Studsvik is a supplier of nuclear analysis software and specialised services to the international nuclear industry. The company is headquartered in Nyköping, Sweden, and has five divisions: Sweden, United Kingdom, Germany, the United States, and Global Services. There are 1,100 employees in 8 countries. The company's shares are listed on the Nasdaq OMX Stockholm.Studsvik offers advanced technical services to the international nuclear power industry in areas including fuel analysis software, waste treatment, decommissioning, engineering and services, and operating efficiency.The company was founded in Stockholm in 1947, as AB Atomenergi, to develop and operate nuclear power stations in Sweden. Originally it was 57% government-owned; during the 1960s, the government acquired the entirety of the shares, but then in the 1970s, government funding was reduced and the company became entirely industry-owned and changed its name to Studsvik Energiteknik AB. Studsvik is the name of the property east of Nyköping, where the business had moved during the 1960s.In 2005, Studsvik's two research reactors, R2 (50 MW) and R2-0 (1 MW), were decommissioned. There was also a third research reactor in Studsvik called FR-0, which was a zero-power fast reactor with low power output. It was operated 1964–1971 and is now dismantled.In the 1990s and 2000s, the company acquired the German SINA Industrieservice GmbH; the nuclear power division of the Norwegian Scandpower AS; the German Industrieanlagen Fritz & Marx; the British Environmental Remediation Services Ltd.; the ISS international health physics group; the American RACE LLC; the German Dr Fary GmbH & Co KG and the English Alpha Engineering Ltd.Studsvik has a radioactive waste processing facility in Erwin, Tennessee in the US, which opened in 2000. The facility operates in partnership with Waste Control Specialists in Andrews County, Texas. With Washington Group Inc., Studsvik co-owns THOR Treatment Technologies, LLC. Studsvik also operates a facility in Memphis, TN on President's Island, which is in the Mississippi River. In January 2008, Studsvik obtained a licence for a radioactive metal recycling plant near Sellafield in England. Studsvik's metal-recycling facility, designed to assist the Nuclear Decommissioning Authority's National LLW Strategy, opened in September 2009. Decontaminating metallic waste both greatly reduces the quantity of waste requiring to be disposed of at the LLW Repository, and recovers valuable metal.Studsvik is the global leader in fuel vendor-independent software for reactor analysis. Studsvik software has been used throughout the world for light water reactor core design, analysis, and operational support.In April 2011, Studsvik praised defeat of a bill in the US State of Tennessee legislature that would have prohibited dumping low-level nuclear waste in landfills. Both of Studsvik's US processing facilities are in Tennessee. Tennessee is one of the few states that allows dumping of low-level nuclear waste in landfills and the only state that allows dumping of nuclear waste on a single-license, rather than seek government approval for each shipment of waste.
Irradiation assisted stress-corrosion cracking (IASCC) is a form of intergranular stress corrosion cracking that occurs in irradiated austenitic alloys. It requires an irradiated microstructure along with high temperature water and stress. The process is ubiquitous in that it occurs in a wide range of austenitic alloys and water chemistries, but only when the alloy is irradiated. Despite evidence of this degradation mode that dates back to the 1960s, the mechanism by which it occurs has remained elusive. Here, using high resolution electron backscattering detection to analyze local stress-strain states, high resolution transmission electron microscopy to identify grain boundary phases at crack tips, and decoupling the roles of stress and grain boundary oxidation, we are able to unfold the complexities of the phenomenon to reveal the mechanism by which IASCC occurs. The significance of the findings impacts the mechanical integrity of core components of both current and advanced nuclear reactor designs worldwide.
This paper presents simulations of the xM3 power ramp with the fuel performance code ALCYONE performed during an international simulation exercise organized within the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Power to Melt and Maneuverability project. The xM3 test involved a large-grain UO2 fuel from Mitsubishi Heavy Industries cladded with Zirlo and pre-irradiated in a Spanish pressurized water reactor up to an average burnup of 27 GWd/tU-1. It was then submitted to a staircase ramp protocol in the R2 reactor at Studsvik (Sweden) with 10 successive steps of 5 kW center dot m-1 up to a ramp terminal level of 70 kW center dot m-1. The fuel rodlet did not fail, and detailed post irradiation examinations performed during the Studsvik Cladding Integrity Project II evidenced recrystallization of the pellet center around a central hole, interpreted as signs of fuel melting.In this paper, simulations with ALCYONE of the xM3 power ramp, including an advanced model for fuel melting based on thermodynamic equilibrium calculations, are detailed. The model relies on the determination of the liquid fuel fraction evolution with temperature that is used to obtain a continuous description of the material properties during phase change. In consequence of the incorporation of rare earths and actinides in the bulk of the fuel, distinct solidus and liquidus temperatures are estimated. It is shown that the observed central hole and recrystallized central part of the pellet could be the consequence of totally melted fuel (liquidus is reached), partially melted fuel (solidus is reached), or pore migration only.
The applicability of the EVENT/HELIOS package to CANDU lattice cell analysis is studied in this paper. A 45-group cross section library is generated using the lattice depletion transport code HELIOS. This library is then used with the 3-D transport code EVENT to compute the pin fission densities and the multiplication constants for six configurations typical of a CANDU cell. The results are compared to those from MCNP with the same multigroup library. Differences of 70-150 pcm in multiplication constant and 0.08-0.95% in pin fission density are found for these cases. It is expected that refining the EVENT calculations can reduce these differences. This gives confidence in applying EVENT to transient analyses at the fuel pin level in a selected part of a CANDU core such as the limiting bundle during a loss of coolant accident (LOCA).
SIMULATE5 is Studsvik Scandpower's next generation nodal code. The core portion of the thermal hydraulic models of PWR and BWRs are treated as essentially identical, with each assembly having an active channel and a number of parallel water channels. In addition, the BWR assembly may be divided into four radial sub-assemblies. For natural circulation reactors, the BWR thermal hydraulic model is capable of modeling an entire vessel loop: core, chimney, upper plenum, standpipes, steam separators, downcomer, recirculation pumps, and lower plenum. This paper presents results of the validation of the BWR thermal hydraulic model against: (1) pressure drop data measured in the Frigg and NUPEC test facilities; (2) void fraction distribution measured in the Frigg and NUPEC loops; (3) quarter-assembly void fraction measured in the NUPEC experiments and (4) natural and forced circulation flow measurements in the Frigg loop.
In this paper, the radial diffusion coefficient for a CANDU cell is calculated by using the Monte Carlo code MCNP [1]. The calculation uses the conventional definition of the diffusion coefficient which is based on the transport cross section. The latter is estimated from an MCNP simulation of the CANDU cell with a reflective boundary condition. These MCNP-based two-group diffusion coefficients are compared to the corresponding result obtained with the transport and depletion code HELIOS [2]. The calculations are performed for two burnup states: 0 and 4 GWd/TU.
