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Nuclear analysis of the chornobyl fuel containing masses with heterogeneous fuel distribution.

Description: Although significant data has been obtained on the condition and composition of the fuel containing masses (FCM) located in the concrete chambers under the Chernobyl Unit 4 reactor cavity, there is still uncertainty regarding the possible recriticality of this material. The high radiation levels make access extremely difficult, and most of the samples are from the FCM surface regions. There is little information on the interior regions of the FCM, and one cannot assume with confidence that the surface measurements are representative of the interior regions. Therefore, reasonable assumptions on the key parameters such as fuel concentration, the concentrations of impurities and neutron poisons (especially boron), the void fraction of the FCM due to its known porosity, and the degrees of fuel heterogeneity, are necessary to evaluate the possibility of recriticality. The void fraction is important since it introduces the possibility of water moderator being distributed throughout the FCM. Calculations indicate that the addition of 10 to 30 volume percent (v/o) water to the FCM has a significant impact on the calculated reactivity of the FCM. Therefore, water addition must be considered carefully. The other possible moderators are graphite and silicone dioxide. As discussed later in this paper, silicone dioxide moderation does not represent a criticality threat. For graphite, both heterogeneous fuel arrangements and very large volume fractions of graphite are necessary for a graphite moderated system to go critical. Based on the observations and measurements of the FCM compositions, these conditions do not appear creditable for the Chernobyl FCM. Therefore, the focus of the analysis reported in this paper will be on reasonable heterogeneous fuel arrangements and water moderation. The analysis will evaluate a range of fuel and diluent compositions.
Date: October 14, 1998
Creator: Turski, R. B.
Partner: UNT Libraries Government Documents Department

Post-initiating phase neutronics analysis of an unprotected LOF event in CRBRP

Description: The reactor system is expected to achieve permanent subcriticality in a loss-of-flow (LOF) event by virtue of fuel removal from the core even under the hypothetical assumption that both shutdown systems fail to function. Based on the analysis performed by S.K. Rhow, et al., adequate fuel removal would occur in the CRBRP heterogeneous core during a meltout period after the initiating phase of the unprotected LOF event. This paper discusses reactivity levels relative to fuel removal in the accident progression beyond the initiating phase for the CRBRP core at the beginning of cycle 1 (BOC-1).
Date: January 1, 1983
Creator: Turski, R.B. & Rhow, S.K.
Partner: UNT Libraries Government Documents Department

Performance and design considerations in metal fueled cores. [LMFBR]

Description: To focus future metal fuel development requirements a study was performed to quantify the relationship between some critical core design parameters. The fuel studied was U-Pu-Zr alloy. Of interest are performance parameters, such as peak Pu enrichment, burnup swing, fast fluence, breeding ratio, and their relation to core parameters such as reactor size, degree of core heterogeneity, pin diameter, and linear heat rating. These performance parameters, while numericaly different from those of ceramic fuels, were found to exhibit the same qualitative dependence on the key design variables.
Date: 1984~
Creator: Orechwa, Y.; Khalil, H. & Turski, R. B.
Partner: UNT Libraries Government Documents Department

Pin diameter optimization in 1200 MWe heterogeneous vs. homogeneous LMFBRs

Description: LMFBRs with internal blankets (heterogeneous reactors) are known for reducing the sodium void reactivity and increasing the breeding ratio. As for homogeneous reactors, the optimization of the fuel pin diameter for heterogeneous reactors is of great interest. The optimum pin diameter is obtained by changing the fuel pin diameter until the homogenized fuel volume fraction is the same as the optimum fuel volume fraction of the homogeneous core. The optimization of the fuel pin diameter with respect to doubling time for a loosely coupled 1200 MWe oxide heterogeneous reactor is described. The results are compared with those of a homogeneous reactor.
Date: January 1, 1977
Creator: Orechwa, Y.; Turski, R.B. & King, M.J.
Partner: UNT Libraries Government Documents Department

Performance of U--Pu--Zr metal fuel in 1000 MWe LMFBRs

Description: The present analyses indicate that very conservatively designed U--Pu--Zr metal-fueled 1000 MW(e) LMFBRs have CSDT in the range of 9 to 10 years, with corresponding specific inventories ranging from 2.5 to 3.5 kg/MWe. Doubling times in the range of 8 years are achievable with only minor changes in the fuel pin design or operating conditions.
Date: January 1, 1979
Creator: Lam, P.S.K.; Turski, R.B. & Barthold, W.P.
Partner: UNT Libraries Government Documents Department

