Calculation results from radiation embrittlement of VVER pressure vessel at Kozloduy NPP

A calculational procedure for the evaluation of the transition temperature shift on the basis of neutron fluence has been applied for assessing the reactor pressure vessel (RPV) embrittlement and life time for a VVER-440/230. The calculated results are lower than the passport values, because the real fuel regimes, the low-leakage schemes and loadings with dummy cassettes have been taken into consideration in neutron fluence calculation. The temperature of the outer wall of the RPV has been measured. No significant deviation between the measurement and the data given in the reactor passport has been observed. This shows the correct application of the calculational procedure.     

Ion traps — recent applications and developments

Paul and Penning traps are now widely applied in chemistry and physics laboratories. They are used as storage devices, as tools for precision spectroscopy and metrology, and as mass spectrometers. Direct mass measurements of short-lived Rb, Sr, Cs, Ba, Fr and Ra isotopes were performed at the on-line mass separator ISOLDE at CERN, Geneva, by means of a tandem Penning trap system. The ions from ISOLDE are captured and cooled in a first trap and trasnferred to a second trap. Here the mass of the trapped ions is determined by measuring their cyclotron frequency. Resolving powers exceeding m/Δm (FWHM) = 10⁶ could be achieved. Mass values of about 60 isotopes have been determined with accuracies of typically δm/m = 10⁻⁷. For the first time in the history of mass spectrometry the isomeric and ground states of a nucleus have been resolved.     

Influence of the depth position on the neutron embrittlement of the VVER reactor pressure vessel steel 15Cr2MoV(A) consequences for the assessment of reactor safety

The dependence of the mechanical properties on the depth position in the unirradiated state and after irradiation up to neutron fluences of approximately 5 × 10¹⁸ and 70 × 10¹⁸ cm⁻² (E > 0.5 MeV) is tested on a forging made out of VVER 440 reactor pressure vessel (RPV) steel 15CrMoV. The near-surface position shows a higher strength and a lower transition temperature than the positions greater than 1/4 wall thickness. Irradiation does not change these differences in a significant manner. The testing of specimens from the 1/4 depth position within the surveillance programme, as normally laid down in the legal rules relating to nuclear power plants, results in a conservative safety assessment against brittle failure up to the EOL fluence. On taking into account fluence attenuation, the flux effect, etc., the toughness gradually increases from the inside to the outside of the wall after longer RPV operating times.     

Neutron fluctuations induced by absorber vibrations: the QMC reactor experiment—I

The dependence of rod vibration induced, neutron density fluctuations on the static neutron gradients was investigated experimentally at the QMC research reactor. By appropriate fuel loading arrangements the rod was made to vibrate (a) within a flat flux and (b) a flux gradient. The neutron density was fluctuating with the frequency of the vibrating absorber only when the static flux gradient in the vicinity of the absorber was not zero. The double frequency effect was not observed.     

Pneumatic pressure tests of steel containment models — Recent developments

This paper describes the results of recent pneumatic pressure tests of steel containment models. These tests are part of the Containment Integrity Program whose objective is the qualification of methods for predicting containment response during severe accidents and extreme environments. Sandia National Laboratories is conducting this combined experimental and analytical program for the U.S. Nuclear Regulatory Commission (NRC). The long-range plans for the program include the following three containment loading conditions: static internal pressurization, dynamic internal pressurization, and seismic loadings. Steel, reinforced concrete, and prestressed concrete containment types are being considered.In the present experimental effort, models of steel containment structures are being subjected to static internal pressurization. The first set of models are about the size of hybrid-steel containments. Tests of these models are nearly finished. Testing of a large steel model, about of full size, will complete the static pressure experiments with steel models. Analysis of the models is paralleling the experimental effort.The Containment Integrity Program is being coordinated with other NRC programs on potential leakage of penetrations in containments. The results from all of the programs should provide a basis for predicting the structural and leakage behavior of containments during temperature and internal pressure loadings.     

Analysis of maximum pressure in VVER1000/V446 reactor containment for LOCA and MSLB

The highest thermal-hydraulic pressure in the containment occurs when reactor coolant in the first loop and steam in the secondary loop discharge simultaneously,and when the maximum amount of energy from reactor unit enters to containment volume.In this paper,we investigate temperature and pressure variations in the VVER1000 containment compartments owing to concurrent break in the pipelines of the primary and secondary loops.A two-phase,multicellular model is applied in the presence of non-condensable gases.Convection and conduction through the main heat structures inside the containment are also considered.The predicted results agree well with available data.Maximum values of pressure and temperature in the containment are then calculated and compared to the design values.If LOCA and MSLB occur simultaneously,the maximum pressure would exceed the design value and integrity of the containment would be threatened.     

