued with the understanding that these items are not yet firmly established in all respects, and therefore, the numbers put down here are subject to change. Consideration of the spectrum of probable changes that might be mademore » leads to the conclusion that the numbers here are close to the final ones and are satisfactory as a basis for the initial stages of design. Some numbers are omitted because of insufficient data at this time.« less

Maintenance method and its critical issues for a fast-ignition laser fusion reactor based on a dry wall chamber

NASA Astrophysics Data System (ADS)

Someya, Y.; Matsumoto, T.; Okano, K.; Asaoka, Y.; Hiwatari, R.; Goto, T.; Ogawa, Y.

2008-05-01

The neutronics analysis has been carried out for feasibility study of the FALCON-D concept by Monte Carlo N-paticle transport code (MCNP), in order to inspect the cooling performance of in-vessel and ex-vessel components, and a connection pipe between Vacuum Vessel and reactor room. The nuclear heating rate in the Vacuum Vessel was at the same level as that of NBI duct of the ITER. The temperature of the connection pipe was found to be 345·, ·which was smaller than the melting point of structure materials (F82H). Moreover, the radiation damage of the final optics was also investigated. We propose a sliding changer concept for replacement. This method could be adapted for the replacement of one FPY cycle in the final optics system.

Final summary report for 1989 inservice inspection (ISI) of SRS (Savannah River Site) 100-P Reactor tank

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morrison, J.M.; Loibl, M.W.

1989-12-15

The integrity of the SRS reactor tanks is a key factor affecting their suitability for continued service since, unlike the external piping system and components, the tanks are virtually irreplaceable. Cracking in various areas of the process water piping systems has occurred beginning in 1960 as a result of several degradation mechanisms, chiefly intergranular stress corrosion cracking (IGSCC) and chloride-induced transgranular cracking. IGSCC, currently the primary degradation mechanism, also occurred in the knuckle'' region (tank wall-to-bottom tube sheet transition piece) unique to C Reactor and was eventually responsible for that reactor being deactivated in 1985. A program of visual examinationsmore » of the SRS reactor tanks was initiated in 1968, which used a specially designed immersible periscope. Under that program the condition of the accessible tank welds and associated heat affected zones (HAZ) was evaluated on a five-year frequency. Prior to 1986, the scope of these inspections comprised approximately 20 percent of the accessible weld area. In late 1986 and early 1987 the scope of the inspections was expanded and a 100 percent visual inspection of accessible welds was performed of the P-, L-, and K-Reactor tanks. Supplemental dye penetrant examinations were performed in L Reactor on selected areas which showed visual indications. No evidence of cracking was detected in any of these inspections of the P-, L-, and K-Reactor tanks. 17 refs., 7 figs.« less

Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mckerley, Bill; Bustamante, Jacqueline M; Costa, David A

2010-01-01

We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is inmore » support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to consolidate the locations of special nuclear material (SNM) to reduce the cost of securing many SNM facilities. The removal of SPR-II from SNL was a significant accomplishment in SNL's de-inventory efforts and played a key role in reducing the number of locations requiring the expensive security measures required for category 1 and 2 SNM facilities. A similar pulse reactor was fabricated at the Y-12 National Security Complex beginning in the late 1960's. This Aberdeen Pulse Reactor (APR) was operated at the Army Pulse Radiation Facility (APRF) located at the Aberdeen Test Center (ATC) in Maryland. When the APRF was shut down in 2003, a portion of the DOE-owned Special Nuclear Material (SNM) was shipped to an interim facility for storage. Subsequently, the DOE determined that the material from both the SPR-II and the APR would be processed in the H-Canyon at the Savannah River Site (SRS). Because of the SRS receipt requirements some of the material was sent to the Los Alamos National Laboratory (LANL) for size-reduction prior to shipment to the SRS for final disposition.« less

The PROSPECT physics program

DOE PAGES

Ashenfelter, J.; Balantekin, A. B.; Band, H. R.; ...

2016-10-17

The precision reactor oscillation and spectrum experiment, PROSPECT, is designed to make a precise measurement of the antineutrino spectrum from a highly-enriched uranium reactor and probe eV-scale sterile neutrinos by searching for neutrino oscillations over a distance of several meters. The subject of this paper, PROSPECT, is conceived as a 2-phase experiment utilizing segmented 6Li-doped liquid scintillator detectors for both efficient detection of reactor antineutrinos through the inverse beta decay reaction and excellent background discrimination. PROSPECT Phase I consists of a movable 3 ton antineutrino detector at distances of 7–12 m from the reactor core. It will probe the best-fitmore » point of the ν e disappearance experiments at 4σ in 1 year and the favored region of the sterile neutrino parameter space at > 3σ in 3 years. With a second antineutrino detector at 15–19 m from the reactor, Phase II of PROSPECT can probe the entire allowed parameter space below 10 eV 2 at 5σ in 3 additional years. Finally, the measurement of the reactor antineutrino spectrum and the search for short-baseline oscillations with PROSPECT will test the origin of the spectral deviations observed in recent θ 13 experiments, search for sterile neutrinos, and conclusively address the hypothesis of sterile neutrinos as an explanation of the reactor anomaly.« less

75 FR 69709 - Office of New Reactors; Notice of Availability of the Final Staff Guidance; Standard Review Plan...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-15

... the Final Staff Guidance; Standard Review Plan, Section 13.6.6, Revision 0 on Cyber Security Plan... Reports for Nuclear Power Plants,'' Section 13.6.6, Revision 0 on ``Cyber Security Plan'' (Agencywide.... Disposition: On May 13, 2010, the NRC staff issued the proposed Revision 0 on SRP Section 13.6.6 on ``Cyber...

Project Luna Succendo: The Lunar Evolutionary Growth-Optimized (LEGO) Reactor

NASA Astrophysics Data System (ADS)

Bess, John Darrell

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched within lunar shipments from the Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides 5 kWe using a free-piston Stirling space converter. The overall envelope for a single unit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. The subunits can be placed with centerline distances of approximately 0.6 m in a hexagonal-lattice pattern to provide sufficient neutronic coupling while allowing room for heat rejection and interstitial control. A lattice of six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network Future improvements include advances in reactor control methods, fuel form and matrix, determination of shielding requirements, as well as power conversion and heat rejection techniques to generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces such as Mars, other moons, and asteroids.

Co-composting of palm oil mill sludge-sawdust.

PubMed

Yaser, Abu Zahrim; Abd Rahman, Rakmi; Kalil, Mohd Sahaid

2007-12-15

Composting of Palm Oil Mill Sludge (POMS) with sawdust was conducted in natural aerated reactor. Composting using natural aerated reactor is cheap and simple. The goal of this study is to observe the potential of composting process and utilizing compost as media for growing Cymbopogun citratus, one of Malaysia herbal plant. The highest maximum temperature achieved is about 40 degrees C and to increase temperature bed, more biodegradable substrate needs to be added. The pH value decrease along the process with final pH compost is acidic (pH 5.7). The highest maximum organic losses are about 50% with final C/N ratio of the compost is about 19. Final compost also showed some fertilizing value but need to be adjusted to obtain an ideal substrate. Addition of about 70% sandy soil causes highest yield and excellent root development for C. citratus in potted media. Beside that, compost from POMS-sawdust also found to have fertilizer value and easy to handle. Composting of POMS with sawdust shows potential as an alternative treatment to dispose and recycle waste components.

Design improvement and performance evaluation of solar photocatalytic reactor for industrial effluent treatment.

PubMed

Nair, Ranjith G; Bharadwaj, P J; Samdarshi, S K

2016-12-01

This work reports the details of the design components and materials used in a linear compound parabolic trough reactor constructed with an aim to use the photocatalyst for solar photocatalytic applications. A compound parabolic trough reactor has been designed and engineered to exploit both UV and visible part of the solar irradiation. The developed compound parabolic trough reactor could receive almost 88% of UV radiation along with a major part of visible radiation. The performance of the reactor has been evaluated in terms of degradation of a probe pollutant using the parameters such as rate constant, residence time and photonic efficiency. An attempt has been made to assess the performance in different ranges of solar spectrum. Finally the developed reactor has been employed for the photocatalytic treatment of a paper mill effluent using Degussa P25 as the photocatalyst. The paper mill effluent collected from Nagaon paper mill, Assam, India has been treated under both batch mode and continuous mode using Degussa P25 photocatalyst under artificial and natural solar radiation, respectively. The photocatalytic degradation kinetics of the paper mill effluent has been determined using the reduction in total organic carbon (TOC) values of the effluent. Copyright © 2015 Elsevier Inc. All rights reserved.

Different cultivation methods to acclimatise ammonia-tolerant methanogenic consortia.

PubMed

Tian, Hailin; Fotidis, Ioannis A; Mancini, Enrico; Angelidaki, Irini

2017-05-01

Bioaugmentation with ammonia tolerant-methanogenic consortia was proposed as a solution to overcome ammonia inhibition during anaerobic digestion process recently. However, appropriate technology to generate ammonia tolerant methanogenic consortia is still lacking. In this study, three basic reactors (i.e. batch, fed-batch and continuous stirred-tank reactors (CSTR)) operated at mesophilic (37°C) and thermophilic (55°C) conditions were assessed, based on methane production efficiency, incubation time, TAN/FAN (total ammonium nitrogen/free ammonia nitrogen) levels and maximum methanogenic activity. Overall, fed-batch cultivation was clearly the most efficient method compared to batch and CSTR. Specifically, by saving incubation time up to 150%, fed-batch reactors were acclimatised to nearly 2-fold higher FAN levels with a 37%-153% methanogenic activity improvement, compared to batch method. Meanwhile, CSTR reactors were inhibited at lower ammonia levels. Finally, specific methanogenic activity test showed that hydrogenotrophic methanogens were more active than aceticlastic methanogens in all FAN levels above 540mgNH 3 -NL -1 . Copyright © 2017 Elsevier Ltd. All rights reserved.

PCI fuel failure analysis: a report on a cooperative program undertaken by Pacific Northwest Laboratory and Chalk River Nuclear Laboratories.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohr, C.L.; Pankaskie, P.J.; Heasler, P.G.

Reactor fuel failure data sets in the form of initial power (P/sub i/), final power (P/sub f/), transient increase in power (..delta..P), and burnup (Bu) were obtained for pressurized heavy water reactors (PHWRs), boiling water reactors (BWRs), and pressurized water reactors (PWRs). These data sets were evaluated and used as the basis for developing two predictive fuel failure models, a graphical concept called the PCI-OGRAM, and a nonlinear regression based model called PROFIT. The PCI-OGRAM is an extension of the FUELOGRAM developed by AECL. It is based on a critical threshold concept for stress dependent stress corrosion cracking. The PROFITmore » model, developed at Pacific Northwest Laboratory, is the result of applying standard statistical regression methods to the available PCI fuel failure data and an analysis of the environmental and strain rate dependent stress-strain properties of the Zircaloy cladding.« less

The Nucifer Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cucoanes, A.S., E-mail: cucoanes@subatech.in2p3.fr

In nuclear reactors, a large number of antineutrinos are generated in the decay chains of the fission products; thus a survey of the antineutrino flux could provide valuable information related to the uranium and plutonium content of the core. This application generated interest by the IAEA in using antineutrino detectors as a potential safeguard tool. Here we present the Nucifer experiment, developed in France, by CEA and CNRS/IN2P3. The design of this new antineutrino detector has focused on safety, size reduction, reliability and high detection efficiency with a good background rejection. The Nucifer detector is currently taking data at themore » OSIRIS research reactor, inside CEA-Saclay. Presently, the ongoing analyses are considering the main sources of background for the antineutrino detection; the first antineutrino result is expected in 2013. A possible contribution to the understanding of the so called “reactor antineutrino anomaly” is also discussed. Finally, we present a brief description of the proposed experiments at very short baselines (VSBL) from reactors in France.« less

Determining Reactor Fuel Type from Continuous Antineutrino Monitoring

NASA Astrophysics Data System (ADS)

Jaffke, Patrick; Huber, Patrick

2017-09-01

We investigate the ability of an antineutrino detector to determine the fuel type of a reactor. A hypothetical 5-ton antineutrino detector is placed 25 m from the core and measures the spectral shape and rate of antineutrinos emitted by fission fragments in the core for a number of 90-d periods. Our results indicate that four major fuel types can be differentiated from the variation of fission fractions over the irradiation time with a true positive probability of detection at approximately 95%. In addition, we demonstrate that antineutrinos can identify the burnup at which weapons-grade mixed-oxide (MOX) fuel would be reduced to reactor-grade MOX, on average, providing assurance that plutonium-disposition goals are met. We also investigate removal scenarios where plutonium is purposefully diverted from a mixture of MOX and low-enriched uranium fuel. Finally, we discuss how our analysis is impacted by a spectral distortion around 6 MeV observed in the antineutrino spectrum measured from commercial power reactors.

Partial nitrification using aerobic granules in continuous-flow reactor: rapid startup.

PubMed

Wan, Chunli; Sun, Supu; Lee, Duu-Jong; Liu, Xiang; Wang, Li; Yang, Xue; Pan, Xiangliang

2013-08-01

This study applied a novel strategy to rapid startup of partial nitrification in continuous-flow reactor using aerobic granules. Mature aerobic granules were first cultivated in a sequencing batch reactor at high chemical oxygen demand in 16 days. The strains including the Pseudoxanthomonas mexicana strain were enriched in cultivated granules to enhance their structural stability. Then the cultivated granules were incubated in a continuous-flow reactor with influent chemical oxygen deamnad being stepped decreased from 1,500 ± 100 (0-19 days) to 750 ± 50 (20-30 days), and then to 350 ± 50 mg l(-1) (31-50 days); while in the final stage 350 mg l(-1) bicarbonate was also supplied. Using this strategy the ammonia-oxidizing bacterium, Nitrosomonas europaea, was enriched in the incubated granules to achieve partial nitrification efficiency of 85-90% since 36 days and onwards. The partial nitrification granules were successfully harvested after 52 days, a period much shorter than those reported in literature. Copyright © 2013 Elsevier Ltd. All rights reserved.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

DOE PAGES

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; ...

2014-06-30

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in ordermore » to reduce the overall numerical uncertainty while leveraging available computational resources. Finally, the coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.« less

Preliminary Tritium Management Design Activities at ORNL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrison, Thomas J.; Felde, David K.; Logsdon, Randall J.

2016-09-01

Interest in salt-cooled and salt-fueled reactors has increased over the last decade (Forsberg et al. 2016). Several private companies and universities in the United States, as well as governments in other countries, are developing salt reactor designs and/or technology. Two primary issues for the development and deployment of many salt reactor concepts are (1) the prevention of tritium generation and (2) the management of tritium to prevent release to the environment. In 2016, the US Department of Energy (DOE) initiated a research project under the Advanced Reactor Technology Program to (1) experimentally assess the feasibility of proposed methods for tritiummore » mitigation and (2) to perform an engineering demonstration of the most promising methods. This document describes results from the first year’s efforts to define, design, and build an experimental apparatus to test potential methods for tritium management. These efforts are focused on producing a final design document as the basis for the apparatus and its scheduled completion consistent with available budget and approvals for facility use.« less

Final report, PT IP-535-C: Test of smaller VSR`s in DR reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaughn, A.D.

1963-04-17

Because of rod-sticking problems at DR Reactor, a knuckle rod of B Reactor design was installed in vertical safety channel number 28. The substitute VSR, which has a smaller diameter than the original DR rod, has been tested for its operational feasibility including both drop time and reactivity effect. The reactivity effect of the rod was estimated by comparison of the reactivity transient caused by insertion of the specific B-type rod after scramming into the pile, with similar transients caused by normal vertical safety rod in an adjacent channel. This document lists the indicated relative control strength of the rodmore » as an empirical basis for future safety calculations. Results indicate that the B-type knuckel rod in DR reactor is about 80% as strong as a normal DR vertical safety rod if used in equivalent flux distribution and location; this makes it reasonable to assume that the local control strength of the B-type knuckel rod is 98 {mu}b.« less

The startup performance and microbial distribution of an anaerobic baffled reactor (ABR) treating medium-strength synthetic industrial wastewater.

PubMed

Jiang, Hao; Nie, Hong; Ding, Jiangtao; Stinner, Walter; Sun, Kaixuan; Zhou, Hongjun

2018-01-02

In this study, an anaerobic baffled reactor (ABR) with seven chambers was applied to treat medium-strength synthetic industrial wastewater (MSIW). The performance of startup and shock test on treating MSIW was investigated. During the acclimation process, the chemical oxygen demand (COD) of MSIW gradually increased from 0 to 2,000 mg L -1 , and the COD removal finally reached 90%. At shock test, the feeding COD concentration increased by one-fifth and the reactor adapted very well with a COD removal of 82%. In a stable state, Comamonas, Smithella, Syntrophomonas and Pseudomonas were the main populations of bacteria, while the predominant methanogen was Methanobacterium. The results of chemical and microbiological analysis indicated the significant advantages of ABR, including buffering shocks, separating stages with matching microorganisms and promoting syntrophism. Meanwhile, the strategies for acclimation and operation were of great importance. Further work can test reactor performance in the treatment of actual industrial wastewater.

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yamada, K.; Aksan, S. N.

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present,more » 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)« less

Investigating the kinetics of the enzymatic depolymerization of polygalacturonic acid in continuous UF-membrane reactors.

PubMed

Gallifuoco, Alberto; Cantarella, Maria; Marucci, Mariagrazia

2007-01-01

A stirred tank membrane reactor is used to study the kinetics of polygalacturonic acid (PGA) enzymatic hydrolysis. The reactor operates in semicontinuous configuration: the native biopolymer is loaded at the initial time and the system is continuously fed with the buffer. The effect of retention time (from 101 to 142 min) and membrane molecular weight cutoff (from 1 to 30 kDa) on the rate of permeable oligomers production is investigated. Reaction products are clustered in two different classes, those sized below the membrane cutoff and those above. The reducing power measured in the permeate is used as an estimate of total product concentration. The characteristic breakdown times range from 40 to 100 min. The overall kinetics obeys a first-order law with a characteristic time estimated to 24 min. New mathematical data handling are developed and illustrated using the experimental data obtained. Finally, the body of the experimental results suggests useful indications (reactor productivity, breakdown induction period) for implementing the bioprocess at the industrial scale.

Study of the Open Loop and Closed Loop Oscillator Techniques

DOE Office of Scientific and Technical Information (OSTI.GOV)

Imel, George R.; Baker, Benjamin; Riley, Tony

This report presents the progress and completion of a five-year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques.The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this report we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign tomore » measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems.« less

Study of the open loop and closed loop oscillator techniques

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baker, Benjamin; Riley, Tony; Langbehn, Adam

This paper presents some aspects of a five year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques. The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this paper we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign tomore » measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems. (authors)« less

Reactor antineutrino shoulder explained by energy scale nonlinearities?

NASA Astrophysics Data System (ADS)

Mention, G.; Vivier, M.; Gaffiot, J.; Lasserre, T.; Letourneau, A.; Materna, T.

2017-10-01

The Daya Bay, Double Chooz and RENO experiments recently observed a significant distortion in their detected reactor antineutrino spectra, being at odds with the current predictions. Although such a result suggests to revisit the current reactor antineutrino spectra modeling, an alternative scenario, which could potentially explain this anomaly, is explored in this letter. Using an appropriate statistical method, a study of the Daya Bay experiment energy scale is performed. While still being in agreement with the γ calibration data and 12B measured spectrum, it is shown that a O (1%) deviation of the energy scale reproduces the distortion observed in the Daya Bay spectrum, remaining within the quoted calibration uncertainties. Potential origins of such a deviation, which challenge the energy calibration of these detectors, are finally discussed.

Use of Gelidium amansii as a promising resource for bioethanol: a practical approach for continuous dilute-acid hydrolysis and fermentation.

PubMed

Park, Jeong-Hoon; Hong, Ji-Yeon; Jang, Hyun Chul; Oh, Seung Geun; Kim, Sang-Hyoun; Yoon, Jeong-Jun; Kim, Yong Jin

2012-03-01

A facile continuous method for dilute-acid hydrolysis of the representative red seaweed species, Gelidium amansii was developed and its hydrolysate was subsequently evaluated for fermentability. In the hydrolysis step, the hydrolysates obtained from a batch reactor and a continuous reactor were systematically compared based on fermentable sugar yield and inhibitor formation. There are many advantages to the continuous hydrolysis process. For example, the low melting point of the agar component in G. amansii facilitates improved raw material fluidity in the continuous reactor. In addition, the hydrolysate obtained from the continuous process delivered a high sugar and low inhibitor concentration, thereby leading to both high yield and high final ethanol titer in the fermentation process. Copyright © 2011 Elsevier Ltd. All rights reserved.

Integral Inherently Safe Light Water Reactor (I 2S-LWR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petrovic, Bojan; Memmott, Matthew; Boy, Guy

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I 2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to addressmore » the preference of some utilities in the US power market for unit power level on the order of 1 GWe.« less

Treatment of fruit-juice industry wastewater in a two-stage anaerobic hybrid (AH) reactor system followed by a sequencing batch reactor (SBR).

PubMed

Tawfik, A; El-Kamah, H

2012-01-01

This study has been carried out to assess the performance of a combined system consisting of an anaerobic hybrid (AH) reactor followed by a sequencing batch reactor (SBR) for treatment of fruit-juice industry wastewater at a temperature of 26 degrees C. Three experimental runs were conducted in this investigation. In the first experiment, a single-stage AH reactor was operated at a hydraulic retention time (HRT) of 10.2 h and organic loading rate (OLR) of 11.8 kg COD m(-3) d(-1). The reactor achieved a removal efficiency of 42% for chemical oxygen demand (COD), 50.8% for biochemical oxygen demand (BOD5), 50.3% for volatile fatty acids (VFA) and 56.4% for total suspended solids (TSS). In the second experiment, two AH reactors connected in series achieved a higher removal efficiency for COD (67.4%), BOD5 (77%), and TSS (71.5%) at a total HRT of 20 h and an OLR of 5.9 kg COD m(-3) d(-1). For removal of the remaining portions of COD, BOD5 and TSS from the effluent of the two-stage AH system, a sequencing batch reactor (SBR) was investigated as a post-treatment unit. The reactor achieved a substantial reduction in total COD, resulting in an average effluent concentration of 50 mg L(-1) at an HRT of 11 h and OLR of 5.3 kg COD m(-3) d(-1). Almost complete removal of total BOD5 and oil and grease was achieved, i.e. 10 mg L(-1) and 1.2 mg L(-1), respectively, remained in the final effluent of the SBR.

Decommissioning of eight surplus production reactors at the Hanford Site, Richland, Washington. Addendum (Final Environmental Impact Statement)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-12-01

The first section of this volume summarizes the content of the draft environmental impact statement (DEIS) and this Addendum, which together constitute the final environmental impact statement (FEIS) prepared on the decommissioning of eight surplus plutonium production reactors at Hanford. The FEIS consists of two volumes. The first volume is the DEIS as written. The second volume (this Addendum) consists of a summary; Chapter 9, which contains comments on the DEIS and provides DOE`s responses to the comments; Appendix F, which provides additional health effects information; Appendix K, which contains costs of decommissioning in 1990 dollars; Appendix L, which containsmore » additional graphite leaching data; Appendix M, which contains a discussion of accident scenarios; Appendix N, which contains errata; and Appendix 0, which contains reproductions of the letters, transcripts, and exhibits that constitute the record for the public comment period.« less

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

Isotopic composition and neutronics of the Okelobondo natural reactor

NASA Astrophysics Data System (ADS)

Palenik, Christopher Samuel

The Oklo-Okelobondo and Bangombe uranium deposits, in Gabon, Africa host Earth's only known natural nuclear fission reactors. These 2 billion year old reactors represent a unique opportunity to study used nuclear fuel over geologic periods of time. The reactors in these deposits have been studied as a means by which to constrain the source term of fission product concentrations produced during reactor operation. The source term depends on the neutronic parameters, which include reactor operation duration, neutron flux and the neutron energy spectrum. Reactor operation has been modeled using a point-source computer simulation (Oak Ridge Isotope Generation and Depletion, ORIGEN, code) for a light water reactor. Model results have been constrained using secondary ionization mass spectroscopy (SIMS) isotopic measurements of the fission products Nd and Te, as well as U in uraninite from samples collected in the Okelobondo reactor zone. Based upon the constraints on the operating conditions, the pre-reactor concentrations of Nd (150 ppm +/- 75 ppm) and Te (<1 ppm) in uraninite were estimated. Related to the burnup measured in Okelobondo samples (0.7 to 13.8 GWd/MTU), the final fission product inventories of Nd (90 to 1200 ppm) and Te (10 to 110 ppm) were calculated. By the same means, the ranges of all other fission products and actinides produced during reactor operation were calculated as a function of burnup. These results provide a source term against which the present elemental and decay abundances at the fission reactor can be compared. Furthermore, they provide new insights into the extent to which a "fossil" nuclear reactor can be characterized on the basis of its isotopic signatures. In addition, results from the study of two other natural systems related to the radionuclide and fission product transport are included. A detailed mineralogical characterization of the uranyl mineralogy at the Bangombe uranium deposit in Gabon, Africa was completed to improve geochemical models of the solubility-limiting phase. A study of the competing effects of radiation damage and annealing in a U-bearing crystal of zircon shows that low temperature annealing in actinide-bearing phases is significant in the annealing of radiation damage.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krieg, R.

