Low-power lead-cooled fast reactor loaded with MOX-fuel

NASA Astrophysics Data System (ADS)

Sitdikov, E. R.; Terekhova, A. M.

2017-01-01

Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

Fuel loading of PeBR for a long operation life on the lunar surface

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

Schriener, T. M.; Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM; El-Genk, M. S.

2012-07-01

The Pellet Bed Reactor (PeBR) power system could provide 99.3 kW e to a lunar outpost for 66 full power years and is designed for no single point failures. The core of this fast energy spectrum reactor consists of three sectors that are neutronically and thermally coupled, but hydraulically independent. Each sector has a separate Closed Brayton Cycle (CBC) loop for energy conversion and separate water heat-pipes radiator panels for heat rejection. He-Xe (40 g/mole) binary gas mixture serves as the reactor coolant and CBC working fluid. On the lunar surface, the emplaced PeBR below grade is loaded with sphericalmore » fuel pellets (1-cm in dia.). It is launched unfueled and the pellets are launched in separate subcritical canisters, one for each core sector. This paper numerically simulates the transient loading of a core sector with fuel pellets on the Moon. The simulation accounts for the dynamic interaction of the pellets during loading and calculates the axial and radial distributions of the volume porosity in the sector. The pellets pack randomly with a volume porosity of 0.39 - 0.41 throughout most of the sector, except near the walls the local porosity is higher. (authors)« less

METHOD AND APPARATUS FOR EARTH PENETRATION

DOEpatents

Adams, W.M.

1963-12-24

A nuclear reactor apparatus for penetrating into the earth's crust is described. The apparatus comprises a cylindrical nuclear core operating at a temperature that is higher than the melting temperature of rock. A high-density ballast member is coupled to the nuclear core such that the overall density of the core-ballast assembly is greater than the density of molten rock. The nuclear core is thermally insulated so that its heat output is constrained to flow axially, with radial heat flow being minimized. In operation, the apparatus is placed in contact with the earth's crust at the point desired to be penetrated. The heat output of the reactor melts the underlying rock, and the apparatus sinks through the resulting magma. The fuel loading of the reactor core determines the ultimate depth of crust penetration. (AEC)

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.

A Methodology for Loading the Advanced Test Reactor Driver Core for Experiment Analysis

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

Cowherd, Wilson M.; Nielsen, Joseph W.; Choe, Dong O.

In support of experiments in the ATR, a new methodology was devised for loading the ATR Driver Core. This methodology will replace the existing methodology used by the INL Neutronic Analysis group to analyze experiments. Studied in this paper was the as-run analysis for ATR Cycle 152B, specifically comparing measured lobe powers and eigenvalue calculations.

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

BNL program in support of LWR degraded-core accident analysis

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

Ginsberg, T.; Greene, G.A.

1982-01-01

Two major sources of loading on dry watr reactor containments are steam generatin from core debris water thermal interactions and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described in this paper. 8 figures.

Structural response of 1/20-scale models of the Clinch River Breeder Reactor to a simulated hypothetical core disruptive accident. Technical report 4

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

Romander, C. M.; Cagliostro, D. J.

Five experiments were performed to help evaluate the structural integrity of the reactor vessel and head design and to verify code predictions. In the first experiment (SM 1), a detailed model of the head was loaded statically to determine its stiffness. In the remaining four experiments (SM 2 to SM 5), models of the vessel and head were loaded dynamically under a simulated 661 MW-sec hypothetical core disruptive accident (HCDA). Models SM 2 to SM 4, each of increasing complexity, systematically showed the effects of upper internals structures, a thermal liner, core support platform, and torospherical bottom on vessel response.more » Model SM 5, identical to SM 4 but more heavily instrumented, demonstrated experimental reproducibility and provided more comprehensive data. The models consisted of a Ni 200 vessel and core barrel, a head with shielding and simulated component masses, an upper internals structure (UIS), and, in the more complex models SM 4 and SM 5, a Ni 200 thermal liner and core support structure. Water simulated the liquid sodium coolant and a low-density explosive simulated the HCDA loads.« less

Study on Ultra-Long Life,Small U-Zr Metallic Fuelled Core With Burnable Poison

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

Kenji Tsuji; Hiromitsu Inagaki; Akira Nishikawa

2002-07-01

A conceptual design for a 50 MWe sodium cooled, U-Pu-Zr metallic fuelled, fast reactor core, which aims at a core lifetime of 30 years, has been performed [1]. As for the compensation for a large burn-up reactivity through 30 years, an axially movable reflector, which is located around the core, carries the major part of it and a burnable poison does the rest. This concept has achieved not only a long core lifetime but also a high discharged burn-up. On this study, a conceptual design for a small fast reactor loading U-Zr metallic fuelled core instead of U-Pu-Zr fuelled coremore » has been conducted, based on the original core arrangement of 4S reactor [2]. Within the range of this study including safety requirements, adopting the burnable poison would be effective to construct a core concept that achieves both a long lifetime and a high discharged burn-up. (authors)« less

BNL severe-accident sequence experiments and analysis program. [PWR; BWR

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

Greene, G.A.; Ginsberg, T.; Tutu, N.K.

1983-01-01

In the analysis of degraded core accidents, the two major sources of pressure loading on light water reactor containments are: steam generation from core debris-water thermal interactions; and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described.

Real-time LMR control parameter generation using advanced adaptive synthesis

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

King, R.W.; Mott, J.E.

1990-01-01

The reactor delta T'', the difference between the average core inlet and outlet temperatures, for the liquid-sodium-cooled Experimental Breeder Reactor 2 is empirically synthesized in real time from, a multitude of examples of past reactor operation. The real-time empirical synthesis is based on reactor operation. The real-time empirical synthesis is based on system state analysis (SSA) technology embodied in software on the EBR 2 data acquisition computer. Before the real-time system is put into operation, a selection of reactor plant measurements is made which is predictable over long periods encompassing plant shutdowns, core reconfigurations, core load changes, and plant startups.more » A serial data link to a personal computer containing SSA software allows the rapid verification of the predictability of these plant measurements via graphical means. After the selection is made, the real-time synthesis provides a fault-tolerant estimate of the reactor delta T accurate to {plus}/{minus}1{percent}. 5 refs., 7 figs.« less

NEUTRONIC REACTORS

DOEpatents

Wigner, E.P.; Young, G.J.

1958-10-14

A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

Effect of volumetric organic loading on the nitrogen removal rate by immobilised activated sludge.

PubMed

Zielinska, M; Wojnowska-Baryla, I

2006-05-01

Activated sludge was immobilised in a porous ceramic carrier to create a stationary core of a bio-reactor. Municipal wastewater was treated in this reactor under varied conditions of volumetric organic loading rate (expressed by chemical oxygen demand (COD)) that were the following: 6.5, 8.0, 20.8, 48.8 g COD l(-1) d(-1). The rate constants of ammonification, nitrification and denitrification under aerobic conditions were determined. All rate constants increased with a growth in volumetric loading rate, but the highest loading value of 48.8 g COD l(-1) d(-1) limited the ammonification and nitrification rates.

Self isolating high frequency saturable reactor

DOEpatents

Moore, James A.

1998-06-23

The present invention discloses a saturable reactor and a method for decoupling the interwinding capacitance from the frequency limitations of the reactor so that the equivalent electrical circuit of the saturable reactor comprises a variable inductor. The saturable reactor comprises a plurality of physically symmetrical magnetic cores with closed loop magnetic paths and a novel method of wiring a control winding and a RF winding. The present invention additionally discloses a matching network and method for matching the impedances of a RF generator to a load. The matching network comprises a matching transformer and a saturable reactor.

Structural response of 1/20-scale models of the Clinch River Breeder Reactor to a simulated hypothetical core-disruptive accident

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

Romander, C M; Cagliostro, D J

Five experiments were performed to help evaluate the structural integrity of the reactor vessel and head design and to verify code predictions. In the first experiment (SM 1), a detailed model of the head was loaded statically to determine its stiffness. In the remaining four experiments (SM 2 to SM 5), models of the vessel and head were loaded dynamically under a simulated 661 MW-s hypothetical core disruptive accident (HCDA). Models SM 2 to SM 4, each of increasing complexity, systematically showed the effects of upper internals structures, a thermal liner, core support platform, and torospherical bottom on vessel response.more » Model SM 5, identical to SM 4 but more heavily instrumented, demonstrated experimental reproducibility and provided more comprehensive data. The models consisted of a Ni 200 vessel and core barrel, a head with shielding and simulated component masses, and an upper internals structure (UIS).« less

Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

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

Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.

2008-06-23

This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been mademore » at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.« less

Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

NASA Astrophysics Data System (ADS)

Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

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

Lell, R. M.; Morman, J. A.; Schaefer, R.W.

ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide,more » U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium component to construct a central core zone with a composition closer to that in an LMFBR core with high burnup. The high {sup 240}Pu configuration was constructed for two reasons. First, the composition of the high {sup 240}Pu zone more closely matched the composition of LMFBR cores anticipated in design work in 1970. Second, comparison of measurements in the ZPR-6/7 uniform core with corresponding measurements in the high {sup 240}Pu zone provided an assessment of some of the effects of long-term {sup 240}Pu buildup in LMFBR cores. The uniform core version of ZPR-6/7 is evaluated in ZPR-LMFR-EXP-001. This document only addresses measurements in the high {sup 240}Pu core version of ZPR-6/7. Many types of measurements were performed as part of the ZPR-6/7 program. Measurements of criticality, sodium void worth, control rod worth and reaction rate distributions in the high {sup 240}Pu core configuration are evaluated here. For each category of measurements, the uncertainties are evaluated, and benchmark model data are provided.« less

Development of 3D pseudo pin-by-pin calculation methodology in ANC

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

Zhang, B.; Mayhue, L.; Huria, H.

2012-07-01

Advanced cores and fuel assembly designs have been developed to improve operational flexibility, economic performance and further enhance safety features of nuclear power plants. The simulation of these new designs, along with strong heterogeneous fuel loading, have brought new challenges to the reactor physics methodologies currently employed in the industrial codes for core analyses. Control rod insertion during normal operation is one operational feature in the AP1000{sup R} plant of Westinghouse next generation Pressurized Water Reactor (PWR) design. This design improves its operational flexibility and efficiency but significantly challenges the conventional reactor physics methods, especially in pin power calculations. Themore » mixture loading of fuel assemblies with significant neutron spectrums causes a strong interaction between different fuel assembly types that is not fully captured with the current core design codes. To overcome the weaknesses of the conventional methods, Westinghouse has developed a state-of-the-art 3D Pin-by-Pin Calculation Methodology (P3C) and successfully implemented in the Westinghouse core design code ANC. The new methodology has been qualified and licensed for pin power prediction. The 3D P3C methodology along with its application and validation will be discussed in the paper. (authors)« less

Internal core tightener

DOEpatents

Brynsvold, Glen V.; Snyder, Jr., Harold J.

1976-06-22

An internal core tightener which is a linear actuated (vertical actuation motion) expanding device utilizing a minimum of moving parts to perform the lateral tightening function. The key features are: (1) large contact areas to transmit loads during reactor operation; (2) actuation cam surfaces loaded only during clamping and unclamping operation; (3) separation of the parts and internal operation involved in the holding function from those involved in the actuation function; and (4) preloaded pads with compliant travel at each face of the hexagonal assembly at the two clamping planes to accommodate thermal expansion and irradiation induced swelling. The latter feature enables use of a "fixed" outer core boundary, and thus eliminates the uncertainty in gross core dimensions, and potential for rapid core reactivity changes as a result of core dimensional change.

Droplet Core Nuclear Rocket (DCNR)

NASA Technical Reports Server (NTRS)

Anghaie, Samim

1991-01-01

The most basic design feature of the droplet core nuclear reactor is to spray liquid uranium into the core in the form of droplets on the order of five to ten microns in size, to bring the reactor to critical conditions. The liquid uranium fuel ejector is driven by hydrogen, and more hydrogen is injected from the side of the reactor to about one and a half meters from the top. High temperature hydrogen is expanded through a nozzle to produce thrust. The hydrogen pressure in the system can be somewhere between 50 and 500 atmospheres; the higher pressure is more desirable. In the lower core region, hydrogen is tangentially injected to serve two purposes: (1) to provide a swirling flow to protect the wall from impingement of hot uranium droplets: (2) to generate a vortex flow that can be used for fuel separation. The reactor is designed to maximize the energy generation in the upper region of the core. The system can result in and Isp of 2000 per second, and a thrust-to-weight ratio of 1.6 for the shielded reactor. The nuclear engine system can reduce the Mars mission duration to less than 200 days. It can reduce the hydrogen consumption by a factor of 2 to 3, which reduces the hydrogen load by about 130 to 150 metric tons.

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

Bi, G.; Liu, C.; Si, S.

This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis ofmore » reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)« less

On Heat Loading, Novel Divertors, and Fusion Reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, Mike

2006-10-01

A new magnetic divertor geometry has been proposed to solve reactor heat exhaust problems, which are far more severe for a reactor than for ITER. Using reactor-compatible coils to generate an extra X-point downstream from the main X-point, the new X-divertor (XD) is shown to greatly expand magnetic flux at the divertor plates. As a result, the heat is distributed over a larger area and the line length is greatly increased. The heat-flux limitations of a standard divertor (SD) force a high core radiation fraction (fRad) in most reactor designs that necessarily have a several times higher ratio of heating power to radius (P/R) than ITER. It is argued that such high values of fRad will probably have serious deleterious consequences on the core confinement and stability of a burning plasma. Operation with internal transport barriers (ITBs) does not appear to overcome this problem. By reducing the core fRad within an acceptable range, the X-divertor is shown to substantially lower the core confinement requirement for a fusion reactor. As a bonus, the XD also enables the use of liquid metals by reducing the MHD drag. A possible series of experiments for an efficient and attractive path to practical fusion power is suggested.

Formation and hydrodynamic characteristics of aerobic granules in an activated sludge system.

PubMed

Ganesan, M V; Saravanan, V; Sreekrishnan, T R

2007-02-01

Development of aerobic granules in the aeration tank of an activated sludge system has been studied. The introduction of activated carbon particles into the aeration tank resulted in the formation of biogranules containing activated carbon as core nuclei. The presence of activated carbon also induced the formation of self-immobilized granules, which did not have any carrier particle at their core. The presence of aerobic granules enhanced the treatment efficiency of the reactor. At an organic loading rate of 32.8 kg COD m(-3)d(-1) and 0.78 h hydraulic retention time (HRT), the reactor showed 96% COD removal efficiency. At an HRT of 0.272 h and organic loading rate of 46.7 kg COD m(-3)d(-1), the reactor outlet COD remained below 100 mg l(-1). Settling velocity studies carried out on the biogranules showed that the drag coefficient of biogranules is greater than that of the rigid particle at the same Reynolds number.

Fuel transfer system

DOEpatents

Townsend, Harold E.; Barbanti, Giancarlo

1994-01-01

A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool.

Fuel transfer system

DOEpatents

Townsend, H.E.; Barbanti, G.

1994-03-01

A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool. 6 figures.

AIR COOLED NEUTRONIC REACTOR

DOEpatents

Fermi, E.; Szilard, L.

1958-05-27

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

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

Chandler, David; Betzler, Ben; Hirtz, Gregory John

2016-09-01

The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length 252Cf production targets, 75Se productionmore » capsules, 63Ni production capsules, a 188W production capsule, and various materials irradiation targets. Fully loaded 238Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.« less

Reactors as a Source of Antineutrinos: Effects of Fuel Loading and Burnup for Mixed-Oxide Fuels

NASA Astrophysics Data System (ADS)

Bernstein, Adam; Bowden, Nathaniel S.; Erickson, Anna S.

2018-01-01

In a conventional light-water reactor loaded with a range of uranium and plutonium-based fuel mixtures, the variation in antineutrino production over the cycle reflects both the initial core fissile inventory and its evolution. Under an assumption of constant thermal power, we calculate the rate at which antineutrinos are emitted from variously fueled cores, and the evolution of that rate as measured by a representative ton-scale antineutrino detector. We find that antineutrino flux decreases with burnup for low-enriched uranium cores, increases for full mixed-oxide (MOX) cores, and does not appreciably change for cores with a MOX fraction of approximately 75%. Accounting for uncertainties in the fission yields in the emitted antineutrino spectra and the detector response function, we show that the difference in corewide MOX fractions at least as small as 8% can be distinguished using a hypothesis test. The test compares the evolution of the antineutrino rate relative to an initial value over part or all of the cycle. The use of relative rates reduces the sensitivity of the test to an independent thermal power measurement, making the result more robust against possible countermeasures. This rate-only approach also offers the potential advantage of reducing the cost and complexity of the antineutrino detectors used to verify the diversion, compared to methods that depend on the use of the antineutrino spectrum. A possible application is the verification of the disposition of surplus plutonium in nuclear reactors.

Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

DOE PAGES

Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

2016-11-18

Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies ismore » a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium disposed of, land use per energy generated, and carbon emissions per energy generated. Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores which means high leakage (there is a radial and axial component), (2) no boron which means heterogeneous core and extensive use of control rods and BPs, and (3) single batch cores. But not all of the SMR designs have all of these traits. As a result, the approach used in this study is therefore a bounding case and not all SMRs may be affected to the same extent.« less

NUCLEAR REACTOR CORE

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

Bell, F.R.

1963-02-01

A nuclear reactor core composed of a number of identical elements of solid moderator material fitted together was designed. Each moderator element is apertured to provide channels for fuel and coolant. The elements have an external shape which permits them to be stacked in layers with similar elements, with the surfaces of adjacent elements fitting and in contact with each other. The cross section of the element is of a general hexagonal shape with identations and protrusions, so that the elements can be fitted together. The described core should not be liable to fracture under transverse loading. Specific arrangements ofmore » moderator elements and fuel and coolant apertures are described. (M.P.G.)« less

Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

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

Petrovic, Bojan; Maldonado, Ivan

2016-04-14

The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed onmore » December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.« less

Feasibility study Part I - Thermal hydraulic analysis of LEU target for {sup 99}Mo production in Tajoura reactor

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

Bsebsu, F.M.; Abotweirat, F.; Elwaer, S.

2008-07-15

The Renewable Energies and Water Desalination Research Center (REWDRC), Libya, will implement the technology for {sup 99}Mo isotope production using LEU foil target, to obtain new revenue streams for the Tajoura nuclear research reactor and desiring to serve the Libyan hospitals by providing the medical radioisotopes. Design information is presented for LEU target with irradiation device and irradiation Beryllium (Be) unit in the Tajoura reactor core. Calculated results for the reactor core with LEU target at different level of power are presented for steady state and several reactivity induced accident situations. This paper will present the steady state thermal hydraulicmore » design and transient analysis of Tajoura reactor was loaded with LEU foil target for {sup 99}Mo production. The results of these calculations show that the reactor with LEU target during the several cases of transient are in safe and no problems will occur. (author)« less

Integral Full Core Multi-Physics PWR Benchmark with Measured Data

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

Forget, Benoit; Smith, Kord; Kumar, Shikhar

In recent years, the importance of modeling and simulation has been highlighted extensively in the DOE research portfolio with concrete examples in nuclear engineering with the CASL and NEAMS programs. These research efforts and similar efforts worldwide aim at the development of high-fidelity multi-physics analysis tools for the simulation of current and next-generation nuclear power reactors. Like all analysis tools, verification and validation is essential to guarantee proper functioning of the software and methods employed. The current approach relies mainly on the validation of single physic phenomena (e.g. critical experiment, flow loops, etc.) and there is a lack of relevantmore » multiphysics benchmark measurements that are necessary to validate high-fidelity methods being developed today. This work introduces a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading and re-loading patterns. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from 58 instrumented assemblies. The benchmark description is now available online and has been used by many groups. However, much work remains to be done on the quantification of uncertainties and modeling sensitivities. This work aims to address these deficiencies and make this benchmark a true non-proprietary international benchmark for the validation of high-fidelity tools. This report details the BEAVRS uncertainty quantification for the first two cycle of operations and serves as the final report of the project.« less

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.

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

DOE PAGES

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; ...

2014-11-04

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Core design of a direct-cycle, supercritical-water-cooled fast breeder reactor

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

Jevremovic, T.; Oka, Yoshiaki; Koshizuka, Seiichi

1994-10-01

The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without muchmore » deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.« less

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

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

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Simultaneous optimization of loading pattern and burnable poison placement for PWRs

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

Alim, F.; Ivanov, K.; Yilmaz, S.

2006-07-01

To solve in-core fuel management optimization problem, GARCO-PSU (Genetic Algorithm Reactor Core Optimization - Pennsylvania State Univ.) is developed. This code is applicable for all types and geometry of PWR core structures with unlimited number of fuel assembly (FA) types in the inventory. For this reason an innovative genetic algorithm is developed with modifying the classical representation of the genotype. In-core fuel management heuristic rules are introduced into GARCO. The core re-load design optimization has two parts, loading pattern (LP) optimization and burnable poison (BP) placement optimization. These parts depend on each other, but it is difficult to solve themore » combined problem due to its large size. Separating the problem into two parts provides a practical way to solve the problem. However, the result of this method does not reflect the real optimal solution. GARCO-PSU achieves to solve LP optimization and BP placement optimization simultaneously in an efficient manner. (authors)« less

Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

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

Beyer, Brian David; Beddingfield, David H; Durst, Philip

2010-01-01

The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguardsmore » criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.« less

EBR-II Data Digitization

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

Yoon, Su-Jong; Rabiti, Cristian; Sackett, John

2014-08-01

1. Objectives To produce a validation database out of those recorded signals it will be necessary also to identify the documents need to reconstruct the status of reactor at the time of the beginning of the recordings. This should comprehends the core loading specification (assemblies type and location and burn-up) along with this data the assemblies drawings and the core drawings will be identified. The first task of the project will be identify the location of the sensors, with respect the reactor plant layout, and the physical quantities recorded by the Experimental Breeder Reactor-II (EBR-II) data acquisition system. This firstmore » task will allow guiding and prioritizing the selection of drawings needed to numerically reproduce those signals. 1.1 Scopes and Deliverables The deliverables of this project are the list of sensors in EBR-II system, the identification of storing location of those sensors, identification of a core isotopic composition at the moment of the start of system recording. Information of the sensors in EBR-II reactor system was summarized from the EBR-II system design descriptions listed in Section 1.2.« less

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

Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies ismore » a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium disposed of, land use per energy generated, and carbon emissions per energy generated. Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores which means high leakage (there is a radial and axial component), (2) no boron which means heterogeneous core and extensive use of control rods and BPs, and (3) single batch cores. But not all of the SMR designs have all of these traits. As a result, the approach used in this study is therefore a bounding case and not all SMRs may be affected to the same extent.« less

A coupled nuclear reactor thermal energy storage system for enhanced load following operation

NASA Astrophysics Data System (ADS)

Alameri, Saeed A.

Nuclear power plants usually provide base-load electric power and operate most economically at a constant power level. In an energy grid with a high fraction of renewable energy sources, future nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling the reactor to a large Thermal Energy Storage (TES) block will allow the reactor to better respond to variable power demands. In the system described in this thesis, a Prismatic-core Advanced High Temperature Reactor (PAHTR) operates at constant power with heat provided to a TES block that supplies power as needed to a secondary energy conversion system. The PAHTR is designed to have a power rating of 300 MW th, with 19.75 wt% enriched Tri-Structural-Isotropic UO 2 fuel and a five year operating cycle. The passive molten salt TES system will operate in the latent heat region with an energy storage capacity of 150 MWd. Multiple smaller TES blocks are used instead of one large block to enhance the efficiency and maintenance complexity of the system. A transient model of the coupled reactor/TES system is developed to study the behavior of the system in response to varying load demands. The model uses six-delayed group point kinetics and decay heat models coupled to thermal-hydraulic and heat transfer models of the reactor and TES system. Based on the transient results, the preferred TES design consists of 1000 blocks, each containing 11000 LiCl phase change material tubes. A safety assessment of major reactor events demonstrates the inherent safety of the coupled system. The loss of forced circulation study determined the minimum required air convection heat removal rate from the reactor core and the lowest possible reduced primary flow rate that can maintain the reactor in a safe condition. The loss of ultimate heat sink study demonstrated the ability of the TES to absorb the decay heat of the reactor fuel while cooling the PAHTR after an emergency shutdown. The simulated reactivity insertion accident assessment determined the maximum allowable reactivity insertion to the PAHTR as a function of shutdown response times.

Coupled field-structural analysis of HGTR fuel brick using ABAQUS

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

Mohanty, S.; Jain, R.; Majumdar, S.

2012-07-01

High-temperature, gas-cooled reactors (HTGRs) are usually helium-gas cooled, with a graphite core that can operate at reactor outlet temperatures much higher than can conventional light water reactors. In HTGRs, graphite components moderate and reflect neutrons. During reactor operation, high temperature and high irradiation cause damage to the graphite crystal and grains and create other defects. This cumulative structural damage during the reactor lifetime leads to changes in graphite properties, which can alter the ability to support the designed loads. The aim of the present research is to develop a finite-element code using commercially available ABAQUS software for the structural integritymore » analysis of graphite core components under extreme temperature and irradiation conditions. In addition, the Reactor Geometry Generator tool-kit, developed at Argonne National Laboratory, is used to generate finite-element mesh for complex geometries such as fuel bricks with multiple pin holes and coolant flow channels. This paper presents the proposed concept and discusses results of stress analysis simulations of a fuel block with H-451 grade material properties. (authors)« less

Benchmark Evaluation of the HTR-PROTEUS Absorber Rod Worths (Core 4)

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

John D. Bess; Leland M. Montierth

2014-06-01

PROTEUS was a zero-power research reactor at the Paul Scherrer Institute (PSI) in Switzerland. The critical assembly was constructed from a large graphite annulus surrounding a central cylindrical cavity. Various experimental programs were investigated in PROTEUS; during the years 1992 through 1996, it was configured as a pebble-bed reactor and designated HTR-PROTEUS. Various critical configurations were assembled with each accompanied by an assortment of reactor physics experiments including differential and integral absorber rod measurements, kinetics, reaction rate distributions, water ingress effects, and small sample reactivity effects [1]. Four benchmark reports were previously prepared and included in the March 2013 editionmore » of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [2] evaluating eleven critical configurations. A summary of that effort was previously provided [3] and an analysis of absorber rod worth measurements for Cores 9 and 10 have been performed prior to this analysis and included in PROTEUS-GCR-EXP-004 [4]. In the current benchmark effort, absorber rod worths measured for Core Configuration 4, which was the only core with a randomly-packed pebble loading, have been evaluated for inclusion as a revision to the HTR-PROTEUS benchmark report PROTEUS-GCR-EXP-002.« less

Switching power pulse system

DOEpatents

Aaland, K.

1983-08-09

A switching system for delivering pulses of power from a source to a load using a storage capacitor charged through a rectifier, and maintained charged to a reference voltage level by a transistor switch and voltage comparator. A thyristor is triggered to discharge the storage capacitor through a saturable reactor and fractional turn saturable transformer having a secondary to primary turn ratio N of n:l/n = n[sup 2]. The saturable reactor functions as a soaker'' while the thyristor reaches saturation, and then switches to a low impedance state. The saturable transformer functions as a switching transformer with high impedance while a load coupling capacitor charges, and then switches to a low impedance state to dump the charge of the storage capacitor into the load through the coupling capacitor. The transformer is comprised of a multilayer core having two secondary windings tightly wound and connected in parallel to add their output voltage and reduce output inductance, and a number of single turn windings connected in parallel at nodes for the primary winding, each single turn winding linking a different one of the layers of the multilayer core. The load may be comprised of a resistive beampipe for a linear particle accelerator and capacitance of a pulse forming network. To hold off discharge of the capacitance until it is fully charged, a saturable core is provided around the resistive beampipe to isolate the beampipe from the capacitance until it is fully charged. 5 figs.

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

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

Permana, Sidik; Novitrian,; Waris, Abdul

Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by conversion ratio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissilemore » material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loading scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.« less

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania`s 14-MW TRIGA reactor

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

Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

1992-12-01

The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, eachmore » containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations.« less

Microbial Population Dynamics and Ecosystem Functions of Anoxic/Aerobic Granular Sludge in Sequencing Batch Reactors Operated at Different Organic Loading Rates

PubMed Central

Szabó, Enikö; Liébana, Raquel; Hermansson, Malte; Modin, Oskar; Persson, Frank; Wilén, Britt-Marie

2017-01-01

The granular sludge process is an effective, low-footprint alternative to conventional activated sludge wastewater treatment. The architecture of the microbial granules allows the co-existence of different functional groups, e.g., nitrifying and denitrifying communities, which permits compact reactor design. However, little is known about the factors influencing community assembly in granular sludge, such as the effects of reactor operation strategies and influent wastewater composition. Here, we analyze the development of the microbiomes in parallel laboratory-scale anoxic/aerobic granular sludge reactors operated at low (0.9 kg m-3d-1), moderate (1.9 kg m-3d-1) and high (3.7 kg m-3d-1) organic loading rates (OLRs) and the same ammonium loading rate (0.2 kg NH4-N m-3d-1) for 84 days. Complete removal of organic carbon and ammonium was achieved in all three reactors after start-up, while the nitrogen removal (denitrification) efficiency increased with the OLR: 0% at low, 38% at moderate, and 66% at high loading rate. The bacterial communities at different loading rates diverged rapidly after start-up and showed less than 50% similarity after 6 days, and below 40% similarity after 84 days. The three reactor microbiomes were dominated by different genera (mainly Meganema, Thauera, Paracoccus, and Zoogloea), but these genera have similar ecosystem functions of EPS production, denitrification and polyhydroxyalkanoate (PHA) storage. Many less abundant but persistent taxa were also detected within these functional groups. The bacterial communities were functionally redundant irrespective of the loading rate applied. At steady-state reactor operation, the identity of the core community members was rather stable, but their relative abundances changed considerably over time. Furthermore, nitrifying bacteria were low in relative abundance and diversity in all reactors, despite their large contribution to nitrogen turnover. The results suggest that the OLR has considerable impact on the composition of the granular sludge communities, but also that the granule communities can be dynamic even at steady-state reactor operation due to high functional redundancy of several key guilds. Knowledge about microbial diversity with specific functional guilds under different operating conditions can be important for engineers to predict the stability of reactor functions during the start-up and continued reactor operation. PMID:28507540

Determining the morphology of polystyrene-block-poly(2-vinylpyridine) micellar reactors for ZnO nanoparticle synthesis.

PubMed

El-Atwani, Osman; El-Atwani, Osman C; Aytun, Taner; Mutaf, Omer Faruk; Srot, Vesna; van Aken, Peter A; Ow-Yang, Cleva W

2010-05-18

We report the use of reverse PS-b-P2VP diblock copolymer micelles as true nanoscale-sized reactor vessels to synthesize ZnO nanoparticles. The reverse micelles were formed in toluene and then sequentially loaded with zinc acetate dihydrate and tetramethylammonium hydroxide reactants. Moreover, high spatial resolution Z-contrast imaging and EDX spectroscopy techniques were used to confirm the segregation of the Zn cation to the core of the loaded micelles. Determining the chemical distribution with high nanoscale spatial resolution is shown to complement the less direct characterization by AFM, DLS and FTIR, thus demonstrating broader implications for the characterization of hybrid nanocomposite systems.

Shear compression testing of glass-fibre steel specimens after 4K reactor irradiation: Present status and facility upgrade

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

Gerstenberg, H.; Kraehling, E.; Katheder, H.

1997-06-01

The shear strengths of various fibre reinforced resins being promising candidate insulators for superconducting coils to be used tinder a strong radiation load, e.g. in future fusion reactors were investigated prior and subsequent to reactor in-core irradiation at liquid helium temperature. A large number of sandwich-like (steel-bonded insulation-steel) specimens representing a widespread variety of materials and preparation techniques was exposed to irradiation doses of up to 5 x 10{sup 7} Gy in form of fast neutrons and {gamma}-radiation. In a systematic study several experimental parameters including irradiation dose, postirradiation storage temperature and measuring temperature were varied before the determination ofmore » the ultimate shear strength. The results obtained from the different tested materials are compared. In addition an upgrade of the in-situ test rig installed at the Munich research reactor is presented, which allows combined shear/compression loading of low temperature irradiated specimens and provides a doubling of the testing rate.« less

Performance of the MTR core with MOX fuel using the MCNP4C2 code.

PubMed

Shaaban, Ismail; Albarhoum, Mohamad

2016-08-01

The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. Copyright © 2016 Elsevier Ltd. All rights reserved.

A NEUTRONIC REACTOR

DOEpatents

Luebke, E.A.; Vandenberg, L.B.

1959-09-01

A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

Nuclear component horizontal seismic restraint

DOEpatents

Snyder, Glenn J.