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.     

The development of radiation embrittlement models for US power reactor pressure vessel steels

A new approach of utilizing information fusion technique is developed to predict the radiation embrittlement of reactor pressure vessel steels. The Charpy transition temperature shift data contained in the Power Reactor Embrittlement Database is used in this study. Six parameters-Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature - are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.     

Current and emerging pressure boundary issues — a US regulatory perspective

Concerns about pressure boundary integrity deal primarily with older plants, and establishing a basis for their continued safe operation. Pressure vessel problems stem from exposure to fast neutrons which changes the Nil-Ductility-Temperature (NDT) and the elevated temperature fracture energy of some vessels. The predicted shift in NDT has increased over the last decade as more has been learned about the effect of impurities (copper) and the synergism between nickel and copper. In PWRs this has lead to concern about excursion in which the a vessel remains at high pressure as the coolant temperature drops rapidly, that is the so-called Pressurized Thermal Shock (PTS) accident. In BWRs one cannot have PTS events, but the more rapid than expected rise in NDT due to irradiation is impacting operations.In another set of PWRs the upper shelf energy of the welds was initially low due to the use of a slag which led to many small inclusions in the weld. Radiation has lowered the Charpy fracture energy of these welds to below the 50 ft lb level at which there is concern that the vessel may undergo low energy ductile failure even if cleavage does not occur.Problems in pressure boundary piping has stemmed primarily from corrosion, that is, IGSCC in BWR recirculation piping, and steam generator tube failures in PWRs. These have made a large contribution to downtime and occupational exposure, but are not seen as significant contributors to risk. There has been some concern about the aging (loss of toughness) of cast stainless components with significant ferrite content, especially because inspection by UT is difficult.     

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.     

The experimental breeder reactor II inherent shutdown and heat removal tests — results and analysis

A test program is being conducted to demonstrate that a power-producing liquid-metal reactor (LMR) can (1) passively remove shutdown heat by natural convection, (2) passively reduce power in response to a loss of reactor flow, and (3) passively reduce power in response to a loss of the balance-of-plant heat sink. Measurements and pretest predictions confirm that natural convection is a reliable, predictable method of shutdown heat removal and suggest that safety-related pumps or pony motors are not necessary for safe shutdown heat removal in an LMR. Measurements from tests in which reactor flow and heat rejection to the balance of plant were perturbed show that reactivity feedbacks can passively control power and temperature. Data from these tests form a basis for additional tests including a complete loss of flow without scram and a complete loss of heat sink without scram.     

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.     

Probability-based load combinations for design of Category I structures — Overview of research program and recent results

This paper discusses the probability-based load combinations for the program dealing with the design of Category I structures, currently being worked on at Brookhaven National Laboratory (BNL) for the Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission (NRC). The objective of this program is to develop a probabilistic approach for the safety evaluations of reactor containments and other seismic Category I structures subjected to multiple static and dynamic loadings. Furthermore, on the basis of the developed probabilistic approach, a load combination methodology for the design of seismic Category I structures will also be established.The major tasks of this program are: (1) establish probabilistic representations for various loads and structural resistance, (2) select appropriate structural analysis methods and identify limit states of structures, (3) develop a reliability analysis method applicable to nuclear structures, (4) apply the developed methodology to existing Category I structures in order to evaluate the reliability levels implied in the current design criteria, and (5) recommend load combination design criteria for Category I structures. When the program is completed, it will be possible to (1) provide a method that can evaluate the safety margins of existing containment and other Category I structures and (2) recommend probability-based load combinations and load factors for the design of Category I structures.At the present time, a reliability analysis method for seismic Category I concrete structures has been completed. By utilizing this method, it is possible to evaluate the safety of structures under various static and dynamic loads. In this paper, results of a reliability analysis of a realistic reinforced concrete containment structure under dead load, accidental pressure, and earthquake ground acceleration are presented to demonstrate the feasibility of the methodology.     