For future pressurized-water reactors, which should be designed against core-meltdown accidents, missiles generated inside the containment present a severe problem for its integrity. The masses and geometries of the missiles, as well as their velocities, may vary to a great extent. Therefore a reliable proof of the containment integrity is very difficult. In this article the potential sources of missiles are discussed, and the conclusion was reached that the generation of heavy missiles must be prevented. Steam explosions must not damage the reactor vessel head. Thus fragments of the head cannot become missiles that endanger the containment shell. Furthermore, duringmore » a melt-through failure of the reactor vessel under high pressure, the resulting forces must not catapult the whole vessel against the containment shell. Only missiles caused by hydrogen explosions may be tolerable, but shielding structures that protect the containment shell may be required. Further investigations are necessary. Finally, measures are described showing that the generation of heavy missiles can indeed be prevented. Investigations are currently being carried out that will confirm the strength of the reactor vessel head. In addition, a device for retaining the fragments of a failing reactor vessel is discussed.« less

Mass transfer study on the electrochemical removal of copper ions from synthetic effluents using reticulated vitreous carbon.

PubMed

Britto-Costa, Pedro H; Ruotolo, Luís Augusto M

2013-01-01

Porous electrodes have been successfully used for metal electrodeposition from diluted aqueous solution due to their high porosity and specific surface area, which lead to high mass transfer rates. This work studies the mass transfer of copper electrodeposition on reticulated vitreous carbon in a flow reactor without membrane. The flow configuration, otherwise the filter-press electrochemical reactors, was designed in order to minimize the pressure drop. The mass transfer coefficient was determined by voltammetric and galvanostatic electrodeposition. In the voltammetric experiments a Luggin capillary was used to measure the current-potential curves and to determine the limiting current (and, consequently, the mass transfer coefficient). In the galvanostatic experiments the concentration-time curves were obtained and considering a limiting current kinetics model, the mass transfer coefficient (k(m)) was determined for different flow velocities. The results showed that both methods give similar values of k(m), thus the voltammetric method can be recommended because it is faster and simpler. Finally, the reactor performance was compared with others from literature, and it was observed that the proposed reactor design has high Sherwood numbers similar to other reactor configurations using membranes and reticulated vitreous carbon electrodes.

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okrent, D.

1997-06-23

This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, withmore » the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.« less

75 FR 30077 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on ESBWR

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-28

... Final Safety Evaluation Report (FSER) associated with the ESBWR design certification. The Subcommittee... from the website cited above or by contacting the identified DFO. Moreover, in view of the possibility...

77 FR 56239 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-12

... Draft Final Revision 1 to Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test Program... inconvenience. If attending this meeting, please enter through the One White Flint North building, 11555...

The effectiveness of power-generating complexes constructed on the basis of nuclear power plants combined with additional sources of energy determined taking risk factors into account

NASA Astrophysics Data System (ADS)

Aminov, R. Z.; Khrustalev, V. A.; Portyankin, A. V.

2015-02-01

The effectiveness of combining nuclear power plants equipped with water-cooled water-moderated power-generating reactors (VVER) with other sources of energy within unified power-generating complexes is analyzed. The use of such power-generating complexes makes it possible to achieve the necessary load pickup capability and flexibility in performing the mandatory selective primary and emergency control of load, as well as participation in passing the night minimums of electric load curves while retaining high values of the capacity utilization factor of the entire power-generating complex at higher levels of the steam-turbine part efficiency. Versions involving combined use of nuclear power plants with hydrogen toppings and gas turbine units for generating electricity are considered. In view of the fact that hydrogen is an unsafe energy carrier, the use of which introduces additional elements of risk, a procedure for evaluating these risks under different conditions of implementing the fuel-and-hydrogen cycle at nuclear power plants is proposed. Risk accounting technique with the use of statistical data is considered, including the characteristics of hydrogen and gas pipelines, and the process pipelines equipment tightness loss occurrence rate. The expected intensities of fires and explosions at nuclear power plants fitted with hydrogen toppings and gas turbine units are calculated. In estimating the damage inflicted by events (fires and explosions) occurred in nuclear power plant turbine buildings, the US statistical data were used. Conservative scenarios of fires and explosions of hydrogen-air mixtures in nuclear power plant turbine buildings are presented. Results from calculations of the introduced annual risk to the attained net annual profit ratio in commensurable versions are given. This ratio can be used in selecting projects characterized by the most technically attainable and socially acceptable safety.

Development of advanced high heat flux and plasma-facing materials

NASA Astrophysics Data System (ADS)

Linsmeier, Ch.; Rieth, M.; Aktaa, J.; Chikada, T.; Hoffmann, A.; Hoffmann, J.; Houben, A.; Kurishita, H.; Jin, X.; Li, M.; Litnovsky, A.; Matsuo, S.; von Müller, A.; Nikolic, V.; Palacios, T.; Pippan, R.; Qu, D.; Reiser, J.; Riesch, J.; Shikama, T.; Stieglitz, R.; Weber, T.; Wurster, S.; You, J.-H.; Zhou, Z.

2017-09-01

Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating tungsten alloys, demonstrating favorable tungsten-like plasma-wall interaction behavior under normal operation conditions, are an intrinsic solution to otherwise catastrophic consequences of loss-of-coolant and air ingress events in a fusion reactor. Permeation barrier layers avoid the escape of tritium into structural and cooling materials, thereby minimizing the release of tritium under normal operation conditions. Finally, solutions for the unique bonding requirements of dissimilar material used in a fusion reactor are demonstrated by describing the current status and prospects of functionally graded materials.

Performance of a full scale prototype detector at the BR2 reactor for the SoLid experiment

NASA Astrophysics Data System (ADS)

Abreu, Y.; Amhis, Y.; Arnold, L.; Ban, G.; Beaumont, W.; Bongrand, M.; Boursette, D.; Castle, B. C.; Clark, K.; Coupé, B.; Cussans, D.; De Roeck, A.; D'Hondt, J.; Durand, D.; Fallot, M.; Ghys, L.; Giot, L.; Guillon, B.; Ihantola, S.; Janssen, X.; Kalcheva, S.; Kalousis, L. N.; Koonen, E.; Labare, M.; Lehaut, G.; Manzanillas, L.; Mermans, J.; Michiels, I.; Moortgat, C.; Newbold, D.; Park, J.; Pestel, V.; Petridis, K.; Piñera, I.; Pommery, G.; Popescu, L.; Pronost, G.; Rademacker, J.; Ryckbosch, D.; Ryder, N.; Saunders, D.; Schune, M.-H.; Simard, L.; Vacheret, A.; Van Dyck, S.; Van Mulders, P.; van Remortel, N.; Vercaemer, S.; Verstraeten, M.; Weber, A.; Yermia, F.

2018-05-01

The SoLid collaboration has developed a new detector technology to detect electron anti-neutrinos at close proximity to the Belgian BR2 reactor at surface level. A 288 kg prototype detector was deployed in 2015 and collected data during the operational period of the reactor and during reactor shut-down. Dedicated calibration campaigns were also performed with gamma and neutron sources. This paper describes the construction of the prototype detector with a high control on its proton content and the stability of its operation over a period of several months after deployment at the BR2 reactor site. All detector cells provide sufficient light yields to achieve a target energy resolution of better than 20%/√E(MeV). The capability of the detector to track muons is exploited to equalize the light response of a large number of channels to a precision of 3% and to demonstrate the stability of the energy scale over time. Particle identification based on pulse-shape discrimination is demonstrated with calibration sources. Despite a lower neutron detection efficiency due to triggering constraints, the main backgrounds at the reactor site were determined and taken into account in the shielding strategy for the main experiment. The results obtained with this prototype proved essential in the design optimization of the final detector.

International Research Reactor Decommissioning Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leopando, Leonardo; Warnecke, Ernst

2008-01-15

Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement tomore » the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.« less

Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G

2014-01-01

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly casesmore » are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.« less

IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brumovsky, M.; Steele, L.E.

1997-02-01

In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracturemore » mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.« less

77 FR 46528 - Sunshine Act Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-03

... Suspend Final Decisions on Reactor License Applications Pending Completion of Remanded Waste Confidence...-nrc/policy-making/schedule.html . * * * * * The NRC provides reasonable accommodation to individuals... Benefits Branch, at 301-415-6200, TDD: 301-415-2100, or by email at [email protected] . Determinations...

Microstructure evolution and tensile properties of Zr-2.5 wt.% Nb pressure tubes processed from billets with different microstructures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kapoor, K.; Saratchandran, N.; Muralidharan, K.

1999-02-01

Pressurized heavy water reactors (PHWR) use zirconium-base alloys for their low neutron-absorption cross section, good mechanical strength, low irradiation creep, and high corrosion resistance in reactor atmospheres. Starting with identical ingots, billets having different microstructures were obtained by three different processing methods for fabrication of Zr-2.5 wt%Nb pressure tubes., The billets were further processed by hot extrusion and cold Pilger tube reducing to the finished product. Microstructural characterization was done at each stage of processing. The effects of the initial billet microstructure on the intermediate and final microstructure and mechanical property results were determined. It was found that the structuremore » at each stage and the final mechanical properties depend strongly on the initial billet microstructure. The structure at the final stage consists of elongated alpha zirconium grains with a network of metastable beta zirconium phase. Some of this metastable phase transforms into stable beta niobium during thermomechanical processing. Billets with quenched structure resulted in less beta niobium at the final stage. The air cooled billets resulted in a large amount of beta niobium. The tensile properties, especially the percentage elongation, were found to vary for the different methods. Higher percentage elongation was observed for billets having quenched structure. Extrusion and forging did not produce any characteristic differences in the properties. The results were used to select a process flow sheet which yields the desired mechanical properties with suitable microstructure in the final product.« less

Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Was, Gary

This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

THE FUEL ELEMENT GRAPHITE. Project DRAGON.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Graham, L.W.; Price, M.S.T.

1963-01-15

The main requirements of a fuel element graphite for reactors based on the Dragon concept are low transmission coefficient for fission products, dimensional stability under service conditions, high strength, high thermal conductivity, high purity, and high resistance to oxidation. Since conclusions reached in early 1960, a considerable amount of information has accumulated concerning the likely behaviour of graphites in high temperature reactor systems, particularly data on dimensional stability under irradiation. The influence of this new knowledge on the development of fuel element graphite with the Dragon Project is discussed in detail in the final section of this paper.

Simulation of chemical-vapor-deposited silicon carbide for a cold wall vertical reactor

NASA Astrophysics Data System (ADS)

Lee, Y. L.; Sanchez, J. M.

1997-07-01

The growth rate of silicon carbide obtained by low-pressure chemical vapor deposition from tetramethylsilane is numerically simulated for a cold wall vertical reactor. The transport equations for momentum, heat, and mass transfer are simultaneously solved by employing the finite volume method. A model for reaction rate is also proposed in order to predict the measured growth rates [A. Figueras, S. Garelik, J. Santiso, R. Rodroguez-Clemente, B. Armas, C. Combescure, R. Berjoan, J.M. Saurel and R. Caplain, Mater. Sci. Eng. B 11 (1992) 83]. Finally, the effects of thermal diffusion on the growth rate are investigated.

Structural behavior of the Bitter plate tf magnet for the Zephyr ignition test reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bobrov, E.S.; Becker, H.

1981-01-01

This paper discusses methods and results of the computer structural analysis of the Bitter plate toroidal field magnet design for the ZEPHYR Ignition Test Reactor. The magnet provides a field of 7.06 T at the center of the bore which is 1.76 m from the major toroidal axis. The ignited plasma is located at a major radius of 1.36 m where the magnetic field is 9.11 T. The plasma is moved to this final position following compression in the major radius. The horizontal bore of the magnet is 1.8 m.

78 FR 7465 - Agency Information Collection Activities: Submission for the Office of Management and Budget (OMB...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-01

..., and applicants for facility (i.e., nuclear power and non-power research and test reactor) operating... the final supporting statement, at the NRC's Public Document Room, Room O-1F21, One White Flint North...

A RADIAL REACTOR FOR TRICHLOROETHYLENE STEAM REFORMING. (R822721C633)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

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

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

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

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

COMPARTMENTED REACTOR FUEL ELEMENT

DOEpatents

Cain, F.M. Jr.

1962-09-11

A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

Self-actuated shutdown system for a commercial size LMFBR. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dupen, C.F.G.

1978-08-01

A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility andmore » reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power.« less

An investigation of the treatment of particulate matter from gasoline engine exhaust using non-thermal plasma.

PubMed

Ye, Dan; Gao, Dengshan; Yu, Gang; Shen, Xianglin; Gu, Fan

2005-12-09

A plasma reactor with catalysts was used to treat exhaust gas from a gasoline engine in order to decrease particulate matter (PM) emissions. The effect of non-thermal plasma (NTP) of the dielectric discharges on the removal of PM from the exhaust gas was investigated experimentally. The removal efficiency of PM was based on the concentration difference in PM for particle diameters ranging from 0.3 to 5.0 microm as measured by a particle counter. Several factors affecting PM conversion, including the density of plasma energy, reaction temperature, flow rate of exhaust gas, were investigated in the experiment. The results indicate that PM removal efficiency ranged approximately from 25 to 57% and increased with increasing energy input in the reactor, reaction temperature and residence time of the exhaust gas in the reactor. Enhanced removal of the PM was achieved by filling the discharge gap of the reactor with Cu-ZSM-5 catalyst pellets. In addition, the removal of unburned hydrocarbons was studied. Finally, available approaches for PM conversion were analyzed involving the interactions between discharge and catalytic reactions.

Novel windows for "solar commodities": a device for CO2 reduction using plasmonic catalyst activation.

PubMed

Navarrete, Alexander; Muñoz, Sergio; Sanz-Moral, Luis M; Brandner, Juergen J; Pfeifer, Peter; Martín, Ángel; Dittmeyer, Roland; Cocero, María J

2015-01-01

A novel plasmonic reactor concept is proposed and tested to work as a visible energy harvesting device while allowing reactions to transform CO2 to be carried out. Particularly the reverse water gas shift (RWGS) reaction has been tested as a means to introduce renewable energy into the economy. The development of the new reactor concept involved the synthesis of a new composite capable of plasmonic activation with light, the development of an impregnation method to create a single catalyst reactor entity, and finally the assembly of a reaction system to test the reaction. The composite developed was based on a Cu/ZnO catalyst dispersed into transparent aerogels. This allows efficient light transmission and a high surface area for the catalyst. An effective yet simple impregnation method was developed that allowed introduction of the composites into glass microchannels. The activation of the reaction was made using LEDs that covered all the sides of the reactor allowing a high power delivery. The results of the reaction show a stable process capable of low temperature transformations.

Small modular reactor: First-of-a-Kind (FOAK) and Nth-of-a-Kind (NOAK) Economic Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boldon, Lauren M.; Sabharwall, Piyush

2014-08-01

Small modular reactors (SMRs) refer to any reactor design in which the electricity generated is less than 300 MWe. Often medium sized reactors with power less than 700 MWe are also grouped into this category. Internationally, the development of a variety of designs for SMRs is booming with many designs approaching maturity and even in or nearing the licensing stage. It is for this reason that a generalized yet comprehensive economic model for first of a kind (FOAK) through nth of a kind (NOAK) SMRs based upon rated power, plant configuration, and the fiscal environment was developed. In the model,more » a particular project’s feasibility is assessed with regards to market conditions and by commonly utilized capital budgeting techniques, such as the net present value (NPV), internal rate of return (IRR), Payback, and more importantly, the levelized cost of energy (LCOE) for comparison to other energy production technologies. Finally, a sensitivity analysis was performed to determine the effects of changing debt, equity, interest rate, and conditions on the LCOE.« less

Sequential two-column electro-Fenton-photolytic reactor for the treatment of winery wastewater.

PubMed

Díez, A M; Sanromán, M A; Pazos, M

2017-01-01

The high amount of winery wastewaters produced each year makes their treatment a priority issue due to their problematic characteristics such as acid pH, high concentration of organic load and colourful compounds. Furthermore, some of these effluents can have dissolved pesticides, due to the previous grape treatments, which are recalcitrant to conventional treatments. Recently, photo-electro-Fenton process has been reported as an effective procedure to mineralize different organic contaminants and a promising technology for the treatment of these complex matrixes. However, the reactors available for applying this process are scarce and they show several limitations. In this study, a sequential two-column reactor for the photo-electro-Fenton treatment was designed and evaluated for the treatment of different pesticides, pirimicarb and pyrimethanil, used in wine production. Both studied pesticides were efficiently removed, and the transformation products were determined. Finally, the treatment of a complex aqueous matrix composed by winery wastewater and the previously studied pesticides was carried out in the designed sequential reactor. The high removals of TOC and COD reached and the low energy consumption demonstrated the efficiency of this new configuration.

Automated in Situ Measurement of Gas Solubility in Liquids with a Simple Tube-in-Tube Reactor.

PubMed

Zhang, Jisong; Teixeira, Andrew R; Zhang, Haomiao; Jensen, Klavs F

2017-08-15

Data on the solubilities of gases in liquids are foundational for assessing a variety of multiphase separations and gas-liquid reactions. Taking advantage of the tube-in-tube reactor design built with semipermeable Teflon AF-2400 tubes, liquids can be rapidly saturated without direct contacting of gas and liquid. The gas solubility can be determined by performing steady-state flux balances of both the gas and liquid flowing into the reactor system. Using this type of reactor, a fully automated strategy has been developed for the rapid in situ measurement of gas solubilities in liquids. The developed strategy enables precise gas solubility measurements within 2-5 min compared with 4-5 h using conventional methods. This technique can be extended to the discrete multipoint steady-state and continuous ramped-multipoint data acquisition methods. The accuracy of this method has been validated against several gas-liquid systems, showing less than 2% deviation from known values. Finally, this strategy has been extended to measure the temperature dependence of gas solubilities in situ and to estimate the local enthalpy of dissolution across a defined temperature range.

Modular fabrication and characterization of complex silicon carbide composite structures Advanced Reactor Technologies (ART) Research Final Report (Feb 2015 – May 2017)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khalifa, Hesham

Advanced ceramic materials exhibit properties that enable safety and fuel cycle efficiency improvements in advanced nuclear reactors. In order to fully exploit these desirable properties, new processing techniques are required to produce the complex geometries inherent to nuclear fuel assemblies and support structures. Through this project, the state of complex SiC-SiC composite fabrication for nuclear components has advanced significantly. New methods to produce complex SiC-SiC composite structures have been demonstrated in the form factors needed for in-core structural components in advanced high temperature nuclear reactors. Advanced characterization techniques have been employed to demonstrate that these complex SiC-SiC composite structures providemore » the strength, toughness and hermeticity required for service in harsh reactor conditions. The complex structures produced in this project represent a significant step forward in leveraging the excellent high temperature strength, resistance to neutron induced damage, and low neutron cross section of silicon carbide in nuclear applications.« less

Conversion of NO with a catalytic packed-bed dielectric barrier discharge reactor

NASA Astrophysics Data System (ADS)

Xu, CAO; Weixuan, ZHAO; Renxi, ZHANG; Huiqi, HOU; Shanping, CHEN; Ruina, ZHANG

2017-11-01

This paper discusses the conversion of nitric oxide (NO) with a low-temperature plasma induced by a catalytic packed-bed dielectric barrier discharge (DBD) reactor. Alumina oxide (Al2O3), glass (SiO2) and zirconium oxide (ZrO2), three different spherical packed materials of the same size, were each present in the DBD reactor. The NO conversion under varying input voltage and specific energy density, and the effects of catalysts (titanium dioxide (TiO2) and manganese oxide (MnO x ) coated on Al2O3) on NO conversion were investigated. The experimental results showed that NO conversion was greatly enhanced in the presence of packed materials in the reactor, and the catalytic packed bed of MnO x /Al2O3 showed better performance than that of TiO2/Al2O3. The surface and crystal structures of the materials and catalysts were characterized through scanning electron microscopy analysis. The final products were clearly observed by a Fourier transform infrared spectrometer and provided a better understanding of NO conversion.

Disposal of radioactive iodine in space

NASA Technical Reports Server (NTRS)

Burns, R. E.; Defield, J. G.

1978-01-01

The possibility of space disposal of iodine waste from nuclear power reactors is investigated. The space transportation system utilized relies upon the space shuttle, a liquid hydrogen/liquid oxygen orbit transfer vehicle, and a solid propellant final stage. The iodine is assumed to be in the form of either an iodide or an iodate, and calculations assume that the final destination is either solar orbit or solar system escape. It is concluded that space disposal of iodine is feasible.

Physical-chemical treatment of wastes: a way to close turnover of elements in LSS

NASA Astrophysics Data System (ADS)

Kudenko, Yu A.; Gribovskaya, I. V.; Zolotukhin, I. G.

2000-05-01

"Man-plants-physical-chemical unit" system designed for space stations or terrestrial ecohabitats to close steady-state mineral, water and gas exchange is proposed. The physical-chemical unit is to mineralize all inedible plant wastes and physiological human wastes (feces, urine, gray water) by electromagnetically activated hydrogen peroxide in an oxidation reactor. The final product is a mineralized solution containing all elements balanced for plants' requirements. The solution has been successfully used in experiments to grow wheat, beans and radish. The solution was reusable: the evaporated moisture was replenished by the phytotron condensate. Sodium salination of plants was precluded by evaporating reactor-mineralized urine to sodium saturation concentration to crystallize out NaCl which can be used as food for the crew. The remaining mineralized product was brought back for nutrition of plants. The gas composition of the reactor comprises O 2, N 2, CO 2, NH 3, H 2. At the reactor's output hydrogen and oxygen were catalyzed into water, NH 3 was converted in a water trap into NH 4 and used for nutrition of plants. A special accessory at the reactor's output may produce hydrogen peroxide from intrasystem water and gas which makes possible to close gas loops between LSS components.

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, Michael M.; Schulz, Terry L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, M.M.; Schulz, T.L.

1993-12-07

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

Dynamic System Simulation of the KRUSTY Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klein, Steven Karl; Kimpland, Robert Herbert

2016-05-09

The proposed KRUSTY experiment is a demonstration of a reactor operating at power. The planned experimental configuration includes a highly enriched uranium (HEU) reflected core, cooled by multiple heat pipes leading to Stirling engines for primary heat rejection. Operating power is expected to be approximately four (4) to five (5) kilowatts with a core temperature above 1,000 K. No data is available on any historical reactor employing HEU metal that operated over the temperature range required for the KRUSTY experiment. Further, no reactor has operated with heat pipes as the primary cooling mechanism. Historic power reactors have employed either naturalmore » or forced convection so data on their operation is not directly applicable to the KRUSTY experiment. The primary purpose of the system model once developed and refined by data from these component experiments, will be used to plan the KRUSTY experiment. This planning will include expected behavior of the reactor from start-up, through various transient conditions where cooling begins to become present and effective, and finally establishment of steady-state. In addition, the model can provide indicators of anticipated off-normal events and appropriate operator response to those conditions. This information can be used to develop specific experiment operating procedures and aids to guide the operators in conduct of the experiment.« less

Application of Radiation Chemistry to Some Selected Technological Issues Related to the Development of Nuclear Energy.

PubMed

Bobrowski, Krzysztof; Skotnicki, Konrad; Szreder, Tomasz

2016-10-01

The most important contributions of radiation chemistry to some selected technological issues related to water-cooled reactors, reprocessing of spent nuclear fuel and high-level radioactive wastes, and fuel evolution during final radioactive waste disposal are highlighted. Chemical reactions occurring at the operating temperatures and pressures of reactors and involving primary transients and stable products from water radiolysis are presented and discussed in terms of the kinetic parameters and radiation chemical yields. The knowledge of these parameters is essential since they serve as input data to the models of water radiolysis in the primary loop of light water reactors and super critical water reactors. Selected features of water radiolysis in heterogeneous systems, such as aqueous nanoparticle suspensions and slurries, ceramic oxides surfaces, nanoporous, and cement-based materials, are discussed. They are of particular concern in the primary cooling loops in nuclear reactors and long-term storage of nuclear waste in geological repositories. This also includes radiation-induced processes related to corrosion of cladding materials and copper-coated iron canisters, dissolution of spent nuclear fuel, and changes of bentonite clays properties. Radiation-induced processes affecting stability of solvents and solvent extraction ligands as well oxidation states of actinide metal ions during recycling of the spent nuclear fuel are also briefly summarized.

Composting in small laboratory pilots: performance and reproducibility.

PubMed

Lashermes, G; Barriuso, E; Le Villio-Poitrenaud, M; Houot, S

2012-02-01

Small-scale reactors (<10 l) have been employed in composting research, but few attempts have assessed the performance of composting considering the transformations of organic matter. Moreover, composting at small scales is often performed by imposing a fixed temperature, thus creating artificial conditions, and the reproducibility of composting has rarely been reported. The objectives of this study are to design an innovative small-scale composting device safeguarding self-heating to drive the composting process and to assess the performance and reproducibility of composting in small-scale pilots. The experimental setup included six 4-l reactors used for composting a mixture of sewage sludge and green wastes. The performance of the process was assessed by monitoring the temperature, O(2) consumption and CO(2) emissions, and characterising the biochemical evolution of organic matter. A good reproducibility was found for the six replicates with coefficients of variation for all parameters generally lower than 19%. An intense self-heating ensured the existence of a spontaneous thermophilic phase in all reactors. The average loss of total organic matter (TOM) was 46% of the initial content. Compared to the initial mixture, the hot water soluble fraction decreased by 62%, the hemicellulose-like fraction by 68%, the cellulose-like fraction by 50% and the lignin-like fractions by 12% in the final compost. The TOM losses, compost stabilisation and evolution of the biochemical fractions were similar to observed in large reactors or on-site experiments, excluding the lignin degradation, which was less important than in full-scale systems. The reproducibility of the process and the quality of the final compost make it possible to propose the use of this experimental device for research requiring a mass reduction of the initial composted waste mixtures. Copyright © 2011 Elsevier Ltd. All rights reserved.