1988-01-01

A nuclear component horizontal seismic restraint. Small gaps limit horizontal displacement of components during a seismic occurrence and therefore reduce dynamic loadings on the free lower end. The reactor vessel and reactor guard vessel use thicker section roll-forged rings welded between the vessel straight shell sections and the bottom hemispherical head sections. The inside of the reactor guard vessel ring forging contains local vertical dovetail slots and upper ledge pockets to mount and retain field fitted and installed blocks. As an option, the horizontal displacement of the reactor vessel core support cone can be limited by including shop fitted/installed local blocks in opposing alignment with the reactor vessel forged ring. Beams embedded in the wall of the reactor building protrude into apertures in the thermal insulation shell adjacent the reactor guard vessel ring and have motion limit blocks attached thereto to provide to a predetermined clearance between the blocks and reactor guard vessel ring.

Nuclear reactor construction with bottom supported reactor vessel

DOEpatents

Sharbaugh, John E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Switching power pulse system

DOEpatents

Aaland, Kristian

1983-01-01

A switching system for delivering pulses of power from a source (10) to a load (20) using a storage capacitor (C3) charged through a rectifier (D1, D2), and maintained charged to a reference voltage level by a transistor switch (Q1) and voltage comparator (12). A thyristor (22) is triggered to discharge the storage capacitor through a saturable reactor (18) and fractional turn saturable transformer (16) having a secondary to primary turn ratio N of n:l/n=n.sup.2. The saturable reactor (18) functions as a "soaker" while the thyristor reaches saturation, and then switches to a low impedance state. The saturable transformer functions as a switching transformer with high impedance while a load coupling capacitor (C4) charges, and then switches to a low impedance state to dump the charge of the storage capacitor (C3) into the load through the coupling capacitor (C4). The transformer is comprised of a multilayer core (26) having two secondary windings (28, 30) tightly wound and connected in parallel to add their output voltage and reduce output inductance, and a number of single turn windings connected in parallel at nodes (32, 34) for the primary winding, each single turn winding linking a different one of the layers of the multilayer core. The load may be comprised of a resistive beampipe (40) for a linear particle accelerator and capacitance of a pulse forming network (42). To hold off discharge of the capacitance until it is fully charged, a saturable core (44) is provided around the resistive beampipe (40) to isolate the beampipe from the capacitance (42) until it is fully charged.

Implementing a Nuclear Power Plant Model for Evaluating Load-Following Capability on a Small Grid

NASA Astrophysics Data System (ADS)

Arda, Samet Egemen

A pressurized water reactor (PWR) nuclear power plant (NPP) model is introduced into Positive Sequence Load Flow (PSLF) software by General Electric in order to evaluate the load-following capability of NPPs. The nuclear steam supply system (NSSS) consists of a reactor core, hot and cold legs, plenums, and a U-tube steam generator. The physical systems listed above are represented by mathematical models utilizing a state variable lumped parameter approach. A steady-state control program for the reactor, and simple turbine and governor models are also developed. Adequacy of the isolated reactor core, the isolated steam generator, and the complete PWR models are tested in Matlab/Simulink and dynamic responses are compared with the test results obtained from the H. B. Robinson NPP. Test results illustrate that the developed models represents the dynamic features of real-physical systems and are capable of predicting responses due to small perturbations of external reactivity and steam valve opening. Subsequently, the NSSS representation is incorporated into PSLF and coupled with built-in excitation system and generator models. Different simulation cases are run when sudden loss of generation occurs in a small power system which includes hydroelectric and natural gas power plants besides the developed PWR NPP. The conclusion is that the NPP can respond to a disturbance in the power system without exceeding any design and safety limits if appropriate operational conditions, such as achieving the NPP turbine control by adjusting the speed of the steam valve, are met. In other words, the NPP can participate in the control of system frequency and improve the overall power system performance.

Evaluation of Non-Oxide Fuel for Fission-based Nuclear Reactors on Spacecraft

DTIC Science & Technology

smaller and potentially lighter core, whichis a significant advantage. The results of this study indicate that use of both UC and UN may result in significant weight savings due tohigher uranium loading density....The goal of this project was to study the performance of atypical uranium-based fuels in a nuclear reactor capable of producing 1 megawattof thermal...UN), or uranium carbide (UC) and compared their performance to uranium oxide (UO2) which is thefuel form used in the vast majority of commercial

Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

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

Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov

One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example.more » Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the size constraints of the MITR water loop are described. The safety case for operation of the high pressure gas-driven bellows mechanism is also presented. Key issues are the design and response of systems to limit gas flow in the event of a high pressure gas leak in the in-core autoclave. Integrity of the autoclave must be maintained and reactivity effects due to voiding of the loop coolant must be shown to be within the reactor technical specifications. The technical development of the crack growth monitor for application in the INL Advanced Test Reactor or the MITR can act as a template for adaptation of this technology in other reactors. (authors)« less

Design principles of a simple and safe 200-MW(thermal) nuclear district heating plant

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

Goetzmann, C.; Bittermann, D.; Gobel, A.

Kraftwerk Union AG has almost completed the development of a dedicated 200-MW(thermal) nuclear district heating plant to provide environmentally clean energy at a predictably low cost. The concept can easily be adapted to meet power requirements within the 100- to 500-MW(thermal) range. This technology is the product of the experience gained with large pressurized water reactor and boiling water reactor power plants, with respect to both plant and fuel performance. The major development task is that of achieving sufficiently low capital cost by tailoring components and systems designed for large plants to the specific requirements of district heating. These requirementsmore » are small absolute power, low temperatures and pressures, and modest load following, all of which result in the characteristics that are summarized. A fully integrated primary system with natural circulation permits a very compact reactor building containing all safety-related systems and components. Plant safety is essentially guaranteed by inherent features. The reactor containment is tightly fitted around the reactor pressure vessel in such a way that, in the event of any postulated coolant leak, the core cannot become uncovered, even temporarily. Shutdown is assured by gravity drop of the control rods mounted above the core. Decay heat is removed from the core by means of natural circulation via dedicated intermediate circuits of external aircoolers.« less

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Dickson, J.J.

1963-09-24

A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

JPRS Report Science and Technology, Japan: Atomic Energy Society 1989 Annual Meeting.

DTIC Science & Technology

1989-10-13

Control Rod Hole in VHTRC-1 Core [F, Akino, T, Yamane, et al.] ,,, 5 Measurement of MEU [Medium Enriched Uranium ] Fuel Element Characteristics in...K. Yoshida, K. Kobayashi, I. Kimura , C. Yamanaka, and S. Nakai, Laser Laboratory,, Osaka University. Nuclear Reactor Laboratory, Kyoto University...1 core loaded with 278 fuel rods (4 percent enriched uranium ). The PNS target was placed at the back center of the 1/2 assembly on the fixed side

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

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

A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis

2015-04-01

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Currentmore » Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.« less

Deterministic Modeling of the High Temperature Test Reactor

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

Ortensi, J.; Cogliati, J. J.; Pope, M. A.

2010-06-01

Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is usedmore » in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.« less

PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

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

Smith, P.; Sheetz, S.

Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply. The primary module in the model compares the consumption of uranium for various production reactors against the base case ofmore » Watts Bar I running a nominal load of 1,696 tritium producing burnable absorber rods (TPBARs) with an average refueling of 41,000 kg low enriched uranium (LEU) on an 18 month cycle. After inputting an initial year, starting inventory of unobligated uranium and tritium production forecast, the model will compare and contrast the depletion rate of the LEU between the entered alternatives. This is an annual tritium production rate of approximately 0.059 grams of tritium per kilogram of LEU (g-T/kg-LEU). To date, the Nuclear Regulatory Commission (NRC) license has not been amended to accept a full load of TPBARs so the nominal tritium production has not yet been achieved. The alternatives currently loaded into the model include the three light water SMRs evaluated in TAMU senior projects including, mPower, Holtec and NuScale designs. Initial evaluations of tritium production in light water reactor (LWR) based SMRs using optimized loads TPBARs is on the order 0.02-0.06 grams of tritium per kilogram of LEU used. The TAMU students also chose to model tritium production in the GE-Hitachi SPRISM, a pooltype sodium fast reactor (SFR) utilizing a modified TPBAR type target. The team was unable to complete their project so no data is available. In order to include results from a fast reactor, the SRNL Technical Advisory Committee (TAC) ran a Monte Carlo N-Particle (MCNP) model of a basic SFR for comparison. A 600MWth core surrounded by a lithium blanket produced approximately 1,000 grams of tritium annually with a 13% enriched, 6 year core. This is similar results to a mid-1990’s study where the Fast Flux Test Facility (FFTF), a 400 MWth reactor at the Idaho National Laboratory (INL), could produce about 1,000 grams with an external lithium target. Normalized to the LWRs values, comparative tritium production for an SFR could be approximately 0.31 g-T/kg LEU.« less

Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

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

Bardet, Philippe; Ricciardi, Guillaume

2016-01-31

Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWRmore » fuel bundle behavior during seismic transients.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

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

John D. Bess; Leland M. Montierth

2013-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters.more » One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering effects during pebble loading. Core 4 was determined to be acceptable benchmark experiment.« less

HTR-proteus pebble bed experimental program core 4: random packing with a 1:1 moderator-to-fuel pebble ratio

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

Bess, John D.; Montierth, Leland M.; Sterbentz, James W.

2014-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters.more » One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering effects during pebble loading. Core 4 was determined to be acceptable benchmark experiment.« less

Deterministic Modeling of the High Temperature Test Reactor with DRAGON-HEXPEDITE

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

J. Ortensi; M.A. Pope; R.M. Ferrer

2010-10-01

The Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine the INL’s current prismatic reactor analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 fuel column thin annular core, and the fully loaded core critical condition with 30 fuel columns. Special emphasis is devoted to physical phenomena and artifacts in HTTR that are similar to phenomena and artifacts in themore » NGNP base design. The DRAGON code is used in this study since it offers significant ease and versatility in modeling prismatic designs. DRAGON can generate transport solutions via Collision Probability (CP), Method of Characteristics (MOC) and Discrete Ordinates (Sn). A fine group cross-section library based on the SHEM 281 energy structure is used in the DRAGON calculations. The results from this study show reasonable agreement in the calculation of the core multiplication factor with the MC methods, but a consistent bias of 2–3% with the experimental values is obtained. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement partially stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.« less

ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

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

Lell, R. M.; Schaefer, R. W.; McKnight, R. D.

Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactormore » physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of demonstration-size LMFBRs. As a benchmark, ZPR-6/7 was devoid of many 'real' reactor features, such as simulated control rods and multiple enrichment zones, in its reference form. Those kinds of features were investigated experimentally in variants of the reference ZPR-6/7 or in other critical assemblies in the Demonstration Reactor Benchmark Program.« less

ZPR-3 Assembly 11 : A cylindrical sssembly of highly enriched uranium and depleted uranium with an average {sup 235}U enrichment of 12 atom % and a depleted uranium reflector.

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

Lell, R. M.; McKnight, R. D.; Tsiboulia, A.

2010-09-30

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physicsmore » benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specificationsa and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 11 began in early January 1958, and the Assembly 11 program ended in late January 1958. The core consisted of highly enriched uranium (HEU) plates and depleted uranium plates loaded into stainless steel drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, six columns of 0.125 in.-wide (3.175 mm) depleted uranium plates and one column of 1.0 in.-wide (25.4 mm) depleted uranium plates. The length of each column was 10 in. (254.0 mm) in each half of the core. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the depleted uranium radial blanket was approximately 14 in. (355.6 mm), and the length of the radial blanket in each half of the matrix was 22 in. (558.8 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/11, the reference critical configuration was loading 10 which was critical on January 21, 1958. Subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/11 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/11 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertainty far beyond that of the original experiment. Such a transformation is described in Section 3. It was obtained using a pair of continuous-energy Monte Carlo calculations. First, the critical configuration was modeled in full detail - every plate, drawer, matrix tube, and air gap was modeled explicitly. Then the regionwise compositions and volumes from the detailed as-built model were used to construct a homogeneous, two-dimensional (RZ) model of ZPR-3/11 that conserved the mass of each nuclide and volume of each region. The simple cylindrical model is the criticality-safety benchmark model. The difference in the calculated k{sub eff} values between the as-built three-dimensional model and the homogeneous two-dimensional benchmark model was used to adjust the measured excess reactivity of ZPR-3/11 loading 10 to obtain the k{sub eff} for the benchmark model.« less

Generating unstructured nuclear reactor core meshes in parallel

DOE PAGES

Jain, Rajeev; Tautges, Timothy J.

2014-10-24

Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

78 FR 56174 - In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-12

...-core thermocouples at different elevations and radial positions throughout the reactor core to enable... different elevations and radial positions throughout the reactor core to enable NPP operators to accurately... NPPs with in-core thermocouples at different elevations and radial positions throughout the reactor...

Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

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

Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.

2013-07-01

REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

Operation of the WWR-S Reactor in Poland and in the Period 1961-1962; DOKLAD OB ISPOL'ZOVANII I EKSPLUATATSII REAKTORA TIPA VVR-C V POL'SHE ZA 1961-1962 GODA

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

Aleksandrowicz, J.

1963-03-01

The experimental equipment used in the work at the horizontal reactor channels is listed. Diagrams of the utilization of the nominal reactor power and core loading are given, the reactivity fractions of the separate fuel assemblies are detonated, together with the diagram of reactivity versus burnup. Reactor channels and space used for sample irradiation and isotope production are described, and the total number of irradiations is given. Results of the measurements connected with the routine reactor operation are quoted, namely: analysis of water purity in the primary circuit, analysis of the work of the ion exchanger and mechanical filter, andmore » analysis of air activity in the special ventilation system. Data are given concerning radiation protection of personnel, including individual monitoring. Leak testing of the fuel elements is discussed. Damage of the reactor equipment and appearance of alarm signals are described. (auth)« less

MYRRHA: A multipurpose nuclear research facility

NASA Astrophysics Data System (ADS)

Baeten, P.; Schyns, M.; Fernandez, Rafaël; De Bruyn, Didier; Van den Eynde, Gert

2014-12-01

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback. As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

The effects of temperatures on the pebble flow in a pebble bed high temperature reactor

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

Sen, R. S.; Cogliati, J. J.; Gougar, H. D.

2012-07-01

The core of a pebble bed high temperature reactor (PBHTR) moves during operation, a feature which leads to better fuel economy (online refueling with no burnable poisons) and lower fuel stress. The pebbles are loaded at the top and trickle to the bottom of the core after which the burnup of each is measured. The pebbles that are not fully burned are recirculated through the core until the target burnup is achieved. The flow pattern of the pebbles through the core is of importance for core simulations because it couples the burnup distribution to the core temperature and power profiles,more » especially in cores with two or more radial burnup 'zones '. The pebble velocity profile is a strong function of the core geometry and the friction between the pebbles and the surrounding structures (other pebbles or graphite reflector blocks). The friction coefficient for graphite in a helium environment is inversely related to the temperature. The Thorium High Temperature Reactor (THTR) operated in Germany between 1983 and 1989. It featured a two-zone core, an inner core (IC) and outer core (OC), with different fuel mixtures loaded in each zone. The rate at which the IC was refueled relative to the OC in THTR was designed to be 0.56. During its operation, however, this ratio was measured to be 0.76, suggesting the pebbles in the inner core traveled faster than expected. It has been postulated that the positive feedback effect between inner core temperature, burnup, and pebble flow was underestimated in THTR. Because of the power shape, the center of the core in a typical cylindrical PBHTR operates at a higher temperature than the region next to the side reflector. The friction between pebbles in the IC is lower than that in the OC, perhaps causing a higher relative flow rate and lower average burnup, which in turn yield a higher local power density. Furthermore, the pebbles in the center region have higher velocities than the pebbles next to the side reflector due to the interaction between the pebbles and the immobile graphite reflector as well as the geometry of the discharge conus near the bottom of the core. In this paper, the coupling between the temperature profile and the pebble flow dynamics was analyzed by using PEBBED/THERMIX and PEBBLES codes by modeling the HTR-10 reactor in China. Two extreme and opposing velocity profiles are used as a starting point for the iterations. The PEBBED/THERMIX code is used to calculate the burnup, power and temperature profiles with one of the velocity profiles as input. The resulting temperature profile is then passed to PEBBLES code to calculate the updated pebble velocity profile taking the new temperature profile into account. If the aforementioned hypothesis is correct, the strong temperature effect upon the friction coefficients would cause the two cases to converge to different final velocity and temperature profiles. The results of this analysis indicates that a single zone pebble bed core is self-stabilizing in terms of the pebble velocity profile and the effect of the temperature profile on the pebble flow is insignificant. (authors)« less

Heat deposition analysis for the High Flux Isotope Reactor’s HEU and LEU core models

DOE PAGES

Davidson, Eva E.; Betzler, Benjamin R.; Chandler, David; ...

2017-08-01

The High Flux Isotope Reactor at Oak Ridge National Laboratory is an 85 MW th pressurized light-water-cooled and -moderated flux-trap type research reactor. The reactor is used to conduct numerous experiments, advancing various scientific and engineering disciplines. As part of an ongoing program sponsored by the US Department of Energy National Nuclear Security Administration Office of Material Management and Minimization, studies are being performed to assess the feasibility of converting the reactor’s highly enriched uranium fuel to low-enriched uranium fuel. To support this conversion project, reference models with representative experiment target loading and explicit fuel plate representation were developed andmore » benchmarked for both fuels to (1) allow for consistent comparison between designs for both fuel types and (2) assess the potential impact of low-enriched uranium conversion. These high-fidelity models were used to conduct heat deposition analyses at the beginning and end of the reactor cycle and are presented herein. This article (1) discusses the High Flux Isotope Reactor models developed to facilitate detailed heat deposition analyses of the reactor’s highly enriched and low-enriched uranium cores, (2) examines the computational approach for performing heat deposition analysis, which includes a discussion on the methodology for calculating the amount of energy released per fission, heating rates, power and volumetric heating rates, and (3) provides results calculated throughout various regions of the highly enriched and low-enriched uranium core at the beginning and end of the reactor cycle. These are the first detailed high-fidelity heat deposition analyses for the High Flux Isotope Reactor’s highly enriched and low-enriched core models with explicit fuel plate representation. Lastly, these analyses are used to compare heat distributions obtained for both fuel designs at the beginning and end of the reactor cycle, and they are essential for enabling comprehensive thermal hydraulics and safety analyses that require detailed estimates of the heat source within all of the reactor’s fuel element regions.« less

Boiling-Water Reactor internals aging degradation study. Phase 1

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

Luk, K.H.

1993-09-01

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor drymore » tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.« less

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

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

Michael A. Pope

2011-10-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

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

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh

2010-09-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

Three dimensional contact/impact methodology

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

Kulak, R.F.

1987-01-01

The simulation of three-dimensional interface mechanics between reactor components and structures during static contact or dynamic impact is necessary to realistically evaluate their structural integrity to off-normal loads. In our studies of postulated core energy release events, we have found that significant structure-structure interactions occur in some reactor vessel head closure designs and that fluid-structure interactions occur within the reactor vessel. Other examples in which three-dimensional interface mechanics play an important role are: (1) impact response of shipping casks containing spent fuel, (2) whipping pipe impact on reinforced concrete panels or pipe-to-pipe impact after a pipe break, (3) aircraft crashmore » on secondary containment structures, (4) missiles generated by turbine failures or tornados, and (5) drops of heavy components due to lifting accidents. The above is a partial list of reactor safety problems that require adequate treatment of interface mechanics and are discussed in this paper.« less

Experimental results from the VENUS-F critical reference state for the GUINEVERE accelerator driven system project

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

Uyttenhove, W.; Baeten, P.; Ban, G.

The GUINEVERE (Generation of Uninterrupted Intense Neutron pulses at the lead Venus Reactor) project was launched in 2006 within the framework of FP6 EUROTRANS in order to validate on-line reactivity monitoring and subcriticality level determination in Accelerator Driven Systems. Therefore the VENUS reactor at SCK.CEN in Mol (Belgium) was modified towards a fast core (VENUS-F) and coupled to the GENEPI-3C accelerator built by CNRS The accelerator can operate in both continuous and pulsed mode. The VENUS-F core is loaded with enriched Uranium and reflected with solid lead. A well-chosen critical reference state is indispensable for the validation of the on-linemore » subcriticality monitoring methodology. Moreover a benchmarking tool is required for nuclear data research and code validation. In this paper the design and the importance of the critical reference state for the GUINEVERE project are motivated. The results of the first experimental phase on the critical core are presented. The control rods worth is determined by the rod drop technique and the application of the Modified Source Multiplication (MSM) method allows the determination of the worth of the safety rods. The results are implemented in the VENUS-F core certificate for full exploitation of the critical core. (authors)« less

Experimental results from the VENUS-F critical reference state for the GUINEVERE accelerator driven system project

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

Uyttenhove, W.; Baeten, P.; Kochetkov, A.

The GUINEVERE (Generation of Uninterrupted Intense Neutron pulses at the lead Venus Reactor) project was launched in 2006 within the framework of FP6 EUROTRANS in order to validate online reactivity monitoring and subcriticality level determination in accelerator driven systems (ADS). Therefore, the VENUS reactor at SCK.CEN in Mol, Belgium, was modified towards a fast core (VENUS-F) and coupled to the GENEPI-3C accelerator built by CNRS. The accelerator can operate in both continuous and pulsed mode. The VENUS-F core is loaded with enriched Uranium and reflected with solid lead. A well-chosen critical reference state is indispensable for the validation of themore » online subcriticality monitoring methodology. Moreover, a benchmarking tool is required for nuclear data research and code validation. In this paper, the design and the importance of the critical reference state for the GUINEVERE project are motivated. The results of the first experimental phase on the critical core are presented. The control rods worth is determined by the positive period method and the application of the Modified Source Multiplication (MSM) method allows the determination of the worth of the safety rods. The results are implemented in the VENUS-F core certificate for full exploitation of the critical core. (authors)« less

Documentation for MeshKit - Reactor Geometry (&mesh) Generator

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

Jain, Rajeev; Mahadevan, Vijay

2015-09-30

This report gives documentation for using MeshKit’s Reactor Geometry (and mesh) Generator (RGG) GUI and also briefly documents other algorithms and tools available in MeshKit. RGG is a program designed to aid in modeling and meshing of complex/large hexagonal and rectilinear reactor cores. RGG uses Argonne’s SIGMA interfaces, Qt and VTK to produce an intuitive user interface. By integrating a 3D view of the reactor with the meshing tools and combining them into one user interface, RGG streamlines the task of preparing a simulation mesh and enables real-time feedback that reduces accidental scripting mistakes that could waste hours of meshing.more » RGG interfaces with MeshKit tools to consolidate the meshing process, meaning that going from model to mesh is as easy as a button click. This report is designed to explain RGG v 2.0 interface and provide users with the knowledge and skills to pilot RGG successfully. Brief documentation of MeshKit source code, tools and other algorithms available are also presented for developers to extend and add new algorithms to MeshKit. RGG tools work in serial and parallel and have been used to model complex reactor core models consisting of conical pins, load pads, several thousands of axially varying material properties of instrumentation pins and other interstices meshes.« less

The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

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

Morreale, A. C.; Ball, M. R.; Novog, D. R.

2012-07-01

The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxidemore » fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)« less

BWR Anticipated Transients Without Scram Leading to Instability

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

Cheng L. Y.; Baek J.; Cuadra, A.

2013-11-10

Anticipated transients without scram (ATWS) in aboiling water reactor (BWR) were simulated in order to understand reactor response and determine the effectiveness of automatic and operator actions to mitigate this beyond-design-basis accident. The events of interest herein are initiated by a turbine trip when the reactor is operating in the expanded operating domainMELLLA+ [maximum extended load line limit plus]. In these events the reactor may initially be at up to 120% of the original licensed thermal power (OLTP) and at flow rates as low as 80% of rated.For these (and similar) ATWS events the concern isthat when the reactor powermore » decreases in response to a dual recirculation pump trip, the core will become unstable and large amplitude oscillations will begin. The occurrence of these power oscillations, if left unmitigated, may result in fuel damage, and the amplitude of the poweroscillations may hamper the effectiveness of the injection of dissolved neutron absorber through the standby liquid control system (SLCS).« less

Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life Bhr Configurations: Designs, Advantages and Limitations

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

Dr. Pavel V. Tsvetkov

2009-05-20

This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologicmore » repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.« less

Dual-mode, high energy utilization system concept for mars missions

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.

2000-01-01

This paper describes a dual-mode, high energy utilization system concept based on the Pellet Bed Reactor (PeBR) to support future manned missions to Mars. The system uses proven Closed Brayton Cycle (CBC) engines to partially convert the reactor thermal power to electricity. The electric power generated is kept the same during the propulsion and the power modes, but the reactor thermal power in the former could be several times higher, while maintaining the reactor temperatures almost constant. During the propulsion mode, the electric power of the system, minus ~1-5 kWe for house keeping, is used to operate a Variable Specific Impulse Magnetoplasma Rocket (VASIMR). In addition, the reactor thermal power, plus more than 85% of the head load of the CBC engine radiators, are used to heat hydrogen. The hot hydrogen is mixed with the high temperature plasma in a VASIMR to provide both high thrust and Isp>35,000 N.s/kg, reducing the travel time to Mars to about 3 months. The electric power also supports surface exploration of Mars. The fuel temperature and the inlet temperatures of the He-Xe working fluid to the nuclear reactor core and the CBC turbine are maintained almost constant during both the propulsion and power modes to minimize thermal stresses. Also, the exit temperature of the He-Xe from the reactor core is kept at least 200 K below the maximum fuel design temperature. The present system has no single point failure and could be tested fully assembled in a ground facility using electric heaters in place of the nuclear reactor. Operation and design parameters of a 40-kWe prototype are presented and discussed to illustrate the operation and design principles of the proposed system. .

ADDITIONAL STRESS AND FRACTURE MECHANICS ANALYSES OF PRESSURIZED WATER REACTOR PRESSURE VESSEL NOZZLES

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

Walter, Matthew; Yin, Shengjun; Stevens, Gary

2012-01-01

In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperaturemore » (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B&PV Code, Section XI, Nonmandatory Appendix G.« less

Search for eV sterile neutrinos at a nuclear reactor — the Stereo project

NASA Astrophysics Data System (ADS)

Haser, J.; Stereo Collaboration

2016-05-01

The re-analyses of the reference spectra of reactor antineutrinos together with a revised neutrino interaction cross section enlarged the absolute normalization of the predicted neutrino flux. The tension between previous reactor measurements and the new prediction is significant at 2.7 σ and is known as “reactor antineutrino anomaly”. In combination with other anomalies encountered in neutrino oscillation measurements, this observation revived speculations about the existence of a sterile neutrino in the eV mass range. Mixing of this light sterile neutrino with the active flavours would lead to a modification of the detected antineutrino flux. An oscillation pattern in energy and space could be resolved by a detector at a distance of few meters from a reactor core: the neutrino detector of the Stereo project will be located at about 10 m distance from the ILL research reactor in Grenoble, France. Lengthwise separated in six target cells filled with 2 m3 Gd-loaded liquid scintillator in total, the experiment will search for a position-dependent distortion in the energy spectrum.

Initial Testing of the Stainless Steel NaK-Cooled Circuit (SNaKC)

NASA Technical Reports Server (NTRS)

Garber, Anne; Godfroy, Thomas

2007-01-01

An actively pumped alkali metal flow circuit, designed and fabricated at the NASA Marshall Space Flight Center, is currently undergoing testing in the Early Flight Fission Test Facility (EFF-TF). Sodium potassium (NaK) was selected as the primary coolant. Basic circuit components include: simulated reactor core, NaK to gas heat exchanger, electromagnetic liquid metal pump, liquid metal flowmeter, load/drain reservoir, expansion reservoir, test section, and instrumentation. Operation of the circuit is based around the 37-pin partial-array core (pin and flow path dimensions are the same as those in a full core), designed to operate at 33 kWt. This presentation addresses the construction, fill and initial testing of the Stainless Steel NaK-Cooled Circuit (SNaKC).

Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop

NASA Technical Reports Server (NTRS)

Clark, John S. (Editor)

1991-01-01

Papers presented at the joint NASA/DOE/DOD workshop on nuclear thermal propulsion are compiled. The following subject areas are covered: nuclear thermal propulsion programs; Rover/NERVA and NERVA systems; Low Pressure Nuclear Thermal Rocket (LPNTR); particle bed reactor nuclear rocket; hybrid propulsion systems; wire core reactor; pellet bed reactor; foil reactor; Droplet Core Nuclear Rocket (DCNR); open cycle gas core nuclear rockets; vapor core propulsion reactors; nuclear light bulb; Nuclear rocket using Indigenous Martian Fuel (NIMF); mission analysis; propulsion and reactor technology; development plans; and safety issues.

Lateral restraint assembly for reactor core

DOEpatents

Gorholt, Wilhelm; Luci, Raymond K.

1986-01-01

A restraint assembly for use in restraining lateral movement of a reactor core relative to a reactor vessel wherein a plurality of restraint assemblies are interposed between the reactor core and the reactor vessel in circumferentially spaced relation about the core. Each lateral restraint assembly includes a face plate urged against the outer periphery of the core by a plurality of compression springs which enable radial preloading of outer reflector blocks about the core and resist low-level lateral motion of the core. A fixed radial key member cooperates with each face plate in a manner enabling vertical movement of the face plate relative to the key member but restraining movement of the face plate transverse to the key member in a plane transverse to the center axis of the core. In this manner, the key members which have their axes transverse to or subtending acute angles with the direction of a high energy force tending to move the core laterally relative to the reactor vessel restrain such lateral movement.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, C.W.

1985-02-19

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, Charles W.

1987-01-01

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Tornado wind-loading requirements based on risk assessment techniques

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

Deobald, T.L.; Coles, G.A.; Smith, G.L.

Regulations require that nuclear power plants be protected from tornado winds. If struck by a tornado, a plant must be capable of safely shutting down and removing decay heat. Probabilistic techniques are used to show that risk to the public from the US Department of Energy (DOE) SP-100 reactor is acceptable without tornado hardening parts of the secondary system. Relaxed requirements for design wind loadings will result in significant cost savings. To demonstrate an acceptable level of risk, this document examines tornado-initiated accidents. The two tornado-initiated accidents examined in detail are loss of cooling resulting in core damage and lossmore » of secondary system boundary integrity leading to sodium release. Loss of core cooling is analyzed using fault/event tree models. Loss of secondary system boundary integrity is analyzed by comparing the consequences to acceptance criteria for the release of radioactive material or alkali metal aerosol. 4 refs., 4 figs.« less

Systems and methods for dismantling a nuclear reactor

DOEpatents

Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

2014-10-28

Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

NASA Astrophysics Data System (ADS)

Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

2016-02-01

The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

Applications of plasma core reactors to terrestrial energy systems

NASA Technical Reports Server (NTRS)

Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

1974-01-01

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

Pressurized-water reactor internals aging degradation study. Phase 1

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

Luk, K.H.

1993-09-01

This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pinsmore » and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.« less

A Review of Gas-Cooled Reactor Concepts for SDI Applications

DTIC Science & Technology

1989-08-01

710 program .) Wire- Core Reactor (proposed by Rockwell). The wire- core reactor utilizes thin fuel wires woven between spacer wires to form an open...reactor is based on results of developmental studies of nuclear rocket propulsion systems. The reactor core is made up of annular fuel assemblies of...XE Addendum to Volume II. NERVA Fuel Development , Westinghouse Astronuclear Laboratory, TNR-230, July 15’ 1972. J I8- Rover Program Reactor Tests

Station blackout calculations for Browns Ferry

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

Ott, L.J.; Weber, C.F.; Hyman, C.R.

1985-01-01

This paper presents the results of calculations performed with the ORNL SASA code suite for the Station Blackout Severe Accident Sequence at Browns Ferry. The accident is initiated by a loss of offsite power combined with failure of all onsite emergency diesel generators to start and load. The Station Blackout is assumed to persist beyond the point of battery exhaustion (at six hours) and without DC power, cooling water could no longer be injected into the reactor vessel. Calculations are continued through the period of core degradation and melting, reactor vessel failure, and the subsequent containment failure. An estimate ofmore » the magnitude and timing of the concomitant fission product releases is also provided.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

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

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

In-situ verification techniques for fast critical assembly cores

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

Brumbach, S.B.; Amundson, P.I.; Roche, C.T.

