new iaea coordinated research project on thermal

NEW IAEA COORDINATED RESEARCH PROJECT ON THERMAL-HYDRAULICS OF SUPERCRITICAL WATER COOLED REACTORS

Katsumi Yamada

International Atomic Energy Agency (IAEA),

Vienna, Austria

Laurence Leung Canadian Nuclear Laboratories

(CNL), Ltd., Chalk River, Ontario, Canada

Walter Ambrosini University of Pisa,

Pisa, Italy

ABSTRACT

In view of the high interest among a number of Member

States in the Supercritical Water Cooled Reactor (SCWR)

concept, the IAEA launched the second Coordinated

Research Project (CRP) on thermal-hydraulics of SCWRs,

entitled “Understanding and Prediction of Thermal-

Hydraulics Phenomena Relevant to SCWRs” in 2014 to

foster international collaboration. The key objectives of this

new CRP are to (i) improve the understanding and prediction

accuracy of thermal-hydraulics phenomena relevant to

SCWRs and (ii) benchmark numerical toolsets for their

analyses. At present, 12 institutes participate in the CRP from

10 IAEA Member States, and the OECD/NEA is in

cooperation, based on a special agreement with the IAEA, to

host a database housing experimental and analytical results

contributed from the CRP participants. The expected

outcomes from this CRP include (i) enhancement of the

understanding of thermal-hydraulics phenomena, (ii) sharing

of experimental and analytical results, and the prediction

methods for key thermal-hydraulics parameters, and (iii)

cross-training of personnel between participating institutes

through their close interactions and collaboration.

This paper describes the plan of the new CRP: overall and

specific research objectives; tasks and sub-tasks; schedule;

and expected outcomes and outputs. It also introduces briefly

other IAEA activities to facilitate and support R&D for

SCWR technology in Member States, which include

technical meetings and training courses.

1. INTRODUCTION

The SCWR is one of the innovative water cooled reactor

(WCR) concepts mainly for large scale production of

electricity. Operated at supercritical pressures, the SCWR can

achieve much higher core outlet coolant temperature, leading

to around 1.3 time higher thermal efficiency than current

WCRs. In addition, the system configuration can be

simplified compared to conventional WCRs. Hence, the

SCWR concepts have the potential for improved economics.

The SCWR concept was selected as one of the six

Generation IV nuclear energy systems and has been

developed in the framework of Generation IV International

Forum (GIF), and it is the only WCR concept among them

[1]. The major technical challenges to develop the SCWR

concept are: thermal-hydraulics of supercritical pressure

water; materials development and water chemistry; and

system integration and assessment [1].

There has been high interest in research and development

of SCWRs in a number of IAEA Member States. In 2008, the

IAEA officially launched the CRP on "Heat Transfer

Behaviour and Thermo-hydraulics Code Testing for SCWRs”,

which promoted international collaboration among 16

institutes from 9 Member States and 2 International

Organizations. The CRP was completed in September 2012.

Information gathered and generated in the CRP was

documented in an IAEA TECDOC [2]. A database of

thermal-hydraulics parameters of interest to SCWR

development was compiled and is currently housed in the

OECD/NEA databank.

Despite of the completion of the CRP, several

collaborations continued between institutes participating in

the CRP. Most of these institutes expressed their strong

interest and support to initiate a new CRP on

thermal-hydraulics of SCWRs to continue the momentum of

international collaborations. Experimental and analytical

information has been generated at those institutes and is

ready to be shared with others to advance the technology.

The IAEA has recently started the second CRP on

thermal-hydraulics of SCWRs. This new CRP is expected to

advance more application-oriented technology for

thermal-hydraulics of SCWRs based on the outcomes of the

previous CRP, which focused on fundamental research to

understand key thermal-hydraulics phenomena of

supercritical fluids. The objective of the new CRP is to

improve the understanding and prediction accuracy of

thermal-hydraulics phenomena relevant to SCWRs and to

benchmark numerical toolsets for their analyses.

Proceedings of the 2016 24th International Conference on Nuclear Engineering ICONE24

June 26-30, 2016, Charlotte, North Carolina

ICONE24-60876

1 Copyright © 2016 by ASME

The new CRP also aims to provide opportunities for

cross-training of personnel between participating institutes

through closer interactions and collaboration in Member

States.