Role of H2O2 in the fluctuating patterns of COD (chemical oxygen demand) during the treatment of palm oil mill effluent (POME) using pilot scale triple frequency ultrasound cavitation reactor.

PubMed

Manickam, Sivakumar; Abidin, Norhaida binti Zainal; Parthasarathy, Shridharan; Alzorqi, Ibrahim; Ng, Ern Huay; Tiong, Timm Joyce; Gomes, Rachel L; Ali, Asgar

2014-07-01

Palm oil mill effluent (POME) is a highly contaminating wastewater due to its high chemical oxygen demand (COD) and biochemical oxygen demand (BOD). Conventional treatment methods require longer residence time (10-15 days) and higher operating cost. Owing to this, finding a suitable and efficient method for the treatment of POME is crucial. In this investigation, ultrasound cavitation technology has been used as an alternative technique to treat POME. Cavitation is the phenomenon of formation, growth and collapse of bubbles in a liquid. The end process of collapse leads to intense conditions of temperature and pressure and shock waves which assist various physical and chemical transformations. Two different ultrasound systems i.e. ultrasonic bath (37 kHz) and a hexagonal triple frequency ultrasonic reactor (28, 40 and 70 kHz) of 15 L have been used. The results showed a fluctuating COD pattern (in between 45,000 and 60,000 mg/L) while using ultrasound bath alone, whereas a non-fluctuating COD pattern with a final COD of 27,000 mg/L was achieved when hydrogen peroxide was introduced. Similarly for the triple frequency ultrasound reactor, coupling all the three frequencies resulted into a final COD of 41,300 mg/L compared to any other individual or combination of two frequencies. With the possibility of larger and continuous ultrasonic cavitational reactors, it is believed that this could be a promising and a fruitful green process engineering technique for the treatment of POME. Copyright © 2014 Elsevier B.V. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Monteleone, S.

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updatedmore » and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.« less

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE PAGES

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

2018-02-02

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Environmental radiation protection studies related to nuclear industries, using AMS

NASA Astrophysics Data System (ADS)

Hellborg, Ragnar; Erlandsson, Bengt; Faarinen, Mikko; Hâkansson, Helena; Hâkansson, Kjell; Kiisk, Madis; Magnusson, Carl-Erik; Persson, Per; Skog, Göran; Stenström, Kristina; Mattsson, Sören; Thornberg, Charlotte

2001-07-01

14C is produced in nuclear reactors during normal operation and part of it is continuously released into the environment. Because of the biological importance of carbon and the long physical half-life of 14C it is of interest to study these releases. The 14C activity concentrations in the air and vegetation around some Swedish as well as foreign nuclear facilities have been measured by accelerator mass spectrometry (AMS). 59Ni is produced by neutron activation in the stainless steel close to the core of a nuclear reactor. The 59Ni levels have been measured in order to be able to classify the different parts of the reactor with respect to their content of long-lived radionuclides before final storage. The technique used to measure 59Ni at a small accelerator such as the Lund facility has been developed over the past few years and material from the Swedish nuclear industry has been analyzed.

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Field Testing of a Pneumatic Regolith Feed System During a 2010 ISRU Field Campaign on Mauna Kea, Hawaii

NASA Technical Reports Server (NTRS)

Craft, Jack; Zacny, Kris; Chu, Philip; Wilson, Jack; Santoro, Chris; Carlson, Lee; Maksymuk, Michael; Townsend, Ivan I.; Mueller, Robert P.; Mantovani, James G.

2010-01-01

Lunar In Situ Resource Utilization (ISRU) consists of a number of tasks starting with mining of lunar regolith, followed by the transfer of regolith to an oxygen extraction reactor and finally processing the regolith and storing of extracted oxygen. The transfer of regolith from the regolith hopper at the ground level to an oxygen extraction reactor many feet above the surface could be accomplished in different ways, including using a mechanical auger, bucket ladder system or a pneumatic system. The latter system is commonly used on earth when moving granular materials since it offers high reliability and simplicity of operation. In this paper, we describe a pneumatic regolith feed system, delivering feedstock to a Carbothermal reactor and lessons learned from deploying the system during the 2010 ISRU field campaign on the Mauna Kea, Hawaii.

Vibration and acoustic noise emitted by dry-type air-core reactors under PWM voltage excitation

NASA Astrophysics Data System (ADS)

Li, Jingsong; Wang, Shanming; Hong, Jianfeng; Yang, Zhanlu; Jiang, Shengqian; Xia, Shichong

2018-05-01

According to coupling way between the magnetic field and the structural order, structure mode is discussed by engaging finite element (FE) method and both natural frequency and modal shape for a dry-type air-core reactor (DAR) are obtained in this paper. On the basis of harmonic response analysis, electromagnetic force under PWM (Pulse Width Modulation) voltage excitation is mapped with the structure mesh, the vibration spectrum is gained and the consequences represents that the whole structure vibration predominates in the radial direction, with less axial vibration. Referring to the test standard of reactor noise, the rules of emitted noise of the DAR are measured and analyzed at chosen switching frequency matches the sample resonant frequency and the methods of active vibration and noise reduction are put forward. Finally, the low acoustic noise emission of a prototype DAR is verified by measurement.

Dismantling of Loop-Type Channel Equipment of MR Reactor in NRC 'Kurchatov Institute' - 13040

DOE Office of Scientific and Technical Information (OSTI.GOV)

Volkov, Victor; Danilovich, Alexey; Zverkov, Yuri

2013-07-01

In 2009 the project of decommissioning of MR and RTF reactors was developed and approved by the Expert Authority of the Russian Federation (Gosexpertiza). The main objective of the decommissioning works identified in this project: - complete dismantling of reactor equipment and systems; - decontamination of reactor premises and site in accordance with the established sanitary and hygienic standards. At the preparatory stage (2008-2010) of the project the following works were executed: loop-type channels' dismantling in the storage pool; experimental fuel assemblies' removal from spent fuel repositories in the central hall; spent fuel assembly removal from the liquid-metal-cooled loop-type channelmore » of the reactor core and its placement into the SNF repository; and reconstruction of engineering support systems to the extent necessary for reactor decommissioning. The project assumes three main phases of dismantling and decontamination: - dismantling of equipment/pipelines of cooling circuits and loop-type channels, and auxiliary reactor equipment (2011-2012); - dismantling of equipment in underground reactor premises and of both MR and RTF in-vessel devices (2013-2014); - decontamination of reactor premises; rehabilitation of the reactor site; final radiation survey of reactor premises, loop-type channels and site; and issuance of the regulatory authorities' de-registration statement (2015). In 2011 the decommissioning license for the two reactors was received and direct MR decommissioning activities started. MR primary pipelines and loop-type facilities situated in the underground reactor hall were dismantled. Works were also launched to dismantle the loop-type channels' equipment in underground reactor premises; reactor buildings were reconstructed to allow removal of dismantled equipment; and the MR/RTF decommissioning sequence was identified. In autumn 2011 - spring 2012 results of dismantling activities performed are: - equipment from underground rooms (No. 66, 66A, 66B, 72, 64, 63) - as well as from water and gas loop corridors - was dismantled, with the total radwaste weight of 53 tons and the total removed activity of 5,0 x 10{sup 10} Bq; - loop-type channel equipment from underground reactor hall premises was dismantled; - 93 loop-type channels were characterized, chopped and removed, with radwaste of 2.6 x 10{sup 13} Bq ({sup 60}Co) and 1.5 x 10{sup 13} Bq ({sup 137}Cs) total activity removed from the reactor pool, fragmented and packaged. Some of this waste was placed into the high-level waste (HLW) repository of the Center. Dismantling works were executed with application of remotely operated mechanisms, which promoted decrease of radiation impact on the personnel. The average individual dose for the personnel was 1.9 mSv/year in 2011, and the collective dose is estimated as 0.0605 man x Sv/year. (authors)« less

Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woods, Brian; Gutowska, Izabela; Chiger, Howard

Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion tomore » ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with an electrical heat source. The peak core region temperature capability is 1400°C. As part of this project, an inventory of test facilities that could be used for these experimental programs was completed. Several of these facilities showed some promise, however, upon further investigation it became clear that only the OSU HTTF had the power and/or peak temperature limits that would allow for the experimental programs envisioned herein. Thus the conceptual design and feasibility study development focused on examining the feasibility of configuring the current HTTF to collect validation data for these experimental programs. In addition to the scaling analyses and conceptual design development, a test plan was developed for the envisioned modified test facility. This test plan included a discussion on an appropriate shakedown test program as well as the specific matrix tests. Finally, a feasibility study was completed to determine the cost and schedule considerations that would be important to any test program developed to investigate these designs and events.« less

FINAL PROJECT REPORT - EVALUATION AND TESTING OF HTGR REACTOR BUILDING RESPONSE TO DEPRESSURIZATION ACCIDENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

ALLIANCE LIMITED, NGNP INDUSTRY

This report provides a description of the project, summarizes each phase of the project, and ends with project conclusions. In addition, the report contains a descriptive index of the technical reports generated during the course of the project.

76 FR 80409 - Draft Supplement 2 to Final Environmental Statement Related to the Operation of Watts Bar Nuclear...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-23

... image files of the NRC's public documents. If you do not have access to ADAMS or if there are problems.... Stephen J. Campbell, Chief, Watts Bar Special Projects Branch, Division of Operating Reactor Licensing...

Fabrication Control Plan for ORNL RH-LOCA ATF Test Specimens to be Irradiated in the ATR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Howard, Richard; Teague, Michael

2014-06-01

The purpose of this fabrication plan is (1) to summarize the design of a set of rodlets that will be fabricated and then irradiated in the Advanced Test Reactor (ATR) and (2) provide requirements for fabrication and acceptance criteria for inspections of the Light Water Reactor (LWR) – Accident Tolerant Fuels (ATF) rodlet components. The functional and operational (F&OR) requirements for the ATF program are identified in the ATF Test Plan. The scope of this document only covers fabrication and inspections of rodlet components detailed in drawings 604496 and 604497. It does not cover the assembly of these items tomore » form a completed test irradiation assembly or the inspection of the final assembly, which will be included in a separate INL final test assembly specification/inspection document. The controls support the requirements that the test irradiations must be performed safely and that subsequent examinations must provide valid results.« less

Composting domestic sewage sludge with natural zeolites in a rotary drum reactor.

PubMed

Villaseñor, J; Rodríguez, L; Fernández, F J

2011-01-01

This work aimed the influence of zeolites addition on a sludge-straw composting process using a pilot-scale rotary drum reactor. The type and concentration of three commercial natural zeolites were considered: a mordenite and two clinoptilolites (Klinolith and Zeocat). Mordenite caused the greatest carbon removal (58%), while the clinoptilolites halved losses of ammonium. All zeolites removed 100% of Ni, Cr, Pb, and significant amounts (more than 60%) of Cu, Zn and Hg. Zeocat displayed the greatest retention of ammonium and metals, and retention efficiencies increased as Zeocat concentration increased. The addition of 10% Zeocat produced compost compliant with Spanish regulations. Zeolites were separated from the final compost, and leaching studies suggested that zeolites leachates contained very low metals concentrations (<1 mg/kg). Thus, the final compost could be applied directly to soil, or metal-polluted zeolites could be separated from the compost prior to application. The different options have been discussed. Copyright © 2010 Elsevier Ltd. All rights reserved.

324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

HUMPHREYS, D C

A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Teammore » counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.« less

Corrosion of carbon steel by bacteria from North Sea offshore seawater injection systems: laboratory investigation.

PubMed

Stipanicev, Marko; Turcu, Florin; Esnault, Loïc; Rosas, Omar; Basseguy, Régine; Sztyler, Magdalena; Beech, Iwona B

2014-06-01

Influence of sulfidogenic bacteria, from a North Sea seawater injection system, on the corrosion of S235JR carbon steel was studied in a flow bioreactor; operating anaerobically for 100days with either inoculated or filtrated seawater. Deposits formed on steel placed in reactors contained magnesium and calcium minerals plus iron sulfide. The dominant biofilm-forming organism was an anaerobic bacterium, genus Caminicella, known to produce hydrogen sulfide and carbon dioxide. Open Circuit Potentials (OCP) of steel in the reactors was, for nearly the entire test duration, in the range -80045), suggested pitting on steel samples within the inoculated environment. However, the actual degree of corrosion could neither be directly correlated with the electrochemical data and nor with the steel corrosion in the filtrated seawater environment. Further laboratory tests are thought to clarify the noticed apparent discrepancies. © 2013.

Cure kinetics, morphologies, and mechanical properties of thermoplastic/MWCNT modified multifunctional glassy epoxies prepared via continuous reaction methods

NASA Astrophysics Data System (ADS)

Cheng, Xiaole

The primary goal of this dissertation is to develop a novel continuous reactor method to prepare partially cured epoxy prepolymers for aerospace prepreg applications with the aim of replacing traditional batch reactors. Compared to batch reactors, the continuous reactor is capable of solubilizing and dispersing a broad range of additives including thermoplastic tougheners, stabilizers, nanoparticles and curatives and advancing epoxy molecular weights and viscosities while reducing energy consumption. In order to prove this concept, polyethersulfone (PES) modified 4, 4'-diaminodiphenylsulfone (44DDS)/tetraglycidyl-4, 4'-diaminodiphenylmethane (TGDDM) epoxy prepolymers were firstly prepared using both continuous reactor and batch reactor methods. Kinetic studies confirmed the chain extension reaction in the continuous reactor is similar to the batch reactor, and the molecular weights and viscosities of prepolymers were readily controlled through reaction kinetics. Atomic force microscopy (AFM) confirmed similar cured network morphologies for formulations prepared from batch and continuous reactors. Additionally tensile strength, tensile modulus and fracture toughness analyses concluded mechanical properties of cured epoxy matrices produced from both reactors were equivalent. Effects of multifunctional epoxy compositions on thermoplastics phase-separated morphologies were systematically studied using a combination of AFM with nanomechanical mapping, spectroscopic and calorimetric techniques to provide new insights to tailor cured reaction induced phase separation (CRIPS) in multifunctional epoxy blend networks. Furthermore, how resultant crosslinked glassy polymer network and phase-separated morphologies correlated with mechanical properties are discussed in detail. Multiwall carbon nanotube (MWCNT)/TGDDM epoxy prepolymers were further prepared by combining the successful strategies for advancing epoxy chemistries and dispersing nanotubes using the continuous reactor. Optical microscopy (OM) and scanning electron microscopy (SEM) were used to characterize the MWCNT dispersion states and stabilization in epoxy prepolymer matrix after continuous process and during curing cycles. Additionally, electrical conductivities and mechanical properties of final cured MWCNT/TGDDM composites were measured and discussed in view of their corresponding MWCNT dispersion states. Ternary blends of MWCNT reinforced thermoplastic/epoxy prepolymers were prepared by the continuous reactor. Influence of MWCNT on the CRIPS mechanism and the cured morphologies were systematically investigated using SEM and rheological analysis. Incorporation of MWCNT in thermoplastic/epoxy matrices can lead to a morphological transformation from phase inverted, to co-continuous, and to droplet dispersed morphology. In additional, dynamic mechanical analysis revealed the heterogeneity of MWCNT dispersion in thermoplastic/thermosets systems.

Artificial neural networks modelling the prednisolone nanoprecipitation in microfluidic reactors.

PubMed

Ali, Hany S M; Blagden, Nicholas; York, Peter; Amani, Amir; Brook, Toni

2009-06-28

This study employs artificial neural networks (ANNs) to create a model to identify relationships between variables affecting drug nanoprecipitation using microfluidic reactors. The input variables examined were saturation levels of prednisolone, solvent and antisolvent flow rates, microreactor inlet angles and internal diameters, while particle size was the single output. ANNs software was used to analyse a set of data obtained by random selection of the variables. The developed model was then assessed using a separate set of validation data and provided good agreement with the observed results. The antisolvent flow rate was found to have the dominant role on determining final particle size.

MCNP6 simulated performance of Micro-Pocket Fission Detectors (MPFDs) in the Transient REActor Test (TREAT) Facility

DOE PAGES

Reichenberger, Michael A.; Patel, Vishal K.; Roberts, Jeremy A.; ...

2017-03-03

Here, Micro-Pocket Fission Detectors (MPFDs) are under development for in-core neutron flux measurements at the Transient REActor Test facility (TREAT) and in other experiments at Idaho National Laboratory (INL). The sensitivity of MPFDs to the energy dependent neutron flux at TREAT has been determined for 0.0300-μm thick active material coatings of 242Pu, 232Th, natural uranium, and 93% enriched 235U. Self-shielding effects in the active material of the MPFD was also confirmed to be negligible. Finally, fission fragment energy deposition was found to be in conformance with previously reported results.

Performance testing and analyses of the VSC-17 ventilated concrete cask. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

1992-05-01

This document details performance test which was conducted on a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask configured for pressurized-water reactor (PWR) spent fuel. The performance test consisted of loading the VSC-17 cask with 17 canisters of consolidated PWR spent fuel from Virginia Power`s Surry and Florida Power & Light Turkey Point reactors. Cask surface, concrete, air channel surfaces, and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in a vertical cask orientation. Data on spent fuel integrity were also obtained.

Diels–Alder reactions of myrcene using intensified continuous-flow reactors

PubMed Central

Álvarez-Diéguez, Miguel Á; Kohl, Thomas M; Tsanaktsidis, John

2017-01-01

This work describes the Diels–Alder reaction of the naturally occurring substituted butadiene, myrcene, with a range of different naturally occurring and synthetic dienophiles. The synthesis of the Diels–Alder adduct from myrcene and acrylic acid, containing surfactant properties, was scaled-up in a plate-type continuous-flow reactor with a volume of 105 mL to a throughput of 2.79 kg of the final product per day. This continuous-flow approach provides a facile alternative scale-up route to conventional batch processing, and it helps to intensify the synthesis protocol by applying higher reaction temperatures and shorter reaction times. PMID:28228853

Characterization of BOR-60 Irradiated 14YWT-NFA1 Tubes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saleh, Tarik A.; Maloy, Stuart Andrew; Aydogan, Eda

2017-02-15

Tubes of FCRD 14YWT-NFA1 Alloy were placed in the BOR-60 reactor and irradiated under a fast flux neutron environment to two conditions: 7 dpa at 360-370 °C and 6 dpa at 385-430 °C. Small sections of the tube were cut and sent to UC Berkeley for nanohardness testing and focused ion beam (FIB) milling of TEM specimens. FIB specimens were sent back to LANL for final FIB milling and TEM imaging. Hardness data and TEM images are presented in this report. This is the first fast reactor neutron irradiated information on the 14YWT-NFA1 alloy.

Selection of alternative central-station technologies for the Satellite Power System (SPS) comparative assessment

NASA Technical Reports Server (NTRS)

Samsa, M.

1980-01-01

An important effort is the Satellite Power System (SPS) comparative Assessment is the selection and characterization of alternative technologies to be compared with the SPS concept. The ground rules, criteria, and screening procedure applied in the selection of those alternative technologies are summarized. The final set of central station alternatives selected for comparison with the SPS concept includes: (1) light water reactor with improved fuel utilization, (2) conventional coal combustion with improved environmental controls, (3) open cycle gas turbine with integral low Btu gasifier, (4) terrestrial photovoltaic, (5) liquid metal fast breeder reactor, and (6) magnetic confinement fusion.

76 FR 5218 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-28

..., October 21, 2010 (74 FR 65038-65039). Thursday, February 10, 2011, Conference Room T2-B1, 11545 Rockville... Evaluation Report Associated with the License Renewal Application for the Palo Verde Nuclear Generating... NRC staff and Arizona Public Service Company regarding the final Safety Evaluation Report associated...

Isotopic Transmutations in Irradiated Beryllium and Their Implications on MARIA Reactor Operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrzejewski, Krzysztof J.; Kulikowska, Teresa A

2004-04-15

Beryllium irradiated by neutrons with energies above 0.7 MeV undergoes (n,{alpha}) and (n,2n) reactions. The Be(n,{alpha}) reaction results in subsequent buildup of {sup 6}Li and {sup 3}He isotopes with large thermal neutron absorption cross sections causing poisoning of irradiated beryllium. The amount of the poison isotopes depends on the neutron flux level and spectrum. The high-flux MARIA reactor operated in Poland since 1975 consists of a beryllium matrix with fuel channels in cutouts of beryllium blocks. As the experimental determination of {sup 6}Li, {sup 3}H, and {sup 3}He content in the operational reactor is impossible, a systematic computational study ofmore » the effect of {sup 3}He and {sup 6}Li presence in beryllium blocks on MARIA reactor reactivity and power density distribution has been undertaken. The analysis of equations governing the transmutation has been done for neutron flux parameters typical for MARIA beryllium blocks. Study of the mutual influence of reactor operational parameters and the buildup of {sup 6}Li, {sup 3}H, and {sup 3}He in beryllium blocks has shown the necessity of a detailed spatial solution of transmutation equations in the reactor, taking into account the whole history of its operation. Therefore, fuel management calculations using the REBUS code with included chains for Be(n,{alpha})-initiated reactions have been done for the whole reactor lifetime. The calculated poisoning of beryllium blocks has been verified against the critical experiment of 1993. Finally, the current {sup 6}Li, {sup 3}H, and {sup 3}He contents, averaged for each beryllium block, have been calculated. The reactivity drop caused by this poisoning is {approx}7%.« less

Advanced Instrumentation for Transient Reactor Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corradini, Michael L.; Anderson, Mark; Imel, George

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and designmore » increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).« less

Calculation and comparison of xenon and samarium reactivities of the HEU, LEU core in the low power research reactor.

PubMed

Dawahra, S; Khattab, K; Saba, G

2015-07-01

Comparative studies for the conversion of the fuel from HEU to LEU in the Miniature Neutron Source Reactor (MNSR) have been performed using the MCNP4C and GETERA codes. The precise calculations of (135)Xe and (149)Sm concentrations and reactivities were carried out and compared during the MNSR operation time and after shutdown for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) in this paper using the MCNP4C and GETERA codes. It was found that the (135)Xe and (149)Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The (149)Sm reactivities could be neglected compared to (135)Xe reactivities during the reactor operating time and after shutdown. The calculations for the UAl4-Al produced the highest (135)Xe reactivity in all the studied fuel group during the reactor operation (0.39 mk) and after the reactor shutdown (0.735 mk), It followed by U3Si-Al (0.34 mk, 0.653 mk), U3Si2-Al (0.33 mk, 0.634 mk), U9Mo-Al (0.3 mk, 0.568 mk) and UO2 (0.24 mk, 0.448 mk) fuels, respectively. Finally, the results showed that the UO2 was the best candidate for fuel conversion to LEU in the MNSR since it gave the lowest (135)Xe reactivity during the reactor operation and after shutdown. Copyright © 2015 Elsevier Ltd. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

During this reporting period, there were three major thrusts in the WVU portion. First, we started a preliminary investigation on the use of a membrane reactor for HAS. Accordingly, the plug-flow reactor which had been isolated from sulfides was substituted by a membrane reactor. The tubular membrane was first characterized in terms of its permeation properties, i.e., the fluxes, permeances and selectivities of the components. After that, a BASF methanol-synthesis catalyst was tested under different conditions on the membrane reactor. The results will be compared with those from a non-permeable stainless steel tubular reactor under the same conditions. Second, wemore » started a detailed study of one of the catalysts tested during the screening runs. Accordingly, a carbon-supported potassium-doped molybdenum-cobalt catalyst was selected to be run in the Rotoberty reactor. Finally, we have started detailed analyses of reaction products from some earlier screening runs in which non-sulfide molybdenum-based catalysts were employed and much more complicated product distributions were generally observed. These products could not hitherto be analyzed using the gas chromatograph which was then available. A Varian gas chromatograph/mass spectrometer (GC/MS) is being used to characterize these liquid products. At UCC, we completed a screening of an Engelhard support impregnated with copper and cesium. We have met or exceeded three of four catalyst development targets. Oxygenate selectivity is our main hurdle. Further, we tested the effect of replacing stainless-steel reactor preheater tubing and fittings with titanium ones. We had hoped to reduce the yield of hydrocarbons which may have been produced at high temperatures due to Fischer-Tropsch catalysis with the iron and nickel in the preheater tube walls. Results showed that total hydrocarbon space time yield was actually increased with the titanium preheater, while total alcohol space time yield was not significantly affected.« less

Treatment of raw and ozonated oil sands process-affected water under decoupled denitrifying anoxic and nitrifying aerobic conditions: a comparative study.

