Materials, Coolants, and Components for Perspective Transmutation Power Units of VVER Spent Fuel

Oldrich Matal, ENERGOVYZKUM, Ltd, Bozetechova 17, 61200 Brno, Czech Republic Pavel Hosnedl, SKODA JS a.s., Orlik 266, 31606 Plzen, Czech Republic

Abstract

As part of the Generation IV International Forum advanced nuclear power generating options for the future have been identified. One of them is the Molten Salt Reactor. Its version for an incineration of VVER nuclear spent fuel has been proposed by a specialist team of the Czech Republic (so called SPHINX concept with plutonium and minor actinides fuel in the fluoride molten salt matrix). Another one is the Advanced High Temperature Reactor proposed by an US team that uses coated-particle graphite matrix fuel and molten fluoride salt coolant. For high temperature operation above 700°C of these new power systems materials, coolants, material-coolant interactions and specific components are key problems. The aim of the contribution is provide information on selection and development of materials and molten salt coolants for high temperature applications for these new nuclear power units and discuss briefly design problems of selected components like reactor and salt-salt and salt-gas heat exchangers.

I Introduction

The growing rate of accumulation of spent fuel and other long-lived wastes will be regarded as unsustainable. The solution could be to develop and implement new reactor systems and associated fuel cycles to use long-lived species such us plutonium, neptunium, americium and curium as a fuel and possibly also the transmutation of the long-lived fission products into nuclides with shorter half-lives. One of important factors influencing the development of reactors and corresponding fuel cycles is to further reduce proliferation risk. This is one of main goals presented by Generation IV Initiative from the US Department of Energy [Ref. 1]. Back-end strategies for the nuclear fuel cycle are receiving increasing attention. Long-term storage of spent fuel and nuclear waste and its disposal in geological repositories on the one hand and partitioning and transmutation of long-lived fission products and minor actinides on the other hand. Regardless of the back-end strategy, fuels incorporating minor actinides need to be studied, their relevant properties determined and tested as well as structural materials of equipments where such fuels have to be handled or operated need to be selected and tested.

II Liquid metal reactor concepts

II.1 Liquid sodium coolant concepts

The objective of projects is to design a Sodium-cooled fast reactor with a high level of safety and fuel recycling of the actinides, it is management of high-level wastes, in particular management of plutonium and other actinides and utilization of almost all the energy in the natural uranium. In the Tab. 1 there are sodium temperatures, water/steam parameters, considered or used structural materials and SG design types of two concepts considered in GEN IV, UK simplified SFR and existing units in operation (BOR 60) [Ref. 2].

The sodium temperatures at the steam generator inlet are 520°C (JSFR project) and below 500°C (K-600 and S-SFR projects and BOR 60) with the aim to simplify the SG design and to reduce temperatures below the creep range of the structural materials. Consequently, the considered SG structural material in GEN IV projects is 12Cr steel and the 2,25Cr-1Mo steel stabilized by Nb in the case of steam generators at BOR 60 (SGs are of the CR design). To verify advanced 2,25Cr-1Mo-Nb steel making and realization of forgings as well as technology of components (tubes, tube sheets etc.) for production and delivery of steam generators for SFRs in the conditions of contemporary Czech Republic industry a R and D project with participation of ENERGOVYZKUM Brno (project leader), ZDAS Žďár nad Sázavou, PBS Třebíč and Institute of Material Physics of the Academy of Sciences. Results of this project as well as very positive experience with sodium heated steam generators developed and manufactured in institutes and companies of the Czech Republic and still successfull operating at the fast reactor facility BOR 60 (Russia)[Ref. 4] provide a good basis for the Czech Republic participation in the SFR project of the Generation IV International Forum [Ref. 3]. Two steam generators of the CR provinience are operating at the BOR 60 facility. The first one is a modular steam generator of the thermal power 28MW with more then 120 000 operating hours in the steam generation mode, the second one is an one modul steam generator of the thermal power rate of 26MW with more then 75 000 operating hours in the steam generation mode. No leak and no negative comments have been reported by BOR 60 operators to both steam generators. In the Fig. 1 there is the front view of the 28MW steam generator operating at BOR 60. To avoid water/steam into sodium leak problems CO2 as working substance in the SFR tertiary circuit with Brayton cycle can be considened. The heat of the sodium in the secondary circuit is transfered to CO2 in a heat exchanger. Studies on Brayton cycle with CO2 at supercritical parameters are subjects of the project supported by the Grand Agency of the Czech Republic [Ref. 5]. Critical parameters of CO2 are: temperature of 30,98°C and pressure of 7,377MPa. Consequently the CO2 parameters at the compresor suction for supercritical parameters could be 30°C and 10MPa. The Brayton cycle parameters could be:

− gas temperature at the gas turbine inlet of 500°C, − compression ratio of 7,5, − cycle thermal efficiency of 34%, − regeneration stage of 85%.

