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Subcritical Limits for Plutonium Systems

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Abstract

As a contribution to a required review of American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, limits for plutonium systems have been recalculated to confirm their subcriticality under the stated conditions or to propose other values. This paper presents calculation methods and results. The validity of each calculation method was established by extensive correlation with critical experiments, and in some cases with experiments performed subsequent to the original limit calculations. 48 refs.

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... The ANS-8.1 SCLs have typically assumed a computational bias and bias uncertainty of about 2% in the keff (0.98). Although it is difficult to verify [12][13][14], it is likely that the final SCLs chosen by past working groups were based on a combination of both computational and experimental data. The results presented here are those with a keff value comparable to the legacy SCLs in the standard (0.98), which corresponds to a bias and bias uncertainty of about 2% in keff. ...
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The ANSI/ANS-8.1 standard, “Safety Standard for Operations with Fissionable Materials Outside Reactors,” has been available since 1964 as ASA N6.1-1964. In 1969, this standard was revised as ANSI N16.1-1969, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.” This version of the standard includes a variety of subcritical limits (SCLs) for uniform aqueous solutions and metals containing fissile nuclides for ²³³ U, ²³⁵ U, and ²³⁹ Pu. Furthermore, SCLs are also included for uranium-water lattices. In the 1983 version of ANSI/ANS-8.1 (a revision of ANSI N16.1-1975), the suite of SCLs in the standard expanded to include ²³⁵ U enrichment limits for homogeneous uranium-water mixtures and dry/damp oxides, uniform aqueous solutions of low-enriched uranium, and uniform aqueous mixtures of Pu(NO 3 ) 4 containing ²⁴⁰ Pu, in addition to the SCLs included in ANSI N16.1-1969. The SCLs have changed little in subsequent revisions (ANSI/ANS-8.1-1998 and ANSI/ANS-8.1-2014). The ANSI/ANS-8.1-2014 standard is currently being revised to include new SCLs (uranium metal and compounds with enrichments up to 20 wt. % ²³⁵ U) and possible updates to the current SCLs already in the standard, although these SCLs will not be available to the nuclear criticality safety community for a number of years. The bases for these SCLs were documented in journal articles such as Nuclear Science and Engineering , and the American Nuclear Society’s meeting transactions; however, the bases were ambiguous enough that sites and regulators in the United States are reluctant to endorse them for safety purposes. The purpose of this paper is to present the results of a comparison study for the SCLs in the ANSI/ANS-8.1-2014 standard using modern codes and cross sections (SCALE/ENDF/B-VIII) to provide some assurance about their quality (bias and bias uncertainty) for use in nuclear criticality safety applications.
... Separated Pu is likely to contain the neutron poison 240 Pu as well as the fissile isotopes 239 Pu and 241 Pu. Although 241 Pu is more reactive than 239 Pu, provided the Pu contains a greater mass of 240 Pu than 241 Pu, it is conservative to model all Pu isotopes as 239 Pu (Clark, 1981). Further, the GCSA (Hicks and Rudge, 2007) presented a series of criticality calculations for a Pu system in which part of the 239 Pu was replaced by an equal mass of 241 Pu (keeping the total mass of 239 Pu and 241 Pu constant), while a mass of 240 Pu equal to that of 241 Pu was added. ...
... This approach assumes that all plutonium in the irradiated fuel is 239 Pu, thereby ignoring the effects of the neutron poison 240 Pu, and that the 239 Pu is represented by a fissile equivalent mass of 235 U. Although 241 Pu is more reactive than 239 Pu, provided the Pu contains a 0560-1 Version 1.1 greater mass of 240 Pu than 241 Pu, it is conservative to model all Pu isotopes as 239 Pu (Clark, 1981). Irradiated Magnox fuel will not contain 233 U (Hicks, 2007a). ...
Article
The first edition of the current American National Standards Institute (ANSI)/American Nuclear Society (ANS) standard ANSI/ANS-8.1-2014, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,” was published in 1964 as ASA N6.1-1964, “Safety Standard for Operations with Fissionable Materials Outside Reactors.” In 1969, that standard was revised as ANSI N16.1-1969, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.” ANSI N16.1-1969 includes a variety of subcritical limits (SCLs) for uniform aqueous solutions and metals containing fissile nuclides of ²³³U, ²³⁵U, and ²³⁹Pu. Furthermore, SCLs are also included for uranium-water lattices. In the 1983 version of ANSI/ANS-8.1 (a revision of ANSI N16.1-1975), the suite of SCLs in the standard was expanded to include ²³⁵U enrichment limits for homogeneous uranium-water mixtures and dry/damp oxides, uniform aqueous solutions of low-enriched uranium, and uniform aqueous mixtures of Pu(NO3)4 containing ²⁴⁰Pu in addition to the SCLs included in ANSI N16.1-1969. The SCLs have changed little in subsequent revisions (ANSI/ANS-8.1-1998 and ANSI/ANS-8.1-2014). The ANSI/ANS-8.1-2014 standard