The identified scope of collaboration covers key

thermal-hydraulics phenomena, both experimentally and

analytically, to the success in developing SCWR concepts

such as heat transfer and pressure drop characteristics of

supercritical-pressure fluids, natural circulation and parallel

channel stability, critical flow and critical heat flux (CHF)

near the critical pressure. It also includes several benchmark

exercises of computational fluid dynamics (CFD) tools and

sub-channel codes, with the objectives to enhance their

prediction accuracy and extend their applicability.

2. SCWR CONCEPTS

The SCWR is a WCR concept that uses water pressurized

above the thermodynamic critical pressure (i.e., 22.1 MPa) as

reactor coolant. At supercritical pressures, water experiences

no phase change between liquid and steam. Reactor coolant

entering the core as low temperature, high density fluid

(liquid-like fluid) comes out of the core as high temperature,

low density fluid (steam-like fluid), which facilitates a

once-through reactor as in a pressurized water reactor (PWR).

Because the outlet temperature of the coolant is high and the

density is low, most of the SCWR concepts are designed as a

direct thermodynamic cycle system like a boiling water

reactor (BWR). The use of supercritical-pressure water as

coolant enables to achieve not only higher thermal efficiency

but also simpler reactor system configuration, which is the

combination of a once-through reactor and a direct cycle

system, than those of conventional WCRs.

A typical SCWR system has operating pressure of 25 MPa

and average core outlet coolant temperature targeted from

500 to 625 °C depending on concepts. Figure 1 compares the

pressure and temperature ranges of a typical SCWR core with

conventional WCR cores.

Fig. 1 Typical pressure and temperature ranges

of reactor coolant at various reactor types

The coolant temperature of conventional WCRs is

generally limited to the saturation temperature of its operating

pressure because heat transfer is decreased drastically when

fuel surfaces are covered with steam (sometimes called

‘boiling crisis’), but the SCWR could overcome the limitation

because no boiling occurs at fuel surfaces although attention

should be paid to heat transfer ‘deterioration’, which could

occur at low flow rate and high heat flux conditions but is

much milder than the boiling crisis.

Table 1 summarizes features of various SCWR design

concepts that are being developed worldwide ([3]-[8]).

The SCWR can be designed as pressure-vessel (PV) type

reactor like conventional light water reactors (LWRs) or as

pressure-tube (PT) type reactor like conventional heavy water

reactors (HWRs). Proposed fuel assemblies are based on

traditional WCR fuels, and the core can be designed as

thermal, fast, or even mixed-spectrum by adjusting the

amount and distribution of moderator in the core. Various

types of SCWR design concepts have been proposed with a

combination of the reactor type and the neutron spectrum.

For the reactor type, Canadian SCWR is pressure tube

type; and the others are pressure vessel type. Regarding the

neutron spectrum, some have thermal spectrum, and others

have fast spectrum. The SCWR-M concept is unique because

it has a mixed neutron spectrum that means the core has both

thermal and fast spectrum regions. The operation pressure,

the core outlet coolant temperature and the thermal efficiency

have been discussed above. For core flow path, most of the

design concept adopt double or triple core flow path that

means the reactor coolant passes the core region twice or

three times between the core inlet and the outlet.

One of the main advantages of the SCWR is that most

systems are based on existing WCR technologies, in which

extensive design, construction and operating experiences

have been accumulated worldwide, and on supercritical fossil

fuel-fired power plant (SC-FPP) technologies, which also

have substantial experiences in constructions and operations.

Proposed SCWR designs are expected to incur low capital

cost attributed to its high thermal efficiency and simplified

reactor system configurations. In addition, the possibility of a

fast neutron spectrum core would enable the SCWR to be a

high converter, or even a breeder, or a burner of

minor-actinides.

To realize the SCWR concept, on the other hand, extensive

and comprehensive R&D is necessary mainly because

operating pressure and temperature conditions are different

from those of conventional WCRs, and its flow channel

configuration is completely different from that of SC-FPP

boilers. One of the most important R&D areas is

thermal-hydraulics of supercritical-pressure water in fuel

assemblies. Precise prediction of fuel cladding temperature is

indispensable to evaluate the fuel rod integrity and safety

margin. Another critical area is materials for in-core

structures, especially for fuel claddings, and water chemistry.