Core design and performance of small inherently safe LMRs

Description: Oxide and metal-fueled core designs at the 900 MWt level and constrained by a requirement for interchangeability are described. The physics parameters of the two cores studied here indicate that metal-fueled cores display attractive economic and safety features and are more flexible than are oxide cores in adapting to currently-changing deployment scenarios.
Date: January 1, 1986
Creator: Orechwa, Y.; Khalil, H.; Turski, R.B. & Fujita, E.K.
Partner: UNT Libraries Government Documents Department

International Nuclear Safety Center (INSC) database

Description: As an integral part of DOE`s International Nuclear Safety Center (INSC) at Argonne National Laboratory, the INSC Database has been established to provide an interactively accessible information resource for the world`s nuclear facilities and to promote free and open exchange of nuclear safety information among nations. The INSC Database is a comprehensive resource database aimed at a scope and level of detail suitable for safety analysis and risk evaluation for the world`s nuclear power plants and facilities. It also provides an electronic forum for international collaborative safety research for the Department of Energy and its international partners. The database is intended to provide plant design information, material properties, computational tools, and results of safety analysis. Initial emphasis in data gathering is given to Soviet-designed reactors in Russia, the former Soviet Union, and Eastern Europe. The implementation is performed under the Oracle database management system, and the World Wide Web is used to serve as the access path for remote users. An interface between the Oracle database and the Web server is established through a custom designed Web-Oracle gateway which is used mainly to perform queries on the stored data in the database tables.
Date: August 1, 1997
Creator: Sofu, T.; Ley, H. & Turski, R.B.
Partner: UNT Libraries Government Documents Department

Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

Description: Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of {minus}3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept.
Date: March 1, 1994
Creator: Wigeland, R. A.; Turski, R. B. & Pizzica, P. A.
Partner: UNT Libraries Government Documents Department

Tradeoff of sodium void worth and burnup reactivity swing: Impacts on balance safety position in metallic-fueled cores

Description: A study has been conducted to investigate the effect of a lower sodium void worth on the consequences of severe accidents in metallic-fueled sodium-cooled reactors. Four 900 MWth designs were used for the study, where all of the reactor cores were designed based on the metallic fuel of the Integral Fast Reactor (IFR) concept. The four core designs each have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation was to determine the differences in severe accident response for the four core designs, in order to estimate the improvement in overall safety that could be achieved from a reduction in the sodium void worth for reactor cores which use a metallic fuel form.
Date: October 1, 1994
Creator: Wigeland, R. A.; Turski, R. B. & Pizzica, P. A.
Partner: UNT Libraries Government Documents Department

Macroscopic cross section generation and application for coupled spatial kinetics and thermal hydraulics analysis with SAS-DIF3DK

Description: This paper discusses the importance of modeling the transient behavior of multigroup cross sections in the context of coupled reactor physics and thermal-hydraulic computations with the SAS-DIF3DK computer code. The MACOEF macroscopic cross section methodology is presented. Results from benchmark verification calculations with a continuous-energy Monte Carlo are reported. Analysis of the Chernobyl accident is made using correlated WIMS-D4M generated group constants with special emphasis placed on the impact of modeling assumptions on the progression of the accident simulation.
Date: August 1, 1997
Creator: Turski, R.B.; Morris, E.E.; Taiwo, T.A. & Cahalan, J.E.
Partner: UNT Libraries Government Documents Department

