Uranium-water Lattices Research Articles

Critical Experiments for Investigation of VVER Type Uranium-Water Lattices with Relatively Highly Enriched Fuel and Erbium Oxide Absorbers

Critical Experiments for Investigation of VVER Type Uranium-Water Lattices with Relatively Highly Enriched Fuel and Erbium Oxide Absorbers

Computational Analysis of Critical Experiments for Uranium-Water Lattices of VVER Type with a Relatively Highly Enriched Fuel and Erbium Oxide Absorbers

Computational Analysis of Critical Experiments for Uranium-Water Lattices of VVER Type with a Relatively Highly Enriched Fuel and Erbium Oxide Absorbers

Verification of Subcritical Limits in ANSI/ANS-8.1-2014

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.

VERIFICATION OF SUBCRITICAL LIMITS IN ANSI/ANS-8.1-2014

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 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 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 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.

Open web system of Virtual labs for nuclear and applied physics

An example of virtual lab work on unique experimental equipment is presented. The virtual lab work is software based on a model of real equipment. Virtual labs can be used for educational process in nuclear safety and analysis field. As an example it includes the virtual lab called “Experimental determination of the material parameter depending on the pitch of a uranium-water lattice”. This paper included general description of this lab. A description of a database on the support of laboratory work on unique experimental equipment which is included this work, its concept development are also presented.

Modeling of critical experiments on the ZR-6 assembly with VVÉR-type uranium-water lattices

An electronic atlas for describing the 2D and 3D mathematical models of critical experiments carried out on the ZR-6 assembly has been developed. A program-generator of the initial data for modeling assemblies using the MCU-RFFI/A program has been produced, by means of which mathematical modeling of 154 configurations can be carried out. The results have been compared with experimental data. For the two-dimensional model the average value of keff over all the configurations amounts to 0.9989±0.0034 (for the KENO-Va program it is 0.9939±0.0048). For the three-dimensional model of 13 configurations an average value of keff of 0.9994±0.0023 has been obtained (for the same assembly in the two-dimensional model the value obtained is 0.99871±0.0030). This enables us to recommend the two-dimensional models investigated for verifying programs of neutron-physical calculations on VVER reactors.

Application of subcriticality measuring methods for studying the temperature effect of the reactivity of uranium-water lattice assemblies

This paper presents some results of investigations of the temperature dependence of the reactivity of the MATR-1 stand of the Physics Power Institute; two techniques were used: a pulsed neutron source and an expelled source. The proposed scheme of the investigations was checked in uranium-water assemblies in which a reflector is important for the reactivity balance. The kinetic parameters were assumed for U 235 in both the experiments and calculations. The calculations were made in the two-group diffusion approximation in two dimensional geometry with the FACTOR and DNESTR programs. The constnats of the thermal neutron group were determined taking into account the heterogeneous core structure by averaging over the space-energy distribution of the neutron flux in modeled Wigner-Seitz cells. Results show that the systematic application of subcriticality measuring methods allows the direct determination of the importance of the core and reflector shapes i nthe development of the temperature dependence of the reactivity.

Multigroup Calculations of the Fast Spectrum and Fast Fission Effect

A multigroup procedure for the calculation of the fast fission phenomena in thermal uranium-water reactors has been developed. The method essentially consists of applying the single-flight collision concept in a manner analogous to the calculation of resonance capture in thermal reactor lattices. The collision and escape probabilities are calculated by numerical integration over the actual neutron paths encountered in a reactor lattice. The multigroup equations are solved by an iterative procedure which converges rapidly. The fast neutron spectrum, δ28 and ɛ can be obtained.Results of calculations are presented in which the value of δ28 homogeneous uranium-water mixtures and for slightly-enriched uranium-water lattices are compared with Monte Carlo calculations and experiment. Very satisfactory agreement has been obtained. Fast neutron spectra in the core of a pool type reactor and in the fuel and moderator regions of a uranium-water lattice, calculated by the present method, are also shown.

A new method of reconstructing true spectra

A new method of solving first order integral equations is proposed, which enables spectra distorted by the measuring instrument to be reconstructed with some meaning. This method is used for processing the results of measuring the energy spectrum of slow neutrons in a uranium-water lattice.

Measurement of neutron spectra in uranium-water and uranium-monoisoprolyldiphenyl lattices

The neutron spectra in a cell of a heterogeneous subcritical system are investigated in detail. The neutron spectra were measured by means of a mechanical velocity selector and a time analyzer. A water-cooled water-moderated reactor (WWR) was used as the neutron source for the subcritical system. The results indicate that the neutron spectrum differs from the Maxwell equilibrium distribution both in the uranium and in the moderator. The neutron spectrum changes sharply at the boundary between the uranium block and the moderator. The spectrum varies only slightly within the block and within the moderator. A simple semiempirical functional relationship is found to exist between the thermal neutron spectra in the moderator and in the block.

Measurements of Relative Pu Fission Rates in Slightly Enriched Uranium—Water Lattices

The relative fission activation integrals of Pu/sup 239/, Pu/sup 2/ /sup 40/, and Pu/sup 241/ are measured. Foils of these isotopes are expo sed to the neutron flux of the TRX reactor, and the fission activation values are found from the relative fission product activities on the foils. (T.F.H.)

