On mechanisms by which a soft neutron spectrum may induce accelerated embrittlement

Both low displacement rates and softened neutron spectrum favor survival of a higher fraction of point defects per displacement for producing micro-structural changes leading to hardening and embrittlement. Low displacement rate results in low bulk recombination rate. A high thermal to fast neutron flux ratio results in a large fraction of point defects produced in small cascades from (n,γ) and (n,) reactions. Defects from such cascades generally avoid in-cascade recombination, while most of the defects created in large cascades produced by fast neutrons are lost to in-cascade recombination. Thus thermal neutrons produce more available defects per unit displacement dose. It is argued that the spectral effect may dominate the accelerated embrittlement observed in ferritic steels at the High Flux Isotope Reactor (HFIR) pressure vessel location. The rate effect is expected to be a secondary factor at temperatures as low as 50°C, where the HFIR data were obtained. Our analysis suggests generally that components subject to neutron environments with high thermal-to-fast ratios and irradiated at low temperatures may be subject to accelerated radiation effects.     

An embrittlement rate effect deduced from HFIR that may impact LWR vessel support life expectancy

Recent results from the High-Flux Isotope Reactor (HFIR) pressure-vessel-material surveillance program indicate that embrittlement rates of the several ferritic carbon steels are substantially greater than were anticipated. The apparent reason is that the fast neutron flux in the HFIR vessel is much less than that in materials testing reactors (factor of 10⁻⁴); that is, there is a fluence-rate effect. It was soon realized that these results may significantly impact the life expectancy for supports of some light-water reactor (LWR) vessels because the temperature and fast-neutron fluxes associated with supports are about the same as for the HFIR vessel.A preliminary evaluation of LWR supports was conducted by estimating the increases in the nil-ductility transition temperatures (NDTT) of supports in several LWR plants. An analysis based on the HFIR data predicts much higher ΔNDTTs than one based on materials-testing-reactor data. This result, combined with assumptions regarding loading conditions and flaws, provides a basis for concerns about support integrity.     

Consolidation of scientific and technological expertise to assess the reliability of reactor pressure vessel embrittlement prediction in particular for the arctic area plant (COBRA)

The evaluation and prognosis of reactor pressure vessel (RPV) material embrittlement in WWERs and the allowable period of their safe operation are performed on the basis of impact test results of irradiated surveillance specimens. The main problem concerns the irradiation conditions (irradiation temperature, neutron flux and neutron spectrum) of the surveillance specimens that have not been determined yet with the necessary accuracy. These conditions could differ from the actual RPV wall condition. In particular, the key issue is the possible difference between the irradiation temperature of the surveillance specimens and the actual RPV wall temperature. It is recognized that the direct measurement of the irradiation temperature by thermocouples during reactor operation is the only way to obtain reliable information. In addition, the neutron field's parameters in the surveillance specimens location have not been determined yet with the necessary accuracy. The use of state of the art dosimeters can provide high accuracy in the determination of the neutron exposure level.The COBRA project, which started in August 2000 and had a duration of 3 years, was designed to solve the above-mentioned problems. Surveillance capsules were manufactured which contained state of art dosimeters and temperature monitors (melting alloys). In addition, thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during reactor operation. The selected reactor for the experiment was the Unit 3 of Kola NPP situated in the arctic area of Russia. Irradiation of capsules and online temperature measurements were performed during one fuel cycle. On the base of statistical processing of thermocouples readings, the temperature of irradiated surveillance specimens in WWER-440/213 reactor can be accepted as 269.5 ± 4 °C. Uncertainties were evaluated also with experimental work carried out in the WWRSZ research reactor and by finite element modelling of surveillance capsules. The results obtained show that there is not need to perform temperature correction when surveillance data of irradiated specimens are used for embrittlement assessment of WWER-440(213) reactor pressure vessels. Maximum neutron flux evaluated using detectors, which were placed in the Charpy specimen simulators, equals 2.7 × 10¹² cm⁻² s⁻¹ with E > 0.5 MeV. It is established that depending on the orientation of the capsules with respect to the core, the detectors of the standard surveillance capsules can give both overestimated and underestimated neutron flux values, as compared to the actual flux received by the surveillance specimens. The overestimation or underestimation can reach 10%.     

Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.     

Irradiation experiments related to reactor safety in particular pressure vessel steels

The dependence of neutron induced embrittlement of reactor pressure vessel steels on irradiation temperature and neutron exposure was investigated for steels with different copper content. A pronounced increase of the ductile to brittle transition temperature shift with decreasing irradiation temperature was found and quantitatively determined. The influence of the neutron energy spectrum and flux density on the embrittlement was not significant.Rigs for irradiating assemblies of fracture mechanics specimens (CT and WOL) up to 100 mm thickness and also for irradiation experiments under cyclic loading were developed. Irradiation experiments with these rigs are in progress.Creep experiments on canning tubes under different load conditions (uniaxial load and biaxial load under internal and external overpressure) as well as an irradiation device for investigating defective PWR fuel rods are briefly reported.     