Gas turbine and compressor are technically feasible for parameters considered above [Ref. 16].

II.2 Lead-bismuth coolant concepts

A lead-bizmuth-cooled fast reactor project was reported in [Ref. 6]. The three-circuit scheme was implemented in the reactor design of the thermal power rate of 900MW (340MW electric) by Russian organizations with a primary system monoblock unit (Pb-Bi coolant, 480/380°C) and steam turbine circuit (460°C, 15MPa steam and 260°C feed water). The melting temperature of the lead-bizmuth eutectic solution is 123,5°C. The coolant temperatures were chosen up to 500°C to solve problems of corrosion resistance of structural materials. The lead-bizmuth coolant heating up in the core (400°C to 500°C only) and outlet temperature (500°C) allows to use 12%Cr ferritic-martensitic steel resistant to radioactive swelling and corrosion (oxygen concentration level in coolant is important). The materials recomended for Pb-Bi facilities are austenite stainless steels for the temperatures up to 400°C and ferritic-martensitic steel for the temperatures above 450°C.

II.3 Lead coolant concept

Russian BREST – type reactor project is one of pure lead coolant usage examples. Pure lead melting temperature is of 327,3°C. The BREST type reactor belongs to advanced concepts requiring additional studies and a number of innovations [Ref. 7].

Lead-cooled fast reactor development was also included into the International Generation IV Forum program.

III CO2-cooled reactor concepts

Gas-cooled fast reactors (GCFR) have some advantages over liquid metal cooled concepts. Readily available and optically transparent gaseous coolant which is compatible with both air and water. The absence of a significant positive coolant void reactivity effect allows the potential for loading greater quantities of minor-actinide isotopes and in terms of minor-actinide incineration harder neutron spectrum [Ref. 8]. To disadvantages of GCFR belong relatively high CO2 pressure of more then 4MPa at high temperatures with the concequence of requirements on reactor pressure vessel design (prestressed concrete vertical cylinder) and materials of components. Nevertheless a preliminary concept for the fast spectrum enhanced gas-cooled reactor has been studied as a possible Generation IV reactor [Ref. 8] for the thermal power level of 3600MW, CO2 temperatures of 525°C/250°C (at the reactor outlet and inlet) and CO2 pressure of 4,2MPa. The CO2 coolant transfers the heat in the steam generator to the water and steam. The steam generator unit is of thermal power of 300MW (12 SG units total) with a platen-style tube bundle. The SG tubes are plain tubes, manufactured in a single material – steel of the type 9%Cr, 1%Mo. Gas fast reactor system with helium direct Brayton cycle or as option He/CO2 (the secondary circuit) two circuit solution both with the helium temperature at the core outlet of 850°C at the pressure of 7MPa are also considered[Ref. 13].

IV Molten-salt fueled and cooled reactor concepts

A molten-salt homogeneous critical reactor is proposed to protect the enviroment from the release of long-lived radwastes (minor actinides and long-lived radioactive fission products), to solve the problems with full closure of the nuclear fuel cycle on long-lived radioactive toxic nuclei and to generate heat for electricity production at the same time [Ref. 9, 10, 11]. The molten-salt reactor was included into the R and D program of the GIF.

