DESCRIPTION OF CASES

The next generation CANDU reactor concept pursued in Canada is the so-called Enhanced CANDU6 or EC6 reactor which evolved from the established CANDU6 technology [40, 41, 78]. The EC6 is a third generation, heavy water cooled and moderated reactor mainly designed for electricity production with an electric power output of 740 MW(e) and a thermal power of 2084 MW(th). In general, CANDU6 is considered the only commercialized reactor with adaptability and flexibility in the fuelling arrangements. Fuel alternatives starts from recovered or reprocessed uranium fuel to advanced fuel like thorium and actinides. Similar to all CANDU reactors, the EC6 design is based on the use of horizontal fuel channels (here 380) arranged in a square pitch. Each fuel channel houses twelve 37-element fuel bundles containing natural uranium fuel and the pressurized D2O coolant. They are mounted in a calandria vessel containing the low-temperature, low-pressure D2O moderator. The fission heat is carried by the reactor coolant to four steam generators provided in the heat transport system producing steam at 260 °C. Major design parameters of the EC6 plant are in Table 65.

Current CANDU reactors (CANDU6 and EC6) produce nuclear heat at ~300 °C. Heat upgrading has to be performed to increase the temperature to the range of operating temperature of current thermochemical cycles. Integration of heat pump with the system to upgrade the nuclear heat is proposed and studied by several researchers. There is potential of internal heat recovery from thermochemical cycles integrated with nuclear power plant [62, 63]. In these studies, the feasibility of a new high temperature heat pump is analysed, which is integrated into a Copper–Chlorine (Cu–Cl) thermochemical water splitting cycle for internal heat recovery, temperature upgrades and hydrogen production. (See Canada country report on the CD-ROM for more details).

FIG. 68. Four technology-based cases of nuclear hydrogen production studied

2246.27 MWe Output to power grid Case B

178.6 MWe Output to power grid

TABLE 65. MAJOR CORE DESIGN PARAMETERS OF THE EC6

Reactor type Horizontal pressure tube

Thermal power 2084 MW(th)

Core power density 11.3 MW(th)/m³

Primary coolant Heavy water

Coolant inlet / outlet temperature 266.3 / 310.0 °C

Coolant pressure 10.1 MPa

Coolant mass flow rate 9200 kg/s

Electric power gross / net 740 / 690 MW(e)

Moderator Heavy water

Moderator temperature 69 °C

Number of fuel channels 380

Number of fuel bundles per channel 12

Channel length 5.94 m

Pressure tube inside diameter 103.4 mm Number of fuel elements per bundle 37 Length of fuel element/bundle ~0.5 m

Fuel UnatO2

Average fuel burnup 7.5 GWd/t-HM

Calandria inside diameter 7595 mm

Calandria wall thickness 28.6 mm

Number of steam generators 4

Steam temperature 260 °C

Steam pressure 4.7 MPa

Steam mass flow rate 1043 kg/s

Design lifetime 60 a

The hydrogen production method considered in Canada is the Copper–Chlorine (Cu–Cl) hybrid cycle which was first developed in 1970s. It is a medium temperature cycle operating around 550°C in three to five steps of thermochemical and electrochemical steps in different configurations. The efficiency of this cycle is calculated at about 40% [61]. Selected for the HEEP study here is only the 5-step Cu–Cl cycle [64].

5.1.2. Case B by China

China’s case is based on the on-going HTR-PM project and research results of the Sulfur–

Iodine process investigated at the Institute of Nuclear and New Energy Technology (INET),

Tsinghua University in Beijing. The HTR-PM is designed for generating an electric power of 210 MW(e) by utilizing two identical reactor units of 250 MW(th) each serving one steam turbine. It is the first commercial pebble-bed modular HTGR in China currently under construction and anticipated to be completed by the end of 2017. The plant concept is such that a high degree of standardization and modularization will be achieved. In this dedicated electricity generating plant, the helium coolant temperature at the exit is 750 °C. In the steam generator, a genuine Chinese development, heat is transferred to the steam cycle. Major design parameters of the HTR-PM are listed in Table 66.

Since the HTGR is the most suitable reactor type for nuclear-assisted hydrogen production, comprehensive investigation on nuclear hydrogen production has been initiated at the INET as part of the R&D objectives of the HTR-PM project. The Sulfur–Iodine (S–I) thermo-chemical cycle for splitting water and the high temperature steam electrolysis (HTSE) process were selected as the most promising processes of nuclear hydrogen production. Since 2005, INET has conducted preliminary studies on the S–I and HTSE processes. A Laboratory with the necessary facilities has been established for process studies of nuclear hydrogen. At the same time, the test reactor HTR-10 located at INET will provide a suitable nuclear facility for future R&D of nuclear hydrogen production technologies.