PubMed

Xue, Jinkai; Zhang, Yanyan; Liu, Yang; Gamal El-Din, Mohamed

2016-11-01

Batch experiments were performed to evaluate biodegradation of raw and ozonated oil sands process-affected water (OSPW) under denitrifying anoxic and nitrifying aerobic conditions for 33 days. The results showed both the anoxic and aerobic conditions are effective in degrading OSPW classical and oxidized naphthenic acids (NAs) with the aerobic conditions demonstrating higher removal efficiency. The reactors under nitrifying aerobic condition reduced the total classical NAs of raw OSPW by 69.1 %, with better efficiency for species of higher hydrophobicity. Compared with conventional aerobic reactor, nitrifying aerobic condition substantially shortened the NA degradation half-life to 16 days. The mild-dose ozonation remarkably accelerated the subsequent aerobic biodegradation of classical NAs within the first 14 days, especially for those with long carbon chains. Moreover, the ozone pretreatment enhanced the biological removal of OSPW classical NAs by leaving a considerably lower final residual concentration of 10.4 mg/L under anoxic conditions, and 5.7 mg/L under aerobic conditions. The combination of ozonation and nitrifying aerobic biodegradation removed total classical NAs by 76.5 % and total oxy-NAs (O3-O6) by 23.6 %. 454 Pyrosequencing revealed that microbial species capable of degrading recalcitrant hydrocarbons were dominant in all reactors. The most abundant genus in the raw and ozonated anoxic reactors was Thauera (~56 % in the raw OSPW anoxic reactor, and ~65 % in the ozonated OSPW anoxic reactor); whereas Rhodanobacter (~40 %) and Pseudomonas (~40 %) dominated the raw and ozonated aerobic reactors, respectively. Therefore, the combination of mild-dose ozone pretreatment and subsequent biological process could be a competent choice for OSPW treatment.

Predictive characterization of aging and degradation of reactor materials in extreme environments. Final report, December 20, 2013 - September 20, 2017

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qu, Jianmin

Understanding of reactor material behavior in extreme environments is vital not only to the development of new materials for the next generation nuclear reactors, but also to the extension of the operating lifetimes of the current fleet of nuclear reactors. To this end, this project conducted a suite of unique experimental techniques, augmented by a mesoscale computational framework, to understand and predict the long-term effects of irradiation, temperature, and stress on material microstructures and their macroscopic behavior. The experimental techniques and computational tools were demonstrated on two distinctive types of reactor materials, namely, Zr alloys and high-Cr martensitic steels. Thesemore » materials are chosen as the test beds because they are the archetypes of high-performance reactor materials (cladding, wrappers, ducts, pressure vessel, piping, etc.). To fill the knowledge gaps, and to meet the technology needs, a suite of innovative in situ transmission electron microscopy (TEM) characterization techniques (heating, heavy ion irradiation, He implantation, quantitative small-scale mechanical testing, and various combinations thereof) were developed and used to elucidate and map the fundamental mechanisms of microstructure evolution in both Zr and Cr alloys for a wide range environmental boundary conditions in the thermal-mechanical-irradiation input space. Knowledge gained from the experimental observations of the active mechanisms and the role of local microstructural defects on the response of the material has been incorporated into a mathematically rigorous and comprehensive three-dimensional mesoscale framework capable of accounting for the compositional variation, microstructural evolution and localized deformation (radiation damage) to predict aging and degradation of key reactor materials operating in extreme environments. Predictions from this mesoscale framework were compared with the in situ TEM observations to validate the model.« less

75 FR 28074 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-19

..., 2009, (74 FR 52829-52830). Wednesday, June 9, 2010, Conference Room T2-B1, Two White Flint North... Regulatory Guide (RG) 1.216, ``Containment Structural Integrity Evaluation for Internal Pressure Loadings... representatives of the NRC staff regarding draft final RG 1.216, ``Containment Structural Integrity Evaluation for...

10 CFR 52.98 - Finality of combined licenses; information requests.

Code of Federal Regulations, 2011 CFR

2011-01-01

... condition of the combined license, the design of the facility, the inspections, tests, analyses, and acceptance criteria contained in the license which are not derived from a referenced standard design... chapter, as applicable. (b) If the combined license does not reference a design certification or a reactor...

10 CFR 52.98 - Finality of combined licenses; information requests.

Code of Federal Regulations, 2010 CFR

2010-01-01

... condition of the combined license, the design of the facility, the inspections, tests, analyses, and acceptance criteria contained in the license which are not derived from a referenced standard design... chapter, as applicable. (b) If the combined license does not reference a design certification or a reactor...

Synthesis of calculational methods for design and analysis of radiation shields for nuclear rocket systems

NASA Technical Reports Server (NTRS)

Capo, M. A.; Disney, R. K.; Jordan, T. A.; Soltesz, R. G.; Woodsum, H. C.

1969-01-01

Eight computer programs make up a nine volume synthesis containing two design methods for nuclear rocket radiation shields. The first design method is appropriate for parametric and preliminary studies, while the second accomplishes the verification of a final nuclear rocket reactor design.

A COST BENEFIT APPROACH TO REACTOR SIZING AND NUTRIENT SUPPLY FOR BIOTRICKLING FILTERS FOR AIR POLLUTION CONTROL. (R825392)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

75 FR 54961 - Final Supplemental Environmental Impact Statement, Single Nuclear Unit at the Bellefonte Plant...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-09

... facility, a wind farm, a methane- gas cofiring facility, and several small solar photovoltaic facilities... maintenance of select plant systems and other regulatory compliance activities. Major buildings and plant... the plant cooling towers and the reactor, auxiliary, control, turbine, office, and service buildings...

EXPERIMENTAL INVESTIGATION OF MERCURY TRANSFORMATIONS IN PILOT-SCALE COMBUSTION SYSTEMS AND A BENCH-SCALE ENTRAINED FLOW REACTOR. (R827649)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

POLYMER PYROLYSIS AND OXIDATION STUDIES IN A CONTINUOUS FEED AND FLOW REACTOR: CELLULOSE AND POLYSTYRENE. (R824970)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

78 FR 45984 - Yankee Atomic Electric Company, Yankee Nuclear Power Station

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-30

... Electric Company, Yankee Nuclear Power Station AGENCY: Nuclear Regulatory Commission. ACTION: Environmental... (YAEC) is the holder of Possession-Only License DPR-3 for the Yankee Nuclear Power Station (YNPS... on the site of any nuclear power reactor. In its Statement of Considerations (SOC) for the Final Rule...

10 CFR 52.98 - Finality of combined licenses; information requests.

Code of Federal Regulations, 2012 CFR

2012-01-01

... condition of the combined license, the design of the facility, the inspections, tests, analyses, and acceptance criteria contained in the license which are not derived from a referenced standard design... chapter, as applicable. (b) If the combined license does not reference a design certification or a reactor...

10 CFR 52.98 - Finality of combined licenses; information requests.

Code of Federal Regulations, 2014 CFR

2014-01-01

... condition of the combined license, the design of the facility, the inspections, tests, analyses, and acceptance criteria contained in the license which are not derived from a referenced standard design... chapter, as applicable. (b) If the combined license does not reference a design certification or a reactor...

10 CFR 52.98 - Finality of combined licenses; information requests.

Code of Federal Regulations, 2013 CFR

2013-01-01

... condition of the combined license, the design of the facility, the inspections, tests, analyses, and acceptance criteria contained in the license which are not derived from a referenced standard design... chapter, as applicable. (b) If the combined license does not reference a design certification or a reactor...

77 FR 3288 - Receipt of Request for Action

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-23

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0004] Receipt of Request for Action Notice is hereby given... Director of the Office of Nuclear Reactor Regulation (NRR). As provided by 10 CFR 2.206, appropriate action... review to inform its final decision on whether to implement the requested actions. A copy of the petition...

Numerical investigation of flow and heat transfer in a novel configuration multi-tubular fixed bed reactor for propylene to acrolein process

NASA Astrophysics Data System (ADS)

Jiang, Bin; Hao, Li; Zhang, Luhong; Sun, Yongli; Xiao, Xiaoming

2015-01-01

In the present contribution, a numerical study of fluid flow and heat transfer performance in a pilot-scale multi-tubular fixed bed reactor for propylene to acrolein oxidation reaction is presented using computational fluid dynamics (CFD) method. Firstly, a two-dimensional CFD model is developed to simulate flow behaviors, catalytic oxidation reaction, heat and mass transfer adopting porous medium model on tube side to achieve the temperature distribution and investigate the effect of operation parameters on hot spot temperature. Secondly, based on the conclusions of tube-side, a novel configuration multi-tubular fixed-bed reactor comprising 790 tubes design with disk-and-doughnut baffles is proposed by comparing with segmental baffles reactor and their performance of fluid flow and heat transfer is analyzed to ensure the uniformity condition using molten salt as heat carrier medium on shell-side by three-dimensional CFD method. The results reveal that comprehensive performance of the reactor with disk-and-doughnut baffles is better than that of with segmental baffles. Finally, the effects of operating conditions to control the hot spots are investigated. The results show that the flow velocity range about 0.65 m/s is applicable and the co-current cooling system flow direction is better than counter-current flow to control the hottest temperature.

NuLat: A Novel Design for a Reactor Anti-Neutrino Detector

NASA Astrophysics Data System (ADS)

Rountree, S. Derek; NuLat Collaboration

2015-04-01

NuLat is a proposed very-short baseline (3-10m) reactor electron antineutrino (anti-νe) experiment that will probe the current best fit for light sterile neutrino mixing, the 5 MeV excess seen in current short baseline reactor experiments, and serve as a portable surface detector for cooperative (~ 30m baseline) surface monitoring of reactors. The NuLat detector will use an optically segmented 3D Raghavan optical lattice (ROL) detector that channels light via total internal reflection from a scintillation event down the 3 primary axes to the detector faces. The high degree of segmentation allows for each voxel's energy to be determined independently of other voxels, thus providing high temporal and spatial resolution and energy reconstruction independent of position. NuLat detects anti-νe via inverse beta decay (IBD), which produces a positron and a neutron. Most of the time, the positron deposits its kinetic energy into a single voxel allowing superior derivation of the incident anti-νe's energy. The final state neutron is captured via (n, α) on 6 Li or 10 B after a characteristic delay time giving a coincidence tag. This talk will discuss the physics reach of NuLat using a solid loaded scintillator, and the timeline of the NuLat reactor anti-νe program. This research has been funded in part by the National Science Foundation on Award Numbers 1001394 and 1001078.

Series-Bosch Technology for Oxygen Recovery During Lunar or Martian Surface Missions

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, J. Matthew; Rabenberg, Ellen; Stanley, Christine M.; Edmunson, Jennifer; Alleman, James E.; Chen, Kevin; Dumez, Sam

2014-01-01

Long-duration surface missions to the Moon or Mars will require life support systems that maximize resource recovery to minimize resupply from Earth. To address this need, NASA previously proposed a Series-Bosch (S-Bosch) oxygen recovery system, based on the Bosch process, which can theoretically recover 100% of the oxygen from metabolic carbon dioxide. Bosch processes have the added benefits of the potential to recover oxygen from atmospheric carbon dioxide and the use of regolith materials as catalysts, thereby eliminating the need for catalyst resupply from Earth. In 2012, NASA completed an initial design for an S-Bosch development test stand that incorporates two catalytic reactors in series including a Reverse Water-Gas Shift (RWGS) Reactor and a Carbon Formation Reactor (CFR). In 2013, fabrication of system components, with the exception of a CFR, and assembly of the test stand was initiated. Stand-alone testing of the RWGS reactor was completed to compare performance with design models. Continued testing of Lunar and Martian regolith simulants provided sufficient data to design a CFR intended to utilize these materials as catalysts. Finally, a study was conducted to explore the possibility of producing bricks from spent regolith catalysts. The results of initial demonstration testing of the RWGS reactor, results of continued catalyst performance testing of regolith simulants, and results of brick material properties testing are reported. Additionally, design considerations for a regolith-based CFR are discussed.

Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

DOE PAGES

Skavdahl, Isaac; Utgikar, Vivek; Christensen, Richard; ...

2016-05-24

We present an alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T co) and the hot outlet temperature of the intermediate heat exchanger (T ho2) by manipulating the hot-side flow rates of the heat exchangers (F h/F h2) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the controlmore » of the cold outlet temperature of the SHX (T co) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The final option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.« less

Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiao, Zhujie; Was, Gary; Bartels, David

2015-04-02

This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that themore » effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.« less

Series-Bosch Technology for Oxygen Recovery During Lunar or Martian Surface Missions

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, James M.; Stanley, Christine; Edmunson, Jennifer; Dumez, Samuel; Chen, Kevin; Alleman, James E.

2014-01-01

Long-duration surface missions to the Moon or Mars will require life support systems that maximize resource recovery to minimize resupply from Earth. To address this need, NASA previously proposed a Series-Bosch (S-Bosch) oxygen recovery system, based on the Bosch process, which can theoretically recover 100% of the oxygen from metabolic carbon dioxide. Bosch processes have the added benefits of the potential to recover oxygen from atmospheric carbon dioxide and the use of regolith materials as catalysts, thereby eliminating the need for catalyst resupply from Earth. In 2012, NASA completed an initial design for an S-Bosch development test stand that incorporates two catalytic reactors in series including a Reverse Water-Gas Shift (RWGS) Reactor and a Carbon Formation Reactor (CFR). In 2013, fabrication of system components, with the exception of a CFR, and assembly of the test stand was initiated. Stand-alone testing of the RWGS reactor was completed to compare performance with design models. Continued testing of Lunar and Martian regolith simulants provided sufficient data to design a CFR intended to utilize these materials as catalysts. Finally, a study was conducted to explore the possibility of producing bricks from spend regolith catalysts. The results of initial demonstration testing of the RWGS reactor, results of continued catalyst performance testing of regolith simulants, and results of brick material properties testing are reported. Additionally, design considerations for a regolith-based CFR are discussed.

A microprocessor tester for the treat upgrade reactor trip system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lenkszus, F.R.; Bucher, R.G.

1985-02-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. To improve the analytical extrapolation of test results to full-size assembly bundles, the facility upgrade will increase the maximum size of the test bundle from 7 to 37 fuel pins. By creating a core convertor zone around the test location, the neutron spectrum incident on the test assembly will be hardened and the maximum energy deposited in the sample will be increased. In addition, a programmable Automated Reactor Control System (ARCS) willmore » permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations. A quantitative reliability analysis of the RTS shows that the unreliability, that is, the probability of failure, is acceptable for a 10 hour mission time or risk interval.« less

Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis.

PubMed

Adam, Zachary R

2016-06-01

Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 10(5)-10(6) years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, D. J.; Varuttamaseni, A.

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portionmore » of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

A reload and startup plan for conversion of the NIST research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. J. Diamond

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reloadmore » portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

Modelling of slaughterhouse solid waste anaerobic digestion: determination of parameters and continuous reactor simulation.

PubMed

López, Iván; Borzacconi, Liliana

2010-10-01

A model based on the work of Angelidaki et al. (1993) was applied to simulate the anaerobic biodegradation of ruminal contents. In this study, two fractions of solids with different biodegradation rates were considered. A first-order kinetic was used for the easily biodegradable fraction and a kinetic expression that is function of the extracellular enzyme concentration was used for the slowly biodegradable fraction. Batch experiments were performed to obtain an accumulated methane curve that was then used to obtain the model parameters. For this determination, a methodology derived from the "multiple-shooting" method was successfully used. Monte Carlo simulations allowed a confidence range to be obtained for each parameter. Simulations of a continuous reactor were performed using the optimal set of model parameters. The final steady-states were determined as functions of the operational conditions (solids load and residence time). The simulations showed that methane flow peaked at a flow rate of 0.5-0.8 Nm(3)/d/m(reactor)(3) at a residence time of 10-20 days. Simulations allow the adequate selection of operating conditions of a continuous reactor. (c) 2010 Elsevier Ltd. All rights reserved.

On the Effectiveness of Wastewater Cylindrical Reactors: an Analysis Through Steiner Symmetrization

NASA Astrophysics Data System (ADS)

Díaz, J. I.; Gómez-Castro, D.

2016-03-01

The mathematical analysis of the shape of chemical reactors is studied in this paper through the research of the optimization of its effectiveness η such as introduced by R. Aris around 1960. Although our main motivation is the consideration of reactors specially designed for the treatment of wastewaters our results are relevant also in more general frameworks. We simplify the modeling by assuming a single chemical reaction with a monotone kinetics leading to a parabolic equation with a non-necessarily differentiable function. In fact we consider here the case of a single, non-reversible catalysis reaction of chemical order q, 00). We assume the chemical reactor of cylindrical shape Ω =G× (0,H) with G and open regular set of {R}2 not necessarily symmetric. We show that among all the sections G with prescribed area the ball is the set of lowest effectiveness η (t,G). The proof uses the notions of Steiner rearrangement. Finally, we show that if the height H is small enough then the effectiveness can be made as close to 1 as desired.

Continuous production of butanol from starch-based packing peanuts.

PubMed

Ezeji, Thaddeus C; Groberg, Marisa; Qureshi, Nasib; Blaschek, Hans P

2003-01-01

Acetone, butanol, ethanol (ABE, or solvents) were produced from starch-based packing peanuts in batch and continuous reactors. In a batch reactor, 18.9 g/L of total ABE was produced from 80 g/L packing peanuts in 110 h of fermentation. The initial and final starch concentrations were 69.6 and 11.1 g/L, respectively. In this fermentation, ABE yield and productivity of 0.32 and 0.17 g/(L h) were obtained, respectively. Compared to the batch fermentation, continuous fermentation of 40 g/L of starchbased packing peanuts in P2 medium resulted in a maximum solvent production of 8.4 g/L at a dilution rate of 0.033 h-1. This resulted in a productivity of 0.27 g/(L h). However, the reactor was not stable and fermentation deteriorated with time. Continuous fermentation of 35 g/L of starch solution resulted in a similar performance. These studies were performed in a vertical column reactor using Clostridium beijerinckii BA101 and P2 medium. It is anticipated that prolonged exposure of culture to acrylamide, which is formed during boiling/autoclaving of starch, affects the fermentation negatively.

Initial development and performance evaluation of a process for formation of dense carbon by pyrolysis of methane

NASA Technical Reports Server (NTRS)

Noyes, G. P.; Cusick, R. J.

1985-01-01

The three steps in pyrolytic carbon formation are: (1) gaseous hydrocarbon polymerization and aromatic formation; (2) gas-phase condensation and surface adsorption/impingement of polyaromatic hydrocarbon; and (3) final dehydration to carbon. The structure of the carbon in the various stages of formation is examined. The apparatuses and experimental procedures for the pyrolysis of methane in a 60 cm long quartz reactor tube at temperatures ranging from 1400-1600 K are described. The percentage of carbon converted and its density are calculated and tabularly presented. The results reveal that dense carbon formation is maximized and soot eliminated by this procedure. It is observed that conversion efficiency depends on the composition of the inlet gas and conversion increases with increasing temperature. Based on the experimental data a three-man carbon reactor subsystem (CRS) is developed; the functions of the Sabatier Methanation Reactor, two carbon formation reactors and fluid handling components of the CRS are analyzed. The CRS forms 16 kg of carbon at a rate of 0.8 kg/day for 20 days in a two percent volume density quartz wool packing at temperature of 1500-1600 K.

Design and scale-up of a semi-industrial downer-reactor for the rounding of irregular polymer particles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sachs, Marius; Schmidt, Jochen; Peukert, Wolfgang

2016-03-09

The recent development of rapid prototyping technologies towards additive manufacturing reveals some major drawbacks of processes such as laser beam melting (LBM). This contribution focuses on the lack of suitable polymer material with a fine particle size and good flowability. Polymer particles obtained by a wet grinding process 1 are treated in a heated downer reactor. This treatment changes the particles’ morphology from a chiselled state towards a spherical form by surface tension forces in a molten state 2 and leads to an improved flowability. To reach the required amount of rounded polymer powder, a downer reactor in semi-industrial scalemore » has been established and will be characterized in this article. For the purpose of particle rounding it is necessary to avoid contact of molten particles with each other and with the hot reactor walls. Furthermore, the heat distribution has been investigated as one of the key parameters of the process. Finally, a proof of concept by rounding wet grinded PBT material was successfully conducted. The product was examined to obtain data about a change in particle size and flowability.« less

Transmutation of Isotopes --- Ecological and Energy Production Aspects

NASA Astrophysics Data System (ADS)

Gudowski, Waclaw

2000-01-01

This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions --- after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors --- are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional nuclear power. In this context a lot of hopes and expectations have been expressed for novel systems called Accelerator-Driven Systems, Accelerator-Driven Transmutation of Waste or just Hybrid Reactors. All these names are used for description of the same nuclear system combining a powerful particle accelerator with a subcritical reactor. A careful analysis of possible environmental impact of ATW together with limitation of this technology is presented also in this paper.

Method of handling radioactive alkali metal waste

DOEpatents

Wolson, Raymond D.; McPheeters, Charles C.

1980-01-01

Radioactive alkali metal is mixed with particulate silica in a rotary drum reactor in which the alkali metal is converted to the monoxide during rotation of the reactor to produce particulate silica coated with the alkali metal monoxide suitable as a feed material to make a glass for storing radioactive material. Silica particles, the majority of which pass through a 95 mesh screen or preferably through a 200 mesh screen, are employed in this process, and the preferred weight ratio of silica to alkali metal is 7 to 1 in order to produce a feed material for the final glass product having a silica to alkali metal monoxide ratio of about 5 to 1.

Method of handling radioactive alkali metal waste

DOEpatents

Wolson, R.D.; McPheeters, C.C.

Radioactive alkali metal is mixed with particulate silica in a rotary drum reactor in which the alkali metal is converted to the monoxide during rotation of the reactor to produce particulate silica coated with the alkali metal monoxide suitable as a feed material to make a glass for storing radioactive material. Silica particles, the majority of which pass through a 95 mesh screen or preferably through a 200 mesh screen, are employed in this process, and the preferred weight ratio of silica to alkali metal is 7 to 1 in order to produce a feed material for the final glass product having a silica to alkali metal monoxide ratio of about 5 to 1.

System Modeling of Lunar Oxygen Production: Mass and Power Requirements

NASA Technical Reports Server (NTRS)

Steffen, Christopher J.; Freeh, Joshua E.; Linne, Diane L.; Faykus, Eric W.; Gallo, Christopher A.; Green, Robert D.

2007-01-01

A systems analysis tool for estimating the mass and power requirements for a lunar oxygen production facility is introduced. The individual modeling components involve the chemical processing and cryogenic storage subsystems needed to process a beneficiated regolith stream into liquid oxygen via ilmenite reduction. The power can be supplied from one of six different fission reactor-converter systems. A baseline system analysis, capable of producing 15 metric tons of oxygen per annum, is presented. The influence of reactor-converter choice was seen to have a small but measurable impact on the system configuration and performance. Finally, the mission concept of operations can have a substantial impact upon individual component size and power requirements.

The benefits of a fast reactor closed fuel cycle in the UK

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gregg, R.; Hesketh, K.

2013-07-01

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size,more » so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)« less

Pyrolysis of tyres. Influence of the final temperature of the process on emissions and the calorific value of the products recovered.

PubMed

Díez, C; Martínez, O; Calvo, L F; Cara, J; Morán, A

2004-01-01

A study was made of the pyrolysis of tyre particles, with the aim of determining the possibilities of using the products resulting from the process as fuel. Three final temperatures were used, determined from thermogravimetric data. The design of the experiment was a horizontal oven containing a reactor into which particles of the original tyre were placed. After the process, a solid fraction (char) remained in the reactor, while the gases generated went through a set of scrubbers where most of the condensable fraction (oils) was retained. Finally, once free of this fraction, the gases were collected in glass ampoules. Solid and liquids fractions were subjected to thermogravimetric analyses in order to study their combustibility. The gas fraction was analysed by means of gas chromatography to establish the content of CO, CO2, H2 and hydrocarbons present in the samples (mainly components of gases produced in the pyrolysis process). A special study was made of the sulphur and chlorine content of all the fractions, as the presence of these elements could be problematic if the products are used as fuel. Tyre pyrolysis engenders a solid carbon residue that concentrates sulphur and chorine, with a relatively high calorific value, although not so high as that of the original tyre. The liquid fraction produced by the process has a high calorific value, which rises with the final temperature, up to 40 MJ/kg. The chlorine content of this fraction is negligible. Over 95% of the gas fraction, regardless of the final temperature, is composed of hydrocarbons of a low molecular weight and hydrogen, this fraction also appearing to be free of chlorine.

Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hawari, Ayman; Ougouag, Abderrafi

2014-07-08

This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can bemore » easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.« less

Aeration costs in stirred-tank and bubble column bioreactors

DOE PAGES

Humbird, D.; Davis, R.; McMillan, J. D.