IV.1 Used and proposed molten-salt fuels and coolants and structural materials

In the Molten Salt Reactor Experiment (MSRE, thermal power 8MW) molten-salt fuel with the content of (mol %) 657LiF-29,1BeF2-57ZrF4-0,9UF4 (melting temperature of 434°C) in the primary circuit and molten-salt coolant of (mol %) 66LiF-34BeF2 in the secondary circuit were used. The fuel-salt of (mol %) 71,77LiF-16BeF2-12ThF4-0,3UF4 (melting temperature of 500°C) in the primary circuit and salt coolant of 92NaBF4-8NaF (melting temperature of 385°C) in the secondary circuit were proposed in the US Molten Salt Breeder Reactor project (MSBR, thermal power of 2250MW). In both systems MSRE and MSBR nickel based alloy (Hastelloy N) were used and proposed as the structural material of components. Similar fuel composition based on the salt coolant of 66LiF-34BeF2 in the primary circuit and salt coolant of 92NaBF4-8NaF in the secondary circuit are considered in the Russian project focused on burning of transuranium nuclides [Ref. 11]. Also the French AMSTER molten salt reactor concept considers fuel-salt composition adopted for the thorium support fuels, namely (mol %) 70LiF-15BeF2-15ThF4 with core inlet and outlet temperature of 550°C and 700°C.

IV.2 Structural material development for MSR concepts

Within R and D program, SKODA Nuclear Machinery (NM) develops a special fluoride-salt resistant high Nickel alloys for molten salt technology. The development of the structural material

and equipment for molten fluoride salt technology plays one of the crucial steps in the Czech National program. The use of molten salts in operation conditions is a material and technological problem and there is no experience with it in our country. Therefore SKODA NM company (Czech leading nuclear manufacturer), Nuclear Research Institute (NRI) (specialist in a fluoride technology) and ENERGOVYZKUM Brno (EVM) all members of the National consorcium realized non active experimental molten fluoride salt loop ADETTE - 0 with natural convection for reactor secondary circuits molten salt fluids by operation temperature to 550°C. The next ADETTE-1 active experimental loop (forced convection, will be made from the high Nickel alloy (MONICR SKODA) is now being prepared for the reactor primary circuits with fluoride molten salt fluids), which should allow performing an extensive irradiations experimental program by operation temperature by 700oC. The both loops are designed and constructed to obtain a sufficient amount of experience on transmutation technology and for the research of the material and equipment. The research and utilization program covers question of corrosion, intergranular corrosion, stress corrosion cracking of structural materials, research of material properties and applied welding technology etc. The loop concept is close to the solution of a compact primary circuit of transmutations demo-unit and a secondary cooling system can be added with a heat-exchanger containing candidate fluoride salts. In this extremely aggressive environment this alloy show very good corrosion resistivity, convenient creep strength and excellent toughness. The first experimental program in the hot neutron environment with special MONICR capsules filled with minor actinide will be realized in the research reactor in Kurchatow Institute (Russia). The Autoclaves experiments will be realized in the new types of SKODA Autoclaves made from MONICR SKODA alloys. A view on the upper part of the experimental apparatus is in Fig. 2. The whole program is based on the generalization of the world know reactor experiments in the USA in early sixties and supposed cooperation with the Russian and American National Laboratories. International cooperation is also starting in the 5FP EU Program as the MOST (Molten Salt Technology) project with Czech participation. A Review of Structural Material for Primary and Secondary Circuit for molten-salt reactor technology was performed in the WP5 of the EC MOST project [Ref. 14]. The objective of this activity is development of the technology necessary for designing and specifying the performance of components and systems, which would constitute molten-salt demonstration reactors. This activity covers all development work associated with reactor systems and components. Although the requirements for components and systems will be specified largely by conceptual design studies for the various candidate fluoride salts and for the various molten-salt reactors, additional conceptual design studies would be carried out as necessary for identifying problem areas in proposed component designs and for properly orienting the component technology development.

IV.3 Experimental studies on material corrosion in fluoride molten-salts

To study interactions (especially corrosion) of selected materials with fluoride molten-salts three stage experimental program has been proposed namely − ampula experiments with stagnant salt melts and temperatures up to 700°C, − experiments using small closed loop with molten salt natural circulation and temperatures up to

750°C, − experiments using small closed loop with forced circulation of the salt by a pump. The first stage of experiments was started in ENERGOVYZKUM and Masaryk University in Brno under the support of the GA CR project [Ref. 12]. A view on the experimental apparatus is in Fig. 3. A vacuum and cover gas system is at the left side and a furnace with ampulas filled with fluoride molten-salt compositions is at the right side. An

example of a corrosive – errosive attack of a fluoride melt on the surface of a stainless steel of the type 18Cr/8Ni is in Fig. 4.