It is necessary to adopt materials that can withstand under

high pressure, high temperature and irradiated water

conditions. And also, development of analysis tools for

system integration and safety assessment is necessary for new

reactor system design. The R&D needs mentioned above are

common to all SCWR concepts, and it makes collaboration

beneficial for the institutes having R&D activities and/or

developing their concepts.


Table 1 Comparison of Various SCWR Design Concepts under Development

Concept Name

[Reference]

Canadian

SCWR [3]

SCWR-M

[4]

CSR 1000

[5]

HPLWR

[6]

Super FR

[7]

VVER-SCP

[8]

Reactor Type1) PT PV PV PV PV PV

Neutron Spectrum2) Thermal Mixed Thermal Thermal Fast Fast

Moderator Heavy

Water

Light

Water

Light

Water

Light

Water (None) (None)

Normal Operation

Pressure (MPa) 25 25 25 25 25 24.5

Core Outlet Coolant

Temperature (deg-C) 625 510 500 500 508 540

Plant Thermal

Efficiency (%) 48 44 43.5 43.5 44 4345

Core Coolant Light

Water

Light

Water

Light

Water

Light

Water

Light

Water

Light

Water

Core Flow Path3) Single Double Double Triple Double Single

/Double

[Notes]

1) Reactor Type: PT means a pressure tube type reactor; and PV means a pressure vessel type reactor.

2) Neutron Spectrum: Mixed neutron spectrum means that the core has both thermal and fast spectrum

regions.

3) Core Flow Path:

Single core flow path means that the coolant passes the core region only once between the core inlet

and the outlet like the conventional PWRs;

Double core flow path means that the coolant passes the core region twice between the core inlet and

the outlet; and

Triple core flow path means that the coolant passes the core region three times between the core inlet

and the outlet.


3. OBJECTIVE OF THE CRP

The overall objective of the CRP is to improve the

understanding of thermal-hydraulics phenomena and

prediction accuracy of thermal-hydraulics parameters related

to SCWRs and to benchmark numerical toolsets for SCWR

thermal-hydraulics analyses.

More specifically, the CRP aims to:

1) Improve predictive capability of heat transfer, pressure

drop, sub-channel mixing, and critical heat flux (CHF)

at near critical pressures for SCWR fuel related

geometries including separate effects, and

2) Improve predictive capability of parallel channel

stability boundary, natural circulation flow and critical

flow.

4. PARTICIPATING INSTITUTES AND RESEARCH

COORDINATION MEETINGS

Twelve (12) institutes from 10 IAEA Member States are

participating in the CRP. These institutes are:

• Canadian Nuclear Laboratories (CNL), Canada;

• University of Ontario Institute of Technology (UOIT),

Canada;

• China Institute for Atomic Energy (CIAE), China;

• Shanghai Jiao Tong University (SJTU), China;

• Karlsruhe Institute of Technology (KIT), Germany;

• Budapest University of Technology & Economics

(BME), Hungary;

• Bhabha Atomic Research Centre (BARC), India;

• University of Pisa, Italy;

• JSC OKB Gidropress, Russian Federation;

• University of Sheffield, UK;

• National Technical University of Ukraine (NTTU),

Ukraine; and

• University of Wisconsin, USA.

The OECD/NEA is supporting the CRP, under a special

agreement with the IAEA, and hosts the thermal-hydraulics

experimental and direct numerical simulation (DNS)

database.

The Chief Scientific Investigators (CSIs) and observers

from participating institutes gathered at the first Research

Coordination Meeting (RCM) held at the IAEA Headquarters

in October 2014 to: 1) share information on R&D activities

so far and future plan related to thermal-hydraulics of

SCWRs at each participating institute; 2) discuss and

establish the tasks and their plan for the CRP; and 3) create

an Integrated Research Plan (IRP) for the 1st CRP period (i.e.,

between the 1st and 2nd RCMs). It was found that a substantial

amount of R&D activities had been performed, is on-going

and/or being planned at the participating institutes. These

activities are either supplementary or complementary to

others, and are of interest to most CSIs. The CSIs discussed

and established the task plan, as described in the next chapter,

was established to achieve the objective of the CRP. It

describes R&D activities in each institute and their common

interests.