Irradiation temperature, flux and spectrum effects

Although great progress has been made in understanding the irradiation behaviour of reactor pressure vessel (RPV) steels, many aspects are still not fully understood. A large amount of data has been generated for understanding the effects of different irradiation conditions on material properties. The data needed for the long term operation of RPVs is almost always created by accelerated irradiations in test reactors, and due to insufficient knowledge on the damage interaction between the material and the high energy neutrons the potential bias of the conclusions on material properties in non-accelerated irradiation conditions can not be excluded. Important parameters for the extrapolation of results from accelerated irradiations to typical power irradiation conditions are the irradiation temperature, the neutron flux and the neutron spectrum. In particular, the effect of neutron flux on embrittlement behaviour is considered a complex phenomenon, and it seems to be dependent on the alloy composition, the neutron fluence range and the irradiation temperature. This paper will present the current knowledge on temperature, flux and spectrum effects, based on a recent literature survey and other relevant publications on the subject. It will explore the implications these effects may have for the safety evaluation of aged RPVs, especially for those exposed to long irradiation periods.     

Observations of a steady state effect limiting radiation damage in reactor vessel steels

Reactor vessel material surveillance capsules which contain specimens of actual material used in the construction of a vessel are contained in nearly all operating reactors. These specimens monitor the changes in properties of the reactor vessel and assure that predicted changes based on trend curves which are used to set operating limits for the plant are conservative. Recently, data has been obtained from the Point Beach Unit No. 1 and Connecticut Yankee reactor vessel surveillance capsules exposed to neutron radiation for much longer periods of time, than those irradiated in test reactors and surveillance capsules which were removed at the first refueling and other early refueling outages. The data from these long time surveillance capsule exposures when compared to data from capsules removed from the same reactors earlier in life indicated that a limiting or steady state condition has resulted rather than a continuous embrittlement as predicted by trend curves. It is believed that the limited embrittlement or steady state condition which occurred from the surveillance capsule tests is due to a combination of relatively low neutron flux compared to that existing in test reactors which were the primary source of data used to establish trend curves and the longer exposure periods in the capsules that led to significant “annealing” during irradiation.     

TEM and PAS study of neutron irradiated VVER-type RPV steels

Conventional transmission electron microscopy and positron lifetime and Doppler broadening positron annihilation spectroscopy techniques have been used to investigate the radiation-induced microstructural changes in surveillance specimens of VVER-type reactor pressure vessel (RPV) steels, and RPV steels irradiated in the research reactor. Defects visible in transmission electron microscopy consist of black dots, dislocation loops and precipitates concentrated along the dislocation substructure. Their size and density depend on the neutron flux and fluence. The parallel set of thermally aged specimens, specimens recovery annealed after irradiation and specimens irradiated in a lower neutron flux was investigated too. No defects discernible in transmission electron microscopy were found after accelerated irradiation in the research reactor. In addition to visible defects, the small-volume vacancy clusters were identified by positron annihilation spectroscopy.     

Positron annihilation study of neutron irradiated model alloys and of a reactor pressure vessel steel

The hardening and embrittlement of reactor pressure vessel steels are of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation-induced damage, and positron annihilation spectroscopy has been shown to be a suitable method for analysing most of these defects. In this paper, this technique (both positron annihilation lifetime spectroscopy and coincidence Doppler broadening) has been used to investigate neutron irradiated model alloys, with increasing chemical complexity and a reactor pressure vessel steel. It is found that the clustering of copper takes place at the very early stages of irradiation using coincidence Doppler broadening, when this element is present in the alloy. On the other hand, considerations based on positron annihilation spectroscopy analyses suggest that the main objects causing hardening are most probably self-interstitial clusters decorated with manganese in Cu-free alloys. In low-Cu reactor pressure vessel steels and in (Fe, Mn, Ni, Cu) alloys, the main effect is still due to Cu-rich precipitates at low doses, but the role of manganese-related features becomes pre-dominant at high doses.     