IV.4 R and D of some components for MSR

A review of system and components for molten-salt reactor technology was performed in the WP4 of the EC MOST project [Ref. 14]. The review of existing or proposed system and components was focused on molten-salt pumps, general and specific requirements to heat exchanges for high temperature applications, primary heat exchangers, steam generators, component configuration concepts etc. A molten-salt pump was developed, designed, manufactured and tested for salt loop applications with designed salt temperature of 550°C and volumetric flow rate of 4liters per second under the support of the GA of the CR. View on the molten-salt pump connected to a test loop with a molten fluoride salt is in Fig. 5 [Ref. 15].

Acknowledgement

The contibution results also from the solution of the Grand Agency of the Czech Republic supported projects Nr. 101/04/0872 and Nr. 101/03/0568.

Tab. 1 Selected data to some existing and considened fast reactors cooled by liquid sodium

Considened (GEN IV) Parameter JSFR K-600 S-SFR In operation

BOR 60

Reactor thermal output, (MW) 1785 1589 3600 60

Reactory type loop pool loop

(no intermediate circuit)

loop

Sodium temperature (°C) at reactor outlet inlet

550 395

510 366

495 310

500 300

Sodium temperature (°C) at steam generator (SG)

outlet inlet

520 335

495 311

495 310

480 280

Steam parameters temperature (°C) pressure (MPa)

495

16,67

463 15

460

460 14

Feed water temperature (°C) 233 230 210 200

SG structural material 12 Cr steel

12 Cr steel 2,25 Cr – 1 Mo – Nb

steel

SG design type shell-tube shell-tube copper bonded Na inside tubes

1, 2, 3 – economiser, evaporator and superheater moduls 4 – sodium inlet pipe 5 – sodium chamber 6 – sodium inlet pipes into a branch 7 – sodium outlet pipes 8 - sodium outlet chamber 9 – N/A 10 – sodium pipe (SG – buffer tank) 11 – buffer tank 12 – feed water chamber

13 – feed water pipes to branches 14 – connecting pipes 15 – steam chamber (evaporator outlet) 16 – steam chamber (separator outlet) 17 – steam outlet (from branch) 18 – steam outlet chamber (superheated steam) 19, 20, 21 – support structure 22 – isolation box 23 – electic heaters 24 – SG concrete cell

Fig. 1: Modular steam generator front view (power rate of 28MW, operating hours cca 120 000)

Fig. 2: View on the upper part of the autoclave SKODA

Fig. 3: View on the experimental apparatus EVM for material corrosion tests in molten fluoride salts

4a) magnitude 30x 4b) magnitude 460x Fig. 4: Surface of a stainless steel tube after exposition in a molten salt flow at cca 400°C

Fig. 5: View on the molten-salt pump of the type EVM-4-MS connected to a test loop

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April 7, 2004, Rez [Ref. 3] O. Matal et al.: Development of material and technology of steam generator components,

Project Nr. FD-K3/078, MIT, Prague 2003 [Ref. 4] O. Matal, A. S. Korolkov: Thirty Years of Operation of the Nuclear Installation BOR 60,

Energetika, 50 (2000), Nr. 11, p. 376 [Ref. 5] J. Fiedler et al.: The tertial circuit of an ADTT system with CO2 coolant at supercritical

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[Ref. 10] J. Uhlir et al.: Current status of molten-salt reactor technology development in the Czech Republic, GIF IV Information Seminar, NRI Rez, April 7, 2004

[Ref. 11] P. N. Alekseen et al.: Molten salt reactor for burning of transuranium nuclides forming in closed nuclear fuel cycle, Workshop Proceedings, Chester, UK, 22-24 Oct. 2001

[Ref. 12] O. Matal et al.: Research into interactions of structural materials with molten fluoride salts, project Nr. 101/04/0872, Grant Agency of the Czech Republic

[Ref. 13] D. Haas: Gas Fast Reactor System: Status and R and D Plant Description, http://www.jrc.cec.eu.irt

[Ref. 14] MOST: Review of Molten Salt Reactor Technology, EC, 5th Euratom Frame Work Programe, contract Nr. FIKW-CT-2001-20183

[Ref. 15] O. Matal et al.: A molten salt pump for a transmuter, project Nr. 101/00/0478, Grand Agency of the Czech Republic

[Ref. 16] J. Fiedler: Private communications, August 2004

Prof. Ing. Oldřich Matal, CSc. ENERGOVÝZKUM, spol s r. o. Božetěchova 17 612 00 Brno tel.: +420 541 214 660 fax.: +420 541 214 659 e-mail: energovyzkum@seznam.cz