The 2nd RCM was convened in Mumbai, India, in

November 2015 to: 1) review the R&D activities related to

the CRP at each participating institute during the 1st period;

2) discuss possible modifications to the CRP plan and tasks;

and 3) create IRP for the 2nd CRP period (i.e., between the 2nd

and 3rd RCMs). The CSIs presented their R&D activities and

discussed the results in detail. Based on the presentations and

discussion, the CSIs modified the task plan and established

the IRP for the 2nd period (i.e., between the 2nd and the 3rd

RCMs).

5. TASK PLAN OF THE CRP

The task plan established by the CSIs consists of eight

tasks and 19 sub-tasks, which are closely related and each led

by one of the CSIs.

5.1 Task 0: SCWR concepts update

Several SCWR concepts have been revised or newly

created since the survey in the previous CRP. This task has

been established to review and update the system parameters

of SCWRs under development such as operation pressure,

reactor coolant flow rate, core inlet and outlet temperatures

and geometries of the core and fuel assembly.

The outcome of this task will be up-to-date descriptions of

SCWR concepts and parametric ranges of interest for

thermal-hydraulics.

5.2 Task 1: Heat transfer in non-bundle geometries

Heat transfer tests in tubes and/or annuli have been

on-going in several participating institutes. In addition, DNS

is planned in an institute. It is expected that information from

the tests and simulation will improve the understanding of

fundamental mechanisms of heat transfer to supercritical

fluids and the prediction accuracy of cladding temperatures.

This task is separated into three sub-tasks: tests and

database update; correlations and scaling; and CFD analyses

and benchmark exercises.

5.3 Task 2: Heat transfer in rod bundles

Heat transfer tests in rod bundles are on-going or planned

in several institutes. Each test bundle has a different

geometry from others. Comparisons of heat-transfer

characteristics of these bundles would improve the

understanding of complex thermal-hydraulics phenomena and

enhance the prediction capability of the analytical toolsets

supporting the design and optimization of SCWR fuel

assemblies. This task is closely related to Task 1 and Task 7.

This task is separated into three sub-tasks: tests and

database construction; correlations for bundle geometries;

and numerical analyses and benchmark exercises.

5.4 Task 3: Pressure drop

Several institutes have obtained or plan to obtain new data

on pressure drop in tubes, annuli and bundles with

supercritical fluids. These data will be compiled into the

existing database to improve the understanding of hydraulics

characteristics and the prediction accuracy of pressure drops

over fuel assemblies.

This task is separated into two sub-tasks: tests and database


update; and correlation assessment, CFD analyses and

benchmark exercises.

5.5 Task 4: Natural circulation and parallel channel

stability

Supercritical pressure fluid flow can be unstable due to the

drastic change of density across the pseudo-critical point.

Single and two parallel channel experiments with natural

circulation of supercritical fluids are on-going or planned at

several participating institutes to improve the understanding

and prediction accuracy of natural circulation heat transfer

and parallel channel stability.


update; and numerical analysis and benchmark exercises.

5.6 Task 5: Critical flow

Understanding critical flow characteristics is essential for

safety analyses and the design of pressure-relieve valves.

Two critical flow experiments with flow restrictions, such as

orifices and valves, are on-going. Data from these

experiments will be applied in model development. In

addition, a numerical analysis is planned.


update; and numerical analyses and model development.

5.7 Task 6: CHF at near critical pressure (sub-critical)

CHF at sub- and near critical pressures is one of the

thermal limitations during pressure transients of SCWRs.

Very few data have been compiled from literature reviews.

CSIs have indicated that four (4) tests are planned with water

or surrogate fluids.


update; and data analyses and correlation development.

5.8 Task 7: Sub-channel model improvement

Performing full-scale fuel assembly experiments is

premature at the conceptual design phase of SCWRs. Fuel

design and optimization are performed using sub-channel

codes. Most sub-channel codes have been qualified for use in

analyses at sub-critical pressures. Further qualifications

/improvements are required to understand and predict the

sub-channel effects at supercritical pressures. This task aims

at improving prediction accuracy of sub-channel models for

supercritical flow using CFD codes and DNS.