2017-08-10

To overcome knowledge gaps in the economics of large-scale aeration for production of commodity products, Aspen Plus is used to simulate steady-state oxygen delivery in both stirred-tank and bubble column bioreactors, using published engineering correlations for oxygen mass transfer as a function of aeration rate and power input, coupled with new equipment cost estimates developed in Aspen Capital Cost Estimator and validated against vendor quotations. Here, these simulations describe the cost efficiency of oxygen delivery as a function of oxygen uptake rate and vessel size, and show that capital and operating costs for oxygen delivery drop considerably moving from standard-sizemore » (200 m 3) to world-class size (500 m 3) reactors, but only marginally in further scaling up to hypothetically large (1000 m 3) reactors. Finally, this analysis suggests bubble-column reactor systems can reduce overall costs for oxygen delivery by 10-20% relative to stirred tanks at low to moderate oxygen transfer rates up to 150 mmol/L-h.« less

Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Benson; J. Cole; J. Jackson

2013-02-01

In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groupsmore » conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.« less

A facile and efficient method of enzyme immobilization on silica particles via Michael acceptor film coatings: immobilized catalase in a plug flow reactor.

PubMed

Bayramoglu, Gulay; Arica, M Yakup; Genc, Aysenur; Ozalp, V Cengiz; Ince, Ahmet; Bicak, Niyazi

2016-06-01

A novel method was developed for facile immobilization of enzymes on silica surfaces. Herein, we describe a single-step strategy for generating of reactive double bonds capable of Michael addition on the surfaces of silica particles. This method was based on reactive thin film generation on the surfaces by heating of impregnated self-curable polymer, alpha-morpholine substituted poly(vinyl methyl ketone) p(VMK). The generated double bonds were demonstrated to be an efficient way for rapid incorporation of enzymes via Michael addition. Catalase was used as model enzyme in order to test the effect of immobilization methodology by the reactive film surface through Michael addition reaction. Finally, a plug flow type immobilized enzyme reactor was employed to estimate decomposition rate of hydrogen peroxide. The highly stable enzyme reactor could operate continuously for 120 h at 30 °C with only a loss of about 36 % of its initial activity.

High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mcwilliams, A. J.

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniquesmore » through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.« less

Atmosphere Processing Module Automation and Catalyst Durability Analysis for Mars ISRU Pathfinder

NASA Technical Reports Server (NTRS)

Petersen, Elspeth M.

2016-01-01

The Mars In-Situ Resource Utilization Pathfinder was designed to create fuel using components found in the planet’s atmosphere and regolith for an ascension vehicle to return a potential sample return or crew return vehicle from Mars. The Atmosphere Processing Module (APM), a subunit of the pathfinder, uses cryocoolers to isolate and collect carbon dioxide from Mars simulant gas. The carbon dioxide is fed with hydrogen into a Sabatier reactor where methane is produced. The APM is currently undergoing the final stages of testing at Kennedy Space Center prior to process integration testing with the other subunits of the pathfinder. The automation software for the APM cryocoolers was tested and found to perform nominally. The catalyst used for the Sabatier reactor was investigated to determine the factors contributing to catalyst failure. The results from the catalyst testing require further analysis, but it appears that the rapid change in temperature during reactor start up or the elevated operating temperature is responsible for the changes observed in the catalyst.

Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, D.D.; Collins, J.L.

One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test usingmore » the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.« less

Nuclear power plant 5,000 to 10,000 kilowatts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

The purpose of this proposal is to present a suggested program for the development of an Aqueous Homogeneous Reactor Power Plant for the production of power in the 5000 to 10,000 kilowatt range under the terms of the Atomic Energy Commission's invitation of September 21, 1955. It envisions a research and development program prior to finalizing fabricating commitments of full scale components for the purpose of proving mechanical and hydraulic operating and chemical processing feasibility with the expectation that such preliminary effort will assure the contruction of the reactor at the lowest cost and successful operation at the earliest date.more » It proposes the construction of a reactor for an eventual net electrical output of ten megawatts but initially in conjunction with a five megawatt turbo-generating unit. This unit would be constructed at the site of the existing Hersey diesel generating plant of the Wolverine Electric Cooperative approximately ten miles north of Big Rapids, Michigan.« less

Aeration costs in stirred-tank and bubble column bioreactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Humbird, D.; Davis, R.; McMillan, J. D.

To overcome knowledge gaps in the economics of large-scale aeration for production of commodity products, Aspen Plus is used to simulate steady-state oxygen delivery in both stirred-tank and bubble column bioreactors, using published engineering correlations for oxygen mass transfer as a function of aeration rate and power input, coupled with new equipment cost estimates developed in Aspen Capital Cost Estimator and validated against vendor quotations. Here, these simulations describe the cost efficiency of oxygen delivery as a function of oxygen uptake rate and vessel size, and show that capital and operating costs for oxygen delivery drop considerably moving from standard-sizemore » (200 m 3) to world-class size (500 m 3) reactors, but only marginally in further scaling up to hypothetically large (1000 m 3) reactors. Finally, this analysis suggests bubble-column reactor systems can reduce overall costs for oxygen delivery by 10-20% relative to stirred tanks at low to moderate oxygen transfer rates up to 150 mmol/L-h.« less

76 FR 17159 - Office of New Reactors; Final Interim Staff Guidance on Standard Review Plan, Section 17.4...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-28

... Interim Staff Guidance on Standard Review Plan, Section 17.4, ``Reliability Assurance Program'' AGENCY... design reliability assurance program (RAP). This ISG updates the guidance provided to the staff in Standard Review Plan (SRP), Section 17.4, ``Reliability Assurance Program,'' of NUREG-0800, ``Standard...

76 FR 23340 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-26

... Thursday, October 21, 2010 (75 FR 65038-65039). Thursday, May 12, 2011, Conference Room T2-B1, 11545... Safety Evaluation Report Associated with the License Renewal Application for the Hope Creek Generating... NRC staff and PSEG Nuclear, LLC regarding the final safety evaluation report (SER) associated with the...

75 FR 81317 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-27

... Thursday, October 21, 2010 (74 FR 65038-65039). Thursday, January 13, 2011, Conference Room T2-B1, Two... Evaluation Report Associated With the Vogtle Units 3 and 4 Combined License Application (Open/Closed)-- The... Nuclear Company, and NuStar Energy regarding the Final Safety Evaluation Report associated with the Vogtle...

78 FR 33120 - Final Interim Staff Guidance LR-ISG-2011-04; Updated Aging Management Criteria for Reactor Vessel...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

..., ``Generic Aging Lessons Learned Report'' (GALL Report), for the aging management of Pressurized Water... communicate insights and lessons learned and to address emergent issues not covered in license renewal... ensure that PWR license renewal applicants will adequately address age-related degradation and aging...

10 CFR 52.157 - Contents of applications; technical information in final safety analysis report.

Code of Federal Regulations, 2010 CFR

2010-01-01

... analysis of the structures, systems, and components of the reactor to be manufactured, with emphasis upon... assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site... structures, systems, and components with the objective of assessing the risk to public health and safety...

Biodrying process: A sustainable technology for treatment of municipal solid waste with high moisture content.

PubMed

Tom, Asha P; Pawels, Renu; Haridas, Ajit

2016-03-01

Municipal solid waste with high moisture content is the major hindrance in the field of waste to energy conversion technologies and here comes the importance of biodrying process. Biodrying is a convective evaporation process, which utilizes the biological heat developed from the aerobic reactions of organic components. The numerous end use possibilities of the output are making the biodrying process versatile, which is possible by achieving the required moisture reduction, volume reduction and bulk density enhancement through the effective utilization of biological heat. In the present case study the detailed research and development of an innovative biodrying reactor has been carried out for the treatment of mixed municipal solid waste with high moisture content. A pilot scale biodrying reactor of capacity 565 cm(3) was designed and set up in the laboratory. The reactor dimensions consisted of an acrylic chamber of 60 cm diameter and 200 cm height, and it was enveloped by an insulation chamber. The insulation chamber was provided to minimise the heat losses through the side walls of the reactor. It simulates the actual condition in scaling up of the reactor, since in bigger scale reactors the heat losses through side walls will be negligible while comparing the volume to surface area ratio. The mixed municipal solid waste with initial moisture content of 61.25% was synthetically prepared in the laboratory and the reactor was fed with 109 kg of this substrate. Aerobic conditions were ensured inside the reactor chamber by providing the air at a constant rate of 40 litre per minute, and the direction of air flow was from the specially designed bottom air chamber to the reactor matrix top. The self heating inside reactor matrix was assumed in the range of 50-60°C during the design stage. Innovative biodrying reactor was found to be efficiently working with the temperature inside the reactor matrix rising to a peak value of 59°C by the fourth day of experiment (the peak observed at a height of 60 cm from the air supply). The process analyses results were promising with a reduction of 56.5% of volume, and an increase of 52% of bulk density of the substrate at the end of 33 days of biodrying. Also the weight of mixed MSW substrate has been reduced by 33.94% in 20 days of reaction and the average moisture reduction of the matrix was 20.81% (reduced from the initial value of 61.25% to final value of 48.5%). The moisture reduction would have been higher, if the condensation of evaporated water at the reactor matrix has been avoided. The non-homogeneous moisture reduction along the height of the reactor is evident and this needs further innovation. The leachate production has been completely eliminated in the innovative biodrying reactor and that is a major achievement in the field of municipal solid waste management technology. Copyright © 2016 Elsevier Ltd. All rights reserved.

Characterization and modelling of the heat transfers in a pilot-scale reactor during composting under forced aeration.

PubMed

de Guardia, A; Petiot, C; Benoist, J C; Druilhe, C

2012-06-01

The paper focused on the modelling of the heat transfers during composting in a pilot-scale reactor under forced aeration. The model took into account the heat production and the transfers by evaporation, convection between material and gas crossing the material, conduction and surface convection between gas and material in bottom and upper parts of the reactor. The model was adjusted thanks to the measurements practised during fifteen composting experiments in which five organic wastes were, each, composted under three constant aeration rates. Heat production was considered proportional to oxygen consumption rate and the enthalpy per mole oxygen consumed was assumed constant. The convective heat transfer coefficients were determined on basis of the continuous measurements of the temperatures of both the lid and the bottom part of the reactor. The model allowed a satisfying prediction of the temperature of the composting material. In most cases, the mean absolute discard between the experimental and the simulated temperatures was inferior to 2.5°C and the peaks of temperature occurred with less than 8h delay. For the half of the experiments the temperature discard between the simulated peak and the experimental one was inferior to 5°C. On basis of the calculation of a stoichiometric production of water through oxidation of the biodegradable organic matter, the simulation of water going out from material as vapour also allowed a rather satisfying prediction of the mass of water in final mixture. The influence of the aeration rate on every type of heat loss was characterized. Finally, the model was used to evaluate the impacts on material temperature caused by the change of the insulation thickness, the ambient temperature, take the lid away, the increase or the decrease of the mass of waste to compost. Copyright © 2012 Elsevier Ltd. All rights reserved.

Characterization and modelling of the heat transfers in a pilot-scale reactor during composting under forced aeration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guardia, A. de, E-mail: amaury.de-guardia@irstea.fr; Universite Europeenne de Bretagne, F-35000 Rennes; Petiot, C.

2012-06-15

The paper focused on the modelling of the heat transfers during composting in a pilot-scale reactor under forced aeration. The model took into account the heat production and the transfers by evaporation, convection between material and gas crossing the material, conduction and surface convection between gas and material in bottom and upper parts of the reactor. The model was adjusted thanks to the measurements practised during fifteen composting experiments in which five organic wastes were, each, composted under three constant aeration rates. Heat production was considered proportional to oxygen consumption rate and the enthalpy per mole oxygen consumed was assumedmore » constant. The convective heat transfer coefficients were determined on basis of the continuous measurements of the temperatures of both the lid and the bottom part of the reactor. The model allowed a satisfying prediction of the temperature of the composting material. In most cases, the mean absolute discard between the experimental and the simulated temperatures was inferior to 2.5 Degree-Sign C and the peaks of temperature occurred with less than 8 h delay. For the half of the experiments the temperature discard between the simulated peak and the experimental one was inferior to 5 Degree-Sign C. On basis of the calculation of a stoichiometric production of water through oxidation of the biodegradable organic matter, the simulation of water going out from material as vapour also allowed a rather satisfying prediction of the mass of water in final mixture. The influence of the aeration rate on every type of heat loss was characterized. Finally, the model was used to evaluate the impacts on material temperature caused by the change of the insulation thickness, the ambient temperature, take the lid away, the increase or the decrease of the mass of waste to compost.« less

Comparative performance of fixed-film biological filters: Application of reactor theory

USGS Publications Warehouse

Watten, B.J.; Sibrell, P.L.

2006-01-01

Nitrification is classified as a two-step consecutive reaction where R1 represents the rate of formation of the intermediate product NO2-N and R2 represents the rate of formation of the final product NO3-N. The relative rates of R1 and R2 are influenced by reactor type characterized hydraulically as plug-flow, plug-flow with dispersion and mixed-flow. We develop substrate conversion models for fixed-film biofilters operating in the first-order kinetic regime based on application of chemical reactor theory. Reactor type, inlet conditions and the biofilm kinetic constants Ki (h-1) are used to predict changes in NH4-N, NO2-N, NO3-N and BOD5. The inhibiting effects of the latter on R1 and R2 were established based on the ?? relation, e.g.:{A formula is presented}where BOD5,max is the concentration that causes nitrification to cease and N is a variable relating Ki to increasing BOD5. Conversion models were incorporated in spreadsheet programs that provided steady-state concentrations of nitrogen and BOD5 at several points in a recirculating aquaculture system operating with input values for fish feed rate, reactor volume, microscreen performance, make-up and recirculating flow rates. When rate constants are standardized, spreadsheet use demonstrates plug-flow reactors provide higher rates of R1 and R2 than mixed-flow reactors thereby reducing volume requirements for target concentrations of NH4-N and NO2-N. The benefit provided by the plug-flow reactor varies with hydraulic residence time t as well as the effective vessel dispersion number, D/??L. Both reactor types are capable of providing net increases in NO2-N during treatment but the rate of decrease in the mixed-flow case falls well behind that predicted for plug-flow operation. We show the potential for a positive net change in NO2-N increases with decreases in the dimensionless ratios K2, (R2 )/K1,( R1 ) and [NO2-N]/[NH4-N] and when the product K1, (R1) t provides low to moderate NH4-N conversions. Maintaining high levels of the latter reduces the effective reactor utilization rate (%) defined here as (RNavg/RNmax)100 where RNavg is the mean reactive nitrogen concentration ([NH4-N] + [NO2-N]) within the reactor, and RNmax represents the feed concentration of the same. Low utilization rates provide a hedge against unexpected increases in substrate loading and reduce water pumping requirements but force use of elevated reactor volumes. Further ?? effects on R1 and R2 can be reduced through use of a tanks-in-series versus a single mixed-flow reactor configuration and by improving the solids removal efficiency of microscreen treatment.

Porous media for catalytic renewable energy conversion

NASA Astrophysics Data System (ADS)

Hotz, Nico

2012-05-01

A novel flow-based method is presented to place catalytic nanoparticles into a reactor by sol-gelation of a porous ceramic consisting of copper-based nanoparticles, silica sand, ceramic binder, and a gelation agent. This method allows for the placement of a liquid precursor containing the catalyst into the final reactor geometry without the need of impregnating or coating of a substrate with the catalytic material. The so generated foam-like porous ceramic shows properties highly appropriate for use as catalytic reactor material, e.g., reasonable pressure drop due to its porosity, high thermal and catalytic stability, and excellent catalytic behavior. The catalytic activity of micro-reactors containing this foam-like ceramic is tested in terms of their ability to convert alcoholic biofuel (e.g. methanol) to a hydrogen-rich gas mixture with low concentrations of carbon monoxide (up to 75% hydrogen content and less than 0.2% CO, for the case of methanol). This gas mixture is subsequently used in a low-temperature fuel cell, converting the hydrogen directly to electricity. A low concentration of CO is crucial to avoid poisoning of the fuel cell catalyst. Since conventional Polymer Electrolyte Membrane (PEM) fuel cells require CO concentrations far below 100 ppm and since most methods to reduce the mole fraction of CO (such as Preferential Oxidation or PROX) have CO conversions of up to 99%, the alcohol fuel reformer has to achieve initial CO mole fractions significantly below 1%. The catalyst and the porous ceramic reactor of the present study can successfully fulfill this requirement.

Evaluation of A Novel Split-Feeding Anaerobic/Oxic Baffled Reactor (A/OBR) For Foodwaste Anaerobic Digestate: Performance, Modeling and Bacterial Community

PubMed Central

Wang, Shaojie; Peng, Liyu; Jiang, Yixin; Gikas, Petros; Zhu, Baoning; Su, Haijia

2016-01-01

To enhance the treatment efficiency from an anaerobic digester, a novel six-compartment anaerobic/oxic baffled reactor (A/OBR) was employed. Two kinds of split-feeding A/OBRs R2 and R3, with influent fed in the 1st, 3rd and 5th compartment of the reactor simultaneously at the respective ratios of 6:3:1 and 6:2:2, were compared with the regular-feeding reactor R1 when all influent was fed in the 1st compartment (control). Three aspects, the COD removal, the hydraulic characteristics and the bacterial community, were systematically investigated, compared and evaluated. The results indicated that R2 and R3 had similar tolerance to loading shock, but the R2 had the highest COD removal of 91.6% with a final effluent of 345 mg/L. The mixing patterns in both split-feeding reactors were intermediate between plug-flow and completely-mixed, with dead spaces between 8.17% and 8.35% compared with a 31.9% dead space in R1. Polymerase chain reaction-denaturing gradient gel electrophoresis (PCR-DGGE) analysis revealed that the split-feeding strategy provided a higher bacterial diversity and more stable bacterial community than that in the regular-feeding strategy. Further analysis indicated that Firmicutes, Bacteroidetes, and Proteobacteria were the dominant bacteria, among which Firmicutes and Bacteroidetes might be responsible for organic matter degradation and Proteobacteria for nitrification and denitrification. PMID:27708368

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mertyurek, Ugur; Gauld, Ian C.

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE PAGES

Mertyurek, Ugur; Gauld, Ian C.

2015-12-24

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Production of aluminum--silicon alloy and ferrosilicon and commercial purity aluminum by the direct reduction process. First interim technical report, Phase B, September 1--November 30, 1978

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bruno, M.J.

1978-12-01

The parameters of charge content, reaction temperatures and residence time were studied in a bench reactor concerning the production of Al--Si and Fe--Si alloys. Results confirmed that minimum final stage reaction temperature is 1950 to 2000/sup 0/C. Residence time varied with initial charge concentration. Fe/sub 2/O/sub 3/ additions to the charge produced a significant increase in metallic yield. A burden preparation procedure was developed for making acceptable agglomerates containing Fe/sub 2/O/sub 3/, bauxite, clay and coke. Particle size distribution of starting materials was correlated with agglomerate strength. A new bench scale reactor was designed and built to facilitate semi-continuous operation,more » using O/sub 2/ injection to burn coke supporting the burden, resulting in burden movement. In a number of runs bridging of the burden material occurred due to condensation of volatilized sub-oxides in the cooler zones of the reactor. The reactor operated smoothly as an iron blast furnace at 1500/sup 0/C, demonstrating the validity of the equipment and test procedures. Initial construction of pilot reactor VSR-1 was completed. Bench scale fractional crystallizer runs were continued to determine the impurity effects of Fe up to 6.9% and Ti up to 1.25% on alloy product concentration and yield. High levels of impurities formed intermetallic complexes with Al and reduced product yield.« less

Ichthyoplankton entrainment study at the SRS Savannah River water intakes for Westinghouse Savannah River Company. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paller, M.

1992-03-26

Cooling water for L and K Reactors and makeup water for Par Pond is pumped from the Savannah River at the 1G, 3G, and 5G pump houses. Ichthyoplankton (drifting fish larvae and eggs) from the river are entrained into the reactor cooling systems with the river water and passed through the reactor`s heat exchangers where temperatures may reach 70{degrees}C during full power operation. Ichthyoplankton mortality under such conditions is assumed to be 100 percent. The number of ichthyoplankton entrained into the cooling system depends on a variety of variables, including time of year, density and distribution of ichthyoplankton in themore » river, discharge levels in the river, and the volume of water withdrawn by the pumps. Entrainment at the 1 G pump house, which is immediately downstream from the confluence of Upper Three Runs Creek and the Savannah River, is also influenced by discharge rates and ichthyoplankton densities in Upper Three Runs Creek. Because of the anticipated restart of several SRS reactors and the growing concern surrounding striped bass and American shad stocks in the Savannah River, the Department of Energy requested that the Environmental Sciences Section (ESS) of the Savannah River Laboratory sample ichthyoplankton at the SRS Savannah River intakes. Dams & Moore, Inc., under a contract with Westinghouse Savannah River Company performed the sampling and data analysis for the ESS.« less

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE PAGES

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.; ...

2017-02-26

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

E. Blanford; E. Keldrauk; M. Laufer

2010-09-20

Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement,more » and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using factory prefabricated structural modules, for application to external event shell and base isolated structures.« less

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

NASA Astrophysics Data System (ADS)

Andrianova, E. A.; Tsibul'skiy, V. F.

2017-12-01

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and 238U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with 238U alone should be reprocessed in the usual way.

Noble gas, iodine, and cesium transport in a postulated loss of decay heat removal accident at Browns Ferry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichner, R.P.; Hodge, S.A.; Weber, C.F.

1984-08-01

This report presents an analysis of the movement of noble gas, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto themore » drywell floor. The analysis of fission product transport presented in this report is based on the no-operator-action sequence and provides an estimate of fission product inventories, as a function of time, within 14 control volumes outside the core, with the atmosphere considered as the final control volume in the transport sequence. As in the case of accident sequences previously studied, we find small barrier for noble gas ejection to air, these gases being effectively purged from the drywell and reactor building by steam and concrete degradation gases. However, significant decay of krypton isotopes occurs during the long delay times involved in this sequence. In contrast, large degrees of holdup for iodine and cesium are projected due to the chemical reactivity of these elements. Only about 2 x 10/sup -4/% of the initial iodine and cesium activity are predicted to be released to the atmosphere. Principal barriers for release are deposition on reactor vessel and containment walls. A significant amount of iodine is captured in the water pool formed in the reactor building basement after actuation of the fire protection system.« less

Electrochemical study of pre- and post-transition corrosion of Zr alloys in PWR coolant

NASA Astrophysics Data System (ADS)

Macák, Jan; Novotný, Radek; Sajdl, Petr; Renčiuková, Veronika; Vrtílková, Věra

Corrosion properties of Zr-Sn and Zr-Nb zirconium alloys were studied under simulated PWR conditions (or, more exactly, VVER conditions — boric acid, potassium hydroxide, lithium hydroxide) at temperatures up to 340°C and 15MPa using in-situ electrochemical impedance spectroscopy (EIS) and polarization measurements. EIS spectra were obtained in a wide range of frequencies (typically 100kHz — 100μHz). It enabled to gain information of both dielectric properties of oxide layers developing on the Zr-alloys surface and of the kinetics of the corrosion process and the associated charge and mass transfer phenomena. Experiments were run for more than 380 days; thus, the study of all the corrosion stages (pre-transition, transition, post-transition) was possible.

Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux

NASA Astrophysics Data System (ADS)

Al Zain, Jamal; El Hajjaji, O.; El Bardouni, T.; Boukhal, H.; Jaï, Otman

2018-06-01

The MNSR is a pool type research reactor, which is difficult to model because of the importance of neutron leakage. The aim of this study is to evaluate a 2-D transport model for the reactor compatible with the latest release of the DRAGON code and 3-D diffusion of the DONJON code. DRAGON code is then used to generate the group macroscopic cross sections needed for full core diffusion calculations. The diffusion DONJON code, is then used to compute the effective multiplication factor (keff), the feedback reactivity coefficients and neutron flux which account for variation in fuel and moderator temperatures as well as the void coefficient have been calculated using the DRAGON and DONJON codes for the MNSR research reactor. The cross sections of all the reactor components at different temperatures were generated using the DRAGON code. These group constants were used then in the DONJON code to calculate the multiplication factor and the neutron spectrum at different water and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. Finally, Good agreements between the calculated and measured have been obtained for every of the feedback reactivity coefficients and neutron flux.

Novel arrangement for an electro-Fenton reactor that does not require addition of iron, acid and a final neutralization stage. Towards the development of a cost-effective technology for the treatment of wastewater.

PubMed

Fernández, Dennys; Robles, Irma; Rodríguez-Valadez, Francisco J; Godínez, Luis A

2018-05-01

A novel arrangement for an electro-Fenton reactor aimed to treat neutral wastewater is presented. The arrangement consists on three-compartments in series, two of them packed with a cation exchange resin and one positioned between these, containing a polarized activated carbon column where the electrochemical generation of the Fenton reagent takes place. While the hydroxyl radicals electrochemically produced in-situ, react with the pollutant species adsorbed on the activated carbon cathode, the resin compartments administrate and collect the iron cation and the hydrated proton species in alternating flow direction cycles. The resulting process is a system that does not require acid or iron chemical addition to the process while at the same time, renders decontaminated water free of iron-dissolved species at neutral pH. The proposed electrochemical reactor arrangement is therefore the basis for the design of commercially viable electro-Fenton reactors in which the addition and subsequent removal of acid and iron chemicals is avoided; two of the currently most limiting features for the development of electro-Fenton technology for treating wastewater. Copyright © 2018 The Authors. Published by Elsevier Ltd.. All rights reserved.

Biocarriers Improve Bioaugmentation Efficiency of a Rapid Sand Filter for the Treatment of 2,6-Dichlorobenzamide-Contaminated Drinking Water.