1979-01-01

Active and passive autoradiographic techniques were used to obtain piece counts of fuel plates in fast critical assembly drawers and to verify the assembly loading pattern. Active autoradiography using prompt-fission and fission-product radiation was more successful with uranium fuel while passive autoradiography was more successful with plutonium fuel. A source multiplication technique was used to measure changes in reactivity when small quantities (2-2.5 kg) of fissile material were removed from a subcritical reference core of the Zero Power Plutonium Reactor. Efforts to compensate for unsuccessful. Some compensation was achieved by replacing U-238 with polyethylene. The sensitivity for detection of partiallymore » compensated fuel removed from minimum worth regions was approximately 2.5 kg (fissile) for a core containing 2600 kg (fissile). Substitution of polyethylene was detected with a spectral index which was the ratio of the rate of the In-115 (n,..gamma..) reaction to the rate of the In-115 (n,n') reaction. This spectral index was sensitive to the presence of an 0.64-cm-thick, 5.08-cm-high polyethylene column 10-15 cm away from the indium foil. The reactivity worth of Pu-239 was also obtained as a function of location in the reactor core with the use of an inverse kinetics technique. Reactivity worths for Pu-239 varied from a maximum of 58.67 Ih/kg near the core center to a minimum of 14.86 Ih/kg at the core edge.« less

Fabrication of TREAT Fuel with Increased Graphite Loading

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

Luther, Erik Paul; Leckie, Rafael M.; Dombrowski, David E.

2014-02-05

As part of the feasibility study exploring the replacement of the HEU fuel core of the TREAT reactor at Idaho National Laboratory with LEU fuel, this study demonstrates that it is possible to increase the graphite content of extruded fuel by reformulation. The extrusion process was use to fabricate the “upgrade” core1 for the TREAT reactor. The graphite content achieved is determined by calculation and has not been measured by any analytical method. In conjunction, a technique, Raman Spectroscopy, has been investigated for measuring the graphite content. This method shows some promise in differentiating between carbon and graphite; however, standardsmore » that would allow the technique to be calibrated to quantify the graphite concentration have yet to be fabricated. Continued research into Raman Spectroscopy is on going. As part of this study, cracking of graphite extrusions due to volatile evolution during heat treatment has been largely eliminated. Continued research to optimize this extrusion method is required.« less

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

Fast-acting nuclear reactor control device

DOEpatents

Kotlyar, Oleg M.; West, Phillip B.

1993-01-01

A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

Optimization of a Boiling Water Reactor Loading Pattern Using an Improved Genetic Algorithm

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

Kobayashi, Yoko; Aiyoshi, Eitaro

2003-08-15

A search method based on genetic algorithms (GA) using deterministic operators has been developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). The search method uses an Improved GA operator, that is, crossover, mutation, and selection. The handling of the encoding technique and constraint conditions is designed so that the GA reflects the peculiar characteristics of the BWR. In addition, some strategies such as elitism and self-reproduction are effectively used to improve the search speed. LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and three-dimensional-dependent constraints have alwaysmore » necessitated the use of three-dimensional core simulators for BWRs, so that an optimization method is required for computational efficiency. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant applying the Haling technique. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.« less

Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

DOEpatents

Peterson, Per F.

2013-05-14

A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

Final Report for the “WSU Neutron Capture Therapy Facility Support”

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

Gerald E. Tripard; Keith G. Fox

2006-08-24

The objective for the cooperative research program for which this report has been written was to provide separate NCT facility user support for the students, faculty and scientists who would be doing the U.S. Department of Energy Office (DOE) of Science supported advanced radiotargeted research at the WSU 1 megawatt TRIGA reactor. The participants were the Idaho National laboratory (INL, P.I., Dave Nigg), the Veterinary Medical Research Center of Washington State University (WSU, Janean Fidel and Patrick Gavin), and the Washington State University Nuclear Radiation Center (WSU, P.I., Gerald Tripard). A significant number of DOE supported modifications were made tomore » the WSU reactor in order to create an epithermal neutron beam while at the same time maintaining the other activities of the 1 MW reactor. These modifications were: (1) Removal of the old thermal column. (2) Construction and insertion of a new epithermal filter, collimator and shield. (3) Construction of a shielded room that could accommodate the very high radiation field created by an intense neutron beam. (4) Removal of the previous reactor core fuel cluster arrangement. (5) Design and loading of the new reactor core fuel cluster arrangement in order to optimize the neutron flux entering the epithermal neutron filter. (6) The integration of the shielded rooms interlocks and radiological controls into the SCRAM chain and operating electronics of the reactor. (7) Construction of a motorized mechanism for moving and remotely controlling the position of the entire reactor bridge. (8) The integration of the reactor bridge control electronics into the SCRAM chain and operating electronics of the reactor. (9) The design, construction and attachment to the support structure of the reactor of an irradiation box that could be inserted into position next to the face of the reactor. (Necessitated by the previously mentioned core rearrangement). All of the above modifications were successfully completed and tested. The resulting epithermal beam of 1 x 10{sup 9} n/sec-cm{sup 2} was measured by Idaho National Laboratory with assistance from WSU's Neutron Activation Analysis Group. The beam is as good as our initial proposals for the project had predicted. In addition to all of the design, construction and insertion of the hardware, shielding, electronics and radiation monitoring systems there was considerable manpower and effort put into changes in the Technical Specifications of the reactor and implementing procedures for use of the new facility. This staff involvement is one of the reasons we requested special facility support from the DOE. Once the facility was competed and all of the recalibrations and measurements made to characterize the differences between this reactor core and the previous core we began to assist INL in making their beam measurements with foils and phantoms. Although we proposed support for only one additional staff position to support this new NCT facility the staff support provided by the WSU Nuclear Radiation Center was greater than had been anticipated by our initial proposal. INL was also assisted in the testing of a heavy water (deuterated water) bladder that can be inserted into the collimator in order to produce an intense, external thermal neutron beam. The external epithermal and/or thermal neutron beam capability remains available for use, if funding becomes available for future research projects.« less

78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

A review of carbide fuel corrosion for nuclear thermal propulsion applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1993-10-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1994-07-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

Large-scale boiling experiments of the flooded cavity concept for in-vessel core retention

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

Chu, T.Y.; Slezak, S.E.; Bentz, J.H.

1994-03-01

This paper presents results of ex-vessel boiling experiments performed in the CYBL (CYlindrical BoiLing) facility. CYBL is a reactor-scale facility for confirmatory research of the flooded cavity concept for accident management. CYBL has a tank-within-a-tank design; the inner tank simulates the reactor vessel and the outer tank simulates the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm{sup 2} across the vessel bottom were performed. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides of the reactormore » vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid/solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion (ejection). The results suggest that under prototypic heat load and heat flux distributions, the flooded cavity in a passive pressurized water reactor like the AP-600 should be capable of cooling the reactor pressure vessel in the central region of the lower head that is addressed by these tests.« less

Sensitivity of the DANSS detector to short range neutrino oscillations

NASA Astrophysics Data System (ADS)

Danilov, Mikhail; DANSS Collaboration

2016-04-01

DANSS is a highly segmented 1 m3 plastic scintillator detector. Its 2500 scintillator strips have a Gd loaded reflective cover. Light is collected with 3 wave length shifting fibers per strip and read out with 50 PMTs and 2500 SiPMs. The DANSS will be installed under the industrial 3 GWth reactor of the Kalinin Nuclear Power Plant at distances varying from 9.7 m to 12.2 m from the reactor core. PMTs and SiPMs collect about 30 photo electrons per MeV distributed approximately equally between two types of the readout. Light collection non-uniformity across and along the strip is about ±13% from maximum to minimum. The resulting energy resolution is modest, σ / E = 15% at 5 MeV. This leads to a smearing of the oscillation pattern comparable with the smearing due to the large size of the reactor core. Nevertheless because of the large counting rate (˜10000/day), small background (< 1%) and good control of systematic uncertainties due to frequent changes of positions, the DANSS is quite sensitive to reactor antineutrino oscillations to hypothetical sterile neutrinos with a mass in eV ballpark suggested recently to explain a so-called reactor anomaly. DANSS will have an elaborated calibration system. The high granularity of the detector allows calibration of every strip with about 40 thousand cosmic muons every day. The expected systematic effects do not reduce much the sensitivity region. Tests of the detector prototype DANSSino demonstrated that in spite of a small size (4% of DANSS), it is quite sensitive to reactor antineutrinos, detecting about 70 Inverse Beta Decay events per day with the signal-to-background ratio of about unity. The prototype tests have demonstrated feasibility to reach the design performance of the DANSS detector.

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

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

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki; Miura, Ryosuke

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design.more » The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.« less

Emergency core cooling system

DOEpatents

Schenewerk, William E.; Glasgow, Lyle E.

1983-01-01

A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.

SPERT Destructive Test - I on Aluminum, Highly Enriched Plate Type Core

ScienceCinema

None

2018-01-16

SPERT - Special Power Excursion Reactor Tests Destructive Test number 1 On Aluminum, Highly Enriched Plate Type Core. A test studying the behavior of the reactor under destructive conditions on a light water moderated pool-type reactor with a plate-type core.

Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

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

Tarchalski, M.; Pytel, K.; Wroblewska, M.

2015-07-01

Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to themore » qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from measurements is of the order of 2.5 W/g at half of the possible MARIA power - 15 MW. The approach and the detailed program for experimental verification of calculations will be presented. The following points will be discussed: - Development of a gamma heating model of MARIA reactor with TRIPOLI 4 (coupled neutron-photon mode) and APOLLO2 model taking into account the key parameters like: configuration of the core, experimental loading, control rod location, reactor power, fuel depletion); - Design of specific measurement tools for MARIA experiments including for instance a new single-cell calorimeter called KAROLINA calorimeter; - MARIA experimental program description and a preliminary analysis of results; - Comparison of calculations for JHR and MARIA cores with experimental verification analysis, calculation behavior and n-γ 'environments'. (authors)« less

Analysis of Radionuclide Releases from the Fukushima Dai-Ichi Nuclear Power Plant Accident Part I

NASA Astrophysics Data System (ADS)

Le Petit, G.; Douysset, G.; Ducros, G.; Gross, P.; Achim, P.; Monfort, M.; Raymond, P.; Pontillon, Y.; Jutier, C.; Blanchard, X.; Taffary, T.; Moulin, C.

2014-03-01

Part I of this publication deals with the analysis of fission product releases consecutive to the Fukushima Dai-ichi accident. Reactor core damages are assessed relying on radionuclide detections performed by the CTBTO radionuclide network, especially at the particulate station located at Takasaki, 210 km away from the nuclear power plant. On the basis of a comparison between the reactor core inventory at the time of reactor shutdowns and the fission product activities measured in air at Takasaki, especially 95Nb and 103Ru, it was possible to show that the reactor cores were exposed to high temperature for a prolonged time. This diagnosis was confirmed by the presence of 113Sn in air at Takasaki. The 133Xe assessed release at the time of reactor shutdown (8 × 1018 Bq) turned out to be in the order of 80 % of the amount deduced from the reactor core inventories. This strongly suggests a broad meltdown of reactor cores.

Evaluation of HFIR LEU Fuel Using the COMSOL Multiphysics Platform

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

Primm, Trent; Ruggles, Arthur; Freels, James D

2009-03-01

A finite element computational approach to simulation of the High Flux Isotope Reactor (HFIR) Core Thermal-Fluid behavior is developed. These models were developed to facilitate design of a low enriched core for the HFIR, which will have different axial and radial flux profiles from the current HEU core and thus will require fuel and poison load optimization. This report outlines a stepwise implementation of this modeling approach using the commercial finite element code, COMSOL, with initial assessment of fuel, poison and clad conduction modeling capability, followed by assessment of mating of the fuel conduction models to a one dimensional fluidmore » model typical of legacy simulation techniques for the HFIR core. The model is then extended to fully couple 2-dimensional conduction in the fuel to a 2-dimensional thermo-fluid model of the coolant for a HFIR core cooling sub-channel with additional assessment of simulation outcomes. Finally, 3-dimensional simulations of a fuel plate and cooling channel are presented.« less

Fission-powered in-core thermoacoustic sensor

DOE PAGES

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.; ...

2016-04-07

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. Furthermore, these signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

Fission-powered in-core thermoacoustic sensor

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

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.

2016-04-04

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

A SURVEY OF CONVENTIONAL STEAM BOILER EXPERIENCE APPLICABLE TO THE HTGR STEAM GENERATORS

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

Paget, J.A.

1959-10-01

BS>The steam generator of a high temperature gas-cooled reactor consists of tubular heating surface inside a shell which forms part of the primary He circuit of the reactor. When a tube fails in such a steam generator, moisture in the form of steam is released into the He steam and is carried through the reactor where it will cause corrosion and mass transfer of C in the core. A paramount consideration in the design of a steam generator for a high temperature gas-cooled reactor is the prevention of tube failures. Preference, therefore, should be given to a forced circulation design.more » The Loeffler Boiler would be the best from this standpoint alone since only steam enters the tubes, and its circulation rate can be maintained at an adequate value to insure cool tubes regardless of load fluctuations. The next type in the order of preference would be the forced recirculation boiler, since at least the boiier tubes always have an adequate cooling flow regardless of output. The third type in order of preference would be a Sulzer Type boiler since it has a separator to remove dissolved material from the water which is comparible in efficiency to a standard boiler drum and although the flow through evaporator and superheater fluctuates with load, the Sulzer Boiler can be operated as a forced recirculation boiler at low loads. The least desirable type would be a Benson or supercritical boiler which is completely dependent on input water purity for its survival. It is not claimed that Benson or supercritical boilers should not or will not be used in the future for gas-cooled reactors, but only that their use would be the least conservative choice from a tube failure standpoint at the present time. (auth)« less

NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1958-04-22

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

MSFR TRU-burning potential and comparison with an SFR

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

Fiorina, C.; Cammi, A.; Franceschini, F.

2013-07-01

The objective of this work is to evaluate the Molten Salt Fast Reactor (MSFR) potential benefits in terms of transuranics (TRU) burning through a comparative analysis with a sodium-cooled FR. The comparison is based on TRU- and MA-burning rates, as well as on the in-core evolution of radiotoxicity and decay heat. Solubility issues limit the TRU-burning rate to 1/3 that achievable in traditional low-CR FRs (low-Conversion-Ratio Fast Reactors). The softer spectrum also determines notable radiotoxicity and decay heat of the equilibrium actinide inventory. On the other hand, the liquid fuel suggests the possibility of using a Pu-free feed composed onlymore » of Th and MA (Minor Actinides), thus maximizing the MA burning rate. This is generally not possible in traditional low-CR FRs due to safety deterioration and decay heat of reprocessed fuel. In addition, the high specific power and the lack of out-of-core cooling times foster a quick transition toward equilibrium, which improves the MSFR capability to burn an initial fissile loading, and makes the MSFR a promising system for a quick (i.e., in a reactor lifetime) transition from the current U-based fuel cycle to a novel closed Th cycle. (authors)« less

Nuclear reactor downcomer flow deflector

DOEpatents

Gilmore, Charles B [Greensburg, PA; Altman, David A [Pittsburgh, PA; Singleton, Norman R [Murrysville, PA

2011-02-15

A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

Spatial Burnout in Water Reactors with Nonuniform Startup Distributions of Uranium and Boron

NASA Technical Reports Server (NTRS)

Fox, Thomas A.; Bogart, Donald

1955-01-01

Spatial burnout calculations have been made of two types of water moderated cylindrical reactor using boron as a burnable poison to increase reactor life. Specific reactors studied were a version of the Submarine Advanced Reactor (sAR) and a supercritical water reactor (SCW) . Burnout characteristics such as reactivity excursion, neutron-flux and heat-generation distributions, and uranium and boron distributions have been determined for core lives corresponding to a burnup of approximately 7 kilograms of fully enriched uranium. All reactivity calculations have been based on the actual nonuniform distribution of absorbers existing during intervals of core life. Spatial burnout of uranium and boron and spatial build-up of fission products and equilibrium xenon have been- considered. Calculations were performed on the NACA nuclear reactor simulator using two-group diff'usion theory. The following reactor burnout characteristics have been demonstrated: 1. A significantly lower excursion in reactivity during core life may be obtained by nonuniform rather than uniform startup distribution of uranium. Results for SCW with uranium distributed to provide constant radial heat generation and a core life corresponding to a uranium burnup of 7 kilograms indicated a maximum excursion in reactivity of 2.5 percent. This compared to a maximum excursion of 4.2 percent obtained for the same core life when w'anium was uniformly distributed at startup. Boron was incorporated uniformly in these cores at startup. 2. It is possible to approach constant radial heat generation during the life of a cylindrical core by means of startup nonuniform radial and axial distributions of uranium and boron. Results for SCW with nonuniform radial distribution of uranium to provide constant radial heat generation at startup and with boron for longevity indicate relatively small departures from the initially constant radial heat generation distribution during core life. Results for SAR with a sinusoidal distribution rather than uniform axial distributions of boron indicate significant improvements in axial heat generation distribution during the greater part of core life. 3. Uranium investments for cylindrical reactors with nonuniform radial uranium distributions which provide constant radial heat generation per unit core volume are somewhat higher than for reactors with uniform uranium concentration at startup. On the other hand, uranium investments for reactors with axial boron distributions which approach constant axial heat generation are somewhat smaller than for reactors with uniform boron distributions at startup.

Magnetic latch trigger for inherent shutdown assembly

DOEpatents

Sowa, Edmund S.

1976-01-01

An inherent shutdown assembly for a nuclear reactor is provided. A neutron absorber is held ready to be inserted into the reactor core by a magnetic latch. The latch includes a magnet whose lines of force are linked by a yoke of material whose Curie point is at the critical temperature of the reactor at which the neutron absorber is to be inserted into the reactor core. The yoke is in contact with the core coolant or fissionable material so that when the coolant or the fissionable material increase in temperature above the Curie point the yoke loses its magnetic susceptibility and the magnetic link is broken, thereby causing the absorber to be released into the reactor core.

Design test request No. 1263 K Reactor graphite key and VSR channel sleeve test

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

Kempf, F.J.

1964-12-10

The objectives of this test were: (1) Determine the coefficient of friction between two adjacent layers of K Reactor graphite at room temperature. (2) Determine the average load required to cause failure of an unirradiated K Reactor side reflector bar, when subjected to tensile loading applied through the reflector keys. (3) Determine the average load at failure and the average deflection at failure of a single VSR channel key when loaded in keyways with clearances equal to those used in original stack construction. (4) Determine the average load and deflection required to break the four K Reactor VSR keys whenmore » loaded simultaneously in both `3-layer` and `7-layer` mockups. Also determine the mode of key failure; i.e., shear, flexure or combined compression and bending. Following these key rupture tests, determine the strength and deflection characteristics of the proposed K Reactor VSR channel sleeve when loaded in a manner identical to that used to fracture the keys. (5) Determine the average load and deflection at failure of both the proposed K Reactor VSR channel sleeves and the proposed C Reactor sleeves when subjected to crushing loads. (6) Determine the extent of damage to the proposed K Reactor VSR channel sleeve when subjected to the following vertical rod loading conditions. (a) Full rod drop in a channel mockup which has been misaligned 2 1/2 inches. (b) Full rod drop in a channel which has been misaligned an amount equal to the maximum flexibility of a `universal` VSR.« less

The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

NASA Astrophysics Data System (ADS)

Cole, Christopher J. P.

Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas that should be investigated. These include developing a detailed point nodel kinetic model coupled with a finite element heat transfer model, undertaking radiation protection shielding calculations in accordance with international and national regulations, and exploring the effects of advanced fuels.

Fission control system for nuclear reactor

DOEpatents

Conley, G.H.; Estes, G.P.

Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

Tory II-A: a nuclear ramjet test reactor

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

Hadley, J.W.

Declassified 28 Nov 1973. The first test reactor in the Pluto program, leading to development of a nuclear ramjet engine, is called Tory II-A. While it is not an actual prototype engine, this reactor embodies a core design which is considered feasible for an engine, and operation of the reactor will provide a test of that core type as well as more generalized values in reactor design and testing. The design of Tory II-A and construction of the reactor and of its test facility are described. Operation of the Tory II-A core at a total power of 160 megawatts, withmore » 800 pounds of air per second passing through the core and emerging at a temperature of 2000 deg F, is the central objective of the test program. All other reactor and facility components exist to support operation of the core, and preliminary steps in the test program itself will be directed primarily toward ensuring attalnment of full-power operation and collection of meaningful data on core behavior during that operation. The core, 3 feet in diameter and 41/2 feet long, will be composed of bundled ceramic tubes whose central holes will provide continuous air passages from end to end of the reactor. These tubes are to be composed of a homogeneous mixture of UO/sub 2/ fuel and BeO moderator, compacted and sintered to achieve high strength and density. (30 references) (auth)« less

Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

NASA Astrophysics Data System (ADS)

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

2018-01-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of neutron flux in the reactor core.

Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

NASA Technical Reports Server (NTRS)

Lantz, E.; Breitwieser, R.; Niederauer, G. F.

1974-01-01

There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

Neutron detection of the Triga Mark III reactor, using nuclear track methodology

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

Espinosa, G., E-mail: espinosa@fisica.unam.mx; Golzarri, J. I.; Raya-Arredondo, R.

Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 10{sup 12} n cm{sup −2} s{sup −1}, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer{sup ®} PADC as neutron detection material, covered by 3 mm Plexiglas{sup ®} as converter. After exposure, plasticmore » detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power.« less

Emergency heat removal system for a nuclear reactor

DOEpatents

Dunckel, Thomas L.

1976-01-01

A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

Capabilities and Testing of the Fission Surface Power Primary Test Circuit (FSP-PTC)

NASA Technical Reports Server (NTRS)

Garber, Anne E.

2007-01-01

An actively pumped alkali metal flow circuit, designed and fabricated at the NASA Marshall Space Flight Center, is currently undergoing testing in the Early Flight Fission Test Facility (EFF-TF). Sodium potassium (NaK), which was used in the SNAP-10A fission reactor, was selected as the primary coolant. Basic circuit components include: simulated reactor core, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, liquid metal flowmeter, load/drain reservoir, expansion reservoir, test section, and instrumentation. Operation of the circuit is based around a 37-pin partial-array core (pin and flow path dimensions are the same as those in a full core), designed to operate at 33 kWt. NaK flow rates of greater than 1 kg/sec may be achieved, depending upon the power applied to the EM pump. The heat exchanger provides for the removal of thermal energy from the circuit, simulating the presence of an energy conversion system. The presence of the test section increases the versatility of the circuit. A second liquid metal pump, an energy conversion system, and highly instrumented thermal simulators are all being considered for inclusion within the test section. This paper summarizes the capabilities and ongoing testing of the Fission Surface Power Primary Test Circuit (FSP-PTC).

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

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

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1961-09-01

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

Radial blanket assembly orificing arrangement

DOEpatents

Patterson, J.F.

1975-07-01

A nuclear reactor core for a liquid metal cooled fast breeder reactor is described in which means are provided for increasing the coolant flow through the reactor fuel assemblies as the reactor ages by varying the coolant flow rate with the changing coolant requirements during the core operating lifetime. (auth)

Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

Adaptive control method for core power control in TRIGA Mark II reactor

NASA Astrophysics Data System (ADS)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

NASA Technical Reports Server (NTRS)

Butler, C.; Albright, D.

2007-01-01

Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

Reactor core isolation cooling system

DOEpatents

Cooke, F.E.

1992-12-08

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

Reactor core isolation cooling system

DOEpatents

Cooke, Franklin E.

1992-01-01

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

DOEpatents

Tokarz, R.D.

1981-10-27

This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

Nuclear reactor apparatus

DOEpatents

Wade, Elman E.

1978-01-01

A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

CALANDRIA TYPE SODIUM GRAPHITE REACTOR

DOEpatents

Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

1964-02-11

A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

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

Jung, Y. S.; Joo, H. G.; Yoon, J. I.

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Nuclear reactor

DOEpatents

Wade, Elman E.

1979-01-01

A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

Automatic safety rod for reactors

DOEpatents

Germer, John H.

1988-01-01

An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Validation of the MCNP computational model for neutron flux distribution with the neutron activation analysis measurement

NASA Astrophysics Data System (ADS)

Tiyapun, K.; Chimtin, M.; Munsorn, S.; Somchit, S.

2015-05-01

The objective of this work is to demonstrate the method for validating the predication of the calculation methods for neutron flux distribution in the irradiation tubes of TRIGA research reactor (TRR-1/M1) using the MCNP computer code model. The reaction rate using in the experiment includes 27Al(n, α)24Na and 197Au(n, γ)198Au reactions. Aluminium (99.9 wt%) and gold (0.1 wt%) foils and the gold foils covered with cadmium were irradiated in 9 locations in the core referred to as CT, C8, C12, F3, F12, F22, F29, G5, and G33. The experimental results were compared to the calculations performed using MCNP which consisted of the detailed geometrical model of the reactor core. The results from the experimental and calculated normalized reaction rates in the reactor core are in good agreement for both reactions showing that the material and geometrical properties of the reactor core are modelled very well. The results indicated that the difference between the experimental measurements and the calculation of the reactor core using the MCNP geometrical model was below 10%. In conclusion the MCNP computational model which was used to calculate the neutron flux and reaction rate distribution in the reactor core can be used for others reactor core parameters including neutron spectra calculation, dose rate calculation, power peaking factors calculation and optimization of research reactor utilization in the future with the confidence in the accuracy and reliability of the calculation.

77 FR 30435 - In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-23

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [Docket No. PRM-50-105; NRC-2012-0056] In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission... of operating licenses for nuclear power plants (``NPP'') to operate NPPs with in-core thermocouples...

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

NASA Astrophysics Data System (ADS)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

Effect of reactor loading on atomic oxygen concentration as measured by NO chemiluminescence

NASA Technical Reports Server (NTRS)

Lerner, N. R.

1989-01-01

It has previously been observed that the etch rate of polyethylene samples in the afterglow of an RF discharge in oxygen increases with reactor loading. This enhancement of the etch rate is attributed to reactive gas phase products of the polymer etching. In the present work, emission spectroscopy is employed to examine the species present in the gas phase during etching of polyethylene. In particular, the concentration of atomic oxygen downstream from the polyethylene samples is studied as a function of the reactor loading. It is found that the concentration of atomic oxygen increases as the reactor loading is increased. The increase of etch rate with increased reactor loading is attributed to the increase of atomic oxygen concentration in the vicinity of the sample.

RADIATION FACILITY FOR NUCLEAR REACTORS

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1961-12-12

A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, M.L.; Rosenstein, R.G.

1998-10-13

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

Mox fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

2001-05-15

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

2001-07-17

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

1998-01-01

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

COOLED NEUTRONIC REACTOR

DOEpatents

Binner, C.R.; Wilkie, C.B.

1958-03-18

This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

NASA Technical Reports Server (NTRS)

Weinstein, H.; Lavan, Z.

1975-01-01

Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

Double-clad nuclear fuel safety rod

DOEpatents

McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

1984-01-01

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Double-clad nuclear-fuel safety rod

DOEpatents

McCarthy, W.H.; Atcheson, D.B.

1981-12-30

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

NASA Astrophysics Data System (ADS)

Alekseev, P. N.; Bobrov, E. A.; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A.

2015-12-01

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U-Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium-plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: 235U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or 233U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

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

Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.

2015-12-15

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no usemore » of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.« less

Current status of SPINNORs designs

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

Su'ud, Zaki

2010-06-22

This study discuss about the SPINNOR (Small Power Reactor, Indonesia, No On-site Refuelling) and the VSPINNOR (Very Small Power Reactor, Indonesia, No On-site Refuelling) which are small lead-bismuth cooled nuclear power reactors with fast neutron spectrum that could be operated for more than 10 or 15 years without on-site refuelling. They are based on the concept of a long-life core reactor developed in Indonesia since early 1990 in collaboration with the Research Laboratory for Nuclear Reactors of the Tokyo Institute of Technology (RLNR TITech). The reactor cores are designed to have near zero (less then one effective delayed neutron fraction)more » burn-up reactivity swing during the whole course of their operation to avoid a possibility of prompt criticality accident. The basic concept is that central region of the reactor core is filled with fertile (blanket) material. During the reactor operation fissile material accumulates in this central region, which helps to compensate fissile material loss in the peripheral core region and also contributes to negative coolant loss reactivity effect. A concept of high fuel volume fraction in the core is applied to achieve smaller size of a critical reactor. In this paper we consider to add Np-237 to the fuel to enhance non proliferation characteristics of the systems. The effect of Np-237 amount variation is discussed.« less

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

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

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

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

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

NASA Technical Reports Server (NTRS)

Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

1991-01-01

The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

DOE PAGES

Bess, John D.; Fujimoto, Nozomu

2014-10-09

Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in themore » experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

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

Dunn, Floyd E.; Hu, Lin-wen; Wilson, Erik

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings onmore » avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.« less

BIOLOGICAL IRRADIATION FACILITY

DOEpatents

McCorkle, W.H.; Cern, H.S.

1962-04-24

A facility for irradiating biological specimens with neutrons is described. It includes a reactor wherein the core is off center in a reflector. A high-exposure room is located outside the reactor on the side nearest the core while a low-exposure room is located on the opposite side. Means for converting thermal neutrons to fast neutrons are movably disposed between the reactor core and the high and low-exposure rooms. (AEC)

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

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

Not Available

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

Reactivity control assembly for nuclear reactor. [LMFBR

DOEpatents

Bollinger, L.R.

1982-03-17

This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

Cooling system for a nuclear reactor

DOEpatents

Amtmann, Hans H.

1982-01-01

A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

Development of a three-dimensional core dynamics analysis program for commercial boiling water reactors

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

Bessho, Yasunori; Yokomizo, Osamu; Yoshimoto, Yuichiro

1997-03-01

Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates calculational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory size as attained by the latest supercomputer technology. The program consists of six major modules, namely a core neutronics module, a fuel heat conduction/transfer module, a fuel channel thermal-hydraulic module, an upper plenum/separator module, a feedwater/recirculation flow module, and amore » control system module. Its core neutronics module is based on the modified one-group neutron kinetics equation with the prompt jump approximation and with six delayed neutron precursor groups. The module is used to analyze one fuel bundle of the reactor core with one mesh (region). The fuel heat conduction/transfer module solves the one-dimensional heat conduction equation in the radial direction with ten nodes in the fuel pin. The fuel channel thermal-hydraulic module is based on separated three-equation, two-phase flow equations with the drift flux correlation, and it analyzes one fuel bundle of the reactor core with one channel to evaluate flow redistribution between channels precisely. Thermal margin is evaluated by using the GEXL correlation, for example, in the module.« less

REACTOR CONTROL DEVICE

DOEpatents

Graham, R.H.

1962-09-01

A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

Optimization of Boiling Water Reactor Loading Pattern Using Two-Stage Genetic Algorithm

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

Kobayashi, Yoko; Aiyoshi, Eitaro

2002-10-15

A new two-stage optimization method based on genetic algorithms (GAs) using an if-then heuristic rule was developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). In the first stage, the LP is optimized using an improved GA operator. In the second stage, an exposure-dependent control rod pattern (CRP) is sought using GA with an if-then heuristic rule. The procedure of the improved GA is based on deterministic operators that consist of crossover, mutation, and selection. The handling of the encoding technique and constraint conditions by that GA reflects the peculiar characteristics of the BWR. In addition, strategies suchmore » as elitism and self-reproduction are effectively used in order to improve the search speed. The LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and constraints dependent on three dimensions have always necessitated the use of three-dimensional core simulators for BWRs, so that optimization of computational efficiency is required. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant in two phases. One phase is only LP optimization applying the Haling technique. The other phase is an LP optimization that considers the CRP during reactor operation. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.« less

Nitrogen removal from high organic loading wastewater in modified Ludzack-Ettinger configuration MBBR system.

PubMed

Torkaman, Mojtaba; Borghei, Seyed Mehdi; Tahmasebian, Sepehr; Andalibi, Mohammad Reza

2015-01-01

A moving bed biofilm reactor with pre-denitrification configuration was fed with a synthetic wastewater containing high chemical oxygen demand (COD) and ammonia. By changing different variables including ammonium and COD loading, nitrification rate in the aerobic reactor and denitrification rate in the anoxic reactor were monitored. Changing the influent loading was achieved via adjusting the inlet COD (956-2,096 mg/L), inlet ammonium (183-438 mg/L), and hydraulic retention time of the aerobic reactor (8, 12, and 18 hours). The overall organic loading rate was in the range of 3.60-17.37 gCOD/m2·day, of which 18.5-91% was removed in the anoxic reactor depending on the operational conditions. Considering the complementary role of the aerobic reactor, the overall COD removal was in the range 87.3-98.8%. In addition, nitrification rate increased with influent ammonium loading, the maximum rate reaching 3.05 gNH4/m2·day. One of the most important factors affecting nitrification rate was influent C:N entering the aerobic reactor, by increasing which nitrification rate decreased asymptotically. Nitrate removal efficiency in the anoxic reactor was also controlled by the inlet nitrate level entering the anoxic reactor. Furthermore, by increasing the nitrate loading rate from 0.91 to 3.49 gNO/m3·day, denitrification rate increased from 0.496 to 2.47 gNO/m3·day.

Fuel management optimization using genetic algorithms and expert knowledge

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

DeChaine, M.D.; Feltus, M.A.

