Engineering damage cross sections for neutron embrittlement of A302-B pressure vessel steel

The continuing reliable operation of reactor pressure vessels requires continuous surveillance over the neutron-induced increase in the steel pressure vessel transition temperature. The increases are plotted versus neutron fluence, but no agreement exists for a meaningful and accurate neutron fluence criterion readily usable by the engineering community. Establishment of such a criterion was the goal of a research study undertaken at the Naval Research Laboratory (NRL); this research also was to support the experimental irradiation effects portion of M.S. thesis work underway by the author at the University of Maryland. Analyses were made of transition temperature increase and neutron fluence data for irradiation of a plate of 6 in. thick (152 mm) A302-B steel. This information plus an accurate neutron spectrum of each irradiation facility was analyzed by an iterative unfolding code to yield two sets of damage cross sections corresponding to embrittlement of A302-B steel at irradiation temperatures of <240 and 550°F (<116 and 288°C).Analyses now completed show that more than 94% of the neutron-induced embrittlement in A302-B steel is caused by neutrons of energies > 0.1 MeV. Furthermore, application of the analysis techniques permits computation of ‘damage-fluence’ values incorporating damage cross sections and fluence > 0.1 MeV that provides the best possible correlation of experimental data regardless of the radiation environment. The technique and cross sections are thought to be sufficiently general that they can be used as a basis for establishing new sets of cross sections for other plates of steel having different chemical composition from that of the original study. This has yet to be proven, however, and will be the subject of future work in this area. In the interests of accuracy of information and uniformity of presentation, it is recommended that these cross sections and the threshold of ‘ > 0.1 MeV’ be adopted for international use in critical engineering assessments of embrittlement of reactor pressure vessel steel.     

Combined effect of thermal and irradiation embrittlement in the reactor pressure vessel CrNiMoV steel

Combined effects of segregation and irradiation embrittlement in reactor pressure vessel CrNiMoV steel were studied. The study deals with an analysis of conditions affecting the 15Ch2NMFA type CrNiMoV steel susceptibility and the development of microsegregation processes in connection with temper brittleness formed on repeated annealing cycles. Microstructural analysis and results of tensile and impact testing for all the treatment conditions are presented.     

Microstructural aspects of neutron embrittlement of reactor pressure vessel steels - A view from positron annihilation spectroscopy

A comprehensive review of positron annihilation studies of Cr---Mo---V reactor pressure vessel (RPV) steels (Soviet type 15Kh2MFA) in unirradiated and neutron irradiated states is presented. The influences of lattice defects, impurity atom distribution, irradiation temperature, flux and fluence of fast neutrons on positron annihilation parameters, especially during isochronal annealing, are discussed in terms of the positron trapping model. In contrast to the literature, where irradiation-enhanced Cu precipitates and solute coated microvoids are thought to be major defect types responsible for strengthening and hence embrittling of RPV steels, we suggest irradiation-induced precipitates, i.e. probably carbides, to play this role. Possibilities to probe this model are suggested.     

A multibarrier safety concept for the reactor pressure vessel of the Stade Nuclear Power Plant

Two specific problems within the safety case of Stade RPV have been analysed: brittle fracture initiation and arrest under strip type emergency core cooling conditions and safety margins against ductile failure from deep cracks as postulated by ASME- and German KTA-rules. For EOL material conditions exclusion of initiation is shown for cracks of more than twice the size which is safely detectable by NDE; for arbitrarily postulated large cracks it is demonstrated that they are arrested well within the allowed depth of of the wall thickness; therefore no critical crack size exists for Stade RPV under strip cooling. Growth in depth of an assumed circumferential flaw in the girth weld embrittled at EOL could occur only at upper shelf temperatures and by loads higher than about twice the service pressure; leak before break was demonstrated in a constraint-modified J − R-curve crack-growth analysis. But neither a transient nor the plant itself would be able to provide the necessary high loads. The LEFM and EPFM proofs are summarized in a multibarrier safety scheme.     