Low sodium void cores

Description: To avoid high energy releases in LMFBR TUC accidents which are accompanied by a failure to scram with a regular shutdown system, various devices have been proposed which would add negative reactivity to the core by either bringing poison material into the core or by creating negative reactivity feedbacks coming from the thermal expansion of the core. While inherent shutdown systems (ISSs) show promise for enhancing safety by adding poison to the reactor, the trigger mechanism and the geometry of the poison are critical design issues. But by postulating an ''unprotected'' accident, an emergency shutdown capability is denied by definition and the core has to rely solely on inherent safety features. Thermal expansion of core components can lead to the addition of negative reactivity to the core, however, dependng on the core restraint system this reactivity margin is usually small and does not prevent sodium boiling. Another approach to enhance the safety of the core is to modify the core design such that the removal of sodium from the core would add only a small amount of positive reactivity or even a negative reactivity. There exists no firm and quantified design goal for the maximum allowable sodium void reactivity. In a simplistic approach, a sodium void reactivity of approximately $2 or less would be desirable to compensate for roughly $1 in Doppler reactivity and $1 as the margin to the prompt-critical stage. But voiding incoherence and fuel motion patterns certainly affect this target value. At this stage of knowledge, it is also not clear if all fuel types would require the same reduction in sodium void reactivity to assure a benign transition phase leading to the termination of a TUC event. A discussion is presented of low sodium void core options which concentrates on the means of lowering sodium void ...
Date: January 1, 1978
Creator: Barthold, W.P.; Beitel, J.C.; Lam, P.S.K.; Orechwa, Y.; Su, S.F. & Turski, R.B.
Partner: UNT Libraries Government Documents Department

Criticality safety strategy for the Fuel Cycle Facility electrorefiner at Argonne National Laboratory, West

Description: The Integral Fast Reactor being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal-cooled reactors and a closed fuel cycle. Presently, the Fuel Cycle Facility (FCF) at ANL-West in Idaho Falls, Idaho is being modified to recycle spent metallic fuel from Experimental Breeder Reactor II as part of a demonstration project sponsored by the Department of Energy. A key component of the FCF is the electrorefiner (ER) in which the actinides are separated from the fission products. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt and refined uranium or uranium/plutonium products are deposited at cathodes. In this report, the criticality safety strategy for the FCF ER is summarized. FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. In order to show criticality safety for the FCF ER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOES) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOES, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that wig verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCF ER operations were found to be safe with respect to criticality.
Date: September 1, 1993
Creator: Mariani, R. D.; Benedict, R. W.; Lell, R. M.; Turski, R. B. & Fujita, E. K.
Partner: UNT Libraries Government Documents Department

Criticality safety evaluation of the fuel cycle facility electrorefiner

Description: The integral Fast Reactor (IFR) being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal cooled reactors and a closed-loop fuel cycle. Some of the primary advantages are passive safety for the reactor and resistance to diversion for the heavy metal in the fuel cycle. in addition, the IFR pyroprocess recycles all the long-lived actinide activation products for casting into new fuel pins so that they may be burned in the reactor. A key component in the Fuel Cycle Facility (FCF) recycling process is the electrorefiner (ER) in which the actinides are separated from the fission products. In the process, the metal fuel is electrochemically dissolved into a high-temperature molten salt, and electrorefined uranium or uranium/plutonium products are deposited at cathodes. This report addresses the new and innovative aspects of the criticality analysis ensuing from processing metallic fuel, rather than metal oxide fuel, and from processing the spent fuel in batch operations. in particular, the criticality analysis employed a mechanistic approach as opposed to a probabilistic one. A probabilistic approach was unsuitable because of a lack of operational experience with some of the processes, rendering the estimation of accident event risk factors difficult. The criticality analysis also incorporated the uncertainties in heavy metal content attending the process items by defining normal operations envelopes (NOES) for key process parameters. The goal was to show that reasonable process uncertainties would be demonstrably safe toward criticality for continuous batch operations provided the key process parameters stayed within their NOES. Consequently the NOEs became the point of departure for accident events in the criticality analysis.
Date: September 1, 1993
Creator: Lell, R. M.; Mariani, R. D.; Fujita, E. K.; Benedict, R. W. & Turski, R. B.
Partner: UNT Libraries Government Documents Department

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

Description: A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance.
Date: November 1, 1979
Creator: Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C. et al.
Partner: UNT Libraries Government Documents Department

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

Description: A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance.
Date: November 1, 1979
Creator: Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C. et al.
Partner: UNT Libraries Government Documents Department

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

Description: A parameter study was conducted to determine the interrelated effects of: loosely of tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance.
Date: November 1, 1979
Creator: Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C. et al.
Partner: UNT Libraries Government Documents Department

Optimization of radially heterogenous 1000-MW(e) LMFBR core configurations. Appendix C. Research project 620-25

Description: A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance.
Date: November 1, 1979
Creator: Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C. et al.
Partner: UNT Libraries Government Documents Department