This task is separated into four sub-tasks: inter-subchannel

mixing; spacer effects on heat transfer; axial/radial power

profile effects on heat transfer; and sub-channel shape effects.

6. EXPECTED OUTCOMES AND OUTPUTS

The expected outcomes of the CRP are:

1) Improved understanding of SCWR thermal-

hydraulic phenomena;

2) Improved accuracy of SCWR thermal-hydraulic

predictions;

3) Coordinated strategy for SCWR thermal-hydraulic

R&D;

4) Enhanced interactions and cooperation among

participating institutes; and

5) Enhanced education and knowledge management.

The outputs from the CRP include:

1) An IAEA TECDOC synthesizing the main results of

the CRP;

2) Joint papers in national/international scientific

journals and conferences; and

3) A thermal-hydraulics database.

7. OTHER IAEA ACTIVITIES RELATED TO SCWRS

In addition to the above-mentioned CRPs, the IAEA have

organized several initiatives regarding the SCWR in

collaboration with the CRP participants and other experts in

Member States.

In order to provide a platform for detailed presentations

and technical discussions, the IAEA holds Technical

Meetings on various areas, which lead to the exchange of

results, the fostering of world-wide collaboration in research

activities, improved communications between industry

(utilities, vendors, etc.), regulatory organizations and research

institutes, and the review and updating of the science and

engineering in the areas of common interest.

A Technical Meeting on “Heat transfer,

Thermal-hydraulics and System Design for SCWRs” was

held at the University of Pisa in Pisa, Italy, July 2010. The

Technical Meeting had the following main objectives with an

emphasis on application and design issues:

• To review progress in the development of correlations,

equations and methods to describe the heat transfer

behaviour with fluids under supercritical pressure

conditions;

• To evaluate comparisons of analyses and numerical

predictions of thermal-hydraulics codes against

theoretical estimates and experimental data;

• To review the status of core design and neutronics

studies for current SCWR concepts; and

• To review the status of current SCWR concepts,

system design and approach to safety.

The second Technical Meeting on the same topic is being

organized in August 2016.

Two Technical Meetings on “Materials and Chemistry for

SCWRs” were held at the Institute for Energy, Joint Research

Centre of the European Commission (EC/JRC) in Petten, the

Netherlands, July 2011, and at the Nuclear power Institute of

China in Chengdu, China, July 2013. Main objectives of

these Technical Meetings were:

• To review progress of international initiatives and

national programs related to R&D of materials for key

components of SCWRs;

• To exchange information related to testing

methodologies and experimental setups for

examination of candidate materials for SCWR

components;

• To review progress of international and national

programs related to the development of water

chemistry control strategies for the SCWR;

• To support further pre-normative research and code

qualification activities and their coupling with

numerical predictions models for the SCWR; and

• To identify the need for a new IAEA Coordinated

Research Project related to SCWR materials and water


chemistry.

As a means to disseminate the knowledge gathered through

the work performed under the CRPs and through the

presentations and discussion at Technical Meetings, the

IAEA has developed a one-week course on "Science and

Technology of SCWRs". The course provides a

comprehensive and up-to-date review of the science and

engineering of SCWR concepts, including thermodynamics,

thermal-hydraulics and heat transfer, neutronics and core

designs, materials and chemistry, system design and safety,

and a description of various SCWR concepts currently under

development in the world. Three courses were held at the

International Centre for Theoretical Physics (ICTP) in Trieste,

Italy, June 2011, at McMaster University in Hamilton,

Canada, July 2012, and at Shanghai Jiao Tong University in

Shanghai, China, August 2013. Participants in these courses

included graduate students, university professors, engineers

and regulators from IAEA Member States.

8. CONCLUSIONS

One of the IAEA’s key functions is to “foster the exchange

of scientific and technical information on peaceful uses of

atomic energy”. The IAEA recently started the CRP on

“Understanding and Prediction of Thermal-Hydraulics

Phenomena Relevant to SCWRs”, to foster international

collaboration in Member States. The identified scope of the

CRP is considered as an application-oriented extension of the

previous CRP on "Heat Transfer Behaviour and

Thermo-hydraulics Code Testing for SCWRs” that was

completed in 2012.