低铜合金反应堆压力容器钢辐照脆化预测评估模型

反应堆压力容器（RPV）材料辐照脆化预测评估对保证核反应堆安全运行、预防重大灾难性事故的发生具有重要意义。通过深入了解RPV材料辐照损伤机理和分析国外较为成熟的RPV辐照脆化预测模型,揭示了国外有关压力容器辐照脆化预测模型对低铜RPV辐照脆化预测的不足及其原因。在此基础上,发展和建立了适用于低铜RPV辐照脆化趋势的预测模型CIAE-2009。利用辐照性能数据对CIAE-2009模型进行了验证。结果表明,CIAE-2009对低铜含量RPV材料辐照脆化趋势预测具有较高的准确性和可靠性。     

反应堆堆外中子剂量测量技术试验验证

堆外中子剂量测量技术在反应堆压力容器辐照监督中有广泛的应用。本文介绍了在国内某试验堆上进行的堆外中子剂量测量技术验证试验及其结果。比较了堆外中子剂量测量所用探测片活度的理论计算值、实测值及解谱计算结果,同时分析对比了试验中各辐照位置处中子能谱的理论计算值及解谱结果。结果表明,基于测量值的解谱结果与理论计算结果符合良好。堆外中子剂量测量技术可以有效完成中子能谱测量。     

Evaluation of irradiation damage effect by applying electric properties based techniques

The most important effect of the degradation by radiation is the decrease in the ductility of the pressure vessel of the reactor (RPV) ferritic steels. The main way to determine the mechanical behaviour of the RPV steels is tensile and impact tests, from which the ductile to brittle transition temperature (DBTT) and its increase due to neutron irradiation can be calculated. These tests are destructive and regularly applied to surveillance specimens to assess the integrity of RPV. The possibility of applying validated non-destructive ageing monitoring techniques would however facilitate the surveillance of the materials that form the reactor vessel.The JRC-IE has developed two devices, focused on the measurement of the electrical properties to assess non-destructively the embrittlement state of materials. The first technique, called Seebeck and Thomson Effects on Aged Material (STEAM), is based on the measurement of the Seebeck coefficient, characteristic of the material and related to the microstructural changes induced by irradiation embrittlement. With the same aim the second technique, named Resistivity Effects on Aged Material (REAM), measures instead the resistivity of the material.The purpose of this research is to correlate the results of the impact tests, STEAM and REAM measurements with the change in the mechanical properties due to neutron irradiation. These results will make possible the improvement of such techniques based on the measurement of material electrical properties for their application to the irradiation embrittlement assessment.     

Sustained load crack growth in A533B-1 steel under neutron irradiation in a water environment

A study of the combined effects of radiation, water and temperature on sustained load crack growth behavior of reactor pressure vessel steel A533B-1 is reported. To complete this study wedge opening loading (WOL) T-type fracture toughness specimens were prepared from a sample of A533B-1 steel which had a copper content of 0.13%. The crack length change was measured after 939 hr of irradiation in a water environment. An electrical potential method was successfully used to measure the crack length of rusted radioactive specimens. Sustained load crack growth occurred at initial stress intensity factor K_Ii as low as . The value of stress corrosion cracking threshold factor K_Iscc after neutron irradiation in a water environment appears to be in the range of . The results of neutron irradiation in a water environment are to apparently increase the susceptibility of A533B-1 steel to stress corrosion cracking and hydrogen embrittlement.     

Reconstruction of metal embrittlement in a reactor vessel of the atomic icebreaker Lenin

Data from a study of radiation damage to the vessel of a reactor from the retired atomic icebreaker Lenin are used to determine the radiation embrittlement characteristics of the metal. Irradiation by a low neutron flux of 10¹⁰–10¹¹ cm⁻²sec⁻¹ at the beginning of operation is found to correspond to more intense embrittlement of the metal. Then, apparently, as harmful elements are depleted in the matrix of the metal, embrittlement is reduced until there is a change in sign relative to the standard curve obtained for neutron fluxes above 10¹³ cm⁻²sec⁻¹. It is proposed that, because of irradiation by low fluxes of neutrons in the peripheral zones of reactor vessels during some stages of operation, these zones may be damaged to a greater extent than those lying closer to the core. The irradiating neutron flux is a factor that influences the embrittlement of reactor vessel materials, so there is some interest in studying how material is damaged in the vessels of power reactors with low radiation loads which are under development. This is also needed in order to evaluate the efficacy of measures undertaken to reduce the effect of neutron irradiation on reactor vessels. Translated from Atomnaya énergiya, Vol. 105, No. 4, pp. 201–205, October, 2008.     