is currently being revised to include new SCLs (uranium metal and compounds with enrichments up to 20 wt% ²³⁵U) and possible updates to the current SCLs already in the standard, although these new/updated SCLs will not be available to the nuclear criticality safety (NCS) community for a number of years. The original bases for these SCLs were documented in papers in journals such as Nuclear Science and Engineering and in ANS meeting transactions; however, these bases are ambiguous enough that sites and regulators in the United States have not widely endorsed them for safety purposes. The purpose of this paper is to present the results of a comparison study for the SCLs in the ANSI/ANS-8.1-2014 standard using modern codes (SCALE and MCNP) and cross sections (ENDF/B-VIII.0) to provide some assurance about their quality (bias and bias uncertainty) for use in NCS applications and for consideration by the ANS-8.1 Working Group as a reference for future revisions.
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In recent years, neural network metamodels have become increasingly popular for reducing the computational burden of performing direct, simulation-based analysis of physical systems. This paper proposes a new methodology for training a neural network metamodel and incorporating it into a Bayesian network-based probabilistic risk assessment. This methodology can be applied to a wide variety of industrial accidents, where there is at least one latent variable that is normally calculated using a physics code. The main benefit of this methodology is that it combines the interpretability and sampling algorithm of a Bayesian network with the high-dimensional, latent variable modeling capability of a neural network metamodel. This paper also provides an example of how this methodology is applied to fissionable material operations in a nuclear facility to estimate process criticality accident risk. Although process criticality accidents are specific to the nuclear industry, the methodology described in this paper can be adapted to other types of industrial accidents and rare events.
Article
Up to now, criticality safety experts used density laws fitted on experimental data and applied them outside the measurement range. Depending on the case, such an approach could be wrong for nitrate solutions. Seven components are concerned: UO2(NO3)2, U(NO3)4, Pu(NO3)4, Pu(NO3)3, Th(NO3)4, Am(NO3)3, and HNO3. To obviate this problem, a new methodology based on the thermodynamic concept of mixtures of binary electrolytes solutions (one electrolyte + water) at constant water activity, a so-called “isopiestic” solution, has been developed by the Institute de Radioprotection et de Sûreté Nucléaire (IRSN) to calculate the nitrate solutions density. This paper presents its qualification by using criticality experiments. The theory and the implementation are also given. Qualification results of the uranyl and plutonium nitrate solutions show that the new density law (also called the isopiestic law) is in good agreement with the benchmarks. Thus, no bias is put into evidence for the uranium solutions, and a small negative bias equal to 0.2% is found for the plutonium solutions. Moreover, the isopiestic law corrects the observed 1% overestimation of keff due to the empirical IRSN Leroy and Jouan density law for uranium solutions and the observed 3.4% underestimation of keff due to the ARH-600 density law for plutonium solutions. The isopiestic density law has been implemented in CIGALES V2.0, the graphical user interface of the French criticality safety package CRISTAL that calculates the atom densities of nuclides (and writes the input file for CRISTAL computations).
Article
Subcommittee 8 of the Standards Committee of the American Nuclear society is revising the Standard for Nuclear Criticality Control and Safety of Homogeneous Plutonium-Uranium Fuel Mixtures Outside Reactors to include limits on heterogeneous systems. In connection with this effort, a number of criticality calculations were completed for mixed-oxide (PuO/sub 2/+UO/sub 2/) fuel pins in water. The concentration of PuO/sub 2/ in the UO/sub 2/ (natural uranium) covered the range from 3.0 to 34 wt%. The isotopic makeup of the plutonium was also varied, up to 25 wt% /sup 240/Pu and 15 wt% /sup 24l/Pu. A search was made on fuel pin diameters and water-to-fuel volume ratios to obtain minimum critical dimensions and masses for a given fuel composition. Calculations made independently by several different members of the Work Group are compiled and compared, together with the proposed subcritical control limits for the Standard. Some difficulties were encountered with calculations pertaining to 30% PuO/sub 2/ at /sup 240/ Pu concentrations at water-to-fuel volume ratios and fuel pin diameters outside the area covered by any critical experiment. For this reason, dimensional limits on heterogeneous systems are not being proposed at this time for the Standard with 30% PuO/sub 2/ at a /sup 240/Pu content of 25%. In general, for a given fuel composition of mixed oxides, a heterogeneous arrangement of fuel pins of optimum diameter in water results in substantially smaller minimum critical dimensions than are obtainable for and aqueous homogeneous plutonium-uranium fuel mixture.
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