Intergranular stress corrosion cracking susceptibility of sensitized type 304 steel in 561 K pure water under γ-ray irradiation

Intergranular stress corrosion cracking (IGSCC) of sensitized type 304 stainless steel has been investigated in 561 K water under γ-ray irradiation at a flux of 2.6 × 10³ C kg⁻¹ h⁻¹ by slow-strain-rate tensile tests. The IGSCC susceptibility was enhanced by γ-ray irradiation in water containing 8 ppm dissolved oxygen (DO). The DO dependence of the IGSCC susceptibility was observed in the water under γ-irradiation. Although slight IGSCC susceptibility was observed even in deaerated water (less than 1 ppb DO) under γ-ray irradiation, the susceptibility was completely suppressed by injection of hydrogen into the water. The enhancement of IGSCC susceptibility seems to be related to the formation of H₂O₂ in high temperature water by radiolysis under γ-ray irradiation and the H₂O₂ formation rate is markedly decreased by hydrogen injection.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

HTGR core auxiliary cooling system availability and reliability

Using fault tree techniques, a quantitative estimate is made to predict both the start-up availability and operational reliability of the core auxiliary cooling system (CACS) of an HTGR following the postulated, simultaneous occurrence of a design basis depressurization accident (DBDA) and the complete loss of main loop cooling (LOMLC). The effects of a postulated, concurrent loss of offsite power are considered as well. Several potential common mode failures are identified. The limited availability of data presents a problem to numerical evaluation and estimates of uncertainty are at best crude. To provide a basis for measure of this uncertainty, the fault trees were solved using, on a consistent basis, either ‘optimistic’ failure rates, ‘pessimistic’ failure rates, or mean values (the geometric mean).Generally, about 80% of the failure rate data was larger than the ‘optimistic’ value, while only 20% was larger than the ‘pessimistic’ value. Predicated on a variety of assumptions, many of which may be unduly pessimistic, the CACS unavailability following a postulated DBDA and LOMLC has been estimated to be between 4 × 10⁻⁷ and 3 × 10⁻⁵ for the 2000 MW (th) HTGR and between 5 × 15⁻⁷ and 5 × 10⁻⁵ for the 3000 MW (th) HTGR. At the end of 300 hr, the estimated probability that the CACS will not leave sufficient core cooling capacity varies between 9 × 10⁻⁵ and 4 × 10⁻² for the smaller plant and 3 × 10⁻⁴ and 6 × 10⁻² for the larger plant. If it is further postulated that offsite power is concurrently lost, then the estimated mean unavailability at start-up is 3 × 10⁻³ for the 2000 MW (th) plant. The estimated mean probability that the CACS of the smaller plant will not be available at start-up and not be operational after 300 hr is 8 × 10⁻².     

JR curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents in oxygen-containing high temperature water at 240°C

J_R curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents (0.003 and 0.011 wt.%) were determined at 240°C in oxygen-containing high temperature water as well as in air. The tests were performed by the single-specimen unloading compliance technique at load line displacement rates from 1 × 10⁻⁴ down to 1 × 10⁻⁶ mm s⁻¹ on 20% side-grooved 2T CT specimens in an autoclave testing facility at an oxygen content of 8 ppm and a pressure of 7 MPa under quasi-stagnant flow conditions.In the case of testing in high temperature water, remarkably lower J_R curves than in air at the same load line displacement rate (1 × 10⁻⁴ mm s⁻¹) were obtained. A decrease in the load line displacement rate as well as an increase in the sulphur content of the steel caused a reduction of the J_R curves. At the fastest load line displacement rate a stretch zone could be detected fractographically on the specimens tested in air and in high temperature water and consequently J_i could be determined. When testing in high temperature water, the J_i value of the higher sulphur material type decreases from 45 N mm⁻¹ in air to 3 N mm⁻¹, much more than that of the optimized material type from 51 N mm⁻¹ in air to 20 N mm⁻¹ at 1 × 10⁻⁴ mm s⁻¹.     