Thermal-hydraulics of supercritical pressure water is one

of the most important areas to be clarified in order to analyze

and integrate the SCWR concept. The scope of this new CRP

covers key thermal-hydraulics phenomena, both

experimentally and analytically, relevant to SCWRs. This

CRP also hosts several benchmark exercises of CFD tools

and sub-channel codes.

The plan of the new CRP has been described. Other IAEA

activities, such as the organization of technical meetings and

training courses relating to SCWR development, have been

presented.

NOMENCLATURE

BWR: Boiling Water Reactor

CHF: Critical Heat Flux

CFD: Computational Fluid Dynamics

CRP: Coordinated Research Project

DNS: Direct Numerical Simulation

GIF: Generation-IV International Forum

HWR: Heavy Water Reactor

IRP: Integrated Research Plan

LWR: Light Water Reactor

PT: Pressure Tube

PV: Pressure Vessel

PWR: Pressurized Water Reactor

RCM: Research Coordination Meeting

SC-FPP: SuperCritical Fossil fuel-fired Power Plant

WCR: Water Cooled Reactor

ACKNOWLEDGEMENTS

The authors express their gratitude to Chief Scientific

Investigators from the participating institutes for their

contribution to the CRP on “Understanding and Prediction of

Thermal-Hydraulics Phenomena Relevant to SCWRs”. They

are: Mr. I. Pioro, University of Ontario Institute of

Technology (UOIT), Canada; Mr. Y. Chen, China Institute of

Atomic Energy (CIAE), China; Mr. X. Liu, Shanghai Jiao

Tong University (SJTU), China; Mr. X. Cheng, Karlsruhe

Institute of Technology (KIT), Germany; Mr. A. Kiss,

Budapest University of Technology & Economics (BME),

Hungary; Mr. P.K. Vijayan, Bhabha Atomic Research Centre

(BARC), India; Mr. A. Churkin, JSC OKB Gidropress,

Russian Federation; Mr. V. Razumovskiy, National Technical

University of Ukraine (NTUU), Ukraine; Mr. S. He,

University of Sheffield, UK; and Mr. M. Anderson,

University of Wisconsin-Madison, USA.

The authors also acknowledge the OECD Nuclear Energy

Agency (NEA) for their cooperation in hosting the database

to collect and share experimental and analytical data for the

CRP.

REFERENCES

1. Generation IV International Forum, “Technology

Roadmap Update for Generation IV Nuclear Energy

Systems”, OECD/NEA, January 2014.

2. IAEA-TECDOC-1746, "Heat Transfer Behaviour and

Thermo-hydraulics Code Testing for Supercritical Water

Cooled Reactors (SCWRs)”, Vienna, 2014.

3. M. Yetsir, M. Gaudet and D. Rhodes, “Development and

Integration of Canadian SCWR Concept with

Counter-Flow Fuel Assembly”, Proceedings of the 6th

International Symposium on Supercritical Water-Cooled

Reactors, Shenzhen, China, March 3-7, 2013.

4. X. Cheng., X.J. Liu, Y.H. Yang, “A Mixed Core for

Supercritical Water-cooled Reactors”, Nuclear

Engineering and Technology, 40(2) (2007) 117–126.

5. Nuclear Power Institute of China, “Status report -

Chinese Supercritical Water-Cooled Reactor

(CSR1000)”, IAEA ARIS Database

(https://aris.iaea.org/), March 2016.

6. T. Schulenberg and J. Starflinger (eds.), “HIGH

PERFORMANCE LIGHT WATER REACTOR Design

and Analysis”, KIT Scientific Publication, 2012.

7. Y. Oka, s. Koshizuka, el at., “Super Light Water

Reactors and Super Fast Reactors: Supercritical-Pressure

Light Water Cooled Reactors”, Springer, July 2010.

8. S.B. Ryzhov, P.L. Kirillov, et al., “Concept of a

single-circuit RP with vessel type supercritical

water-cooled reactor”, Proceedings of the 5th

International Symposium on Supercritical Water-Cooled

Reactors (ISSCWR-5), Vancouver, British Columbia,

Canada, March 13–16, 2011.


new iaea coordinated research project on thermal

Documents