Application of the Seebeck effect for monitoring of neutron embrittlement and low-cycle fatigue in nuclear reactor steel

The monitoring of neutron embrittlement and low-cycle fatigue in nuclear reactor steel is an important topic in lifetime extension of nuclear power plants. Among several material parameters that may change due to material degradation are the thermoelectric properties. Therefore, we investigated the application of the Seebeck effect for determining material degradation of common reactor pressure vessel (RPV) steel. The Seebeck coefficients (SC) of several irradiated Charpy specimens made from Japanese reference steel JRQ were measured. The specimens suffered fluences from 0 up to 4.5 E¹⁹ neutrons/cm², with energies higher than 1 MeV. Measured changes of the SC within this range were about 500 nV/°C, increasing continuously in the range under investigation. Some indications of saturation appeared at fluencies larger than 4.5 E¹⁹ neutrons/cm². We obtained a linear dependency between the SC and the temperature shift ΔT₄₁ of the Charpy energy versus temperature curve, which is widely used to characterize material embrittlement.Similar measurements were performed on fatigue specimens made from the austenitic stainless steel X6CrNiTi18-10 (according to DIN 1.4541) that were fatigued by applying cyclic strain amplitudes of 0.28%. A clear correlation between the change of the SC and the accumulated plastic strain, i.e. number of cycles was obtained.Further investigations were made to quantify the size of the gauge volume in which the thermoelectric power (TEP), also called thermoelectric voltage, is generated. It appeared that the information gathered from a thermoelectric power measurement is very local. This fact can be used to develop a novel TEP-method acting as a thermoelectric scanning microscope (TSM).Finally, we conclude that the change of the SC has a potential for monitoring of material degradation due to neutron irradiation and thermal fatigue, but it has to be taken into account that several influencing parameters could contribute to the TEP in either an additional or extinguishing manner. A disadvantage of the method is the requirement of a clean surface without any oxide layer. This disadvantage can partially be avoided by using the proposed new TSM.     

Small-angle neutron scattering study on the effect of hydrogen in irradiated reactor pressure vessel steels

Hydrogen uptake can enhance the neutron embrittlement of reactor pressure vessel (RPV) steels. This suggests that irradiation defects act as hydrogen traps. The evidence of hydrogen trapping was investigated using the small-angle neutron scattering (SANS) method on four RPV steels. The samples were examined in the unirradiated and irradiated states and both in the as-received condition and after hydrogen charging. Despite the low bulk content of hydrogen achieved after charging with low current densities, an enrichment of hydrogen in small microstructural defects could be identified. Preferential traps were microstructural defects in the size range of ≈ > 10 nm in the unirradiated and irradiated samples. However, the results do not show any evidence for hydrogen trapping in irradiation defects.     

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%.     

Experience of the integrity assessment of the highly embrittled reactor pressure vessel for the life extension

The reactor pressure vessel (RPV) is the key component of pressurized water reactor. It has to comply with various rules and regulatory guides to ensure sufficient safety and operating margins during the plant lifetime. Thus, it is crucial to assure the integrity of RPV for an effective plant lifetime management program. In this paper, the status and the experiences of various integrity issues of the highly embrittled RPV are introduced. A circumferential weld in the beltline region of the Kori Unit 1 RPV was projected to be unable to satisfy the minimum upper-shelf energy requirement and the reference temperature-pressurized thermal shock requirement before the end of 40-year design lifetime. The detailed integrity assessments had been performed to resolve both issues and the results summarized. In addition several actions have been taken as aging management programs to assure the integrity of Kori Unit 1 RPV during the extended operation. Details of the activities such as, redefining initial reference temperature-nil ductility transition temperature, installing ex-vessel dosimetry, and withdrawal and testing of additional surveillance capsule are explained. Finally, the applicability of these and other activities including thermal annealing to mitigate the effects of the irradiation embrittlement are evaluated.     

Toughness investigation on simulated weld HAZs of SQV-2A pressure vessel steel

In order to get detailed information about weld HAZs toughness of SQV-2A steel and determine the optimum welding and heat treatment parameters, the toughness of simulated CGHAZs (coarse grained heat affected zone) and CGHAZs (intercritically reheated CGHAZ) were systematically investigated. The influence of tempering thermal cycles on weld ICCGHAZs toughness was clarified. The effect of post weld heat treatments (PWHT) on weld CGHAZs toughness was also determined. The results showed that high toughness (absorbed energy >200 J) of weld HAZs could be achieved by selecting the optimum welding and PWHT parameters (cooling time Δt_8/5: 6–40 s, PWHT: 893 K, 3.6–7.2 ks). Tempering thermal cycles with peak temperature of above 573 K could remarkably improve the toughness of deteriorated ICCGHAZs and reduce the hardness, when cooling time Δt_8/5(2) of the reheating thermal cycle was 6 s, which implies that welding of SQV-2A without PWHT is possible, provided that low heat input welding is adopted and welding procedure is correctly arranged. Metallography and fractography revealed that M–A constituents in weld HAZs played an important role in controlling weld HAZ toughness.     

