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.     

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.     

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.     

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.     

Design and development of the AHWR—the Indian thorium fuelled innovative nuclear reactor

India has chalked out a nuclear power program based on its domestic resource position of uranium and thorium. The first stage started with setting up the Pressurized Heavy Water Reactors (PHWR) based on natural uranium and pressure tube technology. In the second phase, the fissile material base will be multiplied in Fast Breeder Reactors using the plutonium obtained from the PHWRs. Considering the large thorium reserves in India, the future nuclear power program will be based on thorium–²³³U fuel cycle. However, there is a need for the timely development of thorium-based technologies for the entire fuel cycle. The Advanced Heavy Water Reactor (AHWR) has been designed to fulfill this need. The AHWR is a 300 MW_e, vertical, pressure tube type, heavy water moderated, boiling light water cooled natural circulation reactor. The fuel consists of (Th–Pu)O₂ and (Th–²³³U)O₂ pins. The fuel cluster is designed to generate maximum energy out of ²³³U, which is bred in situ from thorium and has a slightly negative void coefficient of reactivity. For the AHWR, the well-proven pressure tube technology has been adopted and many passive safety features, consistent with the international trend, have been incorporated. A distinguishing feature which makes this reactor unique, from other conventional nuclear power reactors is the fact that it is designed to remove core heat by natural circulation, under normal operating conditions, eliminating the need of pumps. In addition to this passive feature, several innovative passive safety systems have been incorporated in the design, for decay heat removal under shut down condition and mitigation of postulated accident conditions. The design of the reactor has progressively undergone modifications and improvements based on the feedbacks from the analytical and the experimental R&D. This paper gives the details of the current design of the AHWR.     

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%.     

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.     

Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues

A large number of new fast reactors may be needed earlier than foreseen in the Generation IV plans. According to the median forecast of the Special Report on Emission Scenarios commissioned by the Intergovernmental Panel on Climate Control nuclear power will increase by a factor of four by 2050. The drivers for this expected boost are the increasing energy demand in developing countries, energy security, but also climate concerns. However, staying with a once-through cycle will lead to both a substantially increased amount of high-level nuclear waste and an upward pressure on the price of uranium and even concerns about its availability in the coming decades. Therefore, it appears wise to accelerate the development of fast reactors and efficient re-processing technologies.In this paper, two fast reactor systems are discussed—the sodium-cooled fast reactor, which has already been built and can be further improved, and the lead-cooled fast reactor that could be developed relatively soon. An accelerated development of the latter is possible due to the sizeable experience on lead/bismuth eutectic coolant in Russian Alpha-class submarine reactors and the research efforts on accelerator-driven systems in the EU and other countries.First, comparative calculations on critical masses, fissile enrichments and burn-up swings of mid-sized SFRs and LFRs (600 MW_e) are presented. Monte Carlo transport and burn-up codes were used in the analyses. Moreover, Doppler and coolant temperature and axial fuel expansion reactivity coefficients were also evaluated with MCNP and subsequently used in the European Accident Code-2 to calculate reactivity transients and unprotected Loss-of-Flow (ULOF) and Loss-of-Heat Sink (ULOHS) accidents. Further, ULOFs as well as decay heat removal (protected Total Loss-of-Power, TLOP) were calculated with the STAR-CD CFD code for both systems.We show that LFRs and SFRs can be used both as burners and as self-breeders, homogeneously incinerating minor actinides. The tight pin lattice SFRs (P/D = 1.2) appears to have a better neutron economy than wide channel LFRs (P/D = 1.6), resulting in larger BOL actinide inventories and lower burn-up swings for LFRs. The reactivity burn-up swing of an LFR self-breeder employing BeO moderator pins could be limited to 1.3$ in 1 year. For a 600 MW_e LFR burner, LWR-to-burner support ratio was about two for (U, TRU)O₂-fuelled system, while it increased to approximately 2.8 when (Th, TRU)O₂ fuel was employed. The corresponding figures for an SFR were somewhat lower. The calculations revealed that LFRs have an advantage over SFRs in coping with the investigated severe accident initiators (ULOF, ULOHS, TLOP). The reason is better natural circulation behavior of LFR systems and the much higher boiling temperature of lead. A ULOF accident in an LFR only leads to a 220 K coolant outlet temperature increase whereas for an SFR the coolant may boil. Regarding the economics, the LFR seems to have an advantage since it does not require an intermediate coolant circuit. However, it was also proposed to avoid an intermediate coolant circuit in an SFR by using a supercritical CO₂ Brayton cycle. But in an LFR, the reduced concern about air and water ingress may decrease its cost further.     

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.     

ZB2-tests — Description of the intermediate size vessel and AE results

To demonstrate and to extend the performance of acoustic emission testing as a method of detecting and classifying flaws, six institutes conducted acoustic emission measurements in the course of various loading tests on a medium-sized, thickwalled vessel (model of a reactor pressure vessel) containing natural flaws. This paper will present an outline of the working program of the MPA within the framework of this project. The presentation is concentrated on the vessel manufacturing, the preparation of the flaw patch with 14 natural flaws, the performance of the loading tests to simulate pressure test and operating conditions existing in the primary systems of pressurized water reactors, and especially on the conduction of acoustic emission measurements as a method of permanent vessel monitoring during the tests.In the pressure tests conducted with slowly rising pressures, only one flaw was detected unequivocally by the acoustic emission monitoring, although several flaws had grown in the test phases between the pressure tests. In fast pressure tests, flaws in principle were detected slightly better. Cyclic loading over prolonged periods of time produced clear signals of larger flaws, which calculations and subsequent destructive investigations showed to have grown. The small flaws which, most probably, had not changed, could not be detected.     

Core performance of new concept passive-safety reactor “kamado” - safety, burn-up and uranium resource problem -

New concept of a passive-safety reactor “KAMADO” has a negligible possibility of core melting and flexibility of total reactor power. The reactor core of KAMADO consists of fuel elements of graphite blocks, which have UO₂ fuel rods and cooling water holes. These fuel elements are located in a reactor water pool of atmospheric pressure (1 atm) and low temperature (< 60°C). In case of LOCA, decay heat from fuel rods is removed by conduction heat transfer to the reactor water pool. Since the cooling water does not contact a fuel rod directly, core design has much flexibility without considering dry-out limitation and Minimum Critical Power Ratio (MCPR). Additionally an effective use of spent fuel is expected.     

URANUS — A computer programme for the thermal and mechanical analysis of the fuel rods in a nuclear reactor

The URANUS code, a digital computer programme for the thermal and mechanical analysis of integral fuel rods, is described. With this code the fuel rods found in the majority of power reactors can be analyzed. URANUS is built around a quasi two-dimensional analysis of fuel and cladding. The mechanical analysis can accommodate seven components of strain: elastic, time-independent plastic, creep and thermal strains, as well as strains due to swelling, cracking and densification. The heat generation and temperature distribution, cladding/fuel gap closure, pellet cracking and crack healing, fission-gas release, corrosion, O/M-distribution and plutonium redistribution are modelled. Geometric non-linearities (large displacements) are included; steady state or transient loading (pressure, temperature) is possible. In this paper special attention is paid to a theory for determining crack structures. The present status of the URANUS computer programme and a critical comparison with other fuel rod codes as well as sample analyses are given.