1996-09-01

The CIGARO fuel management optimization code based on genetic algorithms is described and tested. The test problem optimized the core lifetime for a pressurized water reactor with a penalty function constraint on the peak normalized power. A bit-string genotype encoded the loading patterns, and genotype bias was reduced with additional bits. Expert knowledge about fuel management was incorporated into the genetic algorithm. Regional crossover exchanged physically adjacent fuel assemblies and improved the optimization slightly. Biasing the initial population toward a known priority table significantly improved the optimization.

On the influence of the americium isotopic vector on the cooling time of minor actinides bearing blankets in fast reactors

NASA Astrophysics Data System (ADS)

Kooyman, Timothée; Buiron, Laurent; Rimpault, Gerald

2018-05-01

In the heterogeneous minor actinides transmutation approach, the nuclei to be transmuted are loaded in dedicated targets often located at the core periphery, so that long-lived heavy nuclides are turned into shorter-lived fission products by fission. To compensate for low flux level at the core periphery, the minor actinides content in the targets is set relatively high (around 20 at.%), which has a negative impact on the reprocessing of the targets due to their important decay heat level. After a complete analysis of the main contributors to the heat load of the irradiated targets, it is shown here that the choice of the reprocessing order of the various feeds of americium from the fuel cycle depends on the actual limit for fuel reprocessing. If reprocessing of hot targets is possible, it is more interesting to reprocess first the americium feed with a high 243Am content in order to limit the total cooling time of the targets, while if reprocessing of targets is limited by their decay heat, it is more interesting to wait for an increase in the 241Am content before loading the americium in the core. An optimization of the reprocessing order appears to lead to a decrease of the total cooling time by 15 years compared to a situation where all the americium feeds are mixed together when two feeds from SFR are considered with a high reprocessing limit.

Particle Filter-Based Recursive Data Fusion With Sensor Indexing for Large Core Neutron Flux Estimation

NASA Astrophysics Data System (ADS)

Tamboli, Prakash Kumar; Duttagupta, Siddhartha P.; Roy, Kallol

2017-06-01

We introduce a sequential importance sampling particle filter (PF)-based multisensor multivariate nonlinear estimator for estimating the in-core neutron flux distribution for pressurized heavy water reactor core. Many critical applications such as reactor protection and control rely upon neutron flux information, and thus their reliability is of utmost importance. The point kinetic model based on neutron transport conveniently explains the dynamics of nuclear reactor. The neutron flux in the large core loosely coupled reactor is sensed by multiple sensors measuring point fluxes located at various locations inside the reactor core. The flux values are coupled to each other through diffusion equation. The coupling facilitates redundancy in the information. It is shown that multiple independent data about the localized flux can be fused together to enhance the estimation accuracy to a great extent. We also propose the sensor anomaly handling feature in multisensor PF to maintain the estimation process even when the sensor is faulty or generates data anomaly.

Magnetic nuclear core restraint and control

DOEpatents

Cooper, Martin H.

1979-01-01

A lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction.

Magnetic nuclear core restraint and control

DOEpatents

Cooper, Martin H.

1978-01-01

A lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction.

Safety and core design of large liquid-metal cooled fast breeder reactors

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

METHOD AND APPARATUS FOR CONTROLLING NEUTRON DENSITY

DOEpatents

Wigner, E.P.; Young, G.J.; Weinberg, A.M.

1961-06-27

A neutronic reactor comprising a moderator containing uniformly sized and spaced channels and uniformly dimensioned fuel elements is patented. The fuel elements have a fissionable core and an aluminum jacket. The cores and the jackets of the fuel elements in the central channels of the reactor are respectively thinner and thicker than the cores and jackets of the fuel elements in the remainder of the reactor, producing a flattened flux.

Space nuclear reactors — A post-operational disposal strategy

NASA Astrophysics Data System (ADS)

Angelo, Joseph A.; Buden, David

If 100-kWe and multimegawatt-electric class space nuclear reactors are to play a significant role in humanity's push into cislunar and heliocentric space in the next millennium, the obvious advantages of space nuclear power plants should not be denied to space mission planners due to a failure to develop internationally-acceptable post-operational disposal strategies for spent reactor cores. This is true whether the space reactor has shut down at the end of its normal mission lifetime or in response to an onboard system failure/emergency which causes a premature mission termination. Up until now the great majority of aerospace nuclear safety efforts have concentrated on prelaunch, launch and reactor startup activities. In fact, with the exception of the development of the "nuclear safe orbit" (NSO) concept, little technical attention has yet been given to the post-operational disposal of future space reactors. This paper describes the technical alternatives available for the safe, acceptable disposal of space reactors that could be used in a wide variety of space applications in the 21st Century. Post-operational core radioactivity levels for typical advanced design (hundred kWe-class) space reactors are presented as a function of decay time and contrasted to the spent core radionuclide inventory of the SNAP-10A system, the only nuclear reactor operated in space by the United States. The role of a permanent space station, smart robotic systems, and an operating lunar base in support of spent core disposal strategies is also presented, including use of a selected portion of the lunar surface as an internationally-designated spent reactor core repository.

Deleterious Thermal Effects Due To Randomized Flow Paths in Pebble Bed, and Particle Bed Style Reactors

NASA Technical Reports Server (NTRS)

Moran, Robert P.

2013-01-01

A review of literature associated with Pebble Bed and Particle Bed reactor core research has revealed a systemic problem inherent to reactor core concepts which utilize randomized rather than structured coolant channel flow paths. For both the Pebble Bed and Particle Bed Reactor designs; case studies reveal that for indeterminate reasons, regions within the core would suffer from excessive heating leading to thermal runaway and localized fuel melting. A thermal Computational Fluid Dynamics model was utilized to verify that In both the Pebble Bed and Particle Bed Reactor concepts randomized coolant channel pathways combined with localized high temperature regions would work together to resist the flow of coolant diverting it away from where it is needed the most to cooler less resistive pathways where it is needed the least. In other words given the choice via randomized coolant pathways the reactor coolant will take the path of least resistance, and hot zones offer the highest resistance. Having identified the relationship between randomized coolant channel pathways and localized fuel melting it is now safe to assume that other reactor concepts that utilize randomized coolant pathways such as the foam core reactor are also susceptible to this phenomenon.

A task-based parallelism and vectorized approach to 3D Method of Characteristics (MOC) reactor simulation for high performance computing architectures

NASA Astrophysics Data System (ADS)

Tramm, John R.; Gunow, Geoffrey; He, Tim; Smith, Kord S.; Forget, Benoit; Siegel, Andrew R.

2016-05-01

In this study we present and analyze a formulation of the 3D Method of Characteristics (MOC) technique applied to the simulation of full core nuclear reactors. Key features of the algorithm include a task-based parallelism model that allows independent MOC tracks to be assigned to threads dynamically, ensuring load balancing, and a wide vectorizable inner loop that takes advantage of modern SIMD computer architectures. The algorithm is implemented in a set of highly optimized proxy applications in order to investigate its performance characteristics on CPU, GPU, and Intel Xeon Phi architectures. Speed, power, and hardware cost efficiencies are compared. Additionally, performance bottlenecks are identified for each architecture in order to determine the prospects for continued scalability of the algorithm on next generation HPC architectures.

PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

NASA Astrophysics Data System (ADS)

Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

2014-06-01

The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

On heat loading, novel divertors, and fusion reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, M.; Valanju, P. M.; Mahajan, S. M.; Wiley, J. C.

2007-07-01

The limited thermal power handling capacity of the standard divertors (used in current as well as projected tokamaks) is likely to force extremely high (˜90%) radiation fractions frad in tokamak fusion reactors that have heating powers considerably larger than ITER [D. J. Campbell, Phys. Plasmas 8, 2041 (2001)]. Such enormous values of necessary frad could have serious and debilitating consequences on the core confinement, stability, and dependability for a fusion power reactor, especially in reactors with Internal Transport Barriers. A new class of divertors, called X-divertors (XD), which considerably enhance the divertor thermal capacity through a flaring of the field lines only near the divertor plates, may be necessary and sufficient to overcome these problems and lead to a dependable fusion power reactor with acceptable economics. X-divertors will lower the bar on the necessary confinement to bring it in the range of the present experimental results. Its ability to reduce the radiative burden imparts the X-divertor with a key advantage. Lower radiation demands allow sharply peaked density profiles that enhance the bootstrap fraction creating the possibility for a highly increased beta for the same beta normal discharges. The X-divertor emerges as a beta-enhancer capable of raising it by up to roughly a factor of 2.

Neutron flux and power in RTP core-15

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

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core withmore » literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.« less

Power monitoring in space nuclear reactors using silicon carbide radiation detectors

NASA Technical Reports Server (NTRS)

Ruddy, Frank H.; Patel, Jagdish U.; Williams, John G.

2005-01-01

Space reactor power monitors based on silicon carbide (SiC) semiconductor neutron detectors are proposed. Detection of fast leakage neutrons using SiC detectors in ex-core locations could be used to determine reactor power: Neutron fluxes, gamma-ray dose rates and ambient temperatures have been calculated as a function of distance from the reactor core, and the feasibility of power monitoring with SiC detectors has been evaluated at several ex-core locations. Arrays of SiC diodes can be configured to provide the required count rates to monitor reactor power from startup to full power Due to their resistance to temperature and the effects of neutron and gamma-ray exposure, SiC detectors can be expected to provide power monitoring information for the fill mission of a space reactor.

HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

DOEpatents

Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

1960-06-14

A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

Two-Dimensional Diffusion Theory Analysis of Reactivity Effects of a Fuel-Plate-Removal Experiment

NASA Technical Reports Server (NTRS)

Gotsky, Edward R.; Cusick, James P.; Bogart, Donald

1959-01-01

Two-dimensional two-group diffusion calculations were performed on the NASA reactor simulator in order to evaluate the reactivity effects of fuel plates removed successively from the center experimental fuel element of a seven- by three-element core loading at the Oak Ridge Bulk Shielding Facility. The reactivity calculations were performed by two methods: In the first, the slowing-down properties of the experimental fuel element were represented by its infinite media parameters; and, in the second, the finite size of the experimental fuel element was recognized, and the slowing-down properties of the surrounding core were attributed to this small region. The latter calculation method agreed very well with the experimented reactivity effects; the former method underestimated the experimental reactivity effects.

The effect of transient loading on the performance of a mesophilic anaerobic contact reactor at constant feed strength.

PubMed

Sentürk, Elif; Ince, Mahir; Engin, Guleda Onkal

2012-12-15

Anaerobic contact reactor is a high rate anaerobic process consisting of an agitated reactor and a solids settling tank for recycling. It was proved earlier that this type of reactor design offers highly efficient performance in the conversion of organic matter to biogas. In this study, the effect of transient loading on reactor performance in terms of a number of key intermediates and parameters such as, COD removal, pH and alkalinity change, VFAs, effluent MLSS concentration and biogas efficiency over time was examined. For this purpose, a step increase of organic loading rate from 3.35kg COD/m(3)day to 15.61kg COD/m(3)day was employed. The hydraulic retention time decreased to a value of 8.42h by an increase in the influent flow-rate during the transient loading. It was observed that the mesophilic anaerobic contact reactor (MACR) was quite resistant to large transient shocks. The reactor recovered back to its baseline performance only in 15h after the shock loading was stopped. Hence, it can be concluded that this type of reactor design has a high potential in treating food processing wastewaters with varying flow characteristics. Copyright © 2012 Elsevier B.V. All rights reserved.

PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

DOEpatents

Puechl, K.H.

1963-09-24

A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

Seed and blanket fuel arrangement for dual-phase nuclear reactors

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

Congdon, S.P.; Fawcett, R.M.

1992-09-22

This patent describes a fuel management method for a dual-phase nuclear reactor, it comprises: installing a fuel bundle at a first core location accessed by coolant through a relatively small aperture, each of the bundles having a predetermined group of fuel elements; operating the reactor a first time; shutting down the reactor; reinstalling the fuel bundle at a second core location accessed by coolant through a relatively large aperture; and operating the reactor a second time.

MERCHANT MARINE SHIP REACTOR

DOEpatents

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Merchant Marine Ship Reactor

DOEpatents

Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

1961-05-01

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

DOEpatents

Moore, W.T.

1958-09-01

This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

[Rapid startup and nitrogen removal characteristic of anaerobic ammonium oxidation reactor in packed bed biofilm reactor with suspended carrier].

PubMed

Chen, Sheng; Sun, De-zhi; Yu, Guang-lu

2010-03-01

Packed bed biofilm reactor with suspended carrier was used to cultivate ANAMMOX bacteria with sludge inoculums from WWTP secondary settler. The startup of ANAMMOX reactor was comparatively studied using high nitrogen loading method and low nitrogen loading method with aerobically biofilmed on the carrier, and the nitrogen removal characteristic was further investigated. The results showed that the reactor could be started up successfully within 90 days using low nitrogen loading method, the removal efficiencies of ammonium and nitrite were nearly 100% and the TN removal efficiencywas over 75% , however, the high nitrogen loading method was proved unsuccessfully for startup of ANAMMOX reactor probably because of the inhibition effect of high concentration of ammonium and nitrite. The pH value of effluent was slightly higher than the influent and the pH value can be used as an indicator for the process of ANAMMOX reaction. The packed bed ANAMMOX reactor with suspended carrier showed good characteristics of high nitrogen loading and high removal efficiency, 100% of removal efficiency could be achieved when the influent ammonium and nitrite concentration was lower than 800 mg/L.

Characteristics of sludge developed under different loading conditions during UASB reactor start-up and granulation.

PubMed

Ghangrekar, M M; Asolekar, S R; Joshi, S G

2005-03-01

Sludge characteristics available inside the reactor are of vital importance to maximize advantages of UASB reactor. The organic loading rate and sludge loading rate applied during start-up are among the important parameters to govern the sludge characteristics. Effects of these loading rates on the characteristics of the sludge developed are evaluated in six laboratory scale UASB reactors. The sludge characteristics considered are VSS/SS ratio of the sludge, sludge volume index, specific gravity, settling velocity and metal contents of the sludge developed under different loading rates. The experimental results indicate that, for developing good characteristics sludge, during primary start-up from flocculent inoculum sludge, organic loading rate and sludge loading rate should be in the range of 2.0-4.5 kg COD/m3 d and 0.1-0.25 kg COD/kg VSS d, respectively (chemical oxygen demand, COD). Proper sludge granulation and higher COD removal efficiency will be achieved by these loading rates.

Nuclear engine flow reactivity shim control

DOEpatents

Walsh, J.M.

1973-12-11

A nuclear engine control system is provided which automatically compensates for reactor reactivity uncertainties at the start of life and reactivity losses due to core corrosion during the reactor life in gas-cooled reactors. The coolant gas flow is varied automatically by means of specially provided control apparatus so that the reactor control drums maintain a predetermined steady state position throughout the reactor life. This permits the reactor to be designed for a constant drum position and results in a desirable, relatively flat temperature profile across the core. (Official Gazette)

Control rod drive for reactor shutdown

DOEpatents

McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

1976-01-20

A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

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

Permana, Sidik; Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132; Sekimoto, Hiroshi

2010-12-23

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period hasmore » been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.« less

REACTOR UNLOADING

DOEpatents

Leverett, M.C.

1958-02-18

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

Nuclear waste disposal utilizing a gaseous core reactor

NASA Technical Reports Server (NTRS)

Paternoster, R. R.

1975-01-01

The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

Gas core reactors for actinide transmutation and breeder applications

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1978-01-01

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

TRACE/PARCS Analysis of ATWS with Instability for a MELLLA+BWR/5

DOE PAGES

L. Y. Cheng; Baek, J. S.; Cuadra, A.; ...

2016-06-06

A TRACE/PARCS model has been developed to analyze anticipated transient without SCRAM (ATWS) events for a boiling water reactor (BWR) operating in the maximum extended load line limit analysis-plus (MELLLA+) expanded operating domain. The MELLLA+ domain expands allowable operation in the power/flow map of a BWR to low flow rates at high power conditions. Such operation exacerbates the likelihood of large amplitude power/flow oscillations during certain ATWS scenarios. The analysis shows that large amplitude power/flow oscillations, both core-wide and out-of-phase, arise following the establishment of natural circulation flow in the reactor pressure vessel (RPV) after the trip of the recirculationmore » pumps and an increase in core inlet subcooling. The analysis also indicates a mechanism by which the fuel may experience heat-up that could result in localized fuel damage. TRACE predicts the heat-up to occur when the cladding surface temperature exceeds the minimum stable film boiling temperature after periodic cycles of dryout and rewet; and the fuel becomes “locked” into a film boiling regime. Further, the analysis demonstrates the effectiveness of the simulated manual operator actions to suppress the instability.« less

NUCLEAR REACTOR CORE

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

Preece, G.E.; Bell, F.R.; Page, R.W.

1963-03-01

A nuclear reactor core is described. It contains fuel in the form of blocks or pellets that have a grooved, wrinkled, or corrugated surface to provide a greater radiating surface area. The surfaces of spaces in the core are correspondingly corrugated for maximum heat exchange area. (C.E.S.)

An approach to model reactor core nodalization for deterministic safety analysis

NASA Astrophysics Data System (ADS)

Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

2016-01-01

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

An approach to model reactor core nodalization for deterministic safety analysis

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

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my; Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to bemore » employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.« less

Automatic safety rod for reactors. [LMFBR

DOEpatents

Germer, J.H.

1982-03-23

An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

78 FR 28896 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-16

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0095] Design Limits and Loading Combinations for Metal... Regulatory Guide (RG) 1.57, ``Design Limits and Loading Combinations for Metal Primary Reactor Containment... the NRC staff considers acceptable for design limits and loading combinations for metal primary...

Nuclear reactor melt-retention structure to mitigate direct containment heating

DOEpatents

Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

1991-01-01

A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

Extending the maximum operation time of the MNSR reactor.

PubMed

Dawahra, S; Khattab, K; Saba, G

2016-09-01

An effective modification to extend the maximum operation time of the Miniature Neutron Source Reactor (MNSR) to enhance the utilization of the reactor has been tested using the MCNP4C code. This modification consisted of inserting manually in each of the reactor inner irradiation tube a chain of three polyethylene-connected containers filled of water. The total height of the chain was 11.5cm. The replacement of the actual cadmium absorber with B(10) absorber was needed as well. The rest of the core structure materials and dimensions remained unchanged. A 3-D neutronic model with the new modifications was developed to compare the neutronic parameters of the old and modified cores. The results of the old and modified core excess reactivities (ρex) were: 3.954, 6.241 mk respectively. The maximum reactor operation times were: 428, 1025min and the safety reactivity factors were: 1.654 and 1.595 respectively. Therefore, a 139% increase in the maximum reactor operation time was noticed for the modified core. This increase enhanced the utilization of the MNSR reactor to conduct a long time irradiation of the unknown samples using the NAA technique and increase the amount of radioisotope production in the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

Fuel handling system for a nuclear reactor

DOEpatents

Saiveau, James G.; Kann, William J.; Burelbach, James P.

1986-01-01

A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

Fuel handling system for a nuclear reactor

DOEpatents

Saiveau, James G.; Kann, William J.; Burelbach, James P.

1986-12-02

A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

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

Trianti, Nuri, E-mail: nuri.trianti@gmail.com; Nurjanah,; Su’ud, Zaki

Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density andmore » inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.« less

Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

NASA Technical Reports Server (NTRS)

Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

1972-01-01

A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (2) Critical experiment of lithium-6 used in LEM and LIM

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

Tsunoda, Hirokazu; Sato, Osamu; Okajima, Shigeaki

2002-07-01

In order to achieve fully automated reactor operation of RAPID-L reactor, innovative reactivity control systems LEM, LIM, and LRM are equipped with lithium-6 as a liquid poison. Because lithium-6 has not been used as a neutron absorbing material of conventional fast reactors, measurements of the reactivity worth of Lithium-6 were performed at the Fast Critical Assembly (FCA) of Japan Atomic Energy Research Institute (JAERI). The FCA core was composed of highly enriched uranium and stainless steel samples so as to simulate the core spectrum of RAPID-L. The samples of 95% enriched lithium-6 were inserted into the core parallel to themore » core axis for the measurement of the reactivity worth at each position. It was found that the measured reactivity worth in the core region well agreed with calculated value by the method for the core designs of RAPID-L. Bias factors for the core design method were obtained by comparing between experimental and calculated results. The factors were used to determine the number of LEM and LIM equipped in the core to achieve fully automated operation of RAPID-L. (authors)« less

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

...The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to regulatory guide (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance for preoperational testing of new pressurized water reactor (PWR) designs.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Reducing numerical costs for core wide nuclear reactor CFD simulations by the Coarse-Grid-CFD

NASA Astrophysics Data System (ADS)

Viellieber, Mathias; Class, Andreas G.

2013-11-01

Traditionally complete nuclear reactor core simulations are performed with subchannel analysis codes, that rely on experimental and empirical input. The Coarse-Grid-CFD (CGCFD) intends to replace the experimental or empirical input with CFD data. The reactor core consists of repetitive flow patterns, allowing the general approach of creating a parametrized model for one segment and composing many of those to obtain the entire reactor simulation. The method is based on a detailed and well-resolved CFD simulation of one representative segment. From this simulation we extract so-called parametrized volumetric forces which close, an otherwise strongly under resolved, coarsely-meshed model of a complete reactor setup. While the formulation so far accounts for forces created internally in the fluid others e.g. obstruction and flow deviation through spacers and wire wraps, still need to be accounted for if the geometric details are not represented in the coarse mesh. These are modelled with an Anisotropic Porosity Formulation (APF). This work focuses on the application of the CGCFD to a complete reactor core setup and the accomplishment of the parametrization of the volumetric forces.

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

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

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

Design Study of a Modular Gas-Cooled, Closed-Brayton Cycle Reactor for Marine Use

DTIC Science & Technology

1989-06-01

materials in the core and surroundings. To investigate this design point in the marine variant I developed the program HEAT.BAS to perform a one-dimensional...helium as the working fluid. The core is a graphite moderated, epithermal spectrum reactor, using TRISO fuel particles in extruded graphite fuel elements...The fuel is highly enriched U2315 . The containment is shaped in an inverted ’T’ with two sections. The upper section contains the reactor core

Horizontal baffle for nuclear reactors

DOEpatents

Rylatt, John A.

1978-01-01

A horizontal baffle disposed in the annulus defined between the core barrel and the thermal liner of a nuclear reactor thereby physically separating the outlet region of the core from the annular area below the horizontal baffle. The horizontal baffle prevents hot coolant that has passed through the reactor core from thermally damaging apparatus located in the annulus below the horizontal baffle by utilizing the thermally induced bowing of the horizontal baffle to enhance sealing while accommodating lateral motion of the baffle base plate.

Convective cooling in a pool-type research reactor

NASA Astrophysics Data System (ADS)

Sipaun, Susan; Usman, Shoaib

2016-01-01

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

DOEpatents

Dickinson, R.W.

1963-03-01

This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

Neutron shielding panels for reactor pressure vessels

DOEpatents

Singleton, Norman R [Murrysville, PA

2011-11-22

In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

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

Monte Carol-based validation of neutronic methodology for EBR-II analyses

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

Liaw, J.R.; Finck, P.J.

1993-01-01

The continuous-energy Monte Carlo code VIM (Ref. 1) has been validated extensively over the years against fast critical experiments and other neutronic analysis codes. A high degree of confidence in VIM for predicting reactor physics parameters has been firmly established. This paper presents a numerical validation of two conventional multigroup neutronic analysis codes, DIF3D (Ref. 4) and VARIANT (Ref. 5), against VIM for two Experimental Breeder Reactor II (EBR-II) core loadings in detailed three-dimensional hexagonal-z geometry. The DIF3D code is based on nodal diffusion theory, and it is used in calculations for day-today reactor operations, whereas the VARIANT code ismore » based on nodal transport theory and is used with increasing frequency for specific applications. Both DIF3D and VARIANT rely on multigroup cross sections generated from ENDF/B-V by the ETOE-2/MC[sup 2]-II/SDX (Ref. 6) code package. Hence, this study also validates the multigroup cross-section processing methodology against the continuous-energy approach used in VIM.« less

AGC-2 Graphite Pre-irradiation Data Package

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

David Swank; Joseph Lord; David Rohrbaugh

2010-08-01

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterizedmore » prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.« less

Can tokamaks PFC survive a single event of any plasma instabilities?

NASA Astrophysics Data System (ADS)

Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

2013-07-01

Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

PROSPECT - A Precision Oscillation and Spectrum Experiment

NASA Astrophysics Data System (ADS)

Zhang, Xianyi; Prospect Collaboration

2017-01-01

PROSPECT, the PRecision Oscillation and SPECTrum Experiment, is a multi-phased short baseline reactor antineutrino experiment that aims to precisely measure the U-235 antineutrino spectrum and prob for oscillation effects involving a possible Δm2 1 eV2 scale sterile neutrino. In PROSPECT Phase-I, an optically segmented Li-6 loaded liquid scintillator detector will be deployed at at the baseline of 7-12m from the High Flux Isotope Reactor at the Oak Ridge National Laboratory. PROSPECT will measure the spectrum of U-235 to aid in resolving the unexplained inconsistency between predictive spectral models and recent experimental measurements using LEU cores, while the oscillation measurement will probe the best fit region suggested by global fitting studies within 1-year data taking. This talk will introduce the design of PROSPECT Phase-I, the discovery potential of the experiment, and the progress the collaboration has made toward realizing PROSPECT Phase-I. Department of Energy

Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report

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

G. S. Chang; M. A. Lillo; R. G. Ambrosek

2008-06-01

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.« less

Inherently Safe Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.; El-Genk, Mohamed S.

2013-09-01

This paper presents the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system for future lunar outposts. The power system nominally provides 38 kWe continuously for 21 years, employs static components and has no single point failures in reactor cooling or power generation. The reactor core has six sectors, each has a separate pair of primary and secondary loops with liquid NaK-56 working fluid, thermoelectric (TE) power conversion and heat-pipes radiator panels. The electromagnetic (EM) pumps in the primary and secondary loops, powered with separate TE power units, ensure operation reliability and passive decay heat removal from the reactor after shutdown. The reactor poses no radiological concerns during launch, and remains sufficiently subcritical, with the radial reflector dissembled, when submerged in wet sand and the core flooded with seawater, following a launch abort accident. After 300 years of storage below grade on the Moon, the total radioactivity in the post-operation reactor drops below 164 Ci, a low enough radioactivity for a recovery and safe handling of the reactor.

Optimization of 200 MWth and 250 MWt Ship Based Small Long Life NPP

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

Fitriyani, Dian; Su'ud, Zaki

2010-06-22

Design optimization of ship-based 200 MWth and 250 MWt nuclear power reactors have been performed. The neutronic and thermo-hydraulic programs of the three-dimensional X-Y-Z geometry have been developed for the analysis of ship-based nuclear power plant. Quasi-static approach is adopted to treat seawater effect. The reactor are loop type lead bismuth cooled fast reactor with nitride fuel and with relatively large coolant pipe above reactor core, the heat from primary coolant system is directly transferred to watersteam loop through steam generators. Square core type are selected and optimized. As the optimization result, the core outlet temperature distribution is changing withmore » the elevation angle of the reactor system and the characteristics are discussed.« less

Dual annular rotating "windowed" nuclear reflector reactor control system

DOEpatents

Jacox, Michael G.; Drexler, Robert L.; Hunt, Robert N. M.; Lake, James A.

1994-01-01

A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

Nuclear design of a vapor core reactor for space nuclear propulsion

NASA Astrophysics Data System (ADS)

Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

1993-01-01

Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

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.

The Chornobyl accident revisited, part II: The state of the nuclear fuel located within the Chornobyl Sarcophagus

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

Borovol, A.A.; Sich, A.R.

1995-10-01

Approximately 135 tonnes of the 190.3-tonne initial core fuel load ({approx}71%) at Chornobyl Unit 4 melted and flowed into the lower regions of the reactor building to form various kinds of the now-solidified lava-like fuel-containing materials (LFCMs) or corium. The results of radiochemical analyses reveal that only 5% of the LFCM inventory of Ru-106 remains, whereas, surprisingly, 35% of the LFCM inventory of Cs-137 remains. Moreover, the results of these analyses support the fact that little if any of the 5020 tonnes of various materials (dropped from helicopters during the active phase of the accident in an attempt to smothermore » the burning graphite) ever made it into the core shaft, where the bulk of the core was located. The results appear to support earlier Western source-term estimates that significantly more volatile radionuclides may have been released as a result of the accident.« less

The Chornobyl accident revisited, Part II: The state of the nuclear fuel located within the Chornobyl sarcophagus

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

Borovoi, A.A.; Sich, A.R.

1995-01-01

Approximately 135 tonnes of the 190.3-tonne initial core fuel load ({approx}71%) at Chernobyl Unit 4 melted and flowed into the lower regions of the reactor building to form various kinds of the now-solidified lava-like fuel-containing materials (LFCMs) or corium. The results of radiochemical analyses reveal that only 5% of the LFCM inventory of Ru-106 remains. whereas, surprisingly, 35% of the LFCM inventory of Cs-137 remains. Moreover, the results of these analyses support the fact that little if any of the 5020 tonnes of various materials (dropped from helicopters during the active phase of the accident in an attempt to smothermore » the burning graphite) ever made it into the core shaft, where the bulk of the core was located. The results appear to support earlier Western source-term estimates that significantly more volatile radionuclides may have been released as a result of the accident. 37 refs., 13 figs., 7 tabs.« less

Nuclear reactor alignment plate configuration

DOEpatents

Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

2014-01-28

An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

Energy production using fission fragment rockets

NASA Astrophysics Data System (ADS)

Chapline, G.; Matsuda, Y.

1991-08-01

Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: approximately twice the efficiency if the fission fragment energy can be directly converted into electricity; reduction of the buildup of a fission fragment inventory in the reactor could avoid a Chernobyl type disaster; and collection of the fission fragments outside the reactor could simplify the waste disposal problem.

Genome-Resolved Meta-Omics Ties Microbial Dynamics to Process Performance in Biotechnology for Thiocyanate Degradation.

PubMed

Kantor, Rose S; Huddy, Robert J; Iyer, Ramsunder; Thomas, Brian C; Brown, Christopher T; Anantharaman, Karthik; Tringe, Susannah; Hettich, Robert L; Harrison, Susan T L; Banfield, Jillian F

2017-03-07

Remediation of industrial wastewater is important for preventing environmental contamination and enabling water reuse. Biological treatment for one industrial contaminant, thiocyanate (SCN - ), relies upon microbial hydrolysis, but this process is sensitive to high loadings. To examine the activity and stability of a microbial community over increasing SCN - loadings, we established and operated a continuous-flow bioreactor fed increasing loadings of SCN - . A second reactor was fed ammonium sulfate to mimic breakdown products of SCN - . Biomass was sampled from both reactors for metagenomics and metaproteomics, yielding a set of genomes for 144 bacteria and one rotifer that constituted the abundant community in both reactors. We analyzed the metabolic potential and temporal dynamics of these organisms across the increasing loadings. In the SCN - reactor, Thiobacillus strains capable of SCN - degradation were highly abundant, whereas the ammonium sulfate reactor contained nitrifiers and heterotrophs capable of nitrate reduction. Key organisms in the SCN - reactor expressed proteins involved in SCN - degradation, sulfur oxidation, carbon fixation, and nitrogen removal. Lower performance at higher loadings was linked to changes in microbial community composition. This work provides an example of how meta-omics can increase our understanding of industrial wastewater treatment and inform iterative process design and development.

MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

DOEpatents

Balent, R.

1963-03-12

This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

Reactor vessel lower head integrity

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

Rubin, A.M.

1997-02-01

On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) inmore » this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.« less

ATR LEU Fuel and Burnable Absorber Neutronics Performance Optimization by Fuel Meat Thickness Variation

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

G. S. Chang

2007-09-01

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.508 mm and the same U-235 enrichment (15.5 wt%) can be used to optimize the radial heat flux profile by varying the fuel plate thickness from 0.254 to 0.457 mm at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, a 0.7g of burnable absorber Boron-10 was added in the inner and outer plates to reduce the initial excess reactivity, and the inner/outer heat flux more effectively. The optimized LEU relative radial fission heat flux profile is bounded by the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Updated Final Safety Analysis Report (UFSAR) safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.« less

Experiment for search for sterile neutrino at SM-3 reactor

NASA Astrophysics Data System (ADS)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

THE COOLING REQUIREMENTS AND PROCESS SYSTEMS OF THE SOUTH AFRICAN RESEARCH REACTOR, SAFARI 1

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

Colley, J.R.

1962-12-01

The SAFARI 1 research reactor is cooled and moderated by light water. There are three process systems, a primary water system which cools the reactor core and surroundings, a pool water system, and a secondary water system which removes the heat from the primary and pool systems. The cooling requirements for the reactor core and experimental facilities are outlined, and the cooling and purification functions of the three process systems are described. (auth)

Neutronics calculation of RTP core

NASA Astrophysics Data System (ADS)

Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

2017-01-01

Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions

NASA Astrophysics Data System (ADS)

Litunovsky, N.; Gervash, A.; Lorenzetto, P.; Mazul, I.; Melder, R.