Gas-chromatographic examination of the accumulation of3H,85Kr,and133Xe in the protective gas,sodium coolant,and constructional materials of the BR-10

Conclusions The tritium level in the sodium coolant in the first loop on the BR-10 attained 6·10^–4 Ci/liter, but this fell to 9·10^–6 Ci/liter when the oxide trap was working. The second loop accumulates tritium up to 3.5·10^–5 Ci/liter with the cold trap switched out. The tritium level in the protective gas in the first loop is low (10^–10–10^–8 Ci/cm³). Fuel-rod sheaths contain³H and⁸⁵Kr. No tritium was detected in the air vented to the stack in the BR-10.Translated from Atomnaya Énergiya, Vol. 47, No. 4, pp. 266–268, October, 1979.     

Irradiation experiments in the testing nuclear power plant VAK

Within the German Research Programme “Integrity of Components” the first two capsules were irradiated in the Testing Nuclear Power Reactor VAK. The materials are of the 22 NiMoCr 3 7 and 20 MnMoNi 5 5 types and represent the lower bound of the base material regarding upper shelf energy and chemical composition (Cu, S, P), as well as a state of material which does not meet both chemical and toughness requirements (low upper shelf test melt). Tensile, Charpy, drop-weight, and fracture mechanics specimens were irradiated up to a range of 1.5 to 2 × 10¹⁹ cm⁻² (E > 1 MeV). Despite the materials being at or beyond the specification limits, the results show irradiation sensitivity which can be predicted from the US Reg. Guide Trend Curves (1.99) and KWU Trend Curves in a conservative manner. The procedure to determine the adjusted reference temperature RT_NDT (adj.) on the basis of ΔT_41J (following ASTM E 185) could also be confirmed as conservative by comparing the different criteria derived from Charpy and drop weight tests in the unirradiated and irradiated condition.The results of fracture mechanics testing in the linear elastic range show a remarkable temperature margin to the K_Ic-curve of ASME XI.Prestrained compact tension specimens CT 40 mm made of 22 NiMoCr 3 7 material with an upper shelf energy of approx. 100 J were wedge loaded in a range up to 30 MPa m and exposed to the water environment during radiation. Macroscopic examination gave no indications of stress corrosion cracking.From tests of these specimens in the linear elastic range, a fracture toughness K_Ic^*, which was not affected by the prestrain and environment history, was found depending only on the overload applied during the prestraining procedure.     

The fracture and aerosol release of impacted HLW-glasses due to accidental canister drops in a disposal site

For the disposal of HLW-canisters in a salt dome, two different accident scenarios have to be considered, canister drops in the reloading hall or in a borehole with drop heights of 10 m and 600 m, and reference drop velocities of 14 m/s and 80 m/s.The experimental program had two parts:

• - Laboratory scale drop tests with bare and canistered waste glass probes (scale: 1:10) to obtain basic data.

• - Full scale drop tests with inactive HLW-canisters, specified as planned for the German salt repository (H = 1.335 m, Ø = 0.43 m, weight: 550 kg, canister: SST 1.4833, wall: 5 mm).

The size distributions of the broken fines were measured by sieving and those of the filtered airborne particles by particle size analysis. The dominating parameter is the impact velocity (i.e. impact energy), further test parameters show no measurable influence, especially the canister influence on the fracture or aerosol release is negligible.Source terms, evaluated for the respirable fraction (particles with d < 10 μ m are between 2 × 10⁻⁴% for a 10 m drop and 0.1% for a 600 m borehole drop.     

Measurement of large negative reactivity in WWER-440. Sources of uncertainty

The sources of uncertainty in measurement of large negative reactivity in WWER-440 by the inverse point kinetics method, are evaluated quantitatively on the example of measurement of the reactivity worth of the shutdown control rod system of WWER-440 at zero power. Considering the specific features of the control rod system of WWER-440, it is demonstrated that using an appropriate formulation of the inverse solution of the equations of point kinetics, the uncertainty of measured reactivity ρ/β introduced by the assumption of constancy of the parameters of kinetics can be reduced to <3–5% for the case of the discussed rod-drop test at zero power. Based on an analysis of both numerically simulated and actual rod-drop transients, it is shown that the uncertainty of measured reactivity ρ/β can be quite considerable due to the underlying delayed neutron data set—the values of ρ/β obtained using different data sets can differ by 15%. Inexact accounting of the share of ²³⁹Pu in the fission neutron source is estimated to contribute to the total uncertainty of measured ρ/β of 1%, whereas possible spatial effects are expected to result in a relative error in ρ/β of 5%.     