PubMed

Horemans, Benjamin; Raes, Bart; Vandermaesen, Johanna; Simanjuntak, Yanti; Brocatus, Hannelore; T'Syen, Jeroen; Degryse, Julie; Boonen, Jos; Wittebol, Janneke; Lapanje, Ales; Sørensen, Sebastian R; Springael, Dirk

2017-02-07

Aminobacter sp. MSH1 immobilized in an alginate matrix in porous stones was tested in a pilot system as an alternative inoculation strategy to the use of free suspended cells for biological removal of micropollutant concentrations of 2,6-dichlorobenzamide (BAM) in drinking water treatment plants (DWTPs). BAM removal rates and MSH1 cell numbers were recorded during operation and assessed with specific BAM degradation rates obtained in lab conditions using either freshly grown cells or starved cells to explain reactor performance. Both reactors inoculated with either suspended or immobilized cells showed immediate BAM removal under the threshold of 0.1 μg/L, but the duration of sufficient BAM removal was 2-fold (44 days) longer for immobilized cells. The longer sufficient BAM removal in case of immobilized cells compared to suspended cells was mainly explained by a lower initial loss of MSH1 cells at operational start due to volume replacement and shear. Overall loss of activity in the reactors though was due to starvation, and final removal rates did not differ between reactors inoculated with immobilized and suspended cells. Management of assimilable organic carbon, in addition to cell immobilization, appears crucial for guaranteeing long-term BAM degradation activity of MSH1 in DWTP units.

Summary of ORSphere critical and reactor physics measurements

NASA Astrophysics Data System (ADS)

Marshall, Margaret A.; Bess, John D.

2017-09-01

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

Radiogenic lead as coolant, reflector and moderator in advanced fast reactors

NASA Astrophysics Data System (ADS)

Kulikov, E. G.

2017-01-01

Main purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors. When performing the study, thermal, physical and neutron-physical properties of natural and radiogenic lead were analyzed. The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and accelerator-driven systems (ADS) makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket, which enables effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.

FROM CONCEPT TO REALITY, IN-SITU DECOMMISSIONING OF THE P AND R REACTORS AT THE SAVANNAH RIVER SITE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Musall, J.; Blankenship, J.; Griffin, W.

2012-01-09

SRS recently completed an approximately three year effort to decommission two SRS reactors: P-Reactor (Building 105-P) and R-Reactor (Building 105-R). Completed in December 2011, the concurrent decommissionings marked the completion of two relatively complex and difficult facility disposition projects at the SRS. Buildings 105-P and 105-R began operating as production reactors in the early 1950s with the mission of producing weapons material (e.g., tritium and plutonium-239). The 'P' Reactor and was shutdown in 1991 while the 'R' Reactor and was shutdown in 1964. In the intervening period between shutdown and deactivation & decommissioning (D&D), Buildings 105-P and 105-R saw limitedmore » use (e.g., storage of excess heavy water and depleted uranium oxide). For Building 105-P, deactivation was initiated in April 2007 and was essentially complete by June 2010. For Building 105-R, deactivation was initiated in August 2008 and was essentially complete by September 2010. For both buildings, the primary objective of deactivation was to remove/mitigate hazards associated with the remaining hazardous materials, and thus prepare the buildings for in-situ decommissioning. Deactivation removed the following hazardous materials to the extent practical: combustibles/flammables, residual heavy water, acids, friable asbestos (as needed to protect workers performing deactivation and decommissioning), miscellaneous chemicals, lead/brass components, Freon(reg sign), oils, mercury/PCB containing components, mold and some radiologically-contaminated equipment. In addition to the removal of hazardous materials, deactivation included the removal of hazardous energy, exterior metallic components (representing an immediate fall hazard), and historical artifacts along with the evaporation of water from the two Disassembly Basins. Finally, so as to facilitate occupancy during the subsequent in-situ decommissioning, deactivation implemented repairs to the buildings and provided temporary power.« less

U.S.-Russian Cooperation in Science and Technology: A Case Study of the TOPAZ Space-Based Nuclear Reactor International Program

NASA Astrophysics Data System (ADS)

Dabrowski, Richard S.

2014-08-01

The TOPAZ International Program (TIP) was the final name given to a series of projects to purchase and test the TOPAZ-II, a space-based nuclear reactor of a type that had been further developed in the Soviet Union than in the United States. In the changing political situation associated with the break-up of the Soviet Union it became possible for the United States to not just purchase the system, but also to employ Russian scientists, engineers and testing facilities to verify its reliability. The lessons learned from the TIP illuminate some of the institutional and cultural challenges to U.S. - Russian cooperation in technology research which remain true today.

Status and improvement of CLAM for nuclear application

NASA Astrophysics Data System (ADS)

Huang, Qunying

2017-08-01

A program for China low activation martensitic steel (CLAM) development has been underway since 2001 to satisfy the material requirements of the test blanket module (TBM) for ITER, China fusion engineering test reactor and China fusion demonstration reactor. It has been undertaken by the Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences under wide domestic and international collaborations. Extensive work and efforts are being devoted to the R&D of CLAM, such as mechanical property evaluation before and after neutron irradiation, fabrication of scaled TBM by welding and additive manufacturing, improvement of its irradiation resistance as well as high temperature properties by precipitate strengthening to achieve its final successful application in fusion systems. The status and improvement of CLAM are introduced in this paper.

Theoretical and experimental study of the dark signal in CMOS image sensors affected by neutron radiation from a nuclear reactor

NASA Astrophysics Data System (ADS)

Xue, Yuanyuan; Wang, Zujun; He, Baoping; Yao, Zhibin; Liu, Minbo; Ma, Wuying; Sheng, Jiangkun; Dong, Guantao; Jin, Junshan

2017-12-01

The CMOS image sensors (CISs) are irradiated with neutron from a nuclear reactor. The dark signal in CISs affected by neutron radiation is studied theoretically and experimentally. The Primary knock-on atoms (PKA) energy spectra for 1 MeV incident neutrons are simulated by Geant4. And the theoretical models for the mean dark signal, dark signal non-uniformity (DSNU) and dark signal distribution versus neutron fluence are established. The results are found to be in good agreement with the experimental outputs. Finally, the dark signal in the CISs under the different neutron fluence conditions is estimated. This study provides the theoretical and experimental evidence for the displacement damage effects on the dark signal CISs.

Experimental Neutrino Physics: Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lane, Charles E.; Maricic, Jelena

2012-09-05

Experimental studies of neutrino properties, with particular emphasis on neutrino oscillation, mass and mixing parameters. This research was pursued by means of underground detectors for reactor anti-neutrinos, measuring the flux and energy spectra of the neutrinos. More recent investigations have been aimed and developing detector technologies for a long-baseline neutrino experiment (LBNE) using a neutrino beam from Fermilab.

Study, Talk, and Action. A Report of a National Conference on Regionalism and Regionalization in American Postsecondary Education.

ERIC Educational Resources Information Center

Martorana, S. V., Ed.; Nespoli, Lawrence A., Ed.

This report of a National Conference on Regionalism and Regionalization in American Postsecondary Education contains an overview and summary of the final project report, a keynote address, four papers on the implications of regionalism, some reactor comments, an essay on leadership, and four descriptive accounts of operational regionalization…

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2010 CFR

2010-01-01

... a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION RULES OF... Work Authorizations Early Partial Decisions on Site Suitability-Combined License Under 10 Cfr Part 52... complete and acceptable for docketing under § 2.101(a)(3), the Director of the Office of New Reactors or...

75 FR 5356 - Office of New Reactors; Final Interim Staff Guidance on Post-Combined License Commitments

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-02

... the guidance provided to the NRC staff in Section 1.0, ``Introduction and Interfaces,'' of NUREG-0800... (COL) applications. In addition, this ISG supplements the guidance provided in Section C.III.4 of... Accession No. ML091671355) to solicit public and industry comment. The NRC staff received comments (ADAMS...

BOILING NUCLEAR SUPERHEATER (BONUS) POWER STATION. Final Summary Design Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1962-05-01

The design and construction of the Boiling Nuclear Superheater (BONUS) Power Station at Punta Higuera on the seacoast at the westernmost tip of Puerto Rico are described. The reactor has an output of 17.5 Mw(e). This report will serve as a source of information for personnel engaged in management, evaluation, and training. (N.W.R.)

Revised FINAL–REPORT NO. 2: INDEPENDENT CONFIRMATORY SURVEY SUMMARY AND RESULTS FOR THE ENRICO FERMI ATOMIC POWER PLANT, UNIT 1, NEWPORT, MICHIGAN (DOCKET NO. 50 16; RFTA 10-004) 2018-SR-02-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Erika Bailey

2011-10-27

The Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor design that was cooled by sodium and operated at essentially atmospheric pressure. On May 10, 1963, the Atomic Energy Commission (AEC) granted an operating license, DPR-9, to the Power Reactor Development Company (PRDC), a consortium specifically formed to own and operate a nuclear reactor at the Fermi 1 site. The reactor was designed for a maximum capability of 430 megawatts (MW); however, the maximum reactor power with the first core loading (Core A) was 200 MW. The primary system was filled with sodium in Decembermore » 1960 and criticality was achieved in August 1963. The reactor was tested at low power during the first couple years of operation. Power ascension testing above 1 MW commenced in December 1965 immediately following the receipt of a high-power operating license. In October 1966 during power ascension, zirconium plates at the bottom of the reactor vessel became loose and blocked sodium coolant flow to some fuel subassemblies. Two subassemblies started to melt and the reactor was manually shut down. No abnormal releases to the environment occurred. Forty-two months later after the cause had been determined, cleanup completed, and the fuel replaced, Fermi 1 was restarted. However, in November 1972, PRDC made the decision to decommission Fermi 1 as the core was approaching its burn-up limit. The fuel and blanket subassemblies were shipped off-site in 1973. Following that, the secondary sodium system was drained and sent off-site. The radioactive primary sodium was stored on-site in storage tanks and 55 gallon (gal) drums until it was shipped off-site in 1984. The initial decommissioning of Fermi 1 was completed in 1975. Effective January 23, 1976, DPR-9 was transferred to the Detroit Edison Company (DTE) as a 'possession only' license (DTE 2010a). This report details the confirmatory activities performed during the second Oak Ridge Institute for Science and Education (ORISE) site visit to Fermi 1 in November 2010. The survey was strategically planned during a Unit 2 (Fermi 2) outage to take advantage of decreased radiation levels that were observed and attributed to Fermi 2 from the operating unit during the first site visit. However, during the second visit there were elevated radiation levels observed and attributed to the partially dismantled Fermi 1 reactor vessel and a waste storage box located on the 3rd floor of the Fermi 1 Turbine Building. Confirmatory surveys (unshielded) performed directly in the line of sight of these areas were affected. The objective of the confirmatory survey was to verify that the final radiological conditions were accurately and adequately described in Final Status Survey (FSS) documentation, relative to the established release criteria. This objective was achieved by performing document reviews, as well as independent measurements and sampling. Specifically, documentation of the planning, implementation, and results of the FSS were evaluated; side-by-side FSS measurement and source comparisons were performed; site areas were evaluated relative to appropriate FSS classification; and areas were assessed for residual, undocumented contamination.« less

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Davenport, Michael; Petti, D. A.; Palmer, Joe

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experimentsmore » are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.« less

Composting in small laboratory pilots: Performance and reproducibility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lashermes, G.; Barriuso, E.; Le Villio-Poitrenaud, M.

2012-02-15

Highlights: Black-Right-Pointing-Pointer We design an innovative small-scale composting device including six 4-l reactors. Black-Right-Pointing-Pointer We investigate the performance and reproducibility of composting on a small scale. Black-Right-Pointing-Pointer Thermophilic conditions are established by self-heating in all replicates. Black-Right-Pointing-Pointer Biochemical transformations, organic matter losses and stabilisation are realistic. Black-Right-Pointing-Pointer The organic matter evolution exhibits good reproducibility for all six replicates. - Abstract: Small-scale reactors (<10 l) have been employed in composting research, but few attempts have assessed the performance of composting considering the transformations of organic matter. Moreover, composting at small scales is often performed by imposing a fixed temperature, thus creatingmore » artificial conditions, and the reproducibility of composting has rarely been reported. The objectives of this study are to design an innovative small-scale composting device safeguarding self-heating to drive the composting process and to assess the performance and reproducibility of composting in small-scale pilots. The experimental setup included six 4-l reactors used for composting a mixture of sewage sludge and green wastes. The performance of the process was assessed by monitoring the temperature, O{sub 2} consumption and CO{sub 2} emissions, and characterising the biochemical evolution of organic matter. A good reproducibility was found for the six replicates with coefficients of variation for all parameters generally lower than 19%. An intense self-heating ensured the existence of a spontaneous thermophilic phase in all reactors. The average loss of total organic matter (TOM) was 46% of the initial content. Compared to the initial mixture, the hot water soluble fraction decreased by 62%, the hemicellulose-like fraction by 68%, the cellulose-like fraction by 50% and the lignin-like fractions by 12% in the final compost. The TOM losses, compost stabilisation and evolution of the biochemical fractions were similar to observed in large reactors or on-site experiments, excluding the lignin degradation, which was less important than in full-scale systems. The reproducibility of the process and the quality of the final compost make it possible to propose the use of this experimental device for research requiring a mass reduction of the initial composted waste mixtures.« less

Advanced Reactor Technologies - Regulatory Technology Development Plan (RTDP)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moe, Wayne L.

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory importance of key DOE reactor research initiatives should be assessed early in the technology development process. Quality assurance requirements supportive of later licensing activities must also be attached to important research activities to ensure resulting data is usable in that context. Early regulatory analysis and licensing approach planning thus provides a significant benefit to the formulation of research plans and also enables the planning and development of a compatible AdvSMR licensing framework, should significant modification be required.« less

Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moe, Wayne Leland

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory importance of key DOE reactor research initiatives should be assessed early in the technology development process. Quality assurance requirements supportive of later licensing activities must also be attached to important research activities to ensure resulting data is usable in that context. Early regulatory analysis and licensing approach planning thus provides a significant benefit to the formulation of research plans and also enables the planning and development of a compatible AdvSMR licensing framework, should significant modification be required.« less

Rapid solar-thermal decarbonization of methane

NASA Astrophysics Data System (ADS)

Dahl, Jaimee Kristen

Due to the ever-increasing demand for energy and the concern over the environmental impact of continuing to produce energy using current methods, there is interest in developing a hydrogen economy. Hydrogen is a desirable energy source because it is abundant in nature and burns cleanly. One method for producing hydrogen is to utilize a renewable energy source to obtain high enough temperatures to decompose a fossil fuel into its elements. This thesis work is directed at developing a solar-thermal aerosol flow reactor to dissociate methane to carbon black and hydrogen. The technology is intended as a "bridge" between current hydrogen production methods, such as conventional steam-methane reformers, and future "zero emission" technology for producing hydrogen, such as dissociating water using a renewable heating source. A solar furnace is used to heat a reactor to temperatures in excess of 2000 K. The final reactor design studied consists of three concentric vertical tubes---an outer quartz protection tube, a middle solid graphite heating tube, and an inner porous graphite reaction tube. A "fluid-wall" is created on the inside wall of the porous reaction tube in order to prevent deposition of the carbon black co-product on the reactor tube wall. The amorphous carbon black produced aids in heating the gas stream by absorbing radiation from the reactor wall. Conversions of 90% are obtained at a reactor wall temperature of 2100 K and an average residence time of 0.01 s. Computer modeling is also performed to study the gas flow and temperature profiles in the reactor as well as the kinetics of the methane dissociation reaction. The simulations indicate that there is little flow of the fluid-wall gas through the porous wall in the hot zone region, but this can be remedied by increasing the inlet temperature of the fluid-wall gas and/or increasing the tube permeability only in the hot zone region of the wall. The following expression describes the kinetics of methane dissociation in a solar-thermal fluid-wall reactor: dXdt=5.8x108 exp-155,600RT 1-X 7.2s-1. The experimental and theoretical work reported in this thesis is the groundwork that will be utilized in scaling up the reactor to produce hydrogen in distributed or centralized facilities.

Removal of highly elevated nitrate from drinking water by pH-heterogenized heterotrophic denitrification facilitated with ferrous sulfide-based autotrophic denitrification.

PubMed

Huang, Bin; Chi, Guangyu; Chen, Xin; Shi, Yi

2011-11-01

The performance of acetic acid-supported pH-heterogenized heterotrophic denitrification (HD) facilitated with ferrous sulfide-based autotrophic denitrification (AD) was investigated in upflow activated carbon-packed column reactors for reliable removal of highly elevated nitrate (42 mg NO(3)-Nl(-1)) in drinking water. The use of acetic acid as substrate provided sufficient internal carbon dioxide to completely eliminate the need of external pH adjustment for HD, but simultaneously created vertically heterogenized pH varying from 4.8 to 7.8 in the HD reactor. After 5-week acclimation, the HD reactor developed a moderate nitrate removal capacity with about one third of nitrate removal occurring in the acidic zone (pH 4.8-6.2). To increase the treatment reliability, acetic acid-supported HD was operated under 10% carbon limitation to remove >85% of nitrate, and ferrous sulfide-based AD was supplementally operated to remove residual nitrate and formed nitrite without excess of soluble organic carbon, nitrite or sulfate in the final effluent. Copyright © 2011 Elsevier Ltd. All rights reserved.

Development of tritium permeation barriers on Al base in Europe

NASA Astrophysics Data System (ADS)

Benamati, G.; Chabrol, C.; Perujo, A.; Rigal, E.; Glasbrenner, H.

The development of the water cooled lithium lead (WCLL) DEMO fusion reactor requires the production of a material capable of acting as a tritium permeation barrier (TPB). In the DEMO blanket reactor permeation barriers on the structural material are required to reduce the tritium permeation from the Pb-17Li or the plasma into the cooling water to acceptable levels (<1 g/d). Because of experimental work previously performed, one of the most promising TPB candidates is A1 base coatings. Within the EU a large R&D programme is in progress to develop a TPB fabrication technique, compatible with the structural materials requirements and capable of producing coatings with acceptable performances. The research is focused on chemical vapour deposition (CVD), hot dipping, hot isostatic pressing (HIP) technology and spray (this one developed also for repair) deposition techniques. The final goal is to select a reference technique to be used in the blanket of the DEMO reactor and in the ITER test module fabrication. The activities performed in four European laboratories are summarised here.

A New Method to Measure Crack Extension in Nuclear Graphite Based on Digital Image Correlation

DOE PAGES

Lai, Shigang; Shi, Li; Fok, Alex; ...

2017-01-01

Graphite components, used as moderators, reflectors, and core-support structures in a High-Temperature Gas-Cooled Reactor, play an important role in the safety of the reactor. Specifically, they provide channels for the fuel elements, control rods, and coolant flow. Fracture is the main failure mode for graphite, and breaching of the above channels by crack extension will seriously threaten the safety of a reactor. In this paper, a new method based on digital image correlation (DIC) is introduced for measuring crack extension in brittle materials. Cross-correlation of the displacements measured by DIC with a step function was employed to identify the advancingmore » crack tip in a graphite beam specimen under three-point bending. The load-crack extension curve, which is required for analyzing the R-curve and tension softening behaviors, was obtained for this material. Furthermore, a sensitivity analysis of the threshold value employed for the cross-correlation parameter in the crack identification process was conducted. Finally, the results were verified using the finite element method.« less

A New Method to Measure Crack Extension in Nuclear Graphite Based on Digital Image Correlation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lai, Shigang; Shi, Li; Fok, Alex

Graphite components, used as moderators, reflectors, and core-support structures in a High-Temperature Gas-Cooled Reactor, play an important role in the safety of the reactor. Specifically, they provide channels for the fuel elements, control rods, and coolant flow. Fracture is the main failure mode for graphite, and breaching of the above channels by crack extension will seriously threaten the safety of a reactor. In this paper, a new method based on digital image correlation (DIC) is introduced for measuring crack extension in brittle materials. Cross-correlation of the displacements measured by DIC with a step function was employed to identify the advancingmore » crack tip in a graphite beam specimen under three-point bending. The load-crack extension curve, which is required for analyzing the R-curve and tension softening behaviors, was obtained for this material. Furthermore, a sensitivity analysis of the threshold value employed for the cross-correlation parameter in the crack identification process was conducted. Finally, the results were verified using the finite element method.« less

Continuous immobilized yeast reactor system for complete beer fermentation using spent grains and corncobs as carrier materials.

PubMed

Brányik, Tomás; Silva, Daniel P; Vicente, António A; Lehnert, Radek; e Silva, João B Almeida; Dostálek, Pavel; Teixeira, José A

2006-12-01

Despite extensive research carried out in the last few decades, continuous beer fermentation has not yet managed to outperform the traditional batch technology. An industrial breakthrough in favour of continuous brewing using immobilized yeast could be expected only on achievement of the following process characteristics: simple design, low investment costs, flexible operation, effective process control and good product quality. The application of cheap carrier materials of by-product origin could significantly lower the investment costs of continuous fermentation systems. This work deals with a complete continuous beer fermentation system consisting of a main fermentation reactor (gas-lift) and a maturation reactor (packed-bed) containing yeast immobilized on spent grains and corncobs, respectively. The suitability of cheap carrier materials for long-term continuous brewing was proved. It was found that by fine tuning of process parameters (residence time, aeration) it was possible to adjust the flavour profile of the final product. Consumers considered the continuously fermented beer to be of a regular quality. Analytical and sensorial profiles of both continuously and batch fermented beers were compared.

Dynamic control of nutrient-removal from industrial wastewater in a sequencing batch reactor, using common and low-cost online sensors.

PubMed

Dries, Jan

2016-01-01

On-line control of the biological treatment process is an innovative tool to cope with variable concentrations of chemical oxygen demand and nutrients in industrial wastewater. In the present study we implemented a simple dynamic control strategy for nutrient-removal in a sequencing batch reactor (SBR) treating variable tank truck cleaning wastewater. The control system was based on derived signals from two low-cost and robust sensors that are very common in activated sludge plants, i.e. oxidation reduction potential (ORP) and dissolved oxygen. The amount of wastewater fed during anoxic filling phases, and the number of filling phases in the SBR cycle, were determined by the appearance of the 'nitrate knee' in the profile of the ORP. The phase length of the subsequent aerobic phases was controlled by the oxygen uptake rate measured online in the reactor. As a result, the sludge loading rate (F/M ratio), the volume exchange rate and the SBR cycle length adapted dynamically to the activity of the activated sludge and the actual characteristics of the wastewater, without affecting the final effluent quality.

In-situ biogas upgrading in thermophilic granular UASB reactor: key factors affecting the hydrogen mass transfer rate.

PubMed

Bassani, Ilaria; Kougias, Panagiotis G; Angelidaki, Irini

2016-12-01

Biological biogas upgrading coupling CO 2 with external H 2 to form biomethane opens new avenues for sustainable biofuel production. For developing this technology, efficient H 2 to liquid transfer is fundamental. This study proposes an innovative setup for in-situ biogas upgrading converting the CO 2 in the biogas into CH 4 , via hydrogenotrophic methanogenesis. The setup consisted of a granular reactor connected to a separate chamber, where H 2 was injected. Different packing materials (rashig rings and alumina ceramic sponge) were tested to increase gas-liquid mass transfer. This aspect was optimized by liquid and gas recirculation and chamber configuration. It was shown that by distributing H 2 through a metallic diffuser followed by ceramic sponge in a separate chamber, having a volume of 25% of the reactor, and by applying a mild gas recirculation, CO 2 content in the biogas dropped from 42 to 10% and the final biogas was upgraded from 58 to 82% CH 4 content. Copyright © 2016 Elsevier Ltd. All rights reserved.

Copper recovery and gold enrichment from waste printed circuit boards by mediated electrochemical oxidation.

PubMed

Fogarasi, Szabolcs; Imre-Lucaci, Florica; Imre-Lucaci, Arpád; Ilea, Petru

2014-05-30

The present study aims to develop an eco-friendly chemical-electrochemical process for the simultaneous recovery of copper and separation of a gold rich residue from waste printed circuit boards (WPCBs). The process was carried out by employing two different types of reactors coupled in series: a leaching reactor with a perforated rotating drum, for the dissolution of base metals and a divided electrochemical reactor for the regeneration of the leaching solution with the parallel electrowinning of copper. The process performances were evaluated on the basis of the dissolution efficiency, current efficiency and specific energy consumptions. Finally a process scale up was realized taking into consideration the optimal values of the operating parameters. The laboratory scale leaching plant allowed the recovery of a high purity copper deposit (99.04wt.%) at a current efficiency of 63.84% and specific energy consumption of 1.75kWh/kg cooper. The gold concentration in the remained solid residue was 25 times higher than the gold concentration in the initial WPCB samples. Copyright © 2014 Elsevier B.V. All rights reserved.