2009-04-01

The paper describes the experimental technique and preliminary results of thermal fatigue testing of ITER first wall (FW) water-cooled mock-ups inside the core of the RBT-6 experimental fission reactor (RIAR, Dimitrovgrad, Russia). This experiment has provided simultaneous effect of neutron fluence and thermal cycling damages on the mock-ups. A PC-controlled high-temperature graphite ohmic heater was applied to provide cyclic thermal load onto the mock-ups surface. This experiment lasted for 309 effective irradiation days with a final damage level (CuCrZr) of 1 dpa in the mock-ups. About 3700 thermal cycles with a heat flux of 0.4-0.5 MW/m 2 onto the mock-ups were realized before the heater fails. Then, irradiation was continued in a non-cycling mode.

Development work for a borax internal core-catcher for a gas-cooled fast reactor

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

Donne, M.D.; Dorner, S.; Schumacher, G.

1978-07-01

Preliminary thermal calculations show that a corecatcher, which is able to cope with the complete meltdown of the core and blankets of a 1000-MW(electric) gas-cooled fast reactor, appears to be feasible. This core-catcher is based on borax (Na/sub 2/B/sub 4/O/sub 7/) dissolving the oxide fuel and the fission products occurring in oxide form. The borax is contained in steel boxes forming a 2.2-m-thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel (PCRV), just underneath the reactor core. After a complete meltdown accident, the fission products, in oxide form, are dispersed in the pool formedmore » by the liquid borax. The metallic fission products are contained in the steel lying below the borax pool and in contact with the water-cooled PCRV liner. The volumetric power density of the molten core is conveniently reduced as it is dissolved in the borax, and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system.« less

Consistent criticality and radiation studies of Swiss spent nuclear fuel: The CS2M approach.

PubMed

Rochman, D; Vasiliev, A; Ferroukhi, H; Pecchia, M

2018-06-15

In this paper, a new method is proposed to systematically calculate at the same time canister loading curves and radiation sources, based on the inventory information from an in-core fuel management system. As a demonstration, the isotopic contents of the assemblies come from a Swiss PWR, considering more than 6000 cases from 34 reactor cycles. The CS 2 M approach consists in combining four codes: CASMO and SIMULATE to extract the assembly characteristics (based on validated models), the SNF code for source emission and MCNP for criticality calculations for specific canister loadings. The considered cases cover enrichments from 1.9 to 5.0% for the UO 2 assemblies and 4.8% for the MOX, with assembly burnup values from 7 to 74 MWd/kgU. Because such a study is based on the individual fuel assembly history, it opens the possibility to optimize canister loadings from the point-of-view of criticality, decay heat and emission sources. Copyright © 2018 Elsevier B.V. All rights reserved.

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

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

Licht, J.; Bergeron, A.; Dionne, B.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimentalmore » device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.« less

Nuclear reactor neutron shielding

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

Speaker, Daniel P; Neeley, Gary W; Inman, James B

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactormore » cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.« less

Estimated inventory of radionuclides in former Soviet Union naval reactors dumped in the Kara Sea

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

Mount, M.E.; Sheaffer, M.K.; Abbott, D.T.

1993-07-01

Radionuclide inventories have been estimated for the reactor cores, reactor components, and primary system corrosion products in the former Soviet Union naval reactors dumped at the Abrosimov Inlet, Tsivolka Inlet, Stepovoy Inlet, Techeniye Inlet, and Novaya Zemlya Depression sites in the Kara Sea between 1965 and 1988. For the time of disposal, the inventories are estimated at 69 to 111 kCi of actinides plus daughters and 3,053 to 7,472 kCi of fission products in the reactor cores, 917 to 1,127 kCi of activation products in the reactor components, and 1.4 to 1.6 kCi of activation products in the primary systemmore » corrosion products. At the present time, the inventories are estimated to have decreased to 23 to 38 kCi of actinides plus daughters and 674 to 708 kCi of fission products in the reactor cores, 124 to 126 kCi of activation products in the reactor components, and 0.16 to 0.17 kCi of activation products in the primary system corrosion products. Twenty years from now, the inventories are projected to be 11 to 18 kCi of actinides plus daughters and 415 to 437 kCi of fission products in the reactor cores, 63.5 to 64 kCi of activation products in the reactor components, and 0.014 to 0.015 kCi of activation products in the primary system corrosion products. All actinide activities are estimated to be within a factor of two.« less

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

Brunett, A. J.; Fei, T.; Strons, P. S.

The Transient Reactor Test Facility (TREAT), located at Idaho National Laboratory (INL), is a test facility designed to evaluate the performance of reactor fuels and materials under transient accident conditions. The facility, an air-cooled, graphite-moderated reactor designed to utilize fuel containing high-enriched uranium (HEU), has been in non-operational standby status since 1994. Currently, in support of the missions of the Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program, a new core design is being developed for TREAT that will utilize low-enriched uranium (LEU). The primary objective of this conversion effort ismore » to design an LEU core that is capable of meeting the performance characteristics of the existing HEU core. Minimal, if any, changes are anticipated for the supporting systems (e.g. reactor trip system, filtration/cooling system, etc.); therefore, the LEU core must also be able to function with the existing supporting systems, and must also satisfy acceptable safety limits. In support of the LEU conversion effort, a range of ancillary safety analyses are required to evaluate the LEU core operation relative to that of the existing facility. These analyses cover neutronics, shielding, and thermal hydraulic topics that have been identified as having the potential to have reduced safety margins due to conversion to LEU fuel, or are required to support the required safety analyses documentation. The majority of these ancillary tasks have been identified in [1] and [2]. The purpose of this report is to document the ancillary safety analyses that have been performed at Argonne National Laboratory during the early stages of the LEU design effort, and to describe ongoing and anticipated analyses. For all analyses presented in this report, methodologies are utilized that are consistent with, or improved from, those used in analyses for the HEU Final Safety Analysis Report (FSAR) [3]. Depending on the availability of historical data derived from HEU TREAT operation, results calculated for the LEU core are compared to measurements obtained from HEU TREAT operation. While all analyses in this report are largely considered complete and have been reviewed for technical content, it is important to note that all topics will be revisited once the LEU design approaches its final stages of maturity. For most safety significant issues, it is expected that the analyses presented here will be bounding, but additional calculations will be performed as necessary to support safety analyses and safety documentation. It should also be noted that these analyses were completed as the LEU design evolved, and therefore utilized different LEU reference designs. Preliminary shielding, neutronic, and thermal hydraulic analyses have been completed and have generally demonstrated that the various LEU core designs will satisfy existing safety limits and standards also satisfied by the existing HEU core. These analyses include the assessment of the dose rate in the hodoscope room, near a loaded fuel transfer cask, above the fuel storage area, and near the HEPA filters. The potential change in the concentration of tramp uranium and change in neutron flux reaching instrumentation has also been assessed. Safety-significant thermal hydraulic items addressed in this report include thermally-induced mechanical distortion of the grid plate, and heating in the radial reflector.« less

Combination pipe-rupture mitigator and in-vessel core catcher. [LMFBR

DOEpatents

Tilbrook, R.W.; Markowski, F.J.

1982-03-09

A device is described which mitigates against the effects of a failed coolant loop in a nuclear reactor by restricting the outflow of coolant from the reactor through the failed loop and by retaining any particulated debris from a molten core which may result from coolant loss or other cause. The device reduces the reverse pressure drop through the failed loop by limiting the access of coolant in the reactor to the inlet of the failed loop. The device also spreads any particulated core debris over a large area to promote cooling.

Combination pipe rupture mitigator and in-vessel core catcher

DOEpatents

Tilbrook, Roger W.; Markowski, Franz J.

1983-01-01

A device which mitigates against the effects of a failed coolant loop in a nuclear reactor by restricting the outflow of coolant from the reactor through the failed loop and by retaining any particulated debris from a molten core which may result from coolant loss or other cause. The device reduces the reverse pressure drop through the failed loop by limiting the access of coolant in the reactor to the inlet of the failed loop. The device also spreads any particulated core debris over a large area to promote cooling.

NEUTRONIC REACTOR

DOEpatents

Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

1959-03-24

A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

Dual annular rotating [open quotes]windowed[close quotes] nuclear reflector reactor control system

DOEpatents

Jacox, M.G.; Drexler, R.L.; Hunt, R.N.M.; Lake, J.A.

1994-03-29

A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core. 4 figures.

In-Pile Instrumentation Multi- Parameter System Utilizing Photonic Fibers and Nanovision

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

Burgett, Eric

2015-10-13

An advanced in-pile multi-parameter reactor monitoring system is being proposed in this funding opportunity. The proposed effort brings cutting edge, high fidelity optical measurement systems into the reactor environment in an unprecedented fashion, including in-core, in-cladding and in-fuel pellet itself. Unlike instrumented leads, the proposed system provides a unique solution to a multi-parameter monitoring need in core while being minimally intrusive in the reactor core. Detector designs proposed herein can monitor fuel compression and expansion in both the radial and axial dimensions as well as monitor linear power profiles and fission rates during the operation of the reactor. In additionmore » to pressure, stress, strain, compression, neutron flux, neutron spectra, and temperature can be observed inside the fuel bundle and fuel rod using the proposed system. The proposed research aims at developing radiation-hard, harsh-environment multi-parameter systems for insertion into the reactor environment. The proposed research holds the potential to drastically increase the fidelity and precision of in-core instrumentation with little or no impact in the neutron economy in the reactor environment while providing a measurement system capable of operation for entire operating cycles.« less

Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

NASA Astrophysics Data System (ADS)

Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

2017-02-01

International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

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

Fitzpatrick, F.C.; Gray, D.D.; Hyndman, J.R.

The thermal, ecological, and social impacts of a 40-reactor NEC are compared to impacts from four 10-reactor NECs and ten 4-reactor power plants. The comparison was made for surrogate sites in western Tennessee. The surrogate site for the 40-reactor NEC is located on Kentucky Lake. A layout is postulated for ten clusters of four reactors each with 2.5-mile spacing between clusters. The plants use natural-draft cooling towers. A transmission system is proposed for delivering the power (48,000 MW) to five load centers. Comparable transmission systems are proposed for the 10-reactor NECs and the 4-reactor dispersed sites delivering power to themore » same load centers. (auth)« less

Convective cooling in a pool-type research reactor

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

Sipaun, Susan, E-mail: susan@nm.gov.my; Usman, Shoaib, E-mail: usmans@mst.edu

2016-01-22

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to passmore » through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.« less

High-temperature, high-pressure bonding of nested tubular metallic components

DOEpatents

Quinby, T.C.

A tool is described for effecting high-temperature, high-compression bonding between the confronting faces of nested, tubular, metallic components. In a typical application, the tool is used to produce tubular target assemblies for irradiation in nuclear reactors or particle accelerators. The target assembly comprising a uranum foil and an aluninum-alloy substrate. The tool is composed of graphite. It comprises a tubular restraining member in which a mechanically expandable tubular core is mounted to form an annulus. The components to be bonded are mounted in nested relation in the annulus. The expandable core is formed of individually movable, axially elongated segments whose outer faces cooperatively define a cylindrical pressing surface and whose inner faces cooperatively define two opposed, inwardly tapered, axial bores. Tapered rams extend into the bores. The loaded tool is mounted in a conventional hot-press provided with evacuation means, heaters for maintaining its interior at bonding temperature, and hydraulic cylinders for maintaining a selected inwardly directed pressure on the tapered rams. With the hot-press evacuated and the loaded tool at the desired temperature, the cylinders are actuated to apply the selected pressure to the rams. The rams in turn expand the segmented core to maintain the nested components in compression against the restraining member. These conditions are maintained until the confronting faces of the nested components are joined in a continuous, uniform bond characterized by high thermal conductivity.

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

NASA Astrophysics Data System (ADS)

Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

2017-07-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

NEUTRONIC REACTOR CONTROL

DOEpatents

Metcalf, H.E.

1958-10-14

Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

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

Ehud Greenspan

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

A feasibility study of reactor-based deep-burn concepts.

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

Kim, T. K.; Taiwo, T. A.; Hill, R. N.

2005-09-16

A systematic assessment of the General Atomics (GA) proposed Deep-Burn concept based on the Modular Helium-Cooled Reactor design (DB-MHR) has been performed. Preliminary benchmarking of deterministic physics codes was done by comparing code results to those from MONTEBURNS (MCNP-ORIGEN) calculations. Detailed fuel cycle analyses were performed in order to provide an independent evaluation of the physics and transmutation performance of the one-pass and two-pass concepts. Key performance parameters such as transuranic consumption, reactor performance, and spent fuel characteristics were analyzed. This effort has been undertaken in close collaborations with the General Atomics design team and Brookhaven National Laboratory evaluation team.more » The study was performed primarily for a 600 MWt reference DB-MHR design having a power density of 4.7 MW/m{sup 3}. Based on parametric and sensitivity study, it was determined that the maximum burnup (TRU consumption) can be obtained using optimum values of 200 {micro}m and 20% for the fuel kernel diameter and fuel packing fraction, respectively. These values were retained for most of the one-pass and two-pass design calculations; variation to the packing fraction was necessary for the second stage of the two-pass concept. Using a four-batch fuel management scheme for the one-pass DB-MHR core, it was possible to obtain a TRU consumption of 58% and a cycle length of 286 EFPD. By increasing the core power to 800 MWt and the power density to 6.2 MW/m{sup 3}, it was possible to increase the TRU consumption to 60%, although the cycle length decreased by {approx}64 days. The higher TRU consumption (burnup) is due to the reduction of the in-core decay of fissile Pu-241 to Am-241 relative to fission, arising from the higher power density (specific power), which made the fuel more reactivity over time. It was also found that the TRU consumption can be improved by utilizing axial fuel shuffling or by operating with lower material temperatures (colder core). Results also showed that the transmutation performance of the one-pass deep-burn concept is sensitive to the initial TRU vector, primarily because longer cooling time reduces the fissile content (Pu-241 specifically.) With a cooling time of 5 years, the TRU consumption increases to 67%, while conversely, with 20-year cooling the TRU consumption is about 58%. For the two-pass DB-MHR (TRU recycling option), a fuel packing fraction of about 30% is required in the second pass (the recycled TRU). It was found that using a heterogeneous core (homogeneous fuel element) concept, the TRU consumption is dependent on the cooling interval before the 2nd pass, again due to Pu-241 decay during the time lag between the first pass fuel discharge and the second pass fuel charge. With a cooling interval of 7 years (5 and 2 years before and after reprocessing) a TRU consumption of 55% is obtained. With an assumed ''no cooling'' interval, the TRU consumption is 63%. By using a cylindrical core to reduce neutron leakage, TRU consumption of the case with 7-year cooling interval increases to 58%. For a two-pass concept using a heterogeneous fuel element (and homogeneous core) with first and second pass volume ratio of 2:1, the TRU consumption is 62.4%. Finally, the repository loading benefits arising from the deep-burn and Inert Matrix Fuel (IMF) concepts were estimated and compared, for the same initial TRU vector. The DB-MHR concept resulted in slightly higher TRU consumption and repository loading benefit compared to the IMF concept (58.1% versus 55.1% for TRU consumption and 2.0 versus 1.6 for estimated repository loading benefit).« less

Eddy current position indicating apparatus for measuring displacements of core components of a liquid metal nuclear reactor

DOEpatents

Day, Clifford K.; Stringer, James L.

1977-01-01

Apparatus for measuring displacements of core components of a liquid metal fast breeder reactor by means of an eddy current probe. The active portion of the probe is located within a dry thimble which is supported on a stationary portion of the reactor core support structure. Split rings of metal, having a resistivity significantly different than sodium, are fixedly mounted on the core component to be monitored. The split rings are slidably positioned around, concentric with the probe and symmetrically situated along the axis of the probe so that motion of the ring along the axis of the probe produces a proportional change in the probes electrical output.

MEANS FOR COOLING REACTORS

DOEpatents

Wheeler, J.A.

1957-11-01

A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

Thermos reactors

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

Labrousse, M.; Lerouge, B.; Dupuy, G.

1978-04-01

THERMOS is a water reactor designed to provide hot water up to 120/sup 0/C for district heating or for desalination applications. It is a 100-MW reactor based on proven technology: oxide fuel plate elements, integrated primary circuit, and reactor vessel located in the bottom of a pool. As in swimming pool reactors, the pool is used for biological shielding, emergency core cooling, and fission product filtering (in case of an accident). Before economics, safety is the main characteristic of the concept: no fuel failure admitted, core under water in any accidental configuration, inspection of every ''nuclear'' component, and double-wall containment.

Deflection Measurements of a Thermally Simulated Nuclear Core Using a High-Resolution CCD-Camera

NASA Technical Reports Server (NTRS)

Stanojev, B. J.; Houts, M.

2004-01-01

Space fission systems under consideration for near-term missions all use compact. fast-spectrum reactor cores. Reactor dimensional change with increasing temperature, which affects neutron leakage. is the dominant source of reactivity feedback in these systems. Accurately measuring core dimensional changes during realistic non-nuclear testing is therefore necessary in predicting the system nuclear equivalent behavior. This paper discusses one key technique being evaluated for measuring such changes. The proposed technique is to use a Charged Couple Device (CCD) sensor to obtain deformation readings of electrically heated prototypic reactor core geometry. This paper introduces a technique by which a single high spatial resolution CCD camera is used to measure core deformation in Real-Time (RT). Initial system checkout results are presented along with a discussion on how additional cameras could be used to achieve a three- dimensional deformation profile of the core during test.

Autonomous Control of Nuclear Power Plants

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

Basher, H.

2003-10-20

A nuclear reactor is a complex system that requires highly sophisticated controllers to ensure that desired performance and safety can be achieved and maintained during its operations. Higher-demanding operational requirements such as reliability, lower environmental impacts, and improved performance under adverse conditions in nuclear power plants, coupled with the complexity and uncertainty of the models, necessitate the use of an increased level of autonomy in the control methods. In the opinion of many researchers, the tasks involved during nuclear reactor design and operation (e.g., design optimization, transient diagnosis, and core reload optimization) involve important human cognition and decisions that maymore » be more easily achieved with intelligent methods such as expert systems, fuzzy logic, neural networks, and genetic algorithms. Many experts in the field of control systems share the idea that a higher degree of autonomy in control of complex systems such as nuclear plants is more easily achievable through the integration of conventional control systems and the intelligent components. Researchers have investigated the feasibility of the integration of fuzzy logic, neural networks, genetic algorithms, and expert systems with the conventional control methods to achieve higher degrees of autonomy in different aspects of reactor operations such as reactor startup, shutdown in emergency situations, fault detection and diagnosis, nuclear reactor alarm processing and diagnosis, and reactor load-following operations, to name a few. With the advancement of new technologies and computing power, it is feasible to automate most of the nuclear reactor control and operation, which will result in increased safety and economical benefits. This study surveys current status, practices, and recent advances made towards developing autonomous control systems for nuclear reactors.« less

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-02-01

A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

Thermionic fast spectrum reactor-converter on the basis of multi-cell TFE

NASA Astrophysics Data System (ADS)

Ponomarev-Stepnoi, N. N.; Kompaniets, G. V.; Poliakov, D. N.; Stepennov, B. S.; Andreev, P. V.; Zhabotinsky, E. E.; Nikolaev, Yu. V.; Lapochkin, N. V.

2001-02-01

Today Russian experts have technological experience in development of in-core thermionic converters for reactors of space nuclear power plants. Such a converter contains nuclear fuel inside and really represents a fuel element of a reactor. Two types of reactors can be considered on the basis of these thermionic fuel elements: with thermal or intermediate neutron spectrum, and with fast neutron spectrum. The first type is characterized by the presence of moderator in core that ensures most economical usage of nuclear fuel. The estimation shows that moderated system is the most effective in the power range of about 5 ... 100 kWe. The power systems of higher level are characterized by larger dimensions due to the presence of moderator. The second type of reactor is considered for higher power levels. This power range is about hundreds kWe. Dimensions of the fast reactor and core configuration are determined by the necessity to ensure the required net output power, on the one hand, and the necessity to ensure critical state on the other hand. In the case of using in-core thermionic fuel elements of the specified design, minimal reactor output power is determined by reactor criticality condition, and maximum reactor power output is determined by specifications and launcher capabilities. In the present paper the effective multiplication factor of a fast spectrum reactor on the basis of a multi-cell TFE developed by ``Lutch'' is considered a function of the total number of TFEs in the reactor. The MCU Monte-Carlo code, developed in Russia (Alekseev, et al., 1991), was used for computations. TFE computational models are placed in the nodes of a uniform triangular lattice and surrounded with pressure vessel and a side reflector. Ordinary fuel pins without thermionic converters were used instead of some TFEs to optimize criticality parameters, dimensions and output power of the reactor. General weight parameters of the reactor are presented in the paper. .

Genome-Resolved Meta-Omics Ties Microbial Dynamics to Process Performance in Biotechnology for Thiocyanate Degradation

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

Kantor, Rose S.; Huddy, Robert J.; Iyer, Ramsunder

Remediation of industrial wastewater is important for preventing environmental contamination and allowing water reuse. Biological treatment for one industrial contaminant, thiocyanate (SCN - ), relies upon microbial hydrolysis, but this process is sensitive to high loadings. To examine the activity and stability of a microbial community over increasing SCN - loadings, we established and operated a continuous-flow bioreactor fed increasing loadings of SCN - . A second reactor was fed ammonium sulfate to mimic breakdown products of SCN - . Biomass was sampled from both reactors for metagenomics and metaproteomics, yielding a set of genomes for 144 bacteria and onemore » rotifer that constituted the abundant community in both reactors. We analyzed the metabolic potential and temporal dynamics of these organisms across the increasing loadings. In the SCN - reactor, Thiobacillus strains capable of SCN - degradation were highly abundant, whereas the ammonium sulfate reactor contained nitrifiers and heterotrophs capable of nitrate reduction. Key organisms in the SCN - reactor expressed proteins involved in SCN - degradation, sulfur oxidation, carbon fixation, and nitrogen removal. Lower performance at higher loadings was linked to changes in microbial community composition. This work provides an example of how meta-omics can increase our understanding of industrial wastewater treatment and inform iterative process design and development.« less

Genome-Resolved Meta-Omics Ties Microbial Dynamics to Process Performance in Biotechnology for Thiocyanate Degradation

DOE PAGES

Kantor, Rose S.; Huddy, Robert J.; Iyer, Ramsunder; ...

2017-01-31

Remediation of industrial wastewater is important for preventing environmental contamination and allowing water reuse. Biological treatment for one industrial contaminant, thiocyanate (SCN - ), relies upon microbial hydrolysis, but this process is sensitive to high loadings. To examine the activity and stability of a microbial community over increasing SCN - loadings, we established and operated a continuous-flow bioreactor fed increasing loadings of SCN - . A second reactor was fed ammonium sulfate to mimic breakdown products of SCN - . Biomass was sampled from both reactors for metagenomics and metaproteomics, yielding a set of genomes for 144 bacteria and onemore » rotifer that constituted the abundant community in both reactors. We analyzed the metabolic potential and temporal dynamics of these organisms across the increasing loadings. In the SCN - reactor, Thiobacillus strains capable of SCN - degradation were highly abundant, whereas the ammonium sulfate reactor contained nitrifiers and heterotrophs capable of nitrate reduction. Key organisms in the SCN - reactor expressed proteins involved in SCN - degradation, sulfur oxidation, carbon fixation, and nitrogen removal. Lower performance at higher loadings was linked to changes in microbial community composition. This work provides an example of how meta-omics can increase our understanding of industrial wastewater treatment and inform iterative process design and development.« less

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

Mechanical design of a light water breeder reactor

DOEpatents

Fauth, Jr., William L.; Jones, Daniel S.; Kolsun, George J.; Erbes, John G.; Brennan, John J.; Weissburg, James A.; Sharbaugh, John E.

1976-01-01

In a light water reactor system using the thorium-232 -- uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements.

A prototype experiment for cooperative monitoring of nuclear reactors with cubic meter scale antineutrino detectors

NASA Astrophysics Data System (ADS)

Bernstein, A.; Allen, M.; Bowden, N.; Brennan, J.; Carr, D. J.; Estrada, J.; Hagmann, C.; Lund, J. C.; Madden, N. W.; Winant, C. D.

2005-09-01

Our Lawrence Livermore National Laboratory/Sandia National Laboratories collaboration has deployed a cubic-meter-scale antineutrino detector to demonstrate non-intrusive and automatic monitoring of the power levels and plutonium content of a nuclear reactor. Reactor monitoring of this kind is required for all non-nuclear weapons states under the Nuclear Nonproliferation Treaty (NPT), and is implemented by the International Atomic Energy Agency (IAEA). Since the antineutrino count rate and energy spectrum depend on the relative yields of fissioning isotopes in the reactor core, changes in isotopic composition can be observed without ever directly accessing the core. Data from a cubic meter scale antineutrino detector, coupled with the well-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being operated in an illegitimate way. Our group has deployed a detector at the San Onofre reactor site in California to demonstrate this concept. This paper describes the concept and shows preliminary results from 8 months of operation.

Fluorescence-based monitoring of tracer and substrate distribution in an UASB reactor.

PubMed

Lou, S J; Tartakovsky, B; Zeng, Y; Wu, P; Guiot, S R

2006-11-01

In this work, rhodamine-related fluorescence was measured on-line at four reactor heights in order to study hydrodynamics within an upflow anaerobic sludge bed reactor. A linear dependence of the dispersion coefficient (D) on the upflow velocity was observed, while the influence of the organic loading rate (OLR) was insignificant. Furthermore, the Bodenstein number of the reactor loaded with granulated sludge was found to be position-dependent with the largest values measured at the bottom of the sludge bed. This trend was not observed in the reactor without sludge. Chemical oxygen demand (COD) and volatile fatty acid (VFA) concentrations were measured at the same reactor heights as in rhodamine tests using conventional off-line analytical methods and on-line multiwavelength fluorometry. Significant spatial COD and VFA gradients were observed at organic loading rates above 6g COD l(R)(-1)d(-1) and linear upflow velocities below 0.8m h(-1).

Neutronic Reactor Structure

DOEpatents

Vernon, H. C.; Weinberg, A. M.

1961-05-30

The neutronic reactor is comprised of a core consisting of natural uranium and heavy water with a K-factor greater than unity. The core is surrounded by a reflector consisting of natural uranium and ordinary water with a Kfactor less than unity. (AEC)

REACTOR CONTROL ROD OPERATING SYSTEM

DOEpatents

Miller, G.

1961-12-12

A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

Target-fueled nuclear reactor for medical isotope production

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

Coats, Richard L.; Parma, Edward J.

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7more » to 21 days.« less

Submersion criticality safety of tungsten-rhenium urania cermet fuel for space propulsion and power applications

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

A.E. Craft; R. C. O'Brien; S. D. Howe

Nuclear thermal rockets are the preferred propulsion technology for a manned mission to Mars, and tungsten–uranium oxide cermet fuels could provide significant performance and cost advantages for nuclear thermal rockets. A nuclear reactor intended for use in space must remain subcritical before and during launch, and must remain subcritical in launch abort scenarios where the reactor falls back to Earth and becomes submerged in terrestrial materials (including seawater, wet sand, or dry sand). Submersion increases reflection of neutrons and also thermalizes the neutron spectrum, which typically increases the reactivity of the core. This effect is typically very significant for compact,more » fast-spectrum reactors. This paper provides a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor with a range of fuel compositions. Each submersion case considers both the rhenium content in the matrix alloy and the uranium oxide volume fraction in the cermet. The inclusion of rhenium significantly improves the submersion criticality safety of the reactor. While increased uranium oxide content increases the reactivity of the core, it does not significantly affect the submersion behavior of the reactor. There is no significant difference in submersion behavior between reactors with rhenium distributed within the cermet matrix and reactors with a rhenium clad in the coolant channels. The combination of the flooding of the coolant channels in submersion scenarios and the presence of a significant amount of spectral shift absorbers (i.e. high rhenium concentration) further decreases reactivity for short reactor cores compared to longer cores.« less

Estimated inventory of radionuclides in Former Soviet Union Naval Reactors dumped in the Kara Sea and their associated health risk

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

Mount, M.E.; Layton, D.W.; Schwertz, N.L.

1993-05-01

Radionuclide inventories have bin estimated for the reactor cores, reactor components, and primary system corrosion products in the former Soviet Union naval reactors dumped at the Abrosimov Inlet, Tsivolka Inlet, Stepovoy Inlet, Techeniye Inlet, and Novaya Zemlya Depression sites in the Kara Sea between 1965 and 1988. For the time of disposal, the inventories are estimated at 17 to 66 kCi of actinides plus daughters and 1695 to 4782 kCi of fission products in the reactor cores, 917 to 1127 kCi of activation products in the reactor components, and 1.4 to 1.6 kCi of activation products in the primary systemmore » corrosion products. At the present time, the inventories are estimated to have decreased to 6 to 24 kCi of actinides plus daughters and 492 to 540 kCi of fission products in the reactor cores, 124 to 126 kCi of activation products in the reactor components, and 0.16 to 0.17 kCi of activation products in the primary system corrosion products. All actinide activities are estimated to be within a factor of two.« less

Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

NASA Technical Reports Server (NTRS)

Martin, James J.; Reid, Robert S.

2004-01-01

A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

Proliferation resistance of small modular reactors fuels

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

Polidoro, F.; Parozzi, F.; Fassnacht, F.

2013-07-01

In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. Inmore » the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.« less

Nuclear reactor internals alignment configuration

DOEpatents

Gilmore, Charles B [Greensburg, PA; Singleton, Norman R [Murrysville, PA

2009-11-10

An alignment system that employs jacking block assemblies and alignment posts around the periphery of the top plate of a nuclear reactor lower internals core shroud to align an upper core plate with the lower internals and the core shroud with the core barrel. The distal ends of the alignment posts are chamfered and are closely received within notches machined in the upper core plate at spaced locations around the outer circumference of the upper core plate. The jacking block assemblies are used to center the core shroud in the core barrel and the alignment posts assure the proper orientation of the upper core plate. The alignment posts may alternately be formed in the upper core plate and the notches may be formed in top plate.

High flux reactor

DOEpatents

Lake, James A.; Heath, Russell L.; Liebenthal, John L.; DeBoisblanc, Deslonde R.; Leyse, Carl F.; Parsons, Kent; Ryskamp, John M.; Wadkins, Robert P.; Harker, Yale D.; Fillmore, Gary N.; Oh, Chang H.

1988-01-01

A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

Adaptive Core Simulation Employing Discrete Inverse Theory - Part I: Theory

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

Abdel-Khalik, Hany S.; Turinsky, Paul J.

2005-07-15

Use of adaptive simulation is intended to improve the fidelity and robustness of important core attribute predictions such as core power distribution, thermal margins, and core reactivity. Adaptive simulation utilizes a selected set of past and current reactor measurements of reactor observables, i.e., in-core instrumentation readings, to adapt the simulation in a meaningful way. A meaningful adaption will result in high-fidelity and robust adapted core simulator models. To perform adaption, we propose an inverse theory approach in which the multitudes of input data to core simulators, i.e., reactor physics and thermal-hydraulic data, are to be adjusted to improve agreement withmore » measured observables while keeping core simulator models unadapted. At first glance, devising such adaption for typical core simulators with millions of input and observables data would spawn not only several prohibitive challenges but also numerous disparaging concerns. The challenges include the computational burdens of the sensitivity-type calculations required to construct Jacobian operators for the core simulator models. Also, the computational burdens of the uncertainty-type calculations required to estimate the uncertainty information of core simulator input data present a demanding challenge. The concerns however are mainly related to the reliability of the adjusted input data. The methodologies of adaptive simulation are well established in the literature of data adjustment. We adopt the same general framework for data adjustment; however, we refrain from solving the fundamental adjustment equations in a conventional manner. We demonstrate the use of our so-called Efficient Subspace Methods (ESMs) to overcome the computational and storage burdens associated with the core adaption problem. We illustrate the successful use of ESM-based adaptive techniques for a typical boiling water reactor core simulator adaption problem.« less

Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

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

R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag

2012-04-01

The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather thanmore » graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water Reactor (PWR) assemblies. In addition to consideration of this 'naive' use of TRISO fuel in LWRs, several refined options are briefly examined and others are identified for further consideration including the use of advanced, high density fuel forms and larger kernel diameters and TRISO packing fractions. The combination of 800 {micro}m diameter kernels of 20% enriched UN and 50% TRISO packing fraction yielded reactivity sufficient to achieve comparable burnup to present-day PWR fuel.« less

77 FR 64148 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-18

... Regulatory Guides (RG) RG 1.79, ````Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors,'' Revision 2 and RG 1.79.1, ``Initial Test Program of Emergency Core Cooling Systems for...