Reactivity feedback coefficients of a material test research reactor fueled with high-density U3Si2 dispersion fuels

The reactivity feedback coefficients of a material test research reactor fueled with high-density U₃Si₂ dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U₃Si₂ LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 °C to 100 °C, at the beginning of life, followed the relationships (in units of Δk/k × 10⁻⁵ K⁻¹) −2.116 − 0.118 ρ_U, 0.713 − 37.309/ρ_U and −12.765 − 34.309/ρ_U, respectively for 4.0 ≤ ρ_U (g/cm³) ≤ 6.0.     

Measurement of keV-neutron capture cross-sections and capture γ-ray spectra of Fe and Fe

The neutron capture cross-sections and the capture γ-ray spectra of ⁵⁶Fe and ⁵⁷Fe have been measured in the neutron energy range from 10 to 90 keV. Pulsed keV-neutrons were produced from the ⁷Li(p,n)⁷Be reaction by bombarding a lithium target with a 1.5-ns bunched proton beam from a 3 MV Pelletron accelerator. The incident neutron spectrum on the capture sample was measured using a time-of-flight method with a ⁶Li-glass detector. The capture γ-rays emitted from an iron or standard gold sample were detected with a large anti-Compton NaI(Tl) spectrometer. The capture yield of the iron or gold sample was obtained by applying a pulse-height weighting technique to the corresponding capture γ-ray pulse-height spectrum. The capture cross-sections of ^56,57Fe were derived with errors less than 5% using the standard capture cross-sections of ¹⁹⁷Au. The capture γ-ray spectra were obtained by unfolding the observed capture γ-ray pulse-height spectra. The present results for the capture cross-sections were compared with the previous measurements and the evaluated values of ENDF/B-VII.0 and JENDL-3.3. The Maxwellian-averaged capture cross-sections of ⁵⁶Fe and ⁵⁷Fe at 30 keV are derived as 12.22 ± 2.06 mb and 44.48 ± 7.56 mb, respectively.     

Individual Injection, Cooling, and Accumulation of Rare Radioactive Ions

A scheme for accumulating radioactive ions, which is geared toward a quasicontinuous low-intensity flux, is discussed. It is based on individual correction of the trajectory and momentum deflection of each ion in the transport channel and individual ion injection into the accumulator ring. The advantages of this scheme are low accumulator acceptance 5–10 ·mm·mrad, high ion accumulation rate – up to 10³ sec^–1 – with beam intensity after the fragment separator 10³–10⁴ sec^–1, and a 10⁴–10⁵-fold decrease of the pulse intensity of the primary beam on the productive target.     

Neutron irradiation effects on optical absorption of KU1 and KS-4V quartz glasses and Infrasil 301

In this work we have compared the effects of neutron (10²¹–10²² n/m² fluences) and gamma irradiation (23.8 MGy dose) on the IR–vis–UV optical absorption spectra of high purity silica with different OH content: KU1 (800 ppm), KS-4V (<0.2 ppm), and commercial silica Infrasil 301 (<8 ppm). The results show that the UV–vis optical degradation of the silica, after neutron irradiation at the highest fluence is similar for the three grades studied, while gamma-induced optical absorption depends on the material grade (KS-4V shows the lowest optical absorption). The effects of both types of radiation on the IR band related with the hydroxyl group (3650 cm⁻¹) depend on the silica grade. For KU1, the shape of this band changes with neutron fluence. For Infrasil 301 gamma and neutron irradiated, this band height increases, possibly due to free molecular or hydrogen atoms. The shift to lower energies observed for the 2260 cm⁻¹ band in the three neutron irradiated silica grades, reflects the changes induced by neutrons in the lattice bonding angle distribution.     

Radiation Resistance of Fused Titanium Ceramic for Actinide Immobilization

The radiation resistance of ceramics proposed for solidification of actinide-containing wastes is studied. Their main concentrators (for uranium) in the samples are phases with fluorite type structure (zirconolite, pyrochlore, murataite), and brannerite. The critical dose for the transition of the crystal lattice of these phases into an amorphous state under irradiation with 1 MeV Kr⁺ at room temperature was 3·10¹⁸ m^–2 for zirconolite, (1.8–2.4)·10¹⁸ m^–2 for pyrochlore, (1.7–1.9)·10¹⁸ m^–2 for murataite, and 1.4·10¹⁸ m^–2 for brannerite. The forms of murataite with 3-, 5-, and 8-fold motif of the fluorite cell all have close radiation resistance. These data make it possible to estimate the transition time of the murataite structure into the amorphous state to be 6–7 hundred yr and 6–7 thousand yr with hypothetical content of ²³⁹Pu in it of 10% and 1%, respectively. The same values of the radiation resistance were obtained previously for titanates with pyrochlore structure, which were proposed by American researchers as matrices for immobilization of excess weapons plutonium.     