Standards and specifications for selecting optimum nuclear reactor pressure boundary metals — A review of US approaches and needs

Standards and specifications which optimize the quality of structural components including pressure boundary materials fill an important place in assuring reactor structural reliability. The author reviews the needs for and demonstrates the way in which research may contribute to improve codes, standards, and specifications for nuclear reactor pressure boundary metals, considering radiation and other service environmental effects. Examples are cited from studies related to both thermal and fast reactor materials. The organisations and their current emphasis for standards development in the US also are reviewed briefly with specific reference boundary materials for commercial nuclear power plants.     

Design and analysis of concrete reactor vessels: New developments, problems and trends

This lecture reviews new developments in analysis and design of prestressed concrete reactor vessels (PCRV). After a brief assessment of the current status and experience, the advantages, disadvantages, and especially the safety features of PCRV, are discussed. Attention is then focused on the design of penetrations and openings, and on the design for high-temperature resistance — areas in which further developments are needed. Various possible designs for high-temperature exposure of concrete in a hypothetical accident are analyzed. Considered are not only PCRVs for gas-cooled reactors (GCR), but also guard vessels for liquid metal fast breeder reactors (LMFBR), for which designs mitigating the adverse effects of molten sodium, molten steel, and core melt are surveyed. Realistic analysis of these problems requires further development in the knowledge of material behavior and its mathematical modeling. Recent advances in the modeling of high-temperature response of concrete, including pore water transfer, pore pressure, creep and shrinkage are outlined. This is followed by a discussion of new developments in the analysis of cracking of concrete, where the need of switching from stress criteria to energy criteria for fracture is emphasized. The lecture concludes with a brief discussion of long-time behavior, the effect of aging, and probabilistic analysis of creep.     

The COVA program—Validation of the fast reactor containment code seurbnuk

Extensive studies are underway in the UK and elsewhere to investigate the consequences of a hypothetical core disruptive accident in a fast reactor. Within these studies there is a continuing need to determine the response of the primary containment system to loads generated by such a major release of energy. A number of fast reactor containment codes are currently being developed for this purpose.The COVA (COva VAlidation) program of experiments is being undertaken to provide high quality data against which the codes can be validated. The tests progress from simple bare vessel assemblies to tests incorporating the main axisymmetric features of loop and pool reactor designs.A companion paper in this journal discusses recent developments in the containment code SEURBNUK which is being jointly developed by the UKAEA and JRC Ispra. This paper describes the status of the validation of SEURBNUK against data from the COVA experiments. Considerable progress has been made in the application of the code to the more complex configurations of the later tests which are more representative of reactor geometries and it is concluded that SEURBNUK has reached a stage where it provides a useful calculational capability for containment analysis. Nevertheless, further work is needed to improve the predictions of vessel deformations and roof loadings in some geometries.     

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%.     

Neutron and capture γ-ray accumulation factors in engineering calculations of reactor shielding

Translated from Atomnaya Énergiya, Vol. 66, No. 5, pp. 321–324, May, 1989.     

Optimizing cell importances using an extension of the DSA — theory, implementation, preliminary results

Starting from the theoretical results of the extension of Dubi's Direct Statistical Approach (DSA) surface parameter model to a cell importance model, a computer code based on MCNP has been written that directly estimates the second moment and time functions. Two versions have been developed: one that estimates the exact functions and the other that estimates the point-surface approximation functions. Preliminary results obtained using the two versions are presented.     

Neutron noise and random trees links between past and present Special Lecture SMORTH VIII — Göteborg, 27–31 May 2002

After reviewing personal reminiscences about the history of reactor noise research, the generalized notion of neutron importance is discussed and advantages of the backward generating function equation are shown by calculating the space-time fluctuations of the neutron density in a simple virtual (one-dimensional) reactor. Similarities between chain reactions and randomly evolving trees are used to study the special properties of branching processes. It is assumed that at t = 0 the tree consists of a single living node called root which, after a certain time τ ≥ 0, may produce η ≥ 0 new living nodes and then becomes dead. τ and η are random variables with known distribution functions. Each new living node evolves further independently of the others as does the root. The time dependence of the expectation value of the living nodes number is determined by the average number q₁ of the new nodes produced by one dying node. Depending on whether q₁ < 1 or q₁ = 1 or q₁ > 1 the randomly evolving tree is called subcritical, critical, and supercritical, respectively. The probability distributions of the tree lifetime and the tree size are determined in two exactly solvable models, and it is proven that a supercritical tree may be finite even at t = ∞ with non-zero probability.