反应堆压力容器中子辐照损伤的分析和评价

对于材料已经确定的反应堆压力容器,其辐照脆化效应的主要因素是快中子积分通量。本文应用中子输运格林函数法验算了秦山核电站压力容器1/4厚度处最大快中子通量。分析和评价结果表明,该压力容器的设计对中子辐照是安全的。     

Neutron embrittlement surveillance of the garigliano reactor vessel steel

The effects of neutron radiation on the pressure vessel of the Garigliano Nuclear Power Station have been analyzed on the basis of results of a reactor vessel material surveillance program of the plant. A high radiation embrittlement sensitivity was determined for the weld metal and for the A336 forging steel of the ring forging course just above the level of the fuel core. Both showed high copper and phosphorus contents, which accounted for the embrittlement sensitivity. The ring forging opposite the fuel core had a low copper and phosphorus content and revealed relatively low embrittlement. A neutron fluence of 6.3 × 10¹⁹n/cm² > 1 MeV was determined for the peak flux plane for 40yr of operation. However, the 40yr fluence for the ring forging at the top of the core level (3.5 × 10¹⁹n/cm² > 1 MeV) resulted in the highest final transition temperature because of the sensitivity of this steel. The measured Charpy-V shelf energy absorption values were plotted against yield stress values for comparable irradiations on the ratio analysis diagram (RAD). The analysis revealed that the pressure vessel steel properties would continue to degrade toward a condition of possible frangibility at the end of its life. This projection is based on an assumption of uniform embrittlement throughout the vessel wall thickness. Such uniformity does not exist; in fact, a sharp gradient exists in the steel such that ductility rises rapidly in the steel toward the outside wall as well as above and below the fuel core. Hence, because of this strong ductility gradient, the Garigliano reactor vessel should be able to operate safely over its intended design lifetime.     

Embrittlement of low-alloy structural steel by neutron irradiation

The radiation embrittlement of reactor vessel materials is a complex process, which depends on the conditions of irradiation and the microstructure and chemical composition of the steel. It is universally acknowledged that phosphorus, copper, and nickel intensify the radiation embrittlement of vessel material the most. It is believed that Mn, N, C, Mo, Si, As, Sn, V, and other elements also influence radiation embrittlement, but their effect has not been definitely established and is much less than the effect of phosphorus, copper, and nickel. The presence of a synergetic interaction of elements in the irradiation process and the complex interaction of metallurgical factors and the irradiation conditions make it difficult to determine the degree to which impurities and alloying elements influence radiation embrittlement. The effect of the chemical composition of steel, as one of the most important parameters determining the radiation service life of vessel material, on radiation embrittlement is studied, 5 figures, 1 table, 20 references. Translated from Atomnaya énergiya. Vol. 88, No. 4, pp. 271–276, April, 2000.     

Effects of Non-uniform Moderator Temperature on Core Power Distribution in Accelerator Driven Reactors

In a thermal reactor with moderators at different temperatures, a difference arises in the average speeds of thermalized neutrons between the high temperature part and the low temperature part of the moderator, and the non-uniformity of the average speed of thermalized neutrons may effect changes in the spatial dependence of the thermal neutron flux in a core. To investigate the thermal neutron flux in the case the average speed of thermalized neutrons is dependent on the position within a core, time-dependent two-group diffusion equations were applied. The influence of a nonuniform moderator temperature on the core power distribution was investigated about a graphite-moderated subcritical reactor driven by periodic injections of pulsed fast neutrons. The cylindrical reactor model by which a high temperature part of a core that has a spallation target at the center is enclosed by a low temperature part of a core was used. Changes in the core power distribution were calculated. It turned out that the momentary increases of a thermal power density caused by periodic injections of pulsed fast neutrons increase as the difference in the average speeds of thermalized neutrons in the high temperature part and the low temperature part of a core increases.     

Correlation between irradiation-induced changes of microstructural parameters and mechanical properties of RPV steels

Radiation hardening, displayed by the yield stress increase, and irradiation embrittlement, described by the Charpy transition temperature shift, were experimentally determined for a broad variety of irradiation specimens machined from different reactor pressure vessel base and weld materials and irradiated in several VVER-type reactors. Additionally, the same specimens were investigated by small angle neutron scattering. The analysis of the neutron scattering data suggests the presence of nano-scaled irradiation defects. The volume fraction of these defects depends on the neutron fluence and the material. Both irradiation hardening and irradiation embrittlement correlate linearly with the square root of the defect volume fraction. However, a generally valid proportionality is only a rough approximation. In detail, chemical composition and technological pretreatment clearly affect the correlation.     

External corrosion of cladding in PWRs

In-reactor corrosion of Zircaloy is strongly influenced by the fast neutron flux and water chemistry of the primary coolant. Under typical PWR coolant conditions with low oxygen content the fast neutron flux increases the corrosion rate only slightly. On the other hand, under fast neutron irradiation at a high oxygen content in the primary coolant the corrosion is accelerated 5- to 10-fold. In addition localized oxide lenses (nodular corrosion) have been observed. However, hydrogen pick-up rates were generally low. The results are discussed in view of life-limiting aspects; under normal operating conditions of a PWR the external corrosion is not life limiting.