The effects of stainless steel radial reflector on core reactivity for small modular reactor

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

Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr; Cho, Sung Ju, E-mail: sungju@knfc.co.kr

Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5more » w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.« less

LOADING DEVICE

DOEpatents

Ohlinger, L.A.

1958-10-01

A device is presented for loading or charging bodies of fissionable material into a reactor. This device consists of a car, mounted on tracks, into which the fissionable materials may be placed at a remote area, transported to the reactor, and inserted without danger to the operating personnel. The car has mounted on it a heavily shielded magazine for holding a number of the radioactive bodies. The magazine is of a U-shaped configuration and is inclined to the horizontal plane, with a cap covering the elevated open end, and a remotely operated plunger at the lower, closed end. After the fissionable bodies are loaded in the magazine and transported to the reactor, the plunger inserts the body at the lower end of the magazine into the reactor, then is withdrawn, thereby allowing gravity to roll the remaining bodies into position for successive loading in a similar manner.

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

Laser anemometry measurements of natural circulation flow in a scale model PWR reactor system. [Pressurized Water Reactor

NASA Technical Reports Server (NTRS)

Kadambi, J. R.; Schneider, S. J.; Stewart, W. A.

1986-01-01

The natural circulation of a single phase fluid in a scale model of a pressurized water reactor system during a postulated grade core accident is analyzed. The fluids utilized were water and SF6. The design of the reactor model and the similitude requirements are described. Four LDA tests were conducted: water with 28 kW of heat in the simulated core, with and without the participation of simulated steam generators; water with 28 kW of heat in the simulated core, with the participation of simulated steam generators and with cold upflow of 12 lbm/min from the lower plenum; and SF6 with 0.9 kW of heat in the simulated core and without the participation of the simulated steam generators. For the water tests, the velocity of the water in the center of the core increases with vertical height and continues to increase in the upper plenum. For SF6, it is observed that the velocities are an order of magnitude higher than those of water; however, the velocity patterns are similar.

Study on vibration characteristics and fault diagnosis method of oil-immersed flat wave reactor in Arctic area converter station

NASA Astrophysics Data System (ADS)

Lai, Wenqing; Wang, Yuandong; Li, Wenpeng; Sun, Guang; Qu, Guomin; Cui, Shigang; Li, Mengke; Wang, Yongqiang

2017-10-01

Based on long term vibration monitoring of the No.2 oil-immersed fat wave reactor in the ±500kV converter station in East Mongolia, the vibration signals in normal state and in core loose fault state were saved. Through the time-frequency analysis of the signals, the vibration characteristics of the core loose fault were obtained, and a fault diagnosis method based on the dual tree complex wavelet (DT-CWT) and support vector machine (SVM) was proposed. The vibration signals were analyzed by DT-CWT, and the energy entropy of the vibration signals were taken as the feature vector; the support vector machine was used to train and test the feature vector, and the accurate identification of the core loose fault of the flat wave reactor was realized. Through the identification of many groups of normal and core loose fault state vibration signals, the diagnostic accuracy of the result reached 97.36%. The effectiveness and accuracy of the method in the fault diagnosis of the flat wave reactor core is verified.

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

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

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

Catalytic wet air oxidation of coke-plant wastewater on ruthenium-based eggshell catalysts in a bubbling bed reactor.

PubMed

Yang, M; Sun, Y; Xu, A H; Lu, X Y; Du, H Z; Sun, C L; Li, C

2007-07-01

Catalytic wet air of coke-plant wastewater was studied in a bubbling bed reactor. Two types of supported Ru-based catalysts, eggshell and uniform catalysts, were employed. Compared with the results in the wet air oxidation of coke-plant wastewater, supported Ru uniform catalysts showed high activity for chemical oxygen demand (COD) and ammonia/ammonium compounds (NH3-N) removal at temperature of 250 degrees C and pressure of 4.8 MPa, and it has been demonstrated that the catalytic activity of uniform catalyst depended strongly on the distribution of active sites of Ru on catalyst. Compared to the corresponding uniform catalysts with the same Ru loading (0.25 wt.% and 0.1 wt.%, respectively), the eggshell catalysts showed higher activities for CODcr removal and much higher activities for NH3-N degradation. The high activity of eggshell catalyst for treatment of coke-plant wastewater can be attributed to the higher density of active Ru sites in the shell layer than that of the corresponding uniform catalyst with the same Ru loading. It has been also evidenced that the active Ru sites in the internal core of uniform catalyst have very little or no contribution to CODcr and NH3-N removal in the total oxidation of coke-plant wastewater.

Status of the MeLoDIE experiment, an advanced device for the study of the irradiation creep of LWR cladding with full online capabilities

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

Guimbal, P.; Huotilainen, S.; Taehtinen, S.

2015-07-01

As a prototype of future instrumented material experiments in the Jules Horowitz Reactor (JHR), the MELODIE project was launched in 2009 by the CEA in collaboration with VTT. Being designed as a biaxial creep experiment with online capability, MELODIE is able to apply an online-controlled biaxial loading on a LWR clad sample up to 120 MPa and to perform an online measurement of its biaxial deformation. An important experimental challenge was to perform reliably accurate measurements under the high nuclear heat load of in-core locations while keeping within their tight space. For that purpose, specific sensors were co-designed with andmore » built by IFE Halden. Manufacturing of the MELODIE components was completed one year ago. The complexity of its in-pile section and of the pressurization system requested a step-by-step tuning of the setup. The toughest part of this process dealt with the Diameter gauge which required a partial redesign to take into account unexpected and unwanted electromagnetic interactions with the hosting device. Final cold performance tests of the on-board instrumentation will be presented. The MELODIE device is now ready and irradiation should start in OSIRIS reactor this spring. (authors)« less

METHOD AND APPARATUS FOR IMPROVING PERFORMANCE OF A FAST REACTOR

DOEpatents

Koch, L.J.

1959-01-20

A specific arrangement of the fertile material and fissionable material in the active portion of a fast reactor to achieve improvement in performance and to effectively lower the operating temperatures in the center of the reactor is described. According to this invention a group of fuel elements containing fissionable material are assembled to form a hollow fuel core. Elements containing a fertile material, such as depleted uranium, are inserted into the interior of the fuel core to form a central blanket. Additional elemenis of fertile material are arranged about the fuel core to form outer blankets which in tunn are surrounded by a reflector. This arrangement of fuel core and blankets results in substantial flattening of the flux pattern.

POWER GENERATION FROM LIQUID METAL NUCLEAR FUEL

DOEpatents

Dwyer, O.E.

1958-12-23

A nuclear reactor system is described wherein the reactor is the type using a liquid metal fuel, such as a dispersion of fissile material in bismuth. The reactor is designed ln the form of a closed loop having a core sectlon and heat exchanger sections. The liquid fuel is clrculated through the loop undergoing flssion in the core section to produce heat energy and transferrlng this heat energy to secondary fluids in the heat exchanger sections. The fission in the core may be produced by a separate neutron source or by a selfsustained chain reaction of the liquid fuel present in the core section. Additional auxiliary heat exchangers are used in the system to convert water into steam which drives a turbine.

Heat dissipating nuclear reactor with metal liner

DOEpatents

Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

1985-11-21

A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Heat dissipating nuclear reactor with metal liner

DOEpatents

Gluekler, Emil L.; Hunsbedt, Anstein; Lazarus, Jonathan D.

1987-01-01

Disclosed is a nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Cavity temperature and flow characteristics in a gas-core test reactor

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Compact power reactor

DOEpatents

Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

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.

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

DOE PAGES

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin; ...

2017-12-22

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

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

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

Planned Destruction of Metal-Core Reactor: Simulation of Catastrophic Accidents and New Experimental Possibilities

NASA Astrophysics Data System (ADS)

Vorontsov, S. V.; Kuvshinov, M. I.; Narozhnyi, A. T.; Popov, V. A.; Solov'ev, V. P.; Yuferev, V. I.

2017-12-01

A reactor with a destructible core (RIR reactor) generating a pulse with an output of 1.5 × 1019 fissions and a full width at half maximum of 2.5 μs was developed and tested at VNIIEF. In the course of investigation, a computational-experimental method for laboratory calibration of the reactor was created and worked out. This method ensures a high accuracy of predicting the energy release in a real experiment with excess reactivity of 3βeff above prompt criticality. A transportable explosion-proof chamber was also developed, which ensures the safe localization of explosion products of the core of small-sized nuclear devices and charges of high explosives with equivalent mass of up to 100 kg of TNT.

An Overview of Reactor Concepts, a Survey of Reactor Designs.

DTIC Science & Technology

1985-02-01

may be very different. HTGRs may use highly enriched uranium, thereby yielding better fuel economy and a reduc- tion of the actual core size for a...specific power level. The HTGR core may have fuel and control rods placed in graphite arrays similar to PWR core con- figuration, or they may have fuel ...rods are pulled out. A Peach Bottom core design is another HTGR design. This design is featured by the fuel pin’s ability to purge itself of fission

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

Soldatov, Alexey I.

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

Thermionic switched self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

1989-01-01

A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium-Plutonium Carbide based fuel and 300 MWt Power

NASA Astrophysics Data System (ADS)

Clief Pattipawaej, Sandro; Su'ud, Zaki

2017-01-01

A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

Effect of non-feeding period length on the intermittent operation of UASB reactors treating dairy effluents.

PubMed

Coelho, N M; Rodrigues, A A; Arroja, L M; Capela, I F

2007-02-01

Recent environmental concerns have prompted a re-evaluation of conventional management strategies and refueled the search of innovative waste management practices. In this sense, the anaerobic digestion of both fat and the remaining complex organic matter present in dairy wastewaters is attractive, although the continuous operation of high rate anaerobic processes treating this type of wastewaters causes the failure of the process. This work accesses the influence of non-feeding period length on the intermittent operation of mesophilic UASB reactors treating dairy wastewater, in order to allow the biological degradation to catch up with adsorption phenomenon. During the experiments, two UASB reactors were subject to three organic loading rates, ranging from 6 to 12 g(COD) x L(-1) x d(-1), with the same daily load applied to both reactors, each one with a different non-feeding period. Both reactors showed good COD removal efficiencies (87-92%). A material balance for COD in the reactors during the feeding and non-feeding periods showed the importance of the feedless period, which allowed the biomass to degrade substrate that was accumulated during the feeding period. The reactor with the longest non-feeding period had a better performance, which resulted in a higher methane production and adsorption capacity for the same organic load applied with a consequent less accumulation of substrate into the biomass. In addition, both reactors had a stable operation for the organic load of 12 g(COD) x L(-1) x d(-1), which is higher than the maximum applicable load reported in literature for continuous systems (3-6 g(COD) x L(-1) x d(-1)). (c) 2006 Wiley Periodicals, Inc.

Dry fermentation of manure with straw in continuous plug flow reactor: Reactor development and process stability at different loading rates.

PubMed

Patinvoh, Regina J; Kalantar Mehrjerdi, Adib; Sárvári Horváth, Ilona; Taherzadeh, Mohammad J

2017-01-01

In this work, a plug flow reactor was developed for continuous dry digestion processes and its efficiency was investigated using untreated manure bedded with straw at 22% total solids content. This newly developed reactor worked successfully for 230days at increasing organic loading rates of 2.8, 4.2 and 6gVS/L/d and retention times of 60, 40 and 28days, respectively. Organic loading rates up to 4.2gVS/L/d gave a better process stability, with methane yields up to 0.163LCH 4 /gVS added /d which is 56% of the theoretical yield. Further increase of organic loading rate to 6gVS/L/d caused process instability with lower volatile solid removal efficiency and cellulose degradation. Copyright © 2016 Elsevier Ltd. All rights reserved.

Intelligent uranium fission converter for neutron production on the periphery of the nuclear reactor core (MARIA reactor in Swierk - Poland)

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

Gryzinski, M.A.; Wielgosz, M.

The multipurpose, high flux research reactor MARIA in Otwock - Swierk is an open-pool type, water and beryllium moderated and graphite reflected. There are two not occupied experimental H1 and H2 horizontal channels with complex of empty rooms beside them. Making use of these two channels is not in conflict with other research or commercial employing channels. They can work simultaneously, moreover commercial channels covers the cost of reactor working. Such conditions give beneficial possibility of creating epithermal neutron stand for researches in various field at the horizontal channel H2 of MARIA reactor (co-organization of research at H1 channel ismore » additionally planned). At the front of experimental channels the neutron flux is strongly thermalized - neutrons with energies above 0.625 eV constitute only ∼2% of the total flux. This thermalized neutron flux will be used to achieve high flux of epithermal neutrons at the level of 2x10{sup 9} n cm{sup -2}s{sup -1} by uranium neutron converter (fast neutron production - conversion of reactor core thermal neutrons to fast neutrons - and then filtering, moderating and finally cutting of unwanted gamma radiation). The intelligent converter will be placed in the reactor pool, near the front of the H2 channel. It will replace one graphite block at the periphery of MARIA graphite reflector. The converter will consist of 20 fuel elements - low enriched uranium plates. A fuel plate will be a part which will measure 110 mm wide by 380 mm long and will consist of a thin layer of uranium sealed between two aluminium plates. These plates, once assembled, form the fuel element used in converter. The plates will be positioned vertically. There are several important requirements which should be taken into account at the converter design stage: -maximum efficiency of the converter for neutrons conversion, -cooling of the converter need to be integrated with the cooling circuit of the reactor pool and if needed equipped with self-cooling system (enhanced comparing to the cooling properties inherent with regular rector pool water flows), -proper cooling conditions can be ensured by an appropriate water flow, so the resistance to flow has to be optimised, -the requirement of the minimum resistance to water flow leads to the openwork design of the fuel element separator, which, on the other hand, has to be strong enough to ensure the needed strength for mechanical load due to the fuel weight and forces associated with the water flow, -the possibility of changing beam and flux qualities by rotating the converter or repositioning the converter plates by moving or replacing with another materials. In order to minimize the neutron activation of the fuel in the converter, the possibility was predicted to remove the converter and to replace it with an aluminium dummy for the time when the beam at the channel H2 is not used. This means that both, the converter and the dummy, have to be easily removable from the converter socket. There has to be also the place in the water pool, near the research stand or in technological pool, where the converter can be safely stored (this place have to be proper for operation with plates i.e. changing amount of plates). Thermal and neutron load of the fuel plates in the converter will be inhomogeneous. In order to equalize these loads, the converter should be designed in such way that it would be possible to change the order of fuel plates. Moreover replacing the amount of the plates gives the opportunity to obtain different fluxes of neutrons (quantitatively and qualitatively i.e. energetically). The project of the converter is based on Monte Carlo calculation concerning neutron production and on Computational Fluid Dynamics (CFD) i.e. modelling of converter for thermodynamical aspects. (authors)« less

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

PubMed

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

PubMed Central

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

Method for automatically scramming a nuclear reactor

DOEpatents

Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.

2005-12-27

An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

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

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« 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.

Demonstrating Hybrid Heat Transport and Energy Conversion System Performance Characterization Using Intelligent Control Systems

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

Ostrum, Lee; Manic, Milos

The debate continues on the magnitude and validity of climate change caused by human activities. However, there is no debate about the need to make buildings, modes of transportation, factories, and homes as energy efficient as possible. Given that climate change could occur with the wasteful use of fossil fuel and the fact that fossil energy costs could and will swing wildly, it is imperative that every effort be made to utilize energy sources to their fullest. Hybrid energy systems (HES) are two or more separate energy producers used together to produce energy commodities. The HES this report focuses onmore » is the use of nuclear reactor waste heat as a source of further energy utilization. Nuclear reactors use a fluid to cool the core and produce the steam needed for the production of electricity. Traditionally this steam, or coolant, is used to convert the energy then cooled elsewhere. The heat is released into the environment without being used further. By adding technologies to nuclear reactors to use the wasted heat, a system can be developed to make more than just electricity and allow for loading following capabilities.« less

Operational performance of the three bean salad control algorithm on the ACRR (Annular Core Research Reactor)

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

Ball, R.M.; Madaras, J.J.; Trowbridge, F.R. Jr.

Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute. 3 refs., 4 figs., 1 tab.

Fluidic self-actuating control assembly

DOEpatents

Grantz, Alan L.

1979-01-01

A fluidic self-actuating control assembly for use in a reactor wherein no external control inputs are required to actuate (scram) the system. The assembly is constructed to scram upon sensing either a sudden depressurization of reactor inlet flow or a sudden increase in core neutron flux. A fluidic control system senses abnormal flow or neutron flux transients and actuates the system, whereupon assembly coolant flow reverses, forcing absorber balls into the reactor core region.

ATR LEU fuel and burnable absorber neutronics performance optimization by fuel meat thickness variation

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

Chang, G.S.

2008-07-15

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U-235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core th and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.381 mm and the same U-235 enrichment (19.7 wt%) can be used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.5 mil) to 0.343 mm (13.5 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, 0.8g of a burnable absorber, Boron-10, was added in the inner and outer plates to reduce the initial excess reactivity, and the inner/outer heat flux more effectively. The optimized LEU relative radial fission heat flux profile is bounded by the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Updated Final Safety Analysis Report (UFSAR) safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores. (author)« less

An underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

NASA Astrophysics Data System (ADS)

Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has been fulfilled. From the result analysis, it can be concluded that the model of calculation result of neutron dose rate for HTGR-10 core has met the required radiation safety standards.

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

Some features of the effect the pH value and the physicochemical properties of boric acid have on mass transfer in a VVER reactor's core

NASA Astrophysics Data System (ADS)

Gavrilov, A. V.; Kritskii, V. G.; Rodionov, Yu. A.; Berezina, I. G.

2013-07-01

Certain features of the effect of boric acid in the reactor coolant of nuclear power installations equipped with a VVER-440 reactor on mass transfer in the reactor core are considered. It is determined that formation of boric acid polyborate complexes begins under field conditions at a temperature of 300°C when the boric acid concentration is equal to around 0.065 mol/L (4 g/L). Operations for decontaminating the reactor coolant system entail a growth of corrosion product concentration in the coolant, which gives rise to formation of iron borates in the zones where subcooled boiling of coolant takes place and to the effect of axial offset anomalies. A model for simulating variation of pressure drop in a VVER-440 reactor's core that has invariable parameters during the entire fuel campaign is developed by additionally taking into account the concentrations of boric acid polyborate complexes and the quantity of corrosion products (Fe, Ni) represented by the ratio of their solubilities.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Daniels, F.

1957-10-15

Gas-cooled solid-moderator type reactors wherein the fissionable fuel and moderator materials are each in the form of solid pebbles, or discrete particles, and are substantially homogeneously mixed in the proper proportion and placed within the core of the reactor are described. The shape of these discrete particles must be such that voids are present between them when mixed together. Helium enters the bottom of the core and passes through the voids between the fuel and moderator particles to absorb the heat generated by the chain reaction. The hot helium gas is drawn off the top of the core and may be passed through a heat exchanger to produce steam.

Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

NASA Astrophysics Data System (ADS)

Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

2017-09-01

VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

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.

Study the Cyclic Plasticity Behavior of 508 LAS under Constant, Variable and Grid-Load-Following Loading Cycles for Fatigue Evaluation of PWR Components

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

Mohanty, Subhasish; Barua, Bipul; Soppet, William K.

This report provides an update of an earlier assessment of environmentally assisted fatigue for components in light water reactors. This report is a deliverable in September 2016 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2016 report, we presented a detailed thermal-mechanical stress analysis model for simulating the stress-strain state of a reactor pressure vessel and its nozzles under grid-load-following conditions. In this report, we provide stress-controlled fatigue test data for 508 LAS base metal alloy under different loading amplitudes (constant, variable, and random grid-load-following) and environmental conditions (in airmore » or pressurized water reactor coolant water at 300°C). Also presented is a cyclic plasticity-based analytical model that can simultaneously capture the amplitude and time dependency of the component behavior under fatigue loading. Results related to both amplitude-dependent and amplitude-independent parameters are presented. The validation results for the analytical/mechanistic model are discussed. This report provides guidance for estimating time-dependent, amplitude-independent parameters related to material behavior under different service conditions. The developed mechanistic models and the reported material parameters can be used to conduct more accurate fatigue and ratcheting evaluation of reactor components.« less

Benchmark gas core critical experiment.

NASA Technical Reports Server (NTRS)

Kunze, J. F.; Lofthouse, J. H.; Cooper, C. G.; Hyland, R. E.

1972-01-01

A critical experiment with spherical symmetry has been conducted on the gas core nuclear reactor concept. The nonspherical perturbations in the experiment were evaluated experimentally and produce corrections to the observed eigenvalue of approximately 1% delta k. The reactor consisted of a low density, central uranium hexafluoride gaseous core, surrounded by an annulus of void or low density hydrocarbon, which in turn was surrounded with a 97-cm-thick heavy water reflector.

Minimizing or eliminating refueling of nuclear reactor

DOEpatents

Doncals, Richard A.; Paik, Nam-Chin; Andre, Sandra V.; Porter, Charles A.; Rathbun, Roy W.; Schwallie, Ambrose L.; Petras, Diane S.

1989-01-01

Demand for refueling of a liquid metal fast nuclear reactor having a life of 30 years is eliminated or reduced to intervals of at least 10 years by operating the reactor at a low linear-power density, typically 2.5 kw/ft of fuel rod, rather than 7.5 or 15 kw/ft, which is the prior art practice. So that power of the same magnitude as for prior art reactors is produced, the volume of the core is increased. In addition, the height of the core and it diameter are dimensioned so that the ratio of the height to the diameter approximates 1 to the extent practicable considering the requirement of control and that the pressure drop in the coolant shall not be excessive. The surface area of a cylinder of given volume is a minimum if the ratio of the height to the diameter is 1. By minimizing the surface area, the leakage of neutrons is reduced. By reducing the linear-power density, increasing core volume, reducing fissile enrichment and optimizing core geometry, internal-core breeding of fissionable fuel is substantially enhanced. As a result, core operational life, limited by control worth requirements and fuel burnup capability, is extended up to 30 years of continuous power operation.

IMPROVEMENTS RELATING TO NUCLEAR REACTOR CORE

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

Bell, F.R.

1963-03-01

A nuclear reactor core composed of a number of stacked horizontal layers is described. Each layer is made up of elements of moderator material of equal height and of generally hexagonal cross-section. Each element has holes containing nuclear fuel and separate ones for coolant. (C.E.S.)

Measurement of neutron spectra in the experimental reactor LR-0

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

Prenosil, Vaclav; Mravec, Filip; Veskrna, Martin

2015-07-01

The measurement of fast neutron fluxes is important in many areas of nuclear technology. It affects the stability of the reactor structural components, performance of fuel, and also the fuel manner. The experiments performed at the LR-0 reactor were in the past focused on the measurement of neutron field far from the core, in reactor pressure vessel simulator or in biological shielding simulator. In the present the measurement in closer regions to core became more important, especially measurements in structural components like reactor baffle. This importance increases with both reactor power increase and also long term operation. Other important taskmore » is an increasing need for the measurement close to the fuel. The spectra near the fuel are aimed due to the planned measurements with the FLIBE salt, in FHR / MSR research, where one of the task is the measurement of the neutron spectra in it. In both types of experiments there is strong demand for high working count rate. The high count rate is caused mainly by high gamma background and by high fluxes. The fluxes in core or in its vicinity are relatively high to ensure safe reactor operation. This request is met in the digital spectroscopic apparatus. All experiments were realized in the LR-0 reactor. It is an extremely flexible light water zero-power research reactor, operated by the Research Center Rez (Czech Republic). (authors)« less

A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors

NASA Technical Reports Server (NTRS)

Anghaie, S.; Chen, G.

1996-01-01

A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high efficiency in the gas core reactors. The model is also used to predict the convective and radiation heat fluxes for the gas core reactors. The maximum value of heat flux occurs at the exit of the reactor core. Radiation heat flux increases with higher wall temperature. This behavior is due to the fact that the radiative heat flux is strongly dependent on wall temperature. This study also found that at temperature close to 3500 K the radiative heat flux is comparable with the convective heat flux in a uranium fluoride failed gas core reactor.

An investigation into the feasibility of thorium fuels utilization in seed-blanket configurations for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

Thorium is one of the elements that needs to be explored for nuclear fuel research and development. One of the popular core configurations of thorium fuel is seed-blanket configuration or also known as Radkowsky Thorium Fuel concept. The seed will act as a supplier of neutrons, which will be placed inside of the core. The blanket, on the other hand, is the consumer of neutrons that is located at outermost of the core. In this work, a neutronic analysis of seed-blanket configuration for the TRIGA PUSPATI Reactor (RTP) is carried out using Monte Carlo method. The reactor, which has been operated since 1982 use uranium zirconium hydride (U-ZrH1.6) as the fuel and have multiple uranium weight which are 8.5, 12 and 20 wt.%. The pool type reactor is one and only research reactor that located in Malaysia. The design of core included the Uranium Zirconium Hydride located at the centre of the core that will act as the seed to supply neutron. The thorium oxide that will act as blanket situated outside of seed region will receive neutron to transmute 232Th to 233U. The neutron multiplication factor or criticality of each configuration is estimated. Results show that the highest initial criticality achieved is 1.30153.

Parallel Algorithms for Monte Carlo Particle Transport Simulation on Exascale Computing Architectures

NASA Astrophysics Data System (ADS)

Romano, Paul Kollath

Monte Carlo particle transport methods are being considered as a viable option for high-fidelity simulation of nuclear reactors. While Monte Carlo methods offer several potential advantages over deterministic methods, there are a number of algorithmic shortcomings that would prevent their immediate adoption for full-core analyses. In this thesis, algorithms are proposed both to ameliorate the degradation in parallel efficiency typically observed for large numbers of processors and to offer a means of decomposing large tally data that will be needed for reactor analysis. A nearest-neighbor fission bank algorithm was proposed and subsequently implemented in the OpenMC Monte Carlo code. A theoretical analysis of the communication pattern shows that the expected cost is O( N ) whereas traditional fission bank algorithms are O(N) at best. The algorithm was tested on two supercomputers, the Intrepid Blue Gene/P and the Titan Cray XK7, and demonstrated nearly linear parallel scaling up to 163,840 processor cores on a full-core benchmark problem. An algorithm for reducing network communication arising from tally reduction was analyzed and implemented in OpenMC. The proposed algorithm groups only particle histories on a single processor into batches for tally purposes---in doing so it prevents all network communication for tallies until the very end of the simulation. The algorithm was tested, again on a full-core benchmark, and shown to reduce network communication substantially. A model was developed to predict the impact of load imbalances on the performance of domain decomposed simulations. The analysis demonstrated that load imbalances in domain decomposed simulations arise from two distinct phenomena: non-uniform particle densities and non-uniform spatial leakage. The dominant performance penalty for domain decomposition was shown to come from these physical effects rather than insufficient network bandwidth or high latency. The model predictions were verified with measured data from simulations in OpenMC on a full-core benchmark problem. Finally, a novel algorithm for decomposing large tally data was proposed, analyzed, and implemented/tested in OpenMC. The algorithm relies on disjoint sets of compute processes and tally servers. The analysis showed that for a range of parameters relevant to LWR analysis, the tally server algorithm should perform with minimal overhead. Tests were performed on Intrepid and Titan and demonstrated that the algorithm did indeed perform well over a wide range of parameters. (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

NASA Technical Reports Server (NTRS)

Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

1996-01-01

Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the strength of the negative reactivity feedback in the UTVR, it is found that external reactivity insertions alone are inadequate for bringing about significant power level changes during normal reactor operations. Additional methods of reactivity control such as variations in the gaseous fuel mass flow rate, are needed to achieve the desired power level oontrol.

Determination and Variation of Core Bacterial Community in a Two-Stage Full-Scale Anaerobic Reactor Treating High-Strength Pharmaceutical Wastewater.

PubMed

Ma, Haijun; Ye, Lin; Hu, Haidong; Zhang, Lulu; Ding, Lili; Ren, Hongqiang

2017-10-28

Knowledge on the functional characteristics and temporal variation of anaerobic bacterial populations is important for better understanding of the microbial process of two-stage anaerobic reactors. However, owing to the high diversity of anaerobic bacteria, close attention should be prioritized to the frequently abundant bacteria that were defined as core bacteria and putatively functionally important. In this study, using MiSeq sequencing technology, the core bacterial community of 98 operational taxonomic units (OTUs) was determined in a two-stage upflow blanket filter reactor treating pharmaceutical wastewater. The core bacterial community accounted for 61.66% of the total sequences and accurately predicted the sample location in the principal coordinates analysis scatter plot as the total bacterial OTUs did. The core bacterial community in the first-stage (FS) and second-stage (SS) reactors were generally distinct, in that the FS core bacterial community was indicated to be more related to a higher-level fermentation process, and the SS core bacterial community contained more microbes in syntrophic cooperation with methanogens. Moreover, the different responses of the FS and SS core bacterial communities to the temperature shock and influent disturbance caused by solid contamination were fully investigated. Co-occurring analysis at the Order level implied that Bacteroidales, Selenomonadales, Anaerolineales, Syneristales, and Thermotogales might play key roles in anaerobic digestion due to their high abundance and tight correlation with other microbes. These findings advance our knowledge about the core bacterial community and its temporal variability for future comparative research and improvement of the two-stage anaerobic system operation.

Anaerobic degradation of coconut husk leachate using UASB-reactor.

PubMed

Neena, C; Ambily, P S; Jisha, M S

2007-07-01

Reffing of coconut husk, the majorprocess in quality coir fibre extraction, causes serious pollution with brackish water lagoons of Kerala. An attempt is made to treat the coconut husk leachate by using a laboratory scale UASB-reactor The experiment was conducted with loading of leachate from 1 kg of fresh coconut husk. The anaerobic treatment was done continuously The parameters like VFA, pH, COD and polyphenols were analysed regularly during the evaluation of the reactor performance. The polyphenol, VFA and COD were diminished gradually with time. The pH of the reactor during the study was found to be in the range of 6-8. The biogas production was increased with loading and about 82% of the total COD/kg husk could be converted to biogas. The maximum polyphenol loading in the reactor was reached to about 298.51 mg/l of husk.

Simulation numerique de l'effet du reflecteur radial sur les cellules rep en utilisant les codes DRAGON et DONJON

NASA Astrophysics Data System (ADS)

Bejaoui, Najoua

The pressurized water nuclear reactors (PWRs) is the largest fleet of nuclear reactors in operation around the world. Although these reactors have been studied extensively by designers and operators using efficient numerical methods, there are still some calculation weaknesses, given the geometric complexity of the core, still unresolved such as the analysis of the neutron flux's behavior at the core-reflector interface. The standard calculation scheme is a two steps process. In the first step, a detailed calculation at the assembly level with reflective boundary conditions, provides homogenized cross-sections for the assemblies, condensed to a reduced number of groups; this step is called the lattice calculation. The second step uses homogenized properties in each assemblies to calculate reactor properties at the core level. This step is called the full-core calculation or whole-core calculation. This decoupling of the two calculation steps is the origin of methodological bias particularly at the interface core reflector: the periodicity hypothesis used to calculate cross section librairies becomes less pertinent for assemblies that are adjacent to the reflector generally represented by these two models: thus the introduction of equivalent reflector or albedo matrices. The reflector helps to slowdown neutrons leaving the reactor and returning them to the core. This effect leads to two fission peaks in fuel assemblies localised at the core/reflector interface, the fission rate increasing due to the greater proportion of reentrant neutrons. This change in the neutron spectrum arises deep inside the fuel located on the outskirts of the core. To remedy this we simulated a peripheral assembly reflected with TMI-PWR reflector and developed an advanced calculation scheme that takes into account the environment of the peripheral assemblies and generate equivalent neutronic properties for the reflector. This scheme is tested on a core without control mechanisms and charged with fresh fuel. The results of this study showed that explicit representation of reflector and calculation of peripheral assembly with our advanced scheme allow corrections to the energy spectrum at the core interface and increase the peripheral power by up to 12% compared with that of the reference scheme.

Thermal margin protection system for a nuclear reactor

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

Musick, C.R.

1974-02-12

A thermal margin protection system for a nuclear reactor is described where the coolant flow flow trip point and the calculated thermal margin trip point are switched simultaneously and the thermal limit locus is made more restrictive as the allowable flow rate is decreased. The invention is characterized by calculation of the thermal limit Locus in response to applied signals which accurately represent reactor cold leg temperature and core power; cold leg temperature being corrected for stratification before being utilized and reactor power signals commensurate with power as a function of measured neutron flux and thermal energy added to themore » coolant being auctioneered to select the more conservative measure of power. The invention further comprises the compensation of the selected core power signal for the effects of core radial peaking factor under maximum coolant flow conditions. (Official Oazette)« less

Gamma thermometer based reactor core liquid level detector

DOEpatents

Burns, Thomas J.