Qualification of heavy water based irradiation device in the JSI TRIGA reactor for irradiations of FT-TIMS samples for nuclear safeguards

NASA Astrophysics Data System (ADS)

Radulović, Vladimir; Kolšek, Aljaž; Fauré, Anne-Laure; Pottin, Anne-Claire; Pointurier, Fabien; Snoj, Luka

2018-03-01

The Fission Track Thermal Ionization Mass Spectrometry (FT-TIMS) method is considered as the reference method for particle analysis in the field of nuclear Safeguards for measurements of isotopic compositions (fissile material enrichment levels) in micrometer-sized uranium particles collected in nuclear facilities. An integral phase in the method is the irradiation of samples in a very well thermalized neutron spectrum. A bilateral collaboration project was carried out between the Jožef Stefan Institute (JSI, Slovenia) and the Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA, France) to determine whether the JSI TRIGA reactor could be used for irradiations of samples for the FT-TIMS method. This paper describes Monte Carlo simulations, experimental activation measurements and test irradiations performed in the JSI TRIGA reactor, firstly to determine the feasibility, and secondly to design and qualify a purpose-built heavy water based irradiation device for FT-TIMS samples. The final device design has been shown experimentally to meet all the required performance specifications.

Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells

DOE Office of Scientific and Technical Information (OSTI.GOV)

E.T. Robinson; John Sirman; Prasad Apte

2005-05-01

This final report summarizes work accomplished in the Program from January 1, 2001 through December 31, 2004. Most of the key technical objectives for this program were achieved. A breakthrough material system has lead to the development of an OTM (oxygen transport membrane) compact planar reactor design capable of producing either syngas or hydrogen. The planar reactor shows significant advantages in thermal efficiency and a step change reduction in costs compared to either autothermal reforming or steam methane reforming with CO{sub 2} recovery. Syngas derived ultra-clean transportation fuels were tested in the Nuvera fuel cell modular pressurized reactor and inmore » International Truck and Engine single cylinder test engines. The studies compared emission and engine performance of conventional base fuels to various formulations of ultra-clean gasoline or diesel fuels. A proprietary BP oxygenate showed significant advantage in both applications for reducing emissions with minimal impact on performance. In addition, a study to evaluate new fuel formulations for an HCCI engine was completed.« less

SLSF in-reactor local fault safety experiment P4. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thompson, D. H.; Holland, J. W.; Braid, T. H.

The Sodium Loop Safety Facility (SLSF), a major facility in the US fast-reactor safety program, has been used to simulate a variety of sodium-cooled fast reactor accidents. SLSF experiment P4 was conducted to investigate the behavior of a "worse-than-case" local fault configuration. Objectives of this experiment were to eject molten fuel into a 37-pin bundle of full-length Fast-Test-Reactor-type fuel pins form heat-generating fuel canisters, to characterize the severity of any molten fuel-coolant interaction, and to demonstrate that any resulting blockage could either be tolerated during continued power operation or detected by global monitors to prevent fuel failure propagation. The designmore » goal for molten fuel release was 10 to 30 g. Explusion of molten fuel from fuel canisters caused failure of adjacent pins and a partial flow channel blockage in the fuel bundle during full-power operation. Molten fuel and fuel debris also lodged against the inner surface of the test subassembly hex-can wall. The total fuel disruption of 310 g evaluated from posttest examination data was in excellent agreement with results from the SLSF delayed neutron detection system, but exceeded the target molten fuel release by an order of magnitude. This report contains a summary description of the SLSF in-reactor loop and support systems and the experiment operations. results of the detailed macro- and microexamination of disrupted fuel and metal and results from the analysis of the on-line experimental data are described, as are the interpretations and conclusions drawn from the posttest evaluations. 60 refs., 74 figs.« less

Manufacturing demonstration of microbially mediated zinc sulfide nanoparticles in pilot-plant scale reactors

DOE PAGES

Moon, Ji-Won; Phelps, Tommy J.; Fitzgerald Jr, Curtis L.; ...

2016-04-27

The thermophilic anaerobic metal-reducing bacterium Thermoanaerobacter sp. X513 efficiently produces zinc sulfide (ZnS) nanoparticles (NPs) in laboratory-scale ( ≤24-L) reactors. To determine whether this process can be up-scaled and adapted for pilot-plant production while maintaining NP yield and quality, a series of meso-scale experiments were performed using 100-l and 900-l reactors. Pasteurization and N 2-sparging replaced autoclaving and boiling for deoxygenating media in the transition from small-scale to pilot-plant reactors. Consecutive 100-L batches using new or recycled media produced ZnS NPs with highly reproducible ~2 nm average crystallite size (ACS) and yields of ~0.5g L -1, similar to small-scale batches.more » The 900-L pilot plant reactor produced ~ 320 g ZnS without process optimization or replacement of used medium; this quantity would be sufficient to form a ZnS thin film with ~120 nm thickness over 0.5 m width 13 km length. At all scales, the bacteria produced significant amounts of acetic, lactic and formic acids, which could be neutralized by the controlled addition of sodium hydroxide without the use of an organic pH buffer, eliminating 98% of the buffer chemical costs. In conclusion, the final NP products were characterized using XRD, ICP-OES, FTIR, DLS, and C/N analyses, which confirmed the growth medium without organic buffer enhanced the ZnS NP properties by reducing carbon and nitrogen surface coatings and supporting better dispersivity with similar ACS.« less

Anaerobic reactor/high rate pond combined technology for sewage treatment in the Mediterranean area.

PubMed

El Hafiane, F; El Hamouri, B

2005-01-01

Two high-rate, anaerobic/aerobic units were used to treat the sewage of the Institut Agronomique st Vétérinaire Hassan II (Morocco) campus in a 1,100 m2-plant designed for 1,500 e.p. and receiving 63 m3 per day. The anaerobic pre-treatment consisted of a two-step up-flow anaerobic reactor (TSUAR) comprising two reactors and one external settler all in series. The aerobic line, or post-treatment, consisted of a high-rate algal pond (HRAP) and one maturation pond in series. The system totalized a hydraulic retention time (HRT) of 9 days. A gravel filter (GF) was constructed behind the TSUAR to trap low-density particles. The TSUAR removed 80% of COD and 90% of SS within 48 h. Solids retention time in the reactors averaged 32 d with a specific sludge production of 0.28 g SS g(-1) COD removed. Almost 93% of the sludge evacuated from the settler was stabilized. Specific biogas production from both reactors was 0.25m3 kg(-1) COD removed. Used in this configuration, the HRAP lost its BOD removal activity and increased its nutrients and pathogens removal capabilities (tertiary treatment). Results showed that 85% of total nitrogen and 48% of total phosphorus were removed by the HRAP. Land area requirement of this combination was less than 1 m2 per capita and filtered final effluent was of excellent quality (COD, 82 mg/l; TKN, 8.3 mg/l; total P, 2.7 mg/l, faecal coliforms, 2.4 10(3)/100 ml and zero helminths eggs).

Technicians Manufacture a Nozzle for the Kiwi B-1-B Engine

NASA Image and Video Library

1964-05-21

Technicians manufacture a nozzle for the Kiwi B-1-B nuclear rocket engine in the Fabrication Shop’s vacuum oven at the National Aeronautics and Space Administration (NASA) Lewis Research Center. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test basic nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The final phase of the program, called Reactor-In-Flight-Test, would be an actual launch test. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The turbopump, which pumped the fuels from the storage tanks to the engine, was the primary tool for restarting the engine. The NERVA had to be able to restart in space on its own using a safe preprogrammed startup system. Lewis researchers endeavored to design and test this system. This non-nuclear Kiwi engine, seen here, was being prepared for tests at Lewis’ High Energy Rocket Engine Research Facility (B-1) located at Plum Brook Station. The tests were designed to start an unfueled Kiwi B-1-B reactor and its Aerojet Mark IX turbopump without any external power.

Manufacturing demonstration of microbially mediated zinc sulfide nanoparticles in pilot-plant scale reactors.

PubMed

Moon, Ji-Won; Phelps, Tommy J; Fitzgerald, Curtis L; Lind, Randall F; Elkins, James G; Jang, Gyoung Gug; Joshi, Pooran C; Kidder, Michelle; Armstrong, Beth L; Watkins, Thomas R; Ivanov, Ilia N; Graham, David E

2016-09-01

The thermophilic anaerobic metal-reducing bacterium Thermoanaerobacter sp. X513 efficiently produces zinc sulfide (ZnS) nanoparticles (NPs) in laboratory-scale (≤ 24-L) reactors. To determine whether this process can be up-scaled and adapted for pilot-plant production while maintaining NP yield and quality, a series of pilot-plant scale experiments were performed using 100-L and 900-L reactors. Pasteurization and N2-sparging replaced autoclaving and boiling for deoxygenating media in the transition from small-scale to pilot plant reactors. Consecutive 100-L batches using new or recycled media produced ZnS NPs with highly reproducible ~2-nm average crystallite size (ACS) and yields of ~0.5 g L(-1), similar to the small-scale batches. The 900-L pilot plant reactor produced ~320 g ZnS without process optimization or replacement of used medium; this quantity would be sufficient to form a ZnS thin film with ~120 nm thickness over 0.5 m width × 13 km length. At all scales, the bacteria produced significant amounts of acetic, lactic, and formic acids, which could be neutralized by the controlled addition of sodium hydroxide without the use of an organic pH buffer, eliminating 98 % of the buffer chemical costs. The final NP products were characterized using XRD, ICP-OES, TEM, FTIR, PL, DLS, HPLC, and C/N analyses, which confirmed that the growth medium without organic buffer enhanced the ZnS NP properties by reducing carbon and nitrogen surface coatings and supporting better dispersivity with similar ACS.

The Bosch Process-Performance of a Developmental Reactor and Experimental Evaluation of Alternative Catalysts

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, J. Matthew

2010-01-01

Bosch-based reactors have been in development at NASA since the 1960's. Traditional operation involves the reduction of carbon dioxide with hydrogen over a steel wool catalyst to produce water and solid carbon. While the system is capable of completely closing the loop on oxygen and hydrogen for Atmosphere Revitalization, steel wool requires a reaction temperature of 650C or higher for optimum performance. The single pass efficiency of the reaction over steel wool has been shown to be less than 10% resulting in a high recycle stream. Finally, the formation of solid carbon on steel wool ultimately fouls the catalyst necessitating catalyst resupply. These factors result in high mass, volume and power demands for a Bosch system. Interplanetary transportation and surface exploration missions of the moon, Mars, and near-earth objects will require higher levels of loop closure than current technology cannot provide. A Bosch system can provide the level of loop closure necessary for these long-term missions if mass, volume, and power can be kept low. The keys to improving the Bosch system lie in reactor and catalyst development. In 2009, the National Aeronautics and Space Administration refurbished a circa 1980's developmental Bosch reactor and built a sub-scale Bosch Catalyst Test Stand for the purpose of reactor and catalyst development. This paper describes the baseline performance of two commercially available steel wool catalysts as compared to performance reported in the 1960's and 80's. Additionally, the results of sub-scale testing of alternative Bosch catalysts, including nickel- and cobalt-based catalysts, are discussed.

76 FR 6638 - Virginia Electric and Power Company D/B/A/ Dominion Virginia Power and Old Dominion Electric...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-02-07

... Company D/B/A/ Dominion Virginia Power and Old Dominion Electric Cooperative, North Anna Power Station... combined license (COL) application to build and operate a new reactor at its North Anna Power Station (NAPS... Combined License (COL) for North Anna Power Station, Unit 3.'' A notice of availability of the final...

9 CFR 147.6 - Procedure for determining the status of flocks reacting to tests for Mycoplasma gallisepticum...

Code of Federal Regulations, 2010 CFR

2010-01-01

... Agency may examine reactors by the in vivo bio-assay, PCR-based procedures, and/or culture procedures before final determination of the flock status is made. (13) If the in vivo bio-assay, PCR-based... classification for which it was tested. (14) If the in vivo bio-assay, PCR-based procedures, or culture...

Scientific and Technical Manpower Requirements of Selected Segments of the Atomic Energy Field. Final Report.

ERIC Educational Resources Information Center

Voight, Keith L.

The primary purpose of the study was to develop a supply/demand ratio for nuclear degree scientists and engineers from July 1969 through 1973. The need by private industry and electric utilities for scientists and engineers with degrees in disciplines other than nuclear science or engineering, as well as for technicians, nuclear reactor operators,…

Browns Ferry Unit-3 cavity neutron spectral analysis. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martin, G.C.

1981-08-01

This report describes neutron dosimetry measurements performed in the Browns Ferry Unit-3 reactor cavity using multiple dosimeter and spectrum unfolding techniques to assess radiation-induced degradation of nuclear plant pressure vessels. Test results and conclusions indicating the feasibility of determining neutron flux spectra and the densities in the pressure vessel cavity region via dosimetric measurements are presented.

Progress on RERTR activities in Argentina

DOE Office of Scientific and Technical Information (OSTI.GOV)

Balart, S.; Calzetta, O.; Cristini, P.

2008-07-15

Since last RERTR meeting, several tasks involving RERTR activities continued deploying in Argentina: through an agreement between CNEA and US-DoE final steps in the RA-6 reactor core conversion from HEU to LEU are taking place; by means of a return campaign of 42 US origin SNF in the frame of the US-SNF FRR program; an effective minimization of HEU inventory is close to be accomplished; development of a LEU dispersed U-Mo fuel prototype, to be irradiated in a high flux reactor in the frame of the ARG/4/092 IAEA's Technical Cooperation project is progressing; very high density monolithic U-Mo miniplates andmore » plates using MEU and LEU fuel with Zry-4 cladding were developed to be irradiated as a part of the RERTR program irradiation experiment; atomistic modeling prediction (BFS techniques and first principles) enabled to find some trends on the interaction phases; diffusion couples tests under X-ray synchrotron analysis allowed the characterization of several phases involving U-Mo(-Zr) / Al(-Si); finally CNEA continued spreading high quality LEU technology for fission RI production by means of agreements with different producers interested on HEU-LEU conversion. (author)« less

RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly Cummins; Igor Bolshinsky; Ken Allen

2009-07-01

In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required tomore » complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.« less

Aging Management Guideline for commercial nuclear power plants: Motor control centers; Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toman, G.; Gazdzinski, R.; O`Hearn, E.

1994-02-01

This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specificmore » aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.« less

Development of a proton-exchange membrane electrochemical reclaimed water post-treatment system

NASA Technical Reports Server (NTRS)

Kaba, Lamine; Verostko, Charles E.; Hitchens, G. D.; Murphy, Oliver J.

1991-01-01

A single-cell electrochemical reactor that utilizes a proton exchange membrane (PEM) as a solid electrolyte is being investigated for posttreatment of reclaimed waste waters with low or negligible electrolyte content. Posttreatment is a final 'polishing' of reclaimed waste waters prior to reuse, and involves removing organic impurities at levels as high as 100 ppm to below 500 ppb total organic carbon (TOC) content to provide disinfection. The system does not utilize or produce either expendable hardware components or chemicals and has no moving parts. Test data and kinetic analysis are presented. The feasibility and application for water reclamation processes in controlled ecological environments (e.g., lunar/Mars habitats) are also presented. Test results show that the electrochemical single cell reactor provides effective posttreatment.

Grid-to-rod flow-induced impact study for PWR fuel in reactor

DOE PAGES

Jiang, Hao; Qu, Jun; Lu, Roger Y.; ...

2016-06-10

The source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR) is the dynamic contact impact from hydraulic flow-induced fuel assembly vibration. In order to support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. Finally, it is an essential and effectivemore » way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE PAGES

Durham, J. M.; Poulson, D.; Bacon, J.; ...

2018-04-10

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

Hydrothermal corrosion of silicon carbide joints without radiation

DOE PAGES

Koyanagi, Takaaki; Katoh, Yutai; Terrani, Kurt A.; ...

2016-09-28

In this paper, hydrothermal corrosion of four types of the silicon carbide (SiC) to SiC plate joints were investigated under pressurized water reactor and boiling water reactor relevant chemical conditions without irradiation. The joints were formed by metal diffusion bonding using molybdenum or titanium interlayer, reaction sintering using Ti—Si—C system, and SiC nanopowder sintering. Most of the joints withstood the corrosion tests for five weeks. The recession of the SiC substrates was limited. Based on the recession of the bonding layers, it was concluded that all the joints except for the molybdenum diffusion bond are promising under the reducing environmentsmore » without radiation. Finally, the SiC nanopowder sintered joint was the most corrosion tolerant under the oxidizing environment among the four joints.« less

Method for producing components with internal architectures, such as micro-channel reactors, via diffusion bonding sheets

DOEpatents

Alman, David E [Corvallis, OR; Wilson, Rick D [Corvallis, OR; Davis, Daniel L [Albany, OR

2011-03-08

This invention relates to a method for producing components with internal architectures, and more particularly, this invention relates to a method for producing structures with microchannels via the use of diffusion bonding of stacked laminates. Specifically, the method involves weakly bonding a stack of laminates forming internal voids and channels with a first generally low uniaxial pressure and first temperature such that bonding at least between the asperites of opposing laminates occurs and pores are isolated in interfacial contact areas, followed by a second generally higher isostatic pressure and second temperature for final bonding. The method thereby allows fabrication of micro-channel devices such as heat exchangers, recuperators, heat-pumps, chemical separators, chemical reactors, fuel processing units, and combustors without limitation on the fin aspect ratio.

Measuring Theta_13 at Daya Bay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lau, Kwong

2014-03-14

We measured the neutrino mixing angle, theta13, presumably related to the preponderance of matter over antimatter in our universe with high precision. We determined theta13 by measuring the disappearance of neutrinos from a group of six nuclear reactors. The target, located inside a mountain at about 2 km from the reactors, is 80 tons of liquid scintillator doped with trace amount of Gadolinium to increase its neutron detection efficiency. The neutrino flux is measured by the inverse beta-decay reaction where the final-state particles are detected by the liquid scintillator. The measured value of theta13, based on data collected over 3more » years, is large, around 8 degrees, rendering the measurement of the parameter related to matter-antimatter asymmetry in future long baseline neutrino experiments easier.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durham, J. M.; Poulson, D.; Bacon, J.

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

Catalytic biomass conversion methods, catalysts, and methods of making the same

DOEpatents

Delgass, William Nicholas; Agrawal, Rakesh; Ribeiro, Fabio Henrique; Saha, Basudeb; Yohe, Sara Lynn; Abu-Omar, Mahdi M; Parsell, Trenton; Dietrich, Paul James; Klein, Ian Michael

2017-10-10

Described herein are processes for one-step delignification and hydrodeoxygenation of lignin fraction a biomass feedstock. The lignin feedstock is derived from by-products of paper production and biorefineries. Additionally described is a process for converting biomass-derived oxygenates to lower oxygen-content compounds and/or hydrocarbons in the liquid or vapor phase in a reactor system containing hydrogen and a catalyst comprised of a hydrogenation function and/or an oxophilic function and/or an acid function. Finally, also described herein is a process for converting biomass-derived oxygenates to lower oxygen-content compounds and/or hydrocarbons in the liquid or vapor phase in a reactor system containing hydrogen and a catalyst comprised of a hydrogenation function and/or an oxophilic function and/or an acid function.

Screening study for evaluation of the potential for system 80+ to consume excess plutonium - Volume 1. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1994-04-30

As part of the U.S. effort to evaluate technologies offering solutions for the safe disposal or utilization of surplus nuclear materials, the fiscal year 1993 Energy and Water Appropriations legislation provided the Department of Energy (DOE) the necessary funds to conduct multi-phased studies to determine the technical feasibility of using reactor technologies for the triple mission of burning weapons grade plutonium, producing tritium for the existing smaller weapons stockpile, and generating commercial electricity. DOE limited the studies to five advanced reactor designs. Among the technologies selected is the ABB-Combustion Engineering (ABB-CE) System 80+. The DOE study, currently in Phase ID,more » is proceeding with a more detailed evaluation of the design`s capability for plutonium disposition.« less

OPERATIONAL CHARACTERISTICS OF THE ARMOUR FISSION GAS GAMMA FACILITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrell, C.W.; McElroy, W.N.

1958-10-31

As the reactor power level is changed frequently, the radiation levels in the gamma facility fluctuate. Data are presented to show the power dependency of the gamma dose rate and the manner of growth and decay. Additional data show the dependercy of the equilibrium gamma activity on the foel temperature and total system pressure. The final phase of the work is directed toward determining an average gamma energy by attenuation measurements with various thicknesses of several materials. The neutrou flux associated with the gas phase activity is determined by foil measurement. From the measurements of dose rate and average gammamore » energy, calculations to determine the number of curies of gas phase decay gamma activity per watt of reactor power are presented. (auth)« less

Fukushima Accident: Sequence of Events and Lessons Learned

NASA Astrophysics Data System (ADS)

Morse, Edward C.

2011-10-01

The Fukushima Dai-Ichi nuclear power station suffered a devastating Richter 9.0 earthquake followed by a 14.0 m tsunami on 11 March 2011. The subsequent loss of power for emergency core cooling systems resulted in damage to the fuel in the cores of three reactors. The relief of pressure from the containment in these three reactors led to sufficient hydrogen gas release to cause explosions in the buildings housing the reactors. There was probably subsequent damage to a spent fuel pool of a fourth reactor caused by debris from one of these explosions. Resultant releases of fission product isotopes in air were significant and have been estimated to be in the 3 . 7 --> 6 . 3 ×1017 Bq range (~10 MCi) for 131I and 137Cs combined, or approximately one tenth that of the Chernobyl accident. A synopsis of the sequence of events leading up to this large release of radioactivity will be presented, along with likely scenarios for stabilization and site cleanup in the future. Some aspects of the isotope monitoring programs, both locally and at large, will also be discussed. An assessment of radiological health risk for the plant workers as well as the general public will also be presented. Finally, the impact of this accident on design and deployment of nuclear generating stations in the future will be discussed.

DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel, G.; Rudisill, T.; Almond, P.

The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Sitemore » (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.« less

Impact of paper and cardboard suppression on OFMSW anaerobic digestion.

PubMed

Fonoll, X; Astals, S; Dosta, J; Mata-Alvarez, J

2016-10-01

Mechanical-biological treatment plants treat municipal solid waste to recover recyclable materials, nutrients and energy. Waste paper and cardboard (WP), the second main compound in municipal solid waste (∼30% in weight basis), is typically used for biogas generation. However, its recovery is gaining attention as it can be used to produce add-value products like bioethanol and residual derived fuel. Nevertheless, WP suppression or replacement will impact anaerobic digestion in terms of biogas production, process stability and digestate management. Two lab-scale reactors were used to assess the impact of WP in anaerobic digestion performance. A control reactor was only fed with biowaste (BioW), while a second reactor was fed with two different mixtures of BioW and WP, i.e. 85/15% and 70/30% (weight basis). Results indicate that either replacing half of the WP by BioW or removing half of the WP has little impact on the methane production. When removing half of the WP, methane production could be sustained by a larger waste biodegradability. The replacement of all WP by BioW increased the reactor methane production (∼37%), while removing all WP would have reduced the methane production about 15%. Finally, replacing WP loading rate by BioW led to a system less tolerant to instability periods and with poorer digestate quality. Copyright © 2016 Elsevier Ltd. All rights reserved.

Summary of ORSphere Critical and Reactor Physics Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, Margaret A.; Bess, John D.

In the early 1970s Dr. John T. Mihalczo (team leader), J. J. Lynn, and J. R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVAmore » I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is summary summarize all the critical and reactor physics measurements evaluations and, when possible, to compare them to GODIVA experiment results.« less

Pressurized diesel fuel processing using hydrogen peroxide for the fuel cell power unit in low-oxygen environments

NASA Astrophysics Data System (ADS)

Lee, Kwangho; Han, Gwangwoo; Cho, Sungbaek; Bae, Joongmyeon

2018-03-01

A novel concept for diesel fuel processing utilizing H2O2 is suggested to obtain the high-purity H2 required for air-independent propulsion using polymer electrolyte membrane fuel cells for use in submarines and unmanned underwater vehicles. The core components include 1) a diesel-H2O2 autothermal reforming (ATR) reactor to produce H2-rich gas, 2) a water-gas shift (WGS) reactor to convert CO to H2, and 3) a H2 separation membrane to separate only high-purity H2. Diesel and H2O2 can easily be pressurized as they are liquids. The application of the H2 separation membrane without a compressor in the middle of the process is thus advantageous. In this paper, the characteristics of pressurized ATR and WGS reactions are investigated according to the operating conditions. In both reactors, the methanation reaction is enhanced as the pressure increases. Then, permeation experiments with a H2 separation membrane are performed while varying the temperature, pressure difference, and inlet gas composition. In particular, approximately 90% of the H2 is recovered when the steam-separated rear gas of the WGS reactor is used in the H2 separation membrane. Finally, based on the experimental results, design points are suggested for maximizing the efficiency of the diesel-H2O2 fuel processor.

Artemisinin production by plant hairy root cultures in gas- and liquid-phase bioreactors.

PubMed

Patra, Nivedita; Srivastava, Ashok K

2016-01-01

Alternative biotechnological protocol for large-scale artemisinin production was established. It featured enhanced growth and artemisinin production by cultivation of hairy roots in nutrient mist bioreactor (NMB) coupled with novel cultivation strategies. Artemisinin is used for the treatment of cerebral malaria. Presently, its main source is from seasonal plant Artemisia annua. This study featured investigation of growth and artemisinin production by A. annua hairy roots (induced by Agrobacterium rhizogenes-mediated genetic transformation of explants) in three bioreactor configurations-bubble column reactor, NMB and modified NMB particularly to establish their suitability for commercial production. It was observed that cultivation of hairy roots in a non-stirred bubble column reactor exhibited a biomass accumulation of 5.68 g/l only while batch cultivation in a custom-made NMB exhibited a higher biomass concentration of 8.52 g/l but relatively lower artemisinin accumulation of 0.22 mg/g was observed in this reactor. A mixture of submerged liquid-phase growth (for 5 days) followed by gas-phase cultivation in nutrient mist reactor operation strategy (for next 15 days) was adopted for hairy root cultivation in this investigation. Reasonably, high (23.02 g/l) final dry weight along with the artemisinin accumulation (1.12 mg/g, equivalent to 25.78 mg/l artemisinin) was obtained in this bioreactor, which is the highest reported artemisinin yield in the gas-phase NMB cultivation.

Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

NASA Astrophysics Data System (ADS)

Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

2018-02-01

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

DOE PAGES

Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; ...

2016-08-15

Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy’s Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representativemore » of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Finally, insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kreitman, Paul J.; Sirianni, Steve R.; Pillard, Mark M.

Entergy recently performed an Extended Power Up-rate (EPU) on their Grand Gulf Nuclear Station, near Port Gibson, Mississippi. To support the EPU, a new Steam Dryer Assembly was installed during the last refueling outage. Due to limited access into the containment, the large Replacement Steam Dryer (RSD) had to be brought into the containment in pieces and then final assembly was completed on the refueling floor before installation into the reactor. Likewise, the highly contaminated Original Steam Dryer (OSD) had to be segmented into manageable sections, loaded into specially designed shielded containers, and rigged out of containment where they willmore » be safely stored until final disposal is accomplished at an acceptable waste repository. Westinghouse Nuclear Services was contracted by Entergy to segment, package and remove the OSD from containment. This work was performed on critical path during the most recent refueling outage. The segmentation was performed underwater to minimize radiation exposure to the workers. Special hydraulic saws were developed for the cutting operations based on Westinghouse designs previously used in Sweden to segment ABB Reactor Internals. The mechanical cutting method was selected because of its proven reliability and the minimal cutting debris that is generated by the process. Maintaining stability of the large OSD sections during cutting was accomplished using a custom built support stand that was installed into the Moisture Separator Pool after the Moisture Separator was installed back in the reactor vessel. The OSD was then moved from the Steam Dryer Pool to the Moisture Separator Pool for segmentation. This scenario resolved the logistical challenge of having two steam dryers and a moisture separator in containment simultaneously. A water filtration/vacuum unit was supplied to maintain water clarity during the cutting and handling operations and to collect the cutting chips. (authors)« less

High power ring methods and accelerator driven subcritical reactor application

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tahar, Malek Haj

2016-08-07

High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g.,more » PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field imperfections on the transverse beam dynamics. The results obtained allow to develop a correction scheme to minimize the tune variations of the FFAG. This is the cornerstone of a new fixed tune non-scaling FFAG that represents a potential candidate for high power applications. As part of the developments towards high power at the KURRI FFAG, beam dynamics studies have to account for space charge effects. In that framework, models have been installed in the tracking code ZGOUBI to account for the self-interaction of the particles in the accelerator. Application to the FFAG studies is shown. Finally, one focused on the ADSR concept as a candidate to solve the problem of nuclear waste. In order to establish the accelerator requirements, one compared the performance of ADSR with other conventional critical reactors by means of the levelized cost of energy. A general comparison between the different accelerator technologies that can satisfy these requirements is finally presented. In summary, the main drawback of the ADSR technology is the high Levelized Cost Of Energy compared to other advanced reactor concepts that do not employ an accelerator. Nowadays, this is a show-stopper for any industrial application aiming at producing energy (without dealing with the waste problem). Besides, the reactor is not intrinsically safer than critical reactor concepts, given the complexity of managing the target interface between the accelerator and the reactor core.« less

75 FR 13 - Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-04

...The Nuclear Regulatory Commission (NRC) is amending its regulations to provide alternate fracture toughness requirements for protection against pressurized thermal shock (PTS) events for pressurized water reactor (PWR) pressure vessels. This final rule provides alternate PTS requirements based on updated analysis methods. This action is desirable because the existing requirements are based on unnecessarily conservative probabilistic fracture mechanics analyses. This action reduces regulatory burden for those PWR licensees who expect to exceed the existing requirements before the expiration of their licenses, while maintaining adequate safety, and may choose to comply with the final rule as an alternative to complying with the existing requirements.

The Outlook for Some Fission Products Utilization with the Aim to Immobilize Long-Lived Radionuclides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pokhitonov, Y.A.

2008-07-01

The prospects for development of nuclear power are intimately associated with solving the problem of safe management and removal from the biosphere of generated radioactive wastes. The most suitable material for fission products and actinides immobilization is the crystalline ceramics. By now numerous literature data are available concerning the synthesis of a large range of various materials with zirconium-based products. It worth mentioning that zirconium is only one of fission products accumulated in the fuel in large amounts. The development of new materials intended for HLW immobilization will allow increasing of radionuclides concentration in solidified product so providing costs reductionmore » at the stage of subsequent storage. At the same time the idea to use for synthesis of compounds, suitable as materials for long-term storage or final disposal of rad-wastes some fission products occurring in spent fuel in considerable amount and capable to form insoluble substances seems to be rather attractive. In authors opinion in the nearest future one can expect the occurrence of publications proposing the techniques allowing the use of 'reactor's zirconium, molybdenum or, perhaps, technetium as well, with the aim of preparing materials suitable for long-lived radionuclides storage or final disposal. The other element, which is generated in the reactor and worth mentioning, is palladium. The prospects for using palladium are defined not only by its higher generation in the reactor, but by a number of its chemical properties as well. It is evident that the use of natural palladium with the purpose of radionuclides immobilization is impossible due to its high cost and deficiency). In author's opinion such materials could be used as targets for long-lived radionuclides transmutation as well. The object of present work was the study on methods that could allow to use 'reactor' palladium with the aim of long-lived radionuclides such as I-129 and TUE immobilization. In the paper the results of experiments on synthesis of matrices with TUE oxides and PdI{sub 2} on palladium base are presented. (authors)« less

Northrop Triga facility decommissioning plan versus actual results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gardner, F.W.

1986-01-01

This paper compares the Triga facility decontamination and decommissioning plan to the actual results and discusses key areas where operational activities were impacted upon by the final US Nuclear Regulatory Commission (NRC)-approved decontamination and decommissioning plan. Total exposures for fuel transfer were a factor of 4 less than planned. The design of the Triga reactor components allowed the majority of the components to be unconditionally released.

The Optimized Integration of the Decontamination Plan and the Radwaste Management Plan into Decommissioning Plan to the VVR-S Research Reactor from Romania

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barariu, G.

2008-07-01

The paper presents the progress of the Decontamination Plan and Radioactive Waste Management Plan which accompanies the Decommissioning Plan for research reactor VVR-S located in Magurele, Ilfov, near Bucharest, Romania. The new variant of the Decommissioning Plan was elaborated taking into account the IAEA recommendation concerning radioactive waste management. A new feasibility study for VVR-S decommissioning was also elaborated. The preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as rehabilitation of the existing Radioactive Waste Treatment Plant and the upgrade of the Radioactive Waste Disposal Facilitymore » at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and reusing of cleaned reactor building is envisaged. An inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the IAEA assistance. Environmental concerns are a part of the radioactive waste management strategy. In conclusion: The current version 8 of the Draft Decommissioning Plan which include the Integrated concept of Decontamination and Decommissioning and Radwaste Management, reflects the substantial work that has been incorporated by IFIN-HH in collaboration with SITON, which has resulted in substantial improvement in document The decommissioning strategy must take into account costs for VVR-S Reactor decommissioning, as well as costs for much needed refurbishments to the radioactive waste treatment plant and the Baita-Bihor waste disposal repository. Several improvements to the Baita-Bihor repository and IFIN-HH waste treatment facility were proposed. The quantities and composition of the radioactive waste generated by VVR-S Reactor dismantling were again estimated by streams and the best demonstrated practicable processing solution was proposed. The estimated quantities of materials to be managed in the near future raise some issues that need to be solved swiftly, such as treatment of aluminum and lead and graphite management. It is envisaged that these materials to be treated to Subsidiary for Nuclear Research (SCN) Pitesti. (authors)« less

Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department ofmore » Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial reactor use, and the high-level waste from the processing will be stabilized and stored for less than 20 years before being sent back to the Czech Republic for final disposition. Finally, the paper contains a section for the summary and conclusions.« less

Combined process of electrocoagulation and photocatalytic degradation for the treatment of olive washing wastewater.

PubMed

Ates, Hasan; Dizge, Nadir; Yatmaz, H Cengiz

2017-01-01

In this study, an electrocoagulation reactor (ECR) and photocatalytic reactor (PCR) were tested to understand the performance of combined electrocoagulation and photocatalytic-degradation of olive washing wastewater (OWW). The effects of initial pH (6.0, 6.9, 8.0, 9.0), applied voltage (10.0, 12.5, 15.0 V), and operating time (30, 60, 90, 120 min) were investigated in the electrocoagulation reactor when aluminum electrodes were used as both anode and cathode. The pH, conductivity, color, chemical oxygen demand (COD), and phenol were measured versus time to determine the efficiency of the ECR and PCR process. It was observed that electrocoagulation as a single treatment process supplied the COD removal of 62.5%, color removal of 98.1%, and total phenol removal of 87% at optimum conditions as pH 6.9, applied voltage of 12.5 V, and operating time of 120 min. Moreover, final pH and conductivity were 7.7 and 980 μS/cm, respectively. On the other hand, the effect of semiconductor catalyst type (TiO 2 and ZnO) and loading (1, 2, 3 g/L) were tested using PCR as a stand-alone technique. It was found that photocatalytic degradation as a single treatment process when using 1 g/L ZnO achieved the COD removal of 46%, color removal of 99% with a total phenol removal of 41% at optimum conditions. Final pH and conductivity were 6.2 and 915 μS/cm, respectively. Among semiconductor catalysts, TiO 2 and ZnO performed identical efficiencies for both COD and total phenol removal. Moreover, combination in which electrochemical degradation was employed as a pre-treatment to the photocatalytic degradation process obtained high COD removal of 88% and total phenol, as well as color removal of 100% for the OWW. The electrochemical treatment alone was not effective, but in combination with the photocatalytic process, led to a high-quality effluent. Finally, sludge collected from the electrocoagulation process was characterized by attenuated total reflection Fourier transform infrared and X-ray powder diffraction analyses.

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

2017-05-01

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

Biodegradation of 2,4,6-trichlorophenol in a packed-bed biofilm reactor equipped with an internal net draft tube riser for aeration and liquid circulation.

PubMed

Jesús, A Gómez-De; Romano-Baez, F J; Leyva-Amezcua, L; Juárez-Ramírez, C; Ruiz-Ordaz, N; Galíndez-Mayer, J

2009-01-30

For the aerobic biodegradation of the fungicide and defoliant 2,4,6-trichlorophenol (2,4,6-TCP), a bench-scale packed-bed bioreactor equipped with a net draft tube riser for liquid circulation and oxygenation (PB-ALR) was constructed. To obtain a high packed-bed volume relative to the whole bioreactor volume, a high A(D)/A(R) ratio was used. Reactor's downcomer was packed with a porous support of volcanic stone fragments. PB-ALR hydrodynamics and oxygen mass transfer behavior was evaluated and compared to the observed behavior of the unpacked reactor operating as an internal airlift reactor (ALR). Overall gas holdup values epsilon(G), and zonal oxygen mass transfer coefficients determined at various airflow rates in the PB-ALR, were higher than those obtained with the ALR. When comparing mixing time values obtained in both cases, a slight increment in mixing time was observed when reactor was operated as a PB-ALR. By using a mixed microbial community, the biofilm reactor was used to evaluate the aerobic biodegradation of 2,4,6-TCP. Three bacterial strains identified as Burkholderia sp., Burkholderia kururiensis and Stenotrophomonas sp. constituted the microbial consortium able to cometabolically degrade the 2,4,6-TCP, using phenol as primary substrate. This consortium removed 100% of phenol and near 99% of 2,4,6-TCP. Mineralization and dehalogenation of 2,4,6-TCP was evidenced by high COD removal efficiencies ( approximately 95%), and by the stoichiometric release of chloride ions from the halogenated compound ( approximately 80%). Finally, it was observed that the microbial consortium was also capable to metabolize 2,4,6-TCP without phenol as primary substrate, with high removal efficiencies (near 100% for 2,4,6-TCP, 92% for COD and 88% for chloride ions).

Nuclear modules for space electric propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1998-01-01

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

Characterization of reactive CaCO 3 crystallization in a fluidized bed reactor as a central process of direct air capture

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burhenne, Luisa; Giacomin, Caroline; Follett, Trevor

A laboratory-scale, fluidized-bed pellet reactor (BPR) was used to investigate a CaCO 3 crystallization process for the recovery of CO 2 in a Direct Air Capture (DAC) process. The BPR performance was validated against data from a pilot-scale unit. Subsequently, the pellet growth under process-relevant conditions was studied over a period of 144 h. The experimental results with the BPR, containing a bed of pellets sized between 0.65 and 0.84 mm, have shown that a calcium retention of 80% can be achieved at a fluidization velocity of 60 m h -1 and a calcium loading rate of 3 mol hmore » -1. This result is consistent with calcium retention observed at pilot scale operation and hence, results from the BPR are considered representative for the pilot scale unit. Starting with a bed of pellets sized between 0.15 and 0.5 mm, the average pellet growth rate, G, at the reactor bottom increased from 8.1E-10 to 11E–10 m s -1 at the onset and decreased to 4.9E–10 m s -1 over the course of a 144 h test. The calcium retention over the course the test showed the same trend (initial increase and final decrease) as the pellet growth rate. A theoretical bed growth model was developed and validated against data from the pilot scale and benchtop pellet reactors. The model was used to calculate the bed porosity and total pellet surface area in each control volume. Lastly, the pellet surface area growth at the bottom of the reactor reproduced the pellet growth and retention data trends.« less

A Methodology for the Integration of a Mechanistic Source Term Analysis in a Probabilistic Framework for Advanced Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grabaskas, Dave; Brunett, Acacia J.; Bucknor, Matthew

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release ofmore » radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.« less

Catalyst and process development for synthesis gas conversion to isobutylene. Final report, September 1, 1990--January 31, 1994

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anthony, R.G.; Akgerman, A.

1994-05-06

Previous work on isosynthesis (conversion of synthesis gas to isobutane and isobutylene) was performed at very low conversions or extreme process conditions. The objectives of this research were (1) determine the optimum process conditions for isosynthesis; (2) determine the optimum catalyst preparation method and catalyst composition/properties for isosynthesis; (3) determine the kinetics for the best catalyst; (4) develop reactor models for trickle bed, slurry, and fixed bed reactors; and (5) simulate the performance of fixed bed trickle flow reactors, slurry flow reactors, and fixed bed gas phase reactors for isosynthesis. More improvement in catalyst activity and selectivity is needed beforemore » isosynthesis can become a commercially feasible (stand-alone) process. Catalysts prepared by the precipitation method show the most promise for future development as compared with those prepared hydrothermally, by calcining zirconyl nitrate, or by a modified sol-gel method. For current catalysts the high temperatures (>673 K) required for activity also cause the production of methane (because of thermodynamics). A catalyst with higher activity at lower temperatures would magnify the unique selectivity of zirconia for isobutylene. Perhaps with a more active catalyst and acidification, oxygenate production could be limited at lower temperatures. Pressures above 50 atm cause an undesirable shift in product distribution toward heavier hydrocarbons. A model was developed that can predict carbon monoxide conversion an product distribution. The rate equation for carbon monoxide conversion contains only a rate constant and an adsorption equilibrium constant. The product distribution was predicted using a simple ratio of the rate of CO conversion. This report is divided into Introduction, Experimental, and Results and Discussion sections.« less

On the use of a molten salt fast reactor to apply an idealized transmutation scenario for the nuclear phase out.

PubMed

Merk, Bruno; Rohde, Ulrich; Glivici-Cotruţă, Varvara; Litskevich, Dzianis; Scholl, Susanne

2014-01-01

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations--a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described.

On the Use of a Molten Salt Fast Reactor to Apply an Idealized Transmutation Scenario for the Nuclear Phase Out

PubMed Central

Merk, Bruno; Rohde, Ulrich; Glivici-Cotruţă, Varvara; Litskevich, Dzianis; Scholl, Susanne

2014-01-01

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations – a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described. PMID:24690768

Hybrid finite-volume/transported PDF method for the simulation of turbulent reactive flows

NASA Astrophysics Data System (ADS)

Raman, Venkatramanan

A novel computational scheme is formulated for simulating turbulent reactive flows in complex geometries with detailed chemical kinetics. A Probability Density Function (PDF) based method that handles the scalar transport equation is coupled with an existing Finite Volume (FV) Reynolds-Averaged Navier-Stokes (RANS) flow solver. The PDF formulation leads to closed chemical source terms and facilitates the use of detailed chemical mechanisms without approximations. The particle-based PDF scheme is modified to handle complex geometries and grid structures. Grid-independent particle evolution schemes that scale linearly with the problem size are implemented in the Monte-Carlo PDF solver. A novel algorithm, in situ adaptive tabulation (ISAT) is employed to ensure tractability of complex chemistry involving a multitude of species. Several non-reacting test cases are performed to ascertain the efficiency and accuracy of the method. Simulation results from a turbulent jet-diffusion flame case are compared against experimental data. The effect of micromixing model, turbulence model and reaction scheme on flame predictions are discussed extensively. Finally, the method is used to analyze the Dow Chlorination Reactor. Detailed kinetics involving 37 species and 158 reactions as well as a reduced form with 16 species and 21 reactions are used. The effect of inlet configuration on reactor behavior and product distribution is analyzed. Plant-scale reactors exhibit quenching phenomena that cannot be reproduced by conventional simulation methods. The FV-PDF method predicts quenching accurately and provides insight into the dynamics of the reactor near extinction. The accuracy of the fractional time-stepping technique in discussed in the context of apparent multiple-steady states observed in a non-premixed feed configuration of the chlorination reactor.

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mamivand, Mahmood; Yang, Ying; Busby, Jeremy T.

The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300–400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10 –7 dpa/s and 390 °C) and thenmore » use the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10 –8 dpa/s and 275 °C). The model accurately predicts the γ' (Ni 3Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γ' for typical CW 316 SS (with 0.054 wt% carbon) under LWR extended life conditions. Finally, this work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.« less

Characterization of reactive CaCO 3 crystallization in a fluidized bed reactor as a central process of direct air capture

DOE PAGES

Burhenne, Luisa; Giacomin, Caroline; Follett, Trevor; ...

2017-10-25

A laboratory-scale, fluidized-bed pellet reactor (BPR) was used to investigate a CaCO 3 crystallization process for the recovery of CO 2 in a Direct Air Capture (DAC) process. The BPR performance was validated against data from a pilot-scale unit. Subsequently, the pellet growth under process-relevant conditions was studied over a period of 144 h. The experimental results with the BPR, containing a bed of pellets sized between 0.65 and 0.84 mm, have shown that a calcium retention of 80% can be achieved at a fluidization velocity of 60 m h -1 and a calcium loading rate of 3 mol hmore » -1. This result is consistent with calcium retention observed at pilot scale operation and hence, results from the BPR are considered representative for the pilot scale unit. Starting with a bed of pellets sized between 0.15 and 0.5 mm, the average pellet growth rate, G, at the reactor bottom increased from 8.1E-10 to 11E–10 m s -1 at the onset and decreased to 4.9E–10 m s -1 over the course of a 144 h test. The calcium retention over the course the test showed the same trend (initial increase and final decrease) as the pellet growth rate. A theoretical bed growth model was developed and validated against data from the pilot scale and benchtop pellet reactors. The model was used to calculate the bed porosity and total pellet surface area in each control volume. Lastly, the pellet surface area growth at the bottom of the reactor reproduced the pellet growth and retention data trends.« less

New reactor cavity cooling system having passive safety features using novel shape for HTGRs and VHTRs

DOE PAGES

Takamatsu, Kuniyoshi; Hu, Rui

2014-11-27

A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features without a requirement for electricity and mechanical drive is proposed for high temperature gas cooled reactors (HTGRs) and very high temperature reactors (VHTRs). The RCCS design consists of continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 (°C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV with radiation and natural convection. Employing the air as the working fluid and the ambient airmore » as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting due to overheating the fuels. The simulation results from a commercial CFD code, STAR-CCM+, show that the temperature distribution of the RCCS is within the temperature limits of the structures, such as the maximum operating temperature of the RPV, 713.15 (K) = 440 (°C), and the heat released from the RPV could be removed safely, even during a loss of coolant accident (LOCA). Finally, when the RCCS can remove 600 (kW) of the rated nominal state even during LOCA, the safety review for building the HTTR could confirm that the temperature distribution of the HTTR is within the temperature limits of the structures to secure structures and fuels after the shutdown because the large heat capacity of the graphite core can absorb heat from the fuel in a short period. Therefore, the capacity of the new RCCS design would be sufficient for decay heat removal.« less

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

DOE PAGES

Mamivand, Mahmood; Yang, Ying; Busby, Jeremy T.; ...

2017-03-11

The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300–400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10 –7 dpa/s and 390 °C) and thenmore » use the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10 –8 dpa/s and 275 °C). The model accurately predicts the γ' (Ni 3Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γ' for typical CW 316 SS (with 0.054 wt% carbon) under LWR extended life conditions. Finally, this work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.« less

Numerical modeling of carrier gas flow in atomic layer deposition vacuum reactor: A comparative study of lattice Boltzmann models

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pan, Dongqing; Chien Jen, Tien; Li, Tao

2014-01-15

This paper characterizes the carrier gas flow in the atomic layer deposition (ALD) vacuum reactor by introducing Lattice Boltzmann Method (LBM) to the ALD simulation through a comparative study of two LBM models. Numerical models of gas flow are constructed and implemented in two-dimensional geometry based on lattice Bhatnagar–Gross–Krook (LBGK)-D2Q9 model and two-relaxation-time (TRT) model. Both incompressible and compressible scenarios are simulated and the two models are compared in the aspects of flow features, stability, and efficiency. Our simulation outcome reveals that, for our specific ALD vacuum reactor, TRT model generates better steady laminar flow features all over the domainmore » with better stability and reliability than LBGK-D2Q9 model especially when considering the compressible effects of the gas flow. The LBM-TRT is verified indirectly by comparing the numerical result with conventional continuum-based computational fluid dynamics solvers, and it shows very good agreement with these conventional methods. The velocity field of carrier gas flow through ALD vacuum reactor was characterized by LBM-TRT model finally. The flow in ALD is in a laminar steady state with velocity concentrated at the corners and around the wafer. The effects of flow fields on precursor distributions, surface absorptions, and surface reactions are discussed in detail. Steady and evenly distributed velocity field contribute to higher precursor concentration near the wafer and relatively lower particle velocities help to achieve better surface adsorption and deposition. The ALD reactor geometry needs to be considered carefully if a steady and laminar flow field around the wafer and better surface deposition are desired.« less

Pratt & Whitney ESCORT derivative for mars surface power

NASA Astrophysics Data System (ADS)

Feller, Gerald J.; Joyner, Russell

1999-01-01

The purpose of this paper is to address the applicability of a common reactor system design from the Pratt & Whitney ESCORT nuclear thermal rocket engine concept to support current NASA mars surface-based power requirements. The ESCORT is a bimodal engine capable of supporting a wide range of propulsive thermal and vehicle electrical power requirements. The ESCORT engine is powered by a fast-spectrum beryllium-reflected CERMET-fueled nuclear reactor. In addition to an expander cycle propulsive mode, the ESCORT is capable of operating in an electrical power mode. In this mode, the reactor is used to heat a mixture of helium and xenon to drive a closed-loop Brayton cycle in order to generate electrical energy. Recent Design Reference Mission requirements (DRM) from NASA Johnson Space Center and NASA Lewis Research Center studies in 1997 and 1998 have detailed upgraded requirements for potential mars transfer missions. The current NASA DRM requires a nuclear thermal propulsion system capable of delivering total mission requirements of 200170 N (45000 lbf) thrust and 50 kWe of spacecraft electrical power. Additionally, these requirements detailed a surface power system capable of providing approximately 160 kW of electrical energy over an approximate 10 year period within a given weight and volume envelope. Current NASA studies use a SP-100 reactor (0.8 MT) and a NERVA derivative (1.6 MT) as baseline systems. A mobile power cart of approximate dimensions 1.7 m×4.5 m×4.4 m has been conceptualized to transport the reactor power system on the Mars Surface. The 63.25 cm diameter and 80.25 cm height of the ESCORT and its 1.3 MT of weight fit well within the current weight and volume target range of the NASA DRM requirements. The modifications required to the ESCORT reactor system to support this upgraded electrical power requirements along with operation in the Martian atmospheric conditions are addressed in this paper. Sufficient excess reactivity and burnup capability were designed into the ESCORT reactor system to support these upgraded requirements. Only slight modifications to reactor hardware were required to address any environmental considerations. These modifications involved sealing any refractory metal alloy components from the CO2 in the Martian Atmosphere. Also, the Brayton cycle Power Conversion Unit (PCU) hardware was modified to support the upgraded requirements. This paper discusses the design analysis performed and provides information on the final common reactor concept to be used on the Mars surface to support manned missions.

Westinghouse Small Modular Reactor nuclear steam supply system design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development andmore » integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)« less