Monte Carlo simulation of the NaI(Tl) detector response to measure gold activated foils

This work deals with the implementation of a NaI(Tl) detector for the assessment of the specific saturation activities of pure gold foils after neutron irradiation. These gold foils can be placed in the centre of a set of polyethylene spheres with different diameters. This configuration, known as a passive Bonner sphere system, is suitable to measure neutron spectra normally extended over a wide energy range containing up to 11 decades (from thermal to a few MeV), at places where the neutron field is very intense, high frequency pulsed or where it is mixed with an important high-energy photon component. The MCNPX code was used to evaluate the NaI(Tl) responses to different incident photon energies in terms of pulse-height distributions. An experimental validation of the calculated NaI(Tl) responses, using certified standard sources at a given measurement arrangement, indicates that MCNPX is a valid tool for routine calibration and benchmarking studies of this detector. A good agreement is found between the measured pulse-height distributions of the certified standard sources and those obtained from MCNPX simulations. As a preliminary application, a bare disc Au foil was directly exposed to a Bremsstrahlung photon beam at the isocentre of an 18 MV medical LINAC, in order to test the suitability of this activation material to measure the photo-neutrons generated in such facility. Two differentiated main photo-peaks, arising from ¹⁹⁶Au and ¹⁹⁸Au predominant γ-ray emissions, were observed. The two isotopes are produced mainly by the photonuclear, ¹⁹⁷Au(γ, n)¹⁹⁶Au, and radiative capture, ¹⁹⁷Au(n, γ)¹⁹⁸Au, reactions of, respectively, high-energy photons and thermal neutrons on the gold foil. From the measured ¹⁹⁸Au saturation activity, a rough estimation of (378 ± 68) × 10⁴ cm⁻² Gy⁻¹ was derived for the thermal neutron flux within the LINAC treatment room. This value, although being very approximate, is comparable to those reported by other authors for similar LINAC facilities but with different treatment room configurations, nominal acceleration potentials and Bremsstrahlung photon irradiation areas.     

Kinetics modelling for three-body sequential reactions with applications to Li(n, n′)αT

An exact formalism is derived for the double differential cross section of a two-stage sequential reaction, which allows fully for the energy-angle correlations at each stage and for the lifetime of intermediate states. Application of the method is made to the two-stage process ⁷Li + n → ⁵He + ³H, ⁵He → ⁴He + n′ which contributes to the reaction ⁷Li(n, n′)αT. for the range of incident energies 5–14 MeV and with calculations of the contributions from the other major channel, ⁷Li + n → ⁷*Li + n′, ^7*Li → ⁴He + ³H, comparisons are made of calculated and measured values of dσ/dΩ and dσ/dE for ⁷Li(n, n′)αT. Whilst reasonable overall agreement is obtained, comparisons with the emitted neutron energy spectrum in the UKNDL and ENDF/BIV files indicate that there is considerably more structure present than is represented in either of these files.     

A simple coincidence method of deuterium profiling using the DHe, α)H reaction

A simple standard coincidence arrangement is described for D(³He, α)H depth profiling experiments of deuterium in solids. Both reaction products are detected in coincidence, the high-energy protons being observed in transmission through a foil target. The energy of the α-particles is converted into the corresponding depth scale, whereas the local deuterium concentration is calculated from the yield of the α-particle spectrum. The measurement of α-particles in coincidence with protons allows a reduction of background arising from Rutherford scattering of ³He and other reaction products. For samples of deuterium implanted into Ta₂O₅ and Er, the method allows a reduction of the backscattering yield by a factor of more than 10³. The residual background is due to accidental coincidences. It can be even more reduced using an accurate experimental geometry. The background reduction is largest for samples with a low content of deuterium, allowing measurements of deuterium profiles of 9 × 10¹³ D⁺/cm² implanted into Ta₂O₅ at an energy of 8 keV. This corresponds to a maximum deuterium concentration of about 5 × 10⁻⁴ D per Ta₂O₅. This method is not restricted to thin films, but it allows measurements of deuterium profiles in a thick sample, e.g. of an implantation profile in a 0.4 mm silicon wafer.     