1983-01-01

A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

Performance assessment of conventional and base-isolated nuclear power plants for earthquake and blast loadings

NASA Astrophysics Data System (ADS)

Huang, Yin-Nan

Nuclear power plants (NPPs) and spent nuclear fuel (SNF) are required by code and regulations to be designed for a family of extreme events, including very rare earthquake shaking, loss of coolant accidents, and tornado-borne missile impacts. Blast loading due to malevolent attack became a design consideration for NPPs and SNF after the terrorist attacks of September 11, 2001. The studies presented in this dissertation assess the performance of sample conventional and base isolated NPP reactor buildings subjected to seismic effects and blast loadings. The response of the sample reactor building to tornado-borne missile impacts and internal events (e.g., loss of coolant accidents) will not change if the building is base isolated and so these hazards were not considered. The sample NPP reactor building studied in this dissertation is composed of containment and internal structures with a total weight of approximately 75,000 tons. Four configurations of the reactor building are studied, including one conventional fixed-base reactor building and three base-isolated reactor buildings using Friction Pendulum(TM), lead rubber and low damping rubber bearings. The seismic assessment of the sample reactor building is performed using a new procedure proposed in this dissertation that builds on the methodology presented in the draft ATC-58 Guidelines and the widely used Zion method, which uses fragility curves defined in terms of ground-motion parameters for NPP seismic probabilistic risk assessment. The new procedure improves the Zion method by using fragility curves that are defined in terms of structural response parameters since damage and failure of NPP components are more closely tied to structural response parameters than to ground motion parameters. Alternate ground motion scaling methods are studied to help establish an optimal procedure for scaling ground motions for the purpose of seismic performance assessment. The proposed performance assessment procedure is used to evaluate the vulnerability of the conventional and base-isolated NPP reactor buildings. The seismic performance assessment confirms the utility of seismic isolation at reducing spectral demands on secondary systems. Procedures to reduce the construction cost of secondary systems in isolated reactor buildings are presented. A blast assessment of the sample reactor building is performed for an assumed threat of 2000 kg of TNT explosive detonated on the surface with a closest distance to the reactor building of 10 m. The air and ground shock waves produced by the design threat are generated and used for performance assessment. The air blast loading to the sample reactor building is computed using a Computational Fluid Dynamics code Air3D and the ground shock time series is generated using an attenuation model for soil/rock response. Response-history analysis of the sample conventional and base isolated reactor buildings to external blast loadings is performed using the hydrocode LS-DYNA. The spectral demands on the secondary systems in the isolated reactor building due to air blast loading are greater than those for the conventional reactor building but much smaller than those spectral demands associated with Safe Shutdown Earthquake shaking. The isolators are extremely effective at filtering out high acceleration, high frequency ground shock loading.

Decentralized Fuzzy MPC on Spatial Power Control of a Large PHWR

NASA Astrophysics Data System (ADS)

Liu, Xiangjie; Jiang, Di; Lee, Kwang Y.

2016-08-01

Reliable power control for stabilizing the spatial oscillations is quite important for ensuring the safe operation of a modern pressurized heavy water reactor (PHWR), since these spatial oscillations can cause “flux tilting” in the reactor core. In this paper, a decentralized fuzzy model predictive control (DFMPC) is proposed for spatial control of PHWR. Due to the load dependent dynamics of the nuclear power plant, fuzzy modeling is used to approximate the nonlinear process. A fuzzy Lyapunov function and “quasi-min-max” strategy is utilized in designing the DFMPC, to reduce the conservatism. The plant-wide stability is achieved by the asymptotically positive realness constraint (APRC) for this decentralized MPC. The solving optimization problem is based on a receding horizon scheme involving the linear matrix inequalities (LMIs) technique. Through dynamic simulations, it is demonstrated that the designed DFMPC can effectively suppress spatial oscillations developed in PHWR, and further, shows the advantages over the typical parallel distributed compensation (PDC) control scheme.

Sterile Neutrino Search at the NEOS Experiment

NASA Astrophysics Data System (ADS)

Ko, Y. J.; Kim, B. R.; Kim, J. Y.; Han, B. Y.; Jang, C. H.; Jeon, E. J.; Joo, K. K.; Kim, H. J.; Kim, H. S.; Kim, Y. D.; Lee, Jaison; Lee, J. Y.; Lee, M. H.; Oh, Y. M.; Park, H. K.; Park, H. S.; Park, K. S.; Seo, K. M.; Siyeon, Kim; Sun, G. M.; NEOS Collaboration

2017-03-01

An experiment to search for light sterile neutrinos is conducted at a reactor with a thermal power of 2.8 GW located at the Hanbit nuclear power complex. The search is done with a detector consisting of a ton of Gd-loaded liquid scintillator in a tendon gallery approximately 24 m from the reactor core. The measured antineutrino event rate is 1976 per day with a signal to background ratio of about 22. The shape of the antineutrino energy spectrum obtained from the eight-month data-taking period is compared with a hypothesis of oscillations due to active-sterile antineutrino mixing. No strong evidence of 3 +1 neutrino oscillation is found. An excess around the 5 MeV prompt energy range is observed as seen in existing longer-baseline experiments. The mixing parameter sin22 θ14 is limited up to less than 0.1 for Δ m412 ranging from 0.2 to 2.3 eV2 with a 90% confidence level.

EBR-II Reactor Physics Benchmark Evaluation Report

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

Pope, Chad L.; Lum, Edward S; Stewart, Ryan

This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

REACTOR FUEL SCAVENGING MEANS

DOEpatents

Coffinberry, A.S.

1962-04-10

A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

The IRIS Spool-Type Reactor Coolant Pump

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

Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safetymore » to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)« less

Power Peaking Effect of OTTO Fuel Scheme Pebble Bed Reactor

NASA Astrophysics Data System (ADS)

Setiadipura, T.; Suwoto; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

Pebble Bed Reactor (PBR) type of Hight Temperature Gas-cooled Reactor (HTGR) is a very interesting nuclear reactor design to fulfill the growing electricity and heat demand with a superior passive safety features. Effort to introduce the PBR design to the market can be strengthen by simplifying its system with the Once-through-then-out (OTTO) cycle PBR in which the pebble fuel only pass the core once. Important challenge in the OTTO fuel scheme is the power peaking effect which limit the maximum nominal power or burnup of the design. Parametric survey is perform in this study to investigate the contribution of different design parameters to power peaking effect of OTTO cycle PBR. PEBBED code is utilized in this study to perform the equilibrium PBR core analysis for different design parameter and fuel scheme. The parameters include its core diameter, height-per-diameter (H/D), power density, and core nominal power. Results of this study show that diameter and H/D effectsare stronger compare to the power density and nominal core power. Results of this study might become an importance guidance for design optimization of OTTO fuel scheme PBR.

Feasibility study of a small, thorium-based fission power system for space and terrestrial applications

NASA Astrophysics Data System (ADS)

Worrall, Michael Jason

One of the current challenges facing space exploration is the creation of a power source capable of providing useful energy for the entire duration of a mission. Historically, radioisotope batteries have been used to provide load power, but this conventional system may not be capable of sustaining continuous power for longer duration missions. To remedy this, many forays into nuclear powered spacecraft have been investigated, but no robust system for long-term power generation has been found. In this study, a novel spin on the traditional fission power system that represents a potential optimum solution is presented. By utilizing mature High Temperature Gas Reactor (HTGR) technology in conjunction with the capabilities of the thorium fuel cycle, we have created a light-weight, long-term power source capable of a continuous electric power output of up to 70kW for over 15 years. This system relies upon a combination of fissile, highly-enriched uranium dioxide and fertile thorium carbide Tri-Structural Isotropic (TRISO) fuel particles embedded in a hexagonal beryllium oxide matrix. As the primary fissile material is consumed, the fertile material breeds new fissile material leading to more steady fuel loading over the lifetime of the core. Reactor control is achieved through an innovative approach to the conventional boron carbide neutron absorber by utilizing sections of borated aluminum placed in rotating control drums within the reflector. Borated aluminum allows for much smaller boron concentrations, thus eliminating the potential for 10B(n,alpha)6Li heating issues that are common in boron carbide systems. A wide range of other reactivity control systems are also investigated, such as a radially-split rotating reflector. Lastly, an extension of the design to a terrestrial based system is investigated. In this system, uranium enrichment is dropped to 20 percent in order to meet current regulations, a solid uranium-zirconium hydride fissile driver replaces the uranium dioxide TRISO particles, and the moderating material is changed from beryllium oxide to graphite. These changes result in an increased core size, but the same long-term power generation potential is achieved. Additionally, small amounts of erbium are added to the hydride matrix to further extend core lifetime.

High-temperature, high-pressure bonding of nested tubular metallic components

DOEpatents

Quinby, Thomas C.

1980-01-01

This invention is a tool for effecting high-temperature, high-compression bonding between the confronting faces of nested, tubular, metallic components. In a typical application, the tool is used to produce tubular target assemblies for irradiation in nuclear reactors or particle accelerators, the target assembly comprising a uranium foil and an aluminum-alloy substrate. The tool preferably is composed throughout of graphite. It comprises a tubular restraining member in which a mechanically expandable tubular core is mounted to form an annulus with the member. The components to be bonded are mounted in nested relation in the annulus. The expandable core is formed of individually movable, axially elongated segments whose outer faces cooperatively define a cylindrical pressing surface and whose inner faces cooperatively define two opposed, inwardly tapered, axial bores. Tapered rams extend respectively into the bores. The loaded tool is mounted in a conventional hot-press provided with evacuation means, heaters for maintaining its interior at bonding temperature, and hydraulic cylinders for maintaining a selected inwardly directed pressure on the tapered rams. With the hot-press evacuated and the loaded tool at the desired temperature, the cylinders are actuated to apply the selected pressure to the rams. The rams in turn expand the segmented core to maintain the nested components in compression against the restraining member. These conditions are maintained until the confronting faces of the nested components are joined in a continuous, uniform bond characterized by high thermal conductivity.

Spring design for use in the core of a nuclear reactor

DOEpatents

Willard, Jr., H. James

1993-01-01

A spring design particularly suitable for use in the core of a nuclear reactor includes one surface having a first material oriented in a longitudinal direction, and another surface having a second material oriented in a transverse direction. The respective surfaces exhibit different amounts of irraditation induced strain.

Isomer Energy Source for Space Propulsion Systems

DTIC Science & Technology

2004-03-01

1,590 Engine F/W (no shield) 3.4 5.0 20.0 A similar core design replacing the fission fuel with the isomer 178Hfm2 is the starting point for this...particles interact and collide with other atoms in the fuel material, reactor core , or coolant, their energy can be transferred to thermal energy...thrust (44). The program produced several reactors that made it all the way through the testing stages of development . The reactors used uranium-235

Using SAFRAN Software to Assess Radiological Hazards from Dismantling of Tammuz-2 Reactor Core at Al-tuwaitha Nuclear Site

NASA Astrophysics Data System (ADS)

Abed Gatea, Mezher; Ahmed, Anwar A.; jundee kadhum, Saad; Ali, Hasan Mohammed; Hussein Muheisn, Abbas

2018-05-01

The Safety Assessment Framework (SAFRAN) software has implemented here for radiological safety analysis; to verify that the dose acceptance criteria and safety goals are met with a high degree of confidence for dismantling of Tammuz-2 reactor core at Al-tuwaitha nuclear site. The activities characterizing, dismantling and packaging were practiced to manage the generated radioactive waste. Dose to the worker was considered an endpoint-scenario while dose to the public has neglected due to that Tammuz-2 facility is located in a restricted zone and 30m berm surrounded Al-tuwaitha site. Safety assessment for dismantling worker endpoint-scenario based on maximum external dose at component position level in the reactor pool and internal dose via airborne activity while, for characterizing and packaging worker endpoints scenarios have been done via external dose only because no evidence for airborne radioactivity hazards outside the reactor pool. The in-situ measurements approved that reactor core components are radiologically activated by Co-60 radioisotope. SAFRAN results showed that the maximum received dose for workers are (1.85, 0.64 and 1.3mSv/y) for activities dismantling, characterizing and packaging of reactor core components respectively. Hence, the radiological hazards remain below the low level hazard and within the acceptable annual dose for workers in radiation field

Using the sound of nuclear energy

DOE PAGES

Garrett, Steven; Smith, James; Smith, Robert; ...

2016-08-01

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

Using the sound of nuclear energy

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

Garrett, Steven; Smith, James; Smith, Robert

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

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

Ariani, Menik; Su'ud, Zaki; Waris, Abdul

2012-06-06

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it ismore » shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.« less

Long-term performance of high-rate anaerobic reactors for the treatment of oily wastewater.

PubMed

Jeganathan, Jeganaesan; Nakhla, George; Bassi, Amarjeet

2006-10-15

Complex oily wastewater from a food industry was treated in three different UASB reactors at different operating conditions. Although all three systems achieved fat, oil, and grease (FOG) and COD removal efficiencies above 80% at an organic loading of 3 kg COD/m3 x d, system performance deteriorated sharply at higher loading rates, and the presence of high FOG caused a severe sludge flotation resulting in failure. Initially, FOG accumulated onto the biomass which led to sludge flotation and washout of biomass. The loss of sludge in the bed increased the FOG loading to the biomass and failure ensued. Contrary to previous findings, accumulation of FOG rather than influent FOG concentrations or volumetric FOG loading rate was the most importantfactor governing the high-rate anaerobic reactor performance. The critical accumulated FOG loading was identified as 1.04 +/- 0.13 g FOG/g VSS for all three reactors. Furthermore, FOG accumulation onto the biomass was identified mainly as palmitic acid (>60%) whereas the feed LCFA contained only 30% of palmitic acid and 50% of oleic acid.

JPRS Report, Science & Technology, China: Energy

DTIC Science & Technology

1988-06-29

capacity. There are currently two types of HTGR reactor designs: the particle-bed core , which uses spherical fuel elements, and the rod type core , in...and trial operating experience with the HTGR reactor. Its main design features are as follows. 1. A particle-bed core , continuous fueling and...Favorable for Development of Small-Scale HTGR (Xu Jiming; HE DONGLI GONGCHENG, Feb 88) 47 ERRATUM: In JPRS-CEN-88-003 of 25 April 1988 in article

Thermo-Physics Technical Note No. 60: thermal analysis of SNAP 10A reactor core during atmospheric reentry and resulting core disintegration and fuel element separation

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

Mouradian, E.M.

1966-02-16

A thermal analysis is carried out to determine the temperature distribution throughout a SNAP 10A reactor core, particularly in the vicinity of the grid plates, during atmospheric reentry. The transient temperatue distribution of the grid plate indicates when sufficient melting occurs so that fuel elements are free to be released and continue their descent individually.

APPARATUS FOR LOADING AND UNLOADING A MACHINE

DOEpatents

Payne, J.H. Jr.

1962-07-17

An arrangement for loading and unloading a nuclear reactor is described. Depleted fuel elements are removed from the reactor through one of a small number of holes in a shielding plug that is rotatably mounted in an eccentric annular plug rotatably mounted in the top of the reactor. The fuel elements removed are stored in a plurality of openings in a rotatable magazine or storage means rotatably mounted over the plugs. (AEC)

VIEW OF 77710A REACTOR WING, LOOKING NORTHEAST,SHOWING LOADING DOOR TO ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

VIEW OF 777-10A REACTOR WING, LOOKING NORTHEAST,SHOWING LOADING DOOR TO THE PROCESS DEVELOPMENT PILE ROOM. BUILDING 305-A IN BACKGROUND ON LEFT - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

System and method for determining coolant level and flow velocity in a nuclear reactor

DOEpatents

Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

2013-09-10

A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

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

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin

In this paper, we present thermal-mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress-strain states are significantly higher in case of presence of crack than without crack. In conclusion, the stress-strain state under grid load following condition are more realistic compared to the stress-strainmore » state estimated assuming simplified transients.« less

Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

DOE PAGES

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin; ...

2016-10-26

In this paper, we present thermal-mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress-strain states are significantly higher in case of presence of crack than without crack. In conclusion, the stress-strain state under grid load following condition are more realistic compared to the stress-strainmore » state estimated assuming simplified transients.« less

Calculation of the neutron diffusion equation by using Homotopy Perturbation Method

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

Koklu, H., E-mail: koklu@gantep.edu.tr; Ozer, O.; Ersoy, A.

The distribution of the neutrons in a nuclear fuel element in the nuclear reactor core can be calculated by the neutron diffusion theory. It is the basic and the simplest approximation for the neutron flux function in the reactor core. In this study, the neutron flux function is obtained by the Homotopy Perturbation Method (HPM) that is a new and convenient method in recent years. One-group time-independent neutron diffusion equation is examined for the most solved geometrical reactor core of spherical, cubic and cylindrical shapes, in the frame of the HPM. It is observed that the HPM produces excellent resultsmore » consistent with the existing literature.« less

Post impact behavior of mobile reactor core containment systems

NASA Technical Reports Server (NTRS)

Puthoff, R. L.; Parker, W. G.; Vanbibber, L. E.

1972-01-01

The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.

Fast breeder reactor protection system

DOEpatents

van Erp, J.B.

1973-10-01

Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

Control rod drive

DOEpatents

Hawke, Basil C.

1986-01-01

A control rod drive uses gravitational forces to insert one or more control rods upwardly into a reactor core from beneath the reactor core under emergency conditions. The preferred control rod drive includes a vertically movable weight and a mechanism operatively associating the weight with the control rod so that downward movement of the weight is translated into upward movement of the control rod. The preferred control rod drive further includes an electric motor for driving the control rods under normal conditions, an electrically actuated clutch which automatically disengages the motor during a power failure and a decelerator for bringing the control rod to a controlled stop when it is inserted under emergency conditions into a reactor core.

Magnetic switch for reactor control rod. [LMFBR

DOEpatents

Germer, J.H.

1982-09-30

A magnetic reed switch assembly is described for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electro-magnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

Magnetic switch for reactor control rod

DOEpatents

Germer, John H.

1986-01-01

A magnetic reed switch assembly for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electromagnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

Treatment of phenolic wastewater in an anaerobic fixed bed reactor (AFBR) - recovery after shock loading.

PubMed

Bajaj, Mini; Gallert, Claudia; Winter, Josef

2009-03-15

An anaerobic fixed bed reactor (AFBR) was run for 550 days with a mixed microbial flora to stabilize synthetic wastewater that contained glucose and phenol as main carbon sources. The influent phenol concentration was gradually increased from 2 to 40 mmol/l within 221 days. The microbial flora was able to adapt to this high phenol concentration with an average of 94% phenol removal. Microbial adaptation at such a high phenol concentration is not reported elsewhere. The maximum phenol removal observed before the phenol shock load was 39.47 mmol/l or 3.7 g phenol/l at a hydraulic retention time (HRT) of 2.5 days and an organic loading rate (OLR) of 5.3 g/l.d which amounts to a phenol removal rate of ca. 15.8 mmol phenol/l.d. The chemical oxygen demand (COD) removal before exposing the reactor to a shock load corresponded with phenol removal. A shock load was induced in the reactor by increasing the phenol concentration from 40 to 50 mmol/l in the influent. The maximum phenol removal rate observed after shock load was 18 mmol/l.d at 5.7 g COD/l.d. But this was not a stable rate and a consistent drop in COD and phenol removal was observed for 1 week, followed by a sharp decline and production of fatty acids. Recovery of the reactor was possible only when no feed was provided to the reactor for 1 month and the phenol concentration was increased gradually. When glucose was omitted from the influent, unknown intermediates of anaerobic phenol metabolism were observed for some time.

Reactor safety method

DOEpatents

Vachon, Lawrence J.

1980-03-11

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Neutron Radiation Damage Estimation in the Core Structure Base Metal of RSG GAS

NASA Astrophysics Data System (ADS)

Santa, S. A.; Suwoto

2018-02-01

Radiation damage in core structure of the Indonesian RGS GAS multi purpose reactor resulting from the reaction of fast and thermal neutrons with core material structure was investigated for the first time after almost 30 years in operation. The aim is to analyze the degradation level of the critical components of the RSG GAS reactor so that the remaining life of its component can be estimated. Evaluation results of critical components remaining life will be used as data ccompleteness for submission of reactor operating permit extension. Material damage analysis due to neutron radiation is performed for the core structure components made of AlMg3 material and bolts reinforcement of core structure made of SUS304. Material damage evaluation was done on Al and Fe as base metal of AlMg3 and SUS304, respectively. Neutron fluences are evaluated based on the assumption that neutron flux calculations of U3Si8-Al equilibrium core which is operated on power rated of 15 MW. Calculation result using SRAC2006 code of CITATION module shows the maximum total neutron flux and flux >0.1 MeV are 2.537E+14 n/cm2/s and 3.376E+13 n/cm2/s, respectively. It was located at CIP core center close to the fuel element. After operating up to the end of #89 core formation, the total neutron fluence and fluence >0.1 MeV were achieved 9.063E+22 and 1.269E+22 n/cm2, respectively. Those are related to material damage of Al and Fe as much as 17.91 and 10.06 dpa, respectively. Referring to the life time of Al-1100 material irradiated in the neutron field with thermal flux/total flux=1.7 which capable of accepting material damage up to 250 dpa, it was concluded that RSG GAS reactor core structure underwent 7.16% of its operating life span. It means that core structure of RSG GAS reactor is still capable to receive the total neutron fluence of 9.637E+22 n/cm2 or fluence >0.1 MeV of 5.672E+22 n/cm2.

Looking Northeast in Oxide Building at Reactors on Second Floor ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

Looking Northeast in Oxide Building at Reactors on Second Floor Including Reactor One (Left) and Reactor Two (Right) - Hematite Fuel Fabrication Facility, Oxide Building & Oxide Loading Dock, 3300 State Road P, Festus, Jefferson County, MO

The WSTIAC Quarterly. Volume 9, Number 3

DTIC Science & Technology

2010-01-25

program .[8] THE THORIUM FUEL CYCLE AND LFTR POWER PLANT The thorium fuel cycle is based on a series of neutron absorp- tion and beta decay processes...the fig- ure is a graphite matrix moderated MSR reactor with fuel salt mixture (ThF4-U233F4) being circulated by a pump through the core and to a...the core as purified salt. As one of the unique safety features, a melt-plug at the reactor bottom would permit the reactor fluid fuel to be drained

System Study: Reactor Core Isolation Cooling 1998-2014

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

Schroeder, John Alton

2015-12-01

This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

COUPLED FAST-THERMAL POWER BREEDER REACTOR

DOEpatents

Avery, R.

1961-07-18

A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept

NASA Technical Reports Server (NTRS)

Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.

1973-01-01

A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

Split-core heat-pipe reactors for out-of-pile thermionic power systems.

NASA Technical Reports Server (NTRS)

Niederauer, G.; Lantz, E.; Breitweiser, R.

1971-01-01

Description of the concept of splitting a heat-pipe reactor for out-of-core thermionics into two identical halves and using the resulting center gap for reactivity control. Short Li-W reactor heat pipes penetrate the axial reflectors and form a heat exchanger with long heat pipes which wind through the shield to the thermionic diodes. With one reactor half anchored to the shield, the other is attached to a long arm with a pivot behind the shield and swings through a small arc for reactivity control. A safety shim prevents large reactivity inputs, and a fueled control arm drive shaft acts as a power stabilizer. Reactors fueled with U-235C and with U-233C have been studied.-

Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

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

Daily, Charles R.

2015-10-01

An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclearmore » Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.« less

Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

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

Cormon, S.; Fallot, M., E-mail: fallot@subatech.in2p3.fr; Bui, V.-M.

This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {supmore » 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.« less

Support arrangement for core modules of nuclear reactors

DOEpatents

Bollinger, Lawrence R.

1987-01-01

A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

NUT SCREW MECHANISMS

DOEpatents

Glass, J.A.F.

1958-07-01

A reactor control mechanism is described wherein the control is achieved by the partial or total withdrawal of the fissile material which is in the form of a fuel rod. The fuel rod is designed to be raised and lowered from the reactor core area by means of two concentric ball nut and screw assemblies that may telescope one within the other. These screw mechanisms are connected through a magnetic clutch to a speed reduction gear and an accurately controllable prime motive source. With the clutch energized, the fuel rod may be moved into the reactor core area, and fine adjustments may be made through the reduction gearing. However, in the event of a power failure or an emergency signal, the magnetic clutch will become deenergized, and the fuel rod will drop out of the core area by the force of gravity, thus shutting down the operation of the reactor.

Support arrangements for core modules of nuclear reactors. [PWR

DOEpatents

Bollinger, L.R.

1983-11-03

A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

Tokamak reactor for treating fertile material or waste nuclear by-products

DOEpatents

Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

2012-10-02

Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

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

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

Nuclear reactor control column

DOEpatents

Bachovchin, Dennis M.

1982-01-01

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

Performance characteristics of anaerobic downflow stationary fixed film reactors

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

van den Berg, L.; Kennedy, K.J.

1982-01-01

Stationary fixed film reactors operated to ensure a net downflow of substrate have several characteristics different from other retained biomass reactors. The active biomass attaches itself to stationary surface and hence is difficult to wash out. Performance is related to the surface-to-volume of the film support as well as to the composition of the support. Methane production rates of up to 8 cym day at loading rates of up to 30 kg COD/m cym day, are possible. Severe hydraulic and organic overloadings can be tolerated with operation back to normal 24 hours following cessation of mistreatment. Reactors can operate withmore » dilute and concentrated wastes (4000-130,000 mg COD/L) and can change readily over from one waste to another. Intermittent loading at high loading rates are possible. Methane production rates and loading rates decreased linearly with temperature (35) to 10); at 10 C they were about 20% of those at 35 C.« less

Biofilm architecture in a novel pressurized biofilm reactor.

PubMed

Jiang, Wei; Xia, Siqing; Duan, Liang; Hermanowicz, Slawomir W

2015-01-01

A novel pure-oxygen pressurized biofilm reactor was operated at different organic loading, mechanical shear and hydrodynamic conditions to understand the relationships between biofilm architecture and its operation. The ultimate goal was to improve the performance of the biofilm reactor. The biofilm was labeled with seven stains and observed with confocal laser scanning microscopy. Unusual biofilm architecture of a ribbon embedded between two surfaces with very few points of attachment was observed. As organic loading increased, the biofilm morphology changed from a moderately rough layer into a locally smoother biomass with significant bulging protuberances, although the chemical oxygen demand (COD) removal efficiency remained unchanged at about 75%. At higher organic loadings, biofilms contained a larger fraction of active cells distributed uniformly within a proteinaceous matrix with decreasing polysaccharide content. Higher hydrodynamic shear in combination with high organic loading resulted in the collapse of biofilm structure and a substantial decrease in reactor performance (a COD removal of 16%). Moreover, the important role of proteins for the spatial distribution of active cells was demonstrated quantitatively.

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.

Pressurized water reactor flow skirt apparatus

DOEpatents

Kielb, John F.; Schwirian, Richard E.; Lee, Naugab E.; Forsyth, David R.

2016-04-05

A pressurized water reactor vessel having a flow skirt formed from a perforated cylinder structure supported in the lower reactor vessel head at the outlet of the downcomer annulus, that channels the coolant flow through flow holes in the wall of the cylinder structure. The flow skirt is supported at a plurality of circumferentially spaced locations on the lower reactor vessel head that are not equally spaced or vertically aligned with the core barrel attachment points, and the flow skirt employs a unique arrangement of hole patterns that assure a substantially balanced pressure and flow of the coolant over the entire underside of the lower core support plate.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

NASA Astrophysics Data System (ADS)

Koshelev, A. S.; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

2016-12-01

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

NEUTRONIC REACTOR DESIGN TO REDUCE NEUTRON LOSS

DOEpatents

Mills, F.T.

1961-05-01

A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall which is surrounded by successive layers of pure fertile material and fertile material having moderator. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. As the steel has a smaller capture cross-section for the fast neutrons, then greater numbers of the neutrons will pass into the blanket thereby increasing the over-all efficiency of the reactor.

Neutronic Reactor Design to Reduce Neutron Loss

DOEpatents

Miles, F. T.

1961-05-01

A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall. The wall is surrounded by successive layers of pure fertile material and moderator containing fertile material. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. Since the steel has a smaller capture cross section for the fast neutrons, greater nunnbers of neutrons will pass into the blanket, thereby increasing the over-all efficiency of the reactor. (AEC)

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

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

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Kovshov, K. N.; Ovchinnikov, M. A.

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Turf soil enhances treatment efficiency and performance of phenolic wastewater in an up-flow anaerobic sludge blanket reactor.

PubMed

Chen, Chunmao; Yao, Xianyang; Li, Qing X; Wang, Qinghong; Liang, Jiahao; Zhang, Simin; Ming, Jie; Liu, Zhiyuan; Deng, Jingmin; Yoza, Brandon A

2018-08-01

Phenols are industrially generated intermediate chemicals found in wastewaters that are considered a class of environmental priority pollutants. Up-flow anaerobic sludge blanket (UASB) reactors are used for phenolic wastewater treatment and exhibit high volume loading capability, favorable granule settling, and tolerance to impact loads. Use of support materials can promote biological productivity and accelerate start-up period of UASB. In the present study, turf soil was used as a support material in a mesophilic UASB reactor for the removal of phenols in wastewater. During sludge acclimatization (45-96 days), COD and phenols in the treatments were both reduced by 97%, whereas these contents in the controls were decreased by 81% and 75%, respectively. The phenol load threshold for the turf soil UASB reactor was greater (1200 mg/L, the equivalent of COD 3000 mg/L) in comparison with the control UASB reactor (900 mg/L, the equivalent of COD 2250 mg/L) and the turf soil UASB reactor was also more resistant to shock loading. Improved sludge settling, shear resistance, and higher biological activity occurred with the turf soil UASB reactor due to the formation of large granular sludge (0.6 mm or larger) in higher relative percentages. Granular sludge size was further enhanced by the colonization of filamentous bacteria on the irregular surface of the turf soil. Copyright © 2018 Elsevier Ltd. All rights reserved.

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

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

Philip E. MacDonald

2005-01-01

The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission ofmore » the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks: Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking Task 3. Plant Engineering and Reactor Safety Analysis. moderator rods. materials.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE PAGES

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; ...

2016-09-07

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

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

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

SPERT I DESTRUCTIVE TEST PROGRAM SAFETY ANALYSIS REPORT

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

Spano, A.H.; Miller, R.W.

1962-06-15

The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor periodmore » range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)« less

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

Burgett, Eric; Al-Sheikhly, Mohamad; Summers, Christopher

An advanced in-pile multi-parameter reactor monitoring system is being proposed in this funding opportunity. The proposed effort brings cutting edge, high-fidelity optical measurement systems into the reactor environment in an unprecedented fashion, including in-core, in-cladding and in-fuel pellet itself. Unlike instrumented leads, the proposed system provides a unique solution to a multi-parameter monitoring need in core while being minimally intrusive in the reactor core. Detector designs proposed herein can monitor fuel compression and expansion in both the radial and axial dimensions as well as monitor linear power profiles and fission rates during the operation of the reactor. In addition tomore » pressure, stress, strain, compression, neutron flux, neutron spectra, and temperature can be observed inside the fuel bundle and fuel rod using the proposed system. The proposed research aims at developing radiation-hard, harsh-environment multi-parameter systems for insertion into the reactor environment. The proposed research holds the potential to drastically increase the fidelity and precision of in-core instrumentation with little or no impact in the neutron economy in the reactor environment while providing a measurement system capable of operation for entire operating cycles. Significant work has been done over the last few years on the use of nanoparticle-based scintillators. Through the use of metamaterials, the PIs aim to develop planar neutron detectors and large-volume neutron detectors. These detectors will have high efficiencies for neutron detection and will have a high gamma discrimination capability.« less

Determination of the neutron activation profile of core drill samples by gamma-ray spectrometry.

PubMed

Gurau, D; Boden, S; Sima, O; Stanga, D

2018-04-01

This paper provides guidance for determining the neutron activation profile of core drill samples taken from the biological shield of nuclear reactors using gamma spectrometry measurements. Thus, it provides guidance for selecting a model of the right form to fit data and using least squares methods for model fitting. The activity profiles of two core samples taken from the biological shield of a nuclear reactor were determined. The effective activation depth and the total activity of core samples along with their uncertainties were computed by Monte Carlo simulation. Copyright © 2017 Elsevier Ltd. All rights reserved.

Passive containment cooling system

DOEpatents

Conway, Lawrence E.; Stewart, William A.

1991-01-01

A containment cooling system utilizes a naturally induced air flow and a gravity flow of water over the containment shell which encloses a reactor core to cool reactor core decay heat in two stages. When core decay heat is greatest, the water and air flow combine to provide adequate evaporative cooling as heat from within the containment is transferred to the water flowing over the same. The water is heated by heat transfer and then evaporated and removed by the air flow. After an initial period of about three to four days when core decay heat is greatest, air flow alone is sufficient to cool the containment.