On the effects of oxygen cross-sections in the fast neutron fluence analysis for a reactor pressure vessel scraping test

Studies in support of the assessment of aging structural materials in pressurized water reactors are being performed at the Paul Scherrer Institut. To that aim, a state-of-the-art methodology based on applying a CASMO-4/SIMULATE-3/MCNPX calculation scheme has been developed. In the frame of the methodology validation, an investigation is currently reported pertaining to the sensitivity of the calculated results, for a specific reactor pressure vessel scraping test, to the nuclear data used with the Monte Carlo code. Thus, the MCNPX-2.4.0 calculations have been carried out using three different data libraries, based on JEF-2.2, ENDF/B-VI.8 and JENDL-3.3 evaluations, respectively.     

Qualification of the LF-eddy current technique for the inspection of stainless steel cladding and applications on the reactor pressure vessel

As part of the re-inspection of the reactor pressure vessel of the nuclear power plant, the low-frequency-eddy current technique was implemented during the 1995 outage. Since then, this inspection technique and the testing equipment have seen steady further development. Therefore, optimization of the entire testing system, including qualification based on the 1995 results, was conducted. The eddy current testing system was designed as a ten-channel test system with sensors having separate transmitter and receiver coils. The first qualification of the testing technique and sensors was performed using a single-channel system; a second qualification was then carried out using the new testing electronics. The sensor design allows for a simultaneous detection of surface and subsurface flaws. This assumes that testing is performed simultaneously using four frequencies. Data analysis and evaluation are performed using a digital multi-frequency regression analysis technique The detection limits determined using this technique led to the definition of the following recording limits for testing in which the required signal-to-noise ratio of 6 dB was reliably observed.

• Detection of surface connected longitudinal and transverse flaws:

• notch, 3 mm deep and 10 mm long, for weave bead cladding;

• notch, 2 mm deep and 20 mm long, for strip weld cladding.

• Detection of embedded planar longitudinal and transverse flaws:

• ligament of 7 mm for 8 mm clad thickness and 3 mm;

• ligament for 4 mm clad thickness, notch starting at the carbon steel base material with a length of 20 mm.

• Detection of embedded volumetric longitudinal and transverse flaws:

• 3 mm diameter side-drilled hole (SDH) for 8 mm clad thickness; ligament, 4 mm. For 4 mm clad thickness: diameter, 2 mm SDH; ligament, 2 mm. All SDHs are 55 mm deep.

Computational methods for fracture analysis of heavy-section steel technology (HSST) pressure vessel experiments

This paper summarizes the capabilities and applications of the general-purpose and special-purpose computer programs that have been developed at the Oak Ridge National Laboratory (ORNL) for use in fracture mechanics analyses of HSST pressure vessel experiments. Emphasis is placed on the OCA/USA code, which is designed for analysis of pressurized-thermal-shock (PTS) conditions, and on the ORMGEN/ADINA/ORVIRT system which is used for more general analysis. Fundamental features of these programs are discussed, along with applications to pressure vessel experiments.     

A thermal-hydraulic code (THYD) for the miniature neutron source reactor thermal-hydraulic transients

The transient thermal-hydraulic problem of MNSR is represented by ten differential equations solved numerically using Runge–Kutta method.Computational results are then compared with experimental measurements. Fuel grids and cooling coil models are incorporated in the model too. Radiating energy from the clad is taken into account in the energy balance in the reactor. The pool is divided into three sections in the model. The effect of the cooling coil of the pool upper section on reactor thermal-hydraulic parameters is discussed. The only input parameter of the reactor is the power temporal distribution. Good agreement between calculated and measured data was obtained.     