Determination of the composition and instability constants of complex Pu+3oxalate ions

The solubility of Pu₂(C₂O₄)₃ · 9H₂O in aqueous solutions of K₂C₂O₄ of various concentrations (0.01–2.4 moles /liter) has been determined at constant ionic strength of the solution at 20. It was found that Pu⁺³ complexes are formed in these solutions. It was found from the results of Pu₂(C₂O₄)₃ · 9H₂O solubility determinations that in the region of K₂C₂O₄ concentrations studied the following complex ions are formed [Pu(C₂O₄)₂]^–, [Pu (C₂O₄)₃]^–3 and [Pu (C₂O₄)₄]^–5, the total instability constants of which are 4.9 · 10^–10; 4.10 · 10^–10 and 11.9 · 10^–11 respectively. The solubility of Pu₂(C₂O₄)₃ · 9H₂O in aqueous (NH₄)₂C₂O₄ solutions has also been determined in the range of ammonium oxalate concentrations from 0.07 to 0.7 mole/liter at 70 °. It is shown that the composition of the complex ions under these conditions corresponds to [Pu(C₂O₄)₂]^–, [Pu(C₂O₄)₃]^–3 and [Pu(C₂O₄)₄]^–5. The calculated total instability constants of these complex ions are 11.6 · 10^–9; 5.6 · 10^–9 and 2.5 · 10^–9 respectively. The heats of formation of complex Pu⁺³ oxalate ions have been calculated for the reaction Pu⁺³ + nC₂O₄ ^–2 [Pu(C₂O₄)_n]^3–2n ¯Q for the [Pu(C₂O₄)₂]^– ion is 1300 cal., for [Pu(C₂O₄)₃]^–3, 1200 cal., and for [Pu(C₂O₄)₄]^–5, 1300 cal.     

New observations on the crack growth rate of low alloy nuclear grade ferritic steels under constant active load in oxygenated high-temperature water

Within the scope of reactor safety research attempts have been made over several decades to determine corrosion-assisted crack growth rates. National and international investigations have been performed on both an experimental and an analytical basis. A compilation of internationally available experimental data for ferritic steels exhibits a scatter of crack growth rates of up to 5 decades. This was one of the reasons for commencing further experimental investigations focused on the evaluation of corrosion-assisted crack growth rates. These experimental studies were performed under constant, active, external load on 2T-CT specimens of the materials 20 MnMoNi 5 5 with 0.009 and 0.020% S (similar to A508 Cl.3), 22 NiMoCr 3 7 with 0.006% S (similar to A508 Cl.2) and 17 MnMoV 6 4 with 0.017% S. The tests were carried out in deionized oxygenated high-temperature water (240°C; 0.4 and 8.0 ppm O₂). For K_I values up to 60 MPa m^1/2, the experimental results showed no significant dependence between corrosion-assisted crack growth rates and the stress intensity factor, the oxygen content of the medium or the sulphur content of the steel. Here it is important to note, that in this K_I region the high crack growth rates after the onset of cracking due to loading are decreasing and finally come to a standstill after a short period of time as compared with operational times of plants. Consequently, the determination of crack growth velocities as corrosion-assisted crack advance divided by the test duration, so far practised worldwide, results in wrong crack growth rate values in the above-mentioned range of loading up to 60 MPa m^1/2. Based on a test duration of 1000 h, the average crack growth rates are below 10⁻⁸ mm s⁻¹ for K_I ≤ 60 MPa m^1/2. When applied to a single start-up and service period of one year, this would formally lead to an average crack growth rate of 2·10⁻⁹ mm s⁻¹ (equivalent to 0.06 mm per year). At K_I values between 60 and 75 MPa m^1/2 the average corrosion-assisted crack growth rates increase significantly. It can be observed experimentally that the crack propagates during the whole period of the test. Consequently the calculation of crack growth velocities as corrosion-assisted crack advance divided by the test duration as mentioned earlier can be applied as a first estimate. Finally, for K_I values ≥ 75 MPa m^1/2 high crack growth rates up to 10⁻⁴ mm s⁻¹ can be observed. In this region the average crack growth rates are also in quite good agreement with a theoretically based crack growth model.