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note: A nuclear reactor... core of a nuclear reactor and capable of withstanding the operating pressure of the primary coolant. (2...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

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

Martin, William R.; Lee, John C.; baxter, Alan

Information and measured data from the intial Fort St. Vrain (FSV) high temperature gas reactor core is used to develop a benchmark configuration to validate computational methods for analysis of a full-core, commercial HTR configuration. Large uncertainties in the geometry and composition data for the FSV fuel and core are identified, including: (1) the relative numbers of fuel particles for the four particle types, (2) the distribution of fuel kernel diameters for the four particle types, (3) the Th:U ratio in the initial FSV core, (4) and the buffer thickness for the fissile and fertile particles. Sensitivity studies were performedmore » to assess each of these uncertainties. A number of methods were developed to assist in these studies, including: (1) the automation of MCNP5 input files for FSV using Python scripts, (2) a simple method to verify isotopic loadings in MCNP5 input files, (3) an automated procedure to conduct a coupled MCNP5-RELAP5 analysis for a full-core FSV configuration with thermal-hydraulic feedback, and (4) a methodology for sampling kernel diameters from arbitrary power law and Gaussian PDFs that preserved fuel loading and packing factor constraints. A reference FSV fuel configuration was developed based on having a single diameter kernel for each of the four particle types, preserving known uranium and thorium loadings and packing factor (58%). Three fuel models were developed, based on representing the fuel as a mixture of kernels with two diameters, four diameters, or a continuous range of diameters. The fuel particles were put into a fuel compact using either a lattice-bsed approach or a stochastic packing methodology from RPI, and simulated with MCNP5. The results of the sensitivity studies indicated that the uncertainties in the relative numbers and sizes of fissile and fertile kernels were not important nor were the distributions of kernel diameters within their diameter ranges. The uncertainty in the Th:U ratio in the intial FSV core was found to be important with a crude study. The uncertainty in the TRISO buffer thickness was estimated to be unimportant but the study was not conclusive. FSV fuel compacts and a regular FSV fuel element were analyzed with MCNP5 and compared with predictions using a modified version of HELIOS that is capable of analyzing TRISO fuel configurations. The HELIOS analyses were performed by SSP. The eigenvalue discrepancies between HELIOS and MCNP5 are currently on the order of 1% but these are still being evaluated. Full-core FSV configurations were developed for two initial critical configurations - a cold, clean critical loading and a critical configuration at 70% power. MCNP5 predictions are compared to experimental data and the results are mixed. Analyses were also done for the pulsed neutron experiments that were conducted by GA for the initial FSV core. MCNP5 was used to model these experiments and reasonable agreement with measured results has been observed.« less

The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

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

Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber

The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactorsmore » with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).« less

A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

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

Senor, David J.; Painter, Chad L.; Geelhood, Ken J.

2007-12-01

Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling,more » core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.« less

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

NASA Astrophysics Data System (ADS)

Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

2016-01-01

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

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

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

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

Pomorski, Michal; Mer-Calfati, Christine; Foulon, Francois

Diamond exhibits a combination of properties which makes it attractive for neutron detection in hostile conditions. In the particular case of detection in a nuclear reactor, it is resilient to radiation, exhibits a natural low sensitivity to gamma rays, and its small size (as compared with that of gas ionisation chambers) enables fluency monitoring with a high position resolution. We report here on the use of synthetic CVD diamond as a solid state micro-fission chamber with U-235 converting material for in-core thermal neutron monitoring. Two types of thin diamond detectors were developed for this application. The first type of detectormore » is fabricated using thin diamond membrane obtained by etching low-cost commercially available single crystal CVD intrinsic diamond, so called 'optical grade' material. Starting from a few hundred of micrometre thick samples, the sample is sliced with a laser and then plasma etched down to a few tenths of micrometre. Here we report the result obtained with a 17 μm thick device. The detection surface of this detector is equal to 1 mm{sup 2}. Detectors with surfaces up to 1 cm{sup 2} can be fabricated with this technique. The second type of detector is fabricated by growing successively two thin films of diamond, by the microwave enhanced chemical vapour deposition technique, on HPHT single crystal diamond. A first, a film of boron doped (p+) single crystal diamond, a few microns thick, is deposited. Then a second film of intrinsic diamond with a thickness of a few tens of microns is deposited. This results in a P doped, Intrinsic, Metal structure (PIM) structure in which the intrinsic volume id the active part of the detector. Here we report the results obtained with a 20 μm thick intrinsic whose detection surface is equal to 0.5 mm{sup 2}, with the possibility to enlarge the surface of the detector up to 1 cm{sup 2}. These two types of detector were tested at the VR-1 research reactor at the Czech Technical University in Prague. The Training Reactor VR-1 is a pool type (light water) reactor based on UO{sub 2} low enriched uranium. It has a nominal power of 1 kW, and can be operated for a short period up to 5 kW. The arrangement of the reactor pool reactor facilitates access to the core, setting and removal of various experimental samples and detectors, and safe and easy handling of fuel assemblies. The reactor is equipped with two horizontal channels (radial and tangential) and 10 vertical channels, of varying diameters, which can be loaded into various core positions, and one pneumatic transfer system. It is also equipped with several specifically designed educational instrumentation systems that can be used to supply complementary measurements and characterization around the reactor. The reactor is operated by the Department of Nuclear Reactors of the Faculty of Nuclear Sciences and Physical Engineering of the Czech Technical University in Prague. The two detectors were placed in-core through one of the vertical insertion channel. They were coupled to remote placed (5 m BNC cable) classical nuclear charge sensitive electronics. Detection properties of both sensors, including: pulse height spectra of U-235 fission fragments (response linearity with neutron flux, count rate, gamma background, were evaluated varying the power of the reactor from 0.005 W to 500 W. The evolution of the counting rate of the thinned optical grade detector as a function of counting rate of a gas ionization chamber used currently for reactor monitoring shows the very good linearity of the detector over the 5 decades. Similar results were obtained with the PIM detector. Additionally fast transient current signals of the detectors were recorded on a digital storage oscilloscope (DSO) using broad-band amplifier and with a simple bias-T, showing potential use of such sensors for neutron counting with no need of an amplification stage, since non-amplified signals from fission fragments exceeded 4 mV in amplitude. Therefore, one can think of simple neutron counting system by feeding diamond detectors signals directly to the low threshold discriminators. The results obtained on the VR1 will be described and discussed in detail in the paper and associated presentation. The results demonstrate that diamond micro-fission chambers can be used for in-core neutron monitoring, where robust, simple and compact devices are required.« less

Comparison of Analysis Results Between 2D/1D Synthesis and RAPTOR-M3G in the Korea Standard Nuclear Plant (KSNP)

NASA Astrophysics Data System (ADS)

Joung Lim, Mi; Maeng, Young Jae; Fero, Arnold H.; Anderson, Stanwood L.

2016-02-01

The 2D/1D synthesis methodology has been used to calculate the fast neutron (E > 1.0 MeV) exposure to the beltline region of the reactor pressure vessel. This method uses the DORT 3.1 discrete ordinates code and the BUGLE-96 cross-section library based on ENDF/B-VI. RAPTOR-M3G (RApid Parallel Transport Of Radiation-Multiple 3D Geometries) which performs full 3D calculations was developed and is based on domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architecture. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor. Both methods are applied to surveillance test results for the Korea Standard Nuclear Plant (KSNP)-OPR (Optimized Power Reactor) 1000 MW. The objective of this paper is to compare the results of the KSNP surveillance program between 2D/1D synthesis and RAPTOR-M3G. Each operating KSNP has a reactor vessel surveillance program consisting of six surveillance capsules located between the core and the reactor vessel in the downcomer region near the reactor vessel wall. In addition to the In-Vessel surveillance program, an Ex-Vessel Neutron Dosimetry (EVND) program has been implemented. In order to estimate surveillance test results, cycle-specific forward transport calculations were performed by 2D/1D synthesis and by RAPTOR-M3G. The ratio between measured and calculated (M/C) reaction rates will be discussed. The current plan is to install an EVND system in all of the Korea PWRs including the new reactor type, APR (Advanced Power Reactor) 1400 MW. This work will play an important role in establishing a KSNP-specific database of surveillance test results and will employ RAPTOR-M3G for surveillance dosimetry location as well as positions in the KSNP reactor vessel.

Mesophilic and thermophilic biotreatment of BTEX-polluted air in reactors.

PubMed

Mohammad, Balsam T; Veiga, María C; Kennes, Christian

2007-08-15

This study compares the removal of a mixture of benzene, toluene, ethylbenzene, and all three xylene isomers (BTEX) in mesophilic and thermophilic (50 degrees C) bioreactors. In the mesophilic reactor fungi became dominant after long-term operation, while bacteria dominated in the thermophilic unit. Microbial acclimation was achieved by exposing the biofilters to initial BTEX loads of 2-15 g m(-3) h(-1), at an empty bed residence time of 96 s. After adaptation, the elimination capacities ranged from 3 to 188 g m(-3) h(-1), depending on the inlet load, for the mesophilic biofilter with removal efficiencies reaching 96%. On the other hand, in the thermophilic reactor the average removal efficiency was 83% with a maximum elimination capacity of 218 g m(-3) h(-1). There was a clear positive relationship between temperature gradients as well as CO(2) production and elimination capacities across the biofilters. The gas phase was sampled at different depths along the reactors observing that the percentage pollutant removal in each section was strongly dependant on the load applied. The fate of individual alkylbenzene compounds was checked, showing the unusually high biodegradation rate of benzene at high loads under thermophilic conditions (100%) compared to its very low removal in the mesophilic reactor at such load (<10%). Such difference was less pronounced for the other pollutants. After 210 days of operation, the dry biomass content for the mesophilic and thermophilic reactors were 0.300 and 0.114 g g(-1) (support), respectively, reaching higher removals under thermophilic conditions with a lower biomass accumulation, that is, lower pressure drop. (c) 2007 Wiley Periodicals, Inc.

Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

NASA Astrophysics Data System (ADS)

Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

2017-09-01

The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

Johnson Noise Thermometry for Advanced Small Modular Reactors

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

Britton Jr, Charles L; Roberts, Michael; Bull, Nora D

Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurementmore » due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.« less

ZPPR-20 phase D : a cylindrical assembly of polyethylene moderated U metal reflected by beryllium oxide and polyethylene.

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

Lell, R.; Grimm, K.; McKnight, R.

The Zero Power Physics Reactor (ZPPR) fast critical facility was built at the Argonne National Laboratory-West (ANL-W) site in Idaho in 1969 to obtain neutron physics information necessary for the design of fast breeder reactors. The ZPPR-20D Benchmark Assembly was part of a series of cores built in Assembly 20 (References 1 through 3) of the ZPPR facility to provide data for developing a nuclear power source for space applications (SP-100). The assemblies were beryllium oxide reflected and had core fuel compositions containing enriched uranium fuel, niobium and rhenium. ZPPR-20 Phase C (HEU-MET-FAST-075) was built as the reference flight configuration.more » Two other configurations, Phases D and E, simulated accident scenarios. Phase D modeled the water immersion scenario during a launch accident, and Phase E (SUB-HEU-MET-FAST-001) modeled the earth burial scenario during a launch accident. Two configurations were recorded for the simulated water immersion accident scenario (Phase D); the critical configuration, documented here, and the subcritical configuration (SUB-HEU-MET-MIXED-001). Experiments in Assembly 20 Phases 20A through 20F were performed in 1988. The reference water immersion configuration for the ZPPR-20D assembly was obtained as reactor loading 129 on October 7, 1988 with a fissile mass of 167.477 kg and a reactivity of -4.626 {+-} 0.044{cents} (k {approx} 0.9997). The SP-100 core was to be constructed of highly enriched uranium nitride, niobium, rhenium and depleted lithium. The core design called for two enrichment zones with niobium-1% zirconium alloy fuel cladding and core structure. Rhenium was to be used as a fuel pin liner to provide shut down in the event of water immersion and flooding. The core coolant was to be depleted lithium metal ({sup 7}Li). The core was to be surrounded radially with a niobium reactor vessel and bypass which would carry the lithium coolant to the forward inlet plenum. Immediately inside the reactor vessel was a rhenium baffle which would act as a neutron curtain in the event of water immersion. A fission gas plenum and coolant inlet plenum were located axially forward of the core. Some material substitutions had to be made in mocking up the SP-100 design. The ZPPR-20 critical assemblies were fueled by 93% enriched uranium metal because uranium nitride, which was the SP-100 fuel type, was not available. ZPPR Assembly 20D was designed to simulate a water immersion accident. The water was simulated by polyethylene (CH{sub 2}), which contains a similar amount of hydrogen and has a similar density. A very accurate transformation to a simplified model is needed to make any of the ZPPR assemblies a practical criticality-safety benchmark. There is simply too much geometric detail in an exact model of a ZPPR assembly, particularly as complicated an assembly as ZPPR-20D. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertainty far beyond that of the original experiment. Such a transformation will be described in a later section. First, Assembly 20D was modeled in full detail--every plate, drawer, matrix tube, and air gap was modeled explicitly. Then the regionwise compositions and volumes from this model were converted to an RZ model. ZPPR Assembly 20D has been determined to be an acceptable criticality-safety benchmark experiment.« less

Diffusion Couple Alloying of Refractory Metals in Austenitic and Ferritic/Martensitic Steels

DTIC Science & Technology

2012-03-01

applications of austenitic stainless steel and ferritic/martensitic steel can vary from structural and support components in the reactor core to reactor fuel ... fuel . It serves as a boundary to prevent both fission products from escaping to the core coolant, and segregates the fuel from the coolant to...uranium oxide (UO2) fuel in the core . It resists corrosion by the fuel matrix on the inner surface of the cladding and the liquid sodium coolant on

Performance and stability of an expanded granular sludge bed reactor modified with zeolite addition subjected to step increases of organic loading rate (OLR) and to organic shock load (OSL).

PubMed

Pérez-Pérez, T; Pereda-Reyes, I; Pozzi, E; Oliva-Merencio, D; Zaiat, M

2018-01-01

This paper shows the effect of organic shock loads (OSLs) on the anaerobic digestion (AD) of synthetic swine wastewater using an expanded granular sludge bed (EGSB) reactor modified with zeolite. Two reactors (R1 and R2), each with an effective volume of 3.04 L, were operated for 180 days at a controlled temperature of 30 °C and hydraulic retention time of 12 h. In the case of R2, 120 g of zeolite was added. The reactors were operated with an up-flow velocity of 6 m/h. The evolution of pH, total Kjeldahl nitrogen, chemical oxygen demand (COD) and volatile fatty acids (VFAs) was monitored during the AD process with OSL and increases in the organic loading rate (OLR). In addition, the microbial composition and changes in the structure of the bacterial and archaeal communities were assessed. The principal results demonstrate that the presence of zeolite in an EGSB reactor provides a more stable process at higher OLRs and after applying OSL, based on both COD and VFA accumulation, which presented with significant differences compared to the control. Denaturing gradient gel electrophoresis band profiles indicated differences in the populations of Bacteria and Archaea between the R1 and R2 reactors, attributed to the presence of zeolite.

Reactivity change in a fast-spectrum space power reactor due to a 328-meter-per-second (1075-ft/sec) impact

NASA Technical Reports Server (NTRS)

Peoples, J. A., Jr.; Puthoff, R. L.

1973-01-01

Application of nuclear reactors in space will present operational problems. One such problem is the possibility of an earth impact at velocities in excess of 305 m/sec (1000 ft/sec). This report shows the results of an impact against concrete at 328 m/sec (1075 ft/sec) and examines the deformed core to estimate the range of activity inserted as a result of the impact. The results of this examination are that the deformation of the reactor core within the containment vessel left only an estimated 2.7 percent void in the core and that the reactivity inserted due to this impact deformation could be from 4.0 to 10.25 dollars.

77 FR 64563 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on November 1-3, 2012, 11545 Rockville...-Term Core Cooling Approach for the Advanced Boiling Water Reactor (ABWR) Design for South Texas Project...

Determination of parameters of a nuclear reactor through noise measurements

DOEpatents

Cohn, C.E.

1975-07-15

A method of measuring parameters of a nuclear reactor by noise measurements is described. Noise signals are developed by the detectors placed in the reactor core. The polarity coincidence between the noise signals is used to develop quantities from which various parameters of the reactor can be calculated. (auth)

Flow characteristics of Korea multi-purpose research reactor

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

Heonil Kim; Hee Taek Chae; Byung Jin Jun

1995-09-01

The construction of Korea Multi-purpose Research Reactor (KMRR), a 30 MW{sub th} open-tank-in-pool type, is completed. Various thermal-hydraulic experiments have been conducted to verify the design characteristics of the KMRR. This paper describes the commissioning experiments to determine the flow distribution of KMRR core and the flow characteristics inside the chimney which stands on top of the core. The core flow is distributed to within {+-}6% of the average values, which is sufficiently flat in the sense that the design velocity in the fueled region is satisfied. The role of core bypass flow to confine the activated core coolant inmore » the chimney structure is confirmed.« less

ETR, TRA642. BENCH MARK AND ELEVATION LOCATIONS, FLOOR LOADING DATA, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. BENCH MARK AND ELEVATION LOCATIONS, FLOOR LOADING DATA, CRANE WORKING AREAS. PHILLIPS PETROLEUM COMPANY ETR-D-1584, 5/1959. INL INDEX NO. 532-0642-00-706-020323, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Experiment neutrino-4 on searching for a sterile neutrino with multisection detector model

NASA Astrophysics Data System (ADS)

Serebrov, A. P.; Ivochkin, V. G.; Samoilov, R. M.; Fomin, A. K.; Zinov'ev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Chernyi, A. V.; Zherebtsov, O. M.; Polyushkin, A. O.; Martem'yanov, V. P.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Izhutov, A. L.; Tuzov, A. A.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanas'ev, V. V.; Zaitsev, M. E.; Chaikovskii, M. E.

2017-02-01

A laboratory for searching for oscillations of reactor antineutrinos has been created based on the SM-3 reactor in order to approach the problem of the possible existence of a sterile neutrino. The multisection detector prototype with a liquid scintillator volume of 350 L was installed in mid-2015. This detector can move inside the passive shield in a range of 6-11 m from the active core of the reactor. The antineutrino flux was measured for the first time at these short distances from the active core of the reactor by the movable detector. The measurements with the multisection detector prototype demonstrated that it is possible to measure the antineutrino flux from the reactor in the complicated conditions of cosmic background on the Earth's surface.

Neutrino-4 experiment on search for sterile neutrino with multi-section model of detector

NASA Astrophysics Data System (ADS)

Serebrov, A.; Ivochkin, V.; Samoilov, R.; Fomin, A.; Polyushkin, A.; Zinoviev, V.; Neustroev, P.; Golovtsov, V.; Chernyj, A.; Zherebtsov, O.; Martemyanov, V.; Tarasenkov, V.; Aleshin, V.; Petelin, A.; Izhutov, A.; Tuzov, A.; Sazontov, S.; Ryazanov, D.; Gromov, M.; Afanasiev, V.; Zaytsev, M.; Chaikovskii, M.

2017-09-01

In order to carry out research in the field of possible existence of a sterile neutrino the laboratory based on SM-3 reactor (Dimitrovgrad, Russia) was created to search for oscillations of reactor antineutrino. The prototype of a multi-section neutrino detector with liquid scintillator volume of 350 l was installed in the middle of 2015. It is a moveable inside the passive shielding detector, which can be set at distance range from 6 to 11 meters from the reactor core. Measurements of antineutrino flux at such small distances from the reactor core are carried out with moveable detector for the first time. The measurements carried out with detector prototype demonstrated a possibility of measuring a reactor antineutrino flux in difficult conditions of cosmic background at Earth surface.

A Boiling-Potassium Fluoride Reactor for an Artificial-Gravity NEP Vehicle

NASA Technical Reports Server (NTRS)

Sorensen, Kirk; Juhasz, Albert

2007-01-01

Several years ago a rotating manned spacecraft employing nuclear-electric propulsion was examined for Mars exploration. The reactor and its power conversion system essentially served as the counter-mass to an inflatable manned module. A solid-core boiling potassium reactor based on the MPRE concept of the 1960s was baselined in that study. This paper proposes the use of a liquid-fluoride reactor, employing direct boiling of potassium in the core, as a means to overcome some of the residual issues with the MPRE reactor concept. Several other improvements to the rotating Mars vehicle are proposed as well, such as Canfield joints to enable the electric engines to track the inertial thrust vector during rotation, and innovative "cold-ion" engine technologies to improve engine performance.

Structural materials issues for the next generation fission reactors

NASA Astrophysics Data System (ADS)

Chant, I.; Murty, K. L.

2010-09-01

Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

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

Physics of reactor safety. Quarterly report, January--March 1977. [LMFBR

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

None

1977-06-01

This report summarizes work done on reactor safety, Monte Carlo analysis of safety-related critical assembly experiments, and planning of DEMI safety-related critical experiments. Work on reactor core thermal-hydraulics is also included.

Reactivity control assembly for nuclear reactor

DOEpatents

Bollinger, Lawrence R.

1984-01-01

Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

TMI-2 (Three Mile Island Unit 2) core region defueling

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

Rodabaugh, J.M.; Cowser, D.K.

1988-01-01

In July of 1982, a video camera was inserted into the Three Mile Island Unit 2 reactor vessel providing the first visual evidence of core damage. This inspection, and numerous subsequent data acquisition tasks, revealed a central void /approx/1.5 m (5 ft) deep. This void region was surrounded by partial length fuel assemblies and ringed on the periphery by /approx/40 full-length, but partial cross-section, fuel assemblies. All of the original 177 fuel assemblies exhibited signs of damage. The bottom of the void cavity was covered with a bed of granular rubble, fuel assembly upper end fittings, control rod spiders, fuelmore » rod fragments, and fuel pellets. It was obvious that the normal plant refueling system not suitable for removing the damaged core. A new system of defueling tools and equipment was necessary to perform this task. Design of the new system was started immediately, followed by >1 yr of fabrication. Delivery and checkout of the defueling system occurred in mid-1985. Actual defueling was initiated in late 1985 with removal of the debris bed at the bottom of the core void. Obstructions to the debris, such as end fittings and fuel rod fragments ere removed first; then /approx/23,000 kg (50,000lb) of granular debris was quickly loaded into canisters. Core region defueling was completed in late 1987, /approx/2 yr after it was initiated.« less

Characterisation of Ceramic-Coated 316LN Stainless Steel Exposed to High-Temperature Thermite Melt and Molten Sodium

NASA Astrophysics Data System (ADS)

Ravi Shankar, A.; Vetrivendan, E.; Shukla, Prabhat Kumar; Das, Sanjay Kumar; Hemanth Rao, E.; Murthy, S. S.; Lydia, G.; Nashine, B. K.; Mallika, C.; Selvaraj, P.; Kamachi Mudali, U.

2017-11-01

Currently, stainless steel grade 316LN is the material of construction widely used for core catcher of sodium-cooled fast reactors. Design philosophy for core catcher demands its capability to withstand corium loading from whole core melt accidents. Towards this, two ceramic coatings were investigated for its application as a layer of sacrificial material on the top of core catcher to enhance its capability. Plasma-sprayed thermal barrier layer of alumina and partially stabilised zirconia (PSZ) with an intermediate bond coat of NiCrAlY are selected as candidate material and deposited over 316LN SS substrates and were tested for their suitability as thermal barrier layer for core catcher. Coated specimens were exposed to high-temperature thermite melt to simulate impingement of molten corium. Sodium compatibility of alumina and PSZ coatings were also investigated by exposing samples to molten sodium at 400 °C for 500 h. The surface morphology of high-temperature thermite melt-exposed samples and sodium-exposed samples was examined using scanning electron microscope. Phase identification of the exposed samples was carried out by x-ray diffraction technique. Observation from sodium exposure tests indicated that alumina coating offers better protection compared to PSZ coating. However, PSZ coating provided better protection against high-temperature melt exposure, as confirmed during thermite melt exposure test.

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

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

Peterson, Per; Greenspan, Ehud

2015-02-09

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. This report also includes results for additional studies relevant to the design and analysis of pebble bed reactor cores including the study of forces on shut down blades inserted directly into a packed bed and pebble flow in a cylindrical hopper that is representative of a small test reactor.« less

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

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

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using themore » WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)« less

HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio

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

Bess, John D.

2014-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

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

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

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

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Preliminary Design of Critical Function Monitoring System of PGSFR

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

NONE

2015-07-01

A PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) is under development at Korea Atomic Energy Research Institute. A critical function monitoring system of the PGSFR is preliminarily studied. The functions of CFMS are to display critical plant variables related to the safety of the plant during normal and accident conditions and guide the operators corrective actions to keep the plant in a safe condition and mitigate the consequences of accidents. The minimal critical functions of the PGSFR are composed of reactivity control, reactor core cooling, reactor coolant system integrity, primary heat transfer system(PHTS) heat removal, sodium water reaction mitigation, radiation controlmore » and containment conditions. The variables and alarm legs of each critical function of the PGSFR are as follows; - Reactivity control: The variables of reactivity control function are power range neutron flux instrumentation, intermediate range neutron flux instrumentation, source range neutron flux instrumentation, and control rod bottom contacts. The alarm leg to display the reactivity controls consists of status of control drop malfunction, high post trip power and thermal reactivity addition. - Reactor core cooling: The variables are PHTS sodium level, hot pool temperature of PHTS, subassembly exit temperature, cold pool temperature of the PHTS, PHTS pump current, and PHTS pump breaker status. The alarm leg consists of high core delta temperature, low sodium level of the PHTS, high subassembly exit temperature, and low PHTS pump load. - Reactor coolant system integrity: The variables are PHTS sodium level, cover gas pressure, and safeguard vessel sodium level. The alarm leg is composed of low sodium level of PHTS, high cover gas pressure and high sodium level of the safety guard vessel. - PHTS heat removal: The variables are PHTS sodium level, hot pool temperature of PHTS, core exit temperature, cold pool temperature of the PHTS, flow rate of passive residual heat removal system, flow rate of active residual heat removal system, and temperatures of air heat exchanger temperature of residual heat removal systems. The alarm legs are composed of two legs of a 'passive residual heat removal system not cooling' and 'active residual heat removal system not cooling'. - Sodium water reaction mitigation: The variables are intermediate heat transfer system(IHTS) pressure, pressure and temperature and level of sodium dump tank, the status of rupture disk, hydrogen concentration in IHTS and direct variable of sodium-water-reaction measure. The alarm leg consists of high IHTS pressure, the status of sodium water reaction mitigation system and the indication of direct measure. - Radiation control: The variables are radiation of PHTS, radiation of IHTS, and radiation of containment purge. The alarm leg is composed of high radiation of PHTS and IHTS, and containment purge system. - Containment condition: The variables are containment pressure, containment isolation status, and sodium fire. The alarm leg consists of high containment pressure, status of containment isolation and status of sodium fire. (authors)« less

Space power reactor in-core thermionic multicell evolutionary (S-prime) design

NASA Astrophysics Data System (ADS)

Determan, William R.; Van Hagan, Tom H.

1993-01-01

A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m2 heat pipe space radiator.

Beta ray flux measuring device

DOEpatents

Impink, Jr., Albert J.; Goldstein, Norman P.

1990-01-01

A beta ray flux measuring device in an activated member in-core instrumentation system for pressurized water reactors. The device includes collector rings positioned about an axis in the reactor's pressure boundary. Activated members such as hydroballs are positioned within respective ones of the collector rings. A response characteristic such as the current from or charge on a collector ring indicates the beta ray flux from the corresponding hydroball and is therefore a measure of the relative nuclear power level in the region of the reactor core corresponding to the specific exposed hydroball within the collector ring.

Interior of the Plum Brook Reactor Facility

NASA Image and Video Library

1961-02-21

A view inside the 55-foot high containment vessel of the National Aeronautics and Space Administration (NASA) Plum Brook Reactor Facility in Sandusky, Ohio. The 60-megawatt test reactor went critical for the first time in 1961 and began its full-power research operations in 1963. From 1961 to 1973, this reactor performed some of the nation’s most advanced nuclear research. The reactor was designed to determine the behavior of metals and other materials after long durations of irradiation. The materials would be used to construct a nuclear-powered rocket. The reactor core, where the chain reaction occurred, sat at the bottom of the tubular pressure vessel, seen here at the center of the shielding pool. The core contained fuel rods with uranium isotopes. A cooling system was needed to reduce the heat levels during the reaction. A neutron-impervious reflector was also employed to send many of the neutrons back to the core. The Plum Brook Reactor Facility was constructed from high-density concrete and steel to prevent the excess neutrons from escaping the facility, but the water in the pool shielded most of the radiation. The water, found in three of the four quadrants served as a reflector, moderator, and coolant. In this photograph, the three 20-ton protective shrapnel shields and hatch have been removed from the top of the pressure tank revealing the reactor tank. An overhead crane could be manipulated to reach any section of this room. It was used to remove the shrapnel shields and transfer equipment.

Upper internals arrangement for a pressurized water reactor

DOEpatents

Singleton, Norman R; Altman, David A; Yu, Ching; Rex, James A; Forsyth, David R

2013-07-09

In a pressurized water reactor with all of the in-core instrumentation gaining access to the core through the reactor head, each fuel assembly in which the instrumentation is introduced is aligned with an upper internals instrumentation guide-way. In the elevations above the upper internals upper support assembly, the instrumentation is protected and aligned by upper mounted instrumentation columns that are part of the instrumentation guide-way and extend from the upper support assembly towards the reactor head in hue with a corresponding head penetration. The upper mounted instrumentation columns are supported laterally at one end by an upper guide tube and at the other end by the upper support plate.

Heat dissipating nuclear reactor

DOEpatents

Hunsbedt, A.; Lazarus, J.D.

1985-11-21

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

Heat dissipating nuclear reactor

DOEpatents

Hunsbedt, Anstein; Lazarus, Jonathan D.

1987-01-01

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

Propellant actuated nuclear reactor steam depressurization valve

DOEpatents

Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

1992-01-01

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

Design, fabrication, and testing of an external fuel (UO2), full-length thermionic converter

NASA Technical Reports Server (NTRS)

Schock, A.; Raab, B.

1971-01-01

The development of a full-length external-fuel thermionic converter for in-pile testing is described. The development program includes out-of-pile performance testing of the fully fueled-converter, using RF-induction heating, before its installation in the in-pile test capsule. The external-fuel converter is cylindrical in shape, and consists of an inner, centrally cooled collector, and an outer emitter surrounded by nuclear fuel. The term full-length denotes that the converter is long enough to extend over the full height of the reactor core. Thus, the converter is not a scaled-down test device, but a full-scale fuel element of the thermionic reactor. The external-fuel converter concept permits a number of different design options, particularly with respect to the fuel composition and shape, and the collector cooling arrangement. The converter described was developed for the Jet Propulsion Laboratory, and is based on their concept for a thermionic reactor with uninsulated collector cooling as previously described. The converter is double-ended, with through-flow cooling, and with ceramic seals and emitter and collector power take-offs at both ends. The design uses a revolver-shaped tungsten emitter body, with the central emitter hole surrounded by six peripheral fuel holes loaded with cylindrical UO2 pellets.

A new approach to nuclear fuel safeguard enhancement through radionuclide profiling

NASA Astrophysics Data System (ADS)

Peterson, Aaron Dawon

The United States has led the effort to promote peaceful use of nuclear power amongst states actively utilizing it as well as those looking to deploy the technology in the near future. With the attraction being demonstrated by various countries towards nuclear power comes the concern that a nation may have military aspirations for the use of nuclear energy. The International Atomic Energy Agency (IAEA) has established nuclear safeguard protocols and procedures to mitigate nuclear proliferation. The work herein proposed a strategy to further enhance existing safeguard protocols by considering safeguard in nuclear fuel design. The strategy involved the use of radionuclides to profile nuclear fuels. Six radionuclides were selected as identifier materials. The decay and transmutation of these radionuclides were analyzed in reactor operation environment. MCNPX was used to simulate a reactor core. The perturbation in reactivity of the core due to the loading of the radionuclides was insignificant. The maximum positive and negative reactivity change induced was at day 1900 with a value of 0.00185 +/- 0.00256 and at day 2000 with -0.00441 +/- 0.00249, respectively. The mass of the radionuclides were practically unaffected by transmutation in the core; the change in radionuclide inventory was dominated by natural decay. The maximum material lost due to transmutation was 1.17% in Eu154. Extraneous signals from fission products identical to the radionuclide compromised the identifier signals. Eu154 saw a maximum intensity change at EOC and 30 days post-irradiation of 1260% and 4545%, respectively. Cs137 saw a minimum change of 12% and 89%, respectively. Mitigation of the extraneous signals is cardinal to the success of the proposed strategy. The predictability of natural decay provides a basis for the characterization of the signals from the radionuclide.

Lessons Learned about Liquid Metal Reactors from FFTF Experience

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

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

2016-09-20

The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens.more » In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports« less

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