What is a Supercritical Water Reactor?

[chivasorn]

Hi there,

I want to know about SCWR(supercritical water reactor) that is a Gen IV reactor.
Has anybody introduce the technology of SCWR & SSCWR.

thanks

 

[Astronuc]

The Gen IV concepts are still in the design phase. None has been built.

http://www.gen-4.org/Technology/systems/scwr.htm

The Supercritical-Water-Cooled Reactor (SCWR) system is a high-temperature, high-pressure water-cooled reactor that operates above the thermodynamic critical point of water (374 degrees Celsius, 22.1 MPa, or 705 degrees Fahrenheit, 3208 psia).

The supercritical water coolant enables a thermal efficiency about one-third higher than current light-water reactors, as well as simplification in the balance of plant. The balance of plant is considerably simplified because the coolant does not change phase in the reactor and is directly coupled to the energy conversion equipment. The reference system is 1,700 MWe with an operating pressure of 25 MPa, and a reactor outlet temperature of 510 degrees Celsius, possibly ranging up to 550 degrees Celsius. The fuel is uranium oxide. Passive safety features are incorporated similar to those of simplified boiling water reactors.

Putting this in perspective, PWRs operate with pressure ~15.5-15.7 MPa and Tout ~310-330°C (590-626°F). BWR cores operate under saturated conditions at ~7.2 MPa and ~286°C (547°F). The exit temperature of a BWR is approximately (or slightly less than) the inlet temperature of a PWR.

 

[chivasorn]

thanks Astronuc,
can you explain about neutronic calculations and the used codes for SCWR?
Whether the existing codes can be used for this type of reactor is?

 

[Astronuc]

I would imagine that one could MCNP.

One needs something beyond two group diffusion theory, and probably a multi-group transport code. I'll see what I can find.

 

[Hologram0110]

can you explain about neutronic calculations and the used codes for SCWR?
Whether the existing codes can be used for this type of reactor is?

Someone please correct me if I'm wrong, but SCWR would behave similar to most other reactors from a neutronics standpoint. The only difference that I see is due to the increased temperatures thermal expansion of core components and Doppler shifting effects may result in different behavior at temperature compared to cool. Presumably, this would not be an insurmountable modeling challenge.

 

[Astronuc]

Apparently there is a fast spectrum version of the SCWR, but perhaps it depends on the fuel-to-moderator ratio. There are square and hexagonal lattice designs.

Neutronically, it seems similar to a BWR, but the moderator temperature is much higher. I expect the resonance broadening to be more of an effect than for a conventional (lower temperature) LWR.

Different groups have used MCNP (coupled with ORIGEN2 for isotopic/depletion calcs) versions or HELIOS. HELIOS has the capability of hexagonal lattices. There is an effort to couple MCNP with a thermal-hydraulics code because of the significant temperature rise and property changes across the core.

Some examples:

http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6V3X-4RKDHR6-2&_user=10&_coverDate=01%2F31%2F2008&_rdoc=1&_fmt=high&_orig=search&_sort=d&_docanchor=&view=c&_searchStrId=1237336133&_rerunOrigin=google&_acct=C000050221&_version=1&_urlVersion=0&_userid=10&md5=30e99a954792b5baed24bda5def707fb

Abstract

The HPLWR (High-Performance Light Water Reactor) is the European version of the SCWR (Supercritical-pressure Water Cooled Reactor) concept, which is one of the Generation IV reactors. In this reactor the primary water enters the core of the HPLWR under supercritical-pressure condition (25 MPa) at a temperature of 280 °C and leaves it at a temperature of up to 510 °C. Due to the significant changes in the physical properties of water at supercritical-pressure the system is susceptible to local temperature, density and power oscillations. This inclination is increased by the pseudocritical transformation of the supercritical-pressure water used as primary coolant.

At the Budapest University of Technology a coupled neutronics–thermohydraulics program system has been developed which is capable of determining the steady-state equilibrium parameters and calculating the related power, temperature, etc. distributions for the above-mentioned reactor. The program system can be used to optimize the axial enrichment profile of the fuel rods (e.g. in order to obtain a uniform power distribution).

Since the power distribution will change with burn-up, which causes a change in the temperature and density distributions, a coupled neutronics – thermohydraulics – burn-up calculation is required. Therefore, the program system has been extended with the ORIGEN-S burn-up calculator.

In the paper the developed program system and its features are presented. Parametric studies for different operating conditions were carried out and the obtained results are discussed.

They use an MCNP module


http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6TXN-4BVPW9D-2&_user=10&_origUdi=B6V3X-4RKDHR6-2&_fmt=high&_coverDate=05%2F01%2F2004&_rdoc=1&_orig=article&_acct=C000050221&_version=1&_urlVersion=0&_userid=10&md5=6efa215f6fc1d9a311dc2f93d6d1cc20

Abstract

A state-of-the-art study was performed to investigate the operational conditions for in-core and out-of-core materials in a high performance light water reactor (HPLWR) and to evaluate the potential of existing structural materials for application in fuel elements, core structures and out-of-core components. In the conventional parts of a HPLWR-plant the approved materials of supercritical fossil power plants (SCFPP) can be used for given temperatures (600 °C) and pressures (˜250 bar). These are either commercial ferritic/martensitic or austenitic stainless steels. Taking the conditions of existing light water reactors (LWR) into account an assessment of potential cladding materials was made, based on existing creep-rupture data, an extensive analysis of the corrosion in conventional steam power plants and available information on material behaviour under irradiation. As a major result it is shown that for an assumed maximum temperature of 650 °C not only Ni-alloys, but also austenitic stainless steels can be used as cladding materials.

A Korean group has used Helios.


Michael Z. Podowski, Thermal-Hydraulic Aspects of SCWR Design
http://www.jstage.jst.go.jp/article/jpes/2/1/352/_pdf

ANNUAL REPORT EXECUTIVE SUMMARY
http://www.osti.gov/bridge/servlets/purl/829883-ujDbxh/native/829883.pdf


http://www.inl.gov/technicalpublications/Documents/2699828.pdf

INEEL Neutronic Analyses. We have used MOCUP, a combination of MCNP4B and Origen2, to model the consumption of fissile material and the buildup of fission products and actinides during irradiation in a prototypical thermal spectrum SCWR with solid moderator material. The model consists of 40 axial fuel zones along the 427 cm rod height. The model includes the axial variation in the coolant density and has a two-zone variation in the fuel enrichment, 4.0 wt % 235U in the lower half and 4.2 wt % in the upper half of the rod.

 

[MCNPX]

There is a http://mcnpx.net/inputs/10-hplwr-fuel-scwr-model" if you are interested.