The Fast Effect in Uranium and Beryllium Systems

A Monte Carlo code is used to calculate the fast fission factor epsilon and the fast fission ratio delta in both homogeneous and heterogeneous systems. The following results are obtained: for an infinite block of natural uranium, delta = 0.39, epsilon = 1.247; for an infinite block of beryllium, epsilon = 1.078; for beryllium oxide, = 1.046. In addition, results are given for homogeneous mixtures of uranium and beryllium. Calculations are also carried out for uranium-water lattices and compared with experimental results for slightly enriched uranium rods and slabs, and uranium oxide rods. Other results show the increase in epsilon when uranium fuel elements are surrounded by beryllium cladding of varying thickness. (auth)

An Experimental Study of the Relative U235Fission Activation as a Function of Energy in Slightly Enriched Uranium-Water Lattices

Measurements of the relative U/sup 235/ fission rates as a function of energy were made for the TRX facility, a slighly enriched uranium, light water moderated critical assembly. The parameter directly measured is the ratio of the activity of a bare U/sup 235/ foil to that of a similar foil enclosed in a box of absorbing material, which was either cadmium, boron, or gadolinium. The energy dependence associated with these ratios was obtained by the introduction to effective cutoff energies for the absorbing shields. A comparison is made with calculated values, based upon a simplified model for the neutron energy spectrum present, and the agreement is considered ade quate. (auth)

Some Monte Carlo and Analytical Results for Resonance Capture in Lattices

Detailed Monte Carlo studies of the resonance capture in the lowest U/ sup 238/ resonances were made for uraniumwater lattices and a uranium-graphite lattice. Direct comparison with calculated values was considered a gocd test of the validity of the assumptions made in obtaining the theoretical estimates. Good agreement was obtained for the over-all capture in a single resonance and for the energy distribution of absorptions. Monte Carlo tests also showed that the asymptotic flux was fairly well reestablished between the two strongest U/sup 238/ resonances, and that the number of neutrons captured in this energy region was only slightly affected by Doppler broadening and interference between resonance and potential scattering. Finally the calculated resonance escape probability in a uranium-water lattice for 26 resolved resonances was compared with the Monte Carlo value. There was quite close agreement. (auth)

Some Refinements in the Calculation of Resonance Integrals

Two basic formulas for resonance absorption applicable both to mixtures and to lumps are considered: the narrow resonance (NR) approximation and the infinite mass (NRIA) approximation. The formulas are shown to be complementary, yielding accurate results when the choice between them is hased on the practical width of the resonance line as originally suggested by Wigner. The formulas are used to calculate resonance integrals for U/sup 238/ and Th/sup 232/. The results yield a low mass absorption term and a surface absorption term proportional to the square root of the surface-to-mass ratio for lumps of practical size, in quantitative agreement with the experimental work of Egiazarov and Hellstrand for U/sup 238/ and with Dayton and Pettus for thorium. Dresner's suggestion that the ratio of the resonance integral to the mass absorption term is independent of the resonance structure is not borne out. Refinement of the basic formulas is discussed. The correction of the NRIA formula for energy degradation is in agreement with Spinney's calculations for U-H mixtures and with Monte Carlo results obtained by Auerbach for uranium-water lattices. Consideration of lumping effects indicates that the basic formulas generally underestimate the resonance absorption. It is therefore recommended that the common usemore » of ill-defined flux disadvantage factors be dropped. (auth)« less

Neutron diffusion in a plane uranium-water lattice

The diffusion lengths for a plane uranium-water lattice are calculated for thermal-neutron diffusion parallel (L∥) and perpendicular (L⊥) to the layers. The calculation of L⊥ is carried out in the diffusion approximation but can be used without the introduction of nondiffusion corrections since it is correct in the limiting cases (thick and thin layers). The calculation of L∥ is carried out both in the diffusion and nondiffusion approximation: in the latter case, however, it is assumed that the absorption is small. In the case of thick layers, the ratio\(\frac{{L_\parallel ^2 }}{{L_ \bot ^2 }}\) may reach 1.9. It is shown that if a uranium-aluminum alloy is used the anisotropy becomes still larger. As the layer thickness is reduced the anisotropy is reduced and finally vanishes.

A determination of the ratio of effective fission cross-sections for Pu 239 and U 235 in slugs in uranium-water lattices

The ratio of the mean fission cross-sections of Pu 239 and U 235 was determined in natural uranium—ordinary water lattices. For comparison the ratio was also measured in a uranium-graphite reactor. Values of σ ̄ f Pu σ ̄ f U in uranium-water lattices with 45-, 50-; 55-, and 60-mm pitch, and in the uranium-graphite reactor with a 200-mm lattice pitch were found to be 2·24, 1·99, 1·88, 1·79, and 1·79, respectively.

Determination of the relative U 235-Pu 239 conversion factors in lattices of natural uranium and ordinary water

The ratios of the initial conversion factors for various uranium-water lattices to the conversion factor in a uranium-graphite lattice have been measured. The results show that in uranium-water lattices of the pitch indicated, the breeding coefficient of plutonium is greater than in a uranium-graphite lattice.

The slowing down of fission neutrons in uranium-water media

The density distribution of resonance neutrons ( E = 1·46 eV), originating from a point source of U 235 fission neutrons, has been measured in three different uranium-water lattices. The uranium was in the form of thick rods (35 mm) of the natural element sheathed in cadmium. It was shown that there was no anisotropy in the neutron distribution, and the value of τ was determined.

The ratio of the average fission cross sections of Pu239 and U235 in uranium-water lattices

The ratio of the average fission cross sections of Pu239 and U235 is determined for lattices consisting of natural uranium and ordinary water. For comparison the same ratio was measured for a uranium-graphite reactor. It was found that the ratio σPu−f/σU−f for uranium-water lattices with spacings of 45, 50, 55 and 60 mm and for a uranium-graphite reactor with 200 mm lattice spacing are 2.24, 1.99, 1.88 and 1.79, respectively.