Implications of the differences in neutron spectra predicted for reactor pressure vessel walls by transport and the P1MG codes

Neutron-energy spectra were calculated for the interface between the vessel wall and cladding of the Army SM-1A Reactor pressure vessel using the transport theory code Program S and the diffusion code P1MG. Different sets of basic nuclear data and microscopic cross sections were used for the two calculations. Spectra were normalized to the same amount of activation in an iron, neutron flux detector. The transport code predicted a higher flux of neutrons in the energy groups between 6 and 10 MeV resulting in a lower overall intensity for the transport theory spectrum versus the P1MG spectrum. This was found to be consistent with the predictions of two transport codes versus the P1MG code for the PM-2A reactor vessel wall and for a simulated reactor vessel wall experiment. Such divergence of results for a given reactor using two different code analysis techniques raises important questions as to their usually unqualified acceptance and use for projecting the lifetime fluence for a reactor pressure vessel. Strong support is thus generated for establishment of one “standard” set of basic nuclear data from which all reactor physics analysts can draw to generate specific cross sections for reactor physics calculations, and for the writing of a new reactor physics spectrum code specifically for deep penetration analysis of reactor pressure vessel walls.     

Influence of long-term thermal aging on the microstructural evolution of nuclear reactor pressure vessel materials: an atom probe study

Atom probe field ion microscopy (APFIM) investigations of the microstructure of unaged (as-fabricated) and long-term thermally-aged (˜100 000 h at 280°C) surveillance materials from commercial reactor pressure vessel steels were performed. This combination of materials and conditions permitted the investigation of potential thermal aging effects. This microstructural study focused on the quantification of the compositions of the matrix and carbides. The APFIM results indicate that there was no significant microstructural evolution after a long-term thermal exposure in weld, plate and forging materials. The matrix depletion of copper that was observed in weld materials was consistent with the copper concentration in the matrix after the stress relief heat treatment. The composition of cementite carbides aged for 100 000 h were compared to the Thermocalc™ prediction. The APFIM comparisons of materials under these conditions are consistent with the measured change in mechanical properties such as the Charpy transition temperature.     

The viablility of the KIa procedure for predicting the arrest of a crack in a nuclear reactor steel pressure vessel

The current ASME Code procedure for predicting crack arrest in a nuclear reactor steel pressure vessel is based on a static linear elastic fracture mechanics analysis: a crack is presumed to arrest when the crack tip stress intensity factor K_I^ST falls below K_Ia, which is assumed to be a material property and is referred to as the arrest toughness. The viability of this procedure has been questioned since the theoretical justification, in the strictest sense, for this very simple K_Ia approach is based on the behaviour of a semi-infinite crack propagating in an unbounded solid due to the application of time-independent loads. Against this background, the present paper examines the effects of initial crack size and crack jump length on the viability of the K_Ia procedure. A theoretical analysis shows that the procedure should give accurate predictions of the crack length at arrest certainly if the crack jump length is less than twice the initial crack size.     

The effect of high pressure test loop (HPTL) on the Reactivity of ET-RR-2 reactor

The second Egyptian Research Reactor ET-RR-2 is a multipurpose research reactor. It is an open pool type, with nominal power of 22 MW water-cooled. The reactor pool is designed to accommodate two fuel test loops mainly 500 and 20 KW loop in the reactor reflector to enable performing experiments on the behavior of fuel rods for nuclear reactors under their operating conditions. For that, inserted high-pressure test loop (HPTL) loaded with suggested CANDU type fuel element in the reactor core is important to achieve the above reason. From the neutronic safety point of view, it is necessary to study the mutual neutronic and reactivity effect between the reactor core and HPTL. This paper aimed at the study of the temperature coefficients of fuel and moderator of the CANDU type fuel element at different ²³⁵U enrichments, and the effect of HPTL on the reactor core reactivity. The effect of flooding the contact second shut down system (SSS) chamber with water and gadolinium nitrate on the reactor core reactivity in the presence of HPTL. All analysis was performed with the WIMSD4 and DIXY2 codes. This study shows that, an unacceptable change of reactor core reactivity was found due to the presence of the HPTL and the maximum inserted reactivity does not exceed 527 pcm at high possible ²³⁵U enrichment (10%).