TABLE 66. MAJOR CORE DESIGN PARAMETERS OF THE BASELINE HTR-PM

Thermal power (two units) 2×250 MW(th)

Average thermal power density 3.2 MW(th)/m³

Primary coolant Helium

Coolant inlet / outlet temperature 250 / 750 °C

Coolant pressure 7.0 MPa

Coolant mass flow rate (per unit) 96 kg/s

Electric power production 210 MW(e)

Active reactor core diameter / height 3.0 / 11.0 m Number of spherical fuel elements (per unit) 420 000

Average / maximum fuel burnup 90 / 100 GWd/t

RPV inner diameter/height 5.7/24.9 m

Power conversion efficiency 42 %

Steam temperature at turbine inlet 566°C

Steam pressure at turbine inlet/outlet 13.2 MPa / 4.5 kPa Flow rate of superheated steam per unit 96 kg/s

5.1.3. Case C by Germany

The baseline concept for a German small modular HTGR is the electricity producing 200 MW(th) HTR-Modul pebble bed reactor designed by the former German company

SIEMENISNTERATOM [34]. It is characterized by a tall and slim core which ensures — in combination with a low power density — that even in hypothetical accidents, the release of fission products from the core will remain sufficiently low to cause no harm to people or environment. Consequently, a process heat variant of the HTR-Modul reactor [36] has been developed, for which — in comparison to the electricity generating plant — several modifications were necessary. The principal cornerstones of the process heat version are a thermal power of 170 MW and a helium outlet temperature of 950 °C to deliver process heat for the SMR process. A reduced system pressure of 5 MPa was chosen as compromise between a high pressure desired for its favourable effect on operating and accident conditions of the nuclear reactor and a low pressure desired for chemical process reasons in the secondary and tertiary circuit to enhance the conversion rate for maximal hydrogen production. Major design parameters of the HTR-Modul are listed in Table 67.

TABLE 67. MAJOR CORE DESIGN PARAMETERS OF THE PROCESS HEAT HTR-MODUL

Thermal power 170 MW(th)

Thermal power density 2.55 MW(th)/m³

Primary coolant Helium

Coolant inlet / outlet temperature 300 / 950 °C

Coolant pressure 5.0 MPa

Coolant mass flow rate 50.3 kg/s

Active core diameter / height 3000 / 9430 mm Number of spherical fuel elements 360 000

Average fuel burnup 80 GWd/t-HM

Coolant temperature at SR outlet 680 °C Coolant temperature at SG outlet 293 °C

Process gas temperature 810 °C

Process gas pressure 5.2 MPa

Steam temperature 540 °C

Steam pressure 11.5 MPa

Steam mass flow rate 37.6 kg/s

H2 + CO production rate 25.6 m³/s

The steam reformer uses the temperature of the helium between 950 and 700 °C, while the steam generator is using that part of heat between 700 and 250°C. The feed gas mixture with an H2O/CH4 ratio of ~ 3 is preheated up to around 500°C and reformed at a maximum process temperature of 800 °C. A fraction of 85 % of the methane is then converted in this first step.

Utilization of the heat of the reformer gas for preheating the feed gas, shift conversion, and methanation are the steps following the reformer to finally get the product hydrogen. The steam generator supplies the steam needed for the reforming process and power generation.

The overall energy balance delivers roughly the following numbers:

CH4 as raw material is completely converted to hydrogen; the total efficiency including the nuclear heat is around 65%. A complete life cycle analysis has even revealed that depending on operating conditions, about 40% savings of natural gas feedstock could be achieved, if nuclear is selected the primary energy source [79].

Without employing an IHX (which was deemed feasible and licensable at that time), the hot helium coolant is directly fed to the steam reformer as a new nuclear component which is a bundle consisting of straight splitting tubes with a length of 14 m. The reformer consumes 71 MW(th), while the steam generator is operated with 99 MW(th). From the total heat transferred into the steam reformer, 85% are used for the reforming process, with the remaining 15% being taken to heat up the feed gas.

5.1.4. Case D by Japan

The JAEA reference concept for commercial nuclear hydrogen production in Japan is based on the GTHTR300C (C = cogeneration) reactor [28, 80] to be connected to an Sulfur–Iodine thermochemical water splitting process.

The GTHTR300C design is based on a prismatic VHTR. The reactor is rated at 600 MW(th) thermal power and 950 °C coolant outlet temperature. Coolant pressure is 5.1 MPa, a reduced value compared to the electricity-only variant. The intermediate heat exchanger (IHX) used to deliver 900 °C helium as nuclear heat source to the hydrogen process is designed based the helical He-to-He counter-flow tube and shell heat exchanger, the same type operated in the HTTR. The heat capacity of the IHX is 170 MW(th). The gas turbine is designed to produce 300 MW(e) maximum in standalone power generation and 204 MW(e) when hydrogen is being cogenerated. Major design parameters of the GTHTR300C are listed in Table 68.

TABLE 68. MAJOR CORE DESIGN PARAMETERS OF THE GTHTR300C FOR HYDROGEN PRODUCTION

Thermal power 600 MW(th)

Average thermal power density 5.8 MW(th)/m³

Primary coolant Helium

Coolant inlet / outlet temperature 594 / 950 °C

Coolant pressure 5.1 MPa

Coolant mass flow rate 322 kg/s

Electric power production 204 MW(e)

Reactor core equivalent inner–outer radius / height 3600–5500 / 8000 mm

Number of fuel blocks 720 (in 90 columns)

Average fuel burnup 120 GWd/t-HM

Helium temperatures at IHX inlet / outlet 950 / 556 °C Secondary helium IHX inlet / outlet temperature 900 / 850 °C

Hydrogen conversion process S–I thermochemical cycle Efficient thermal power input to H2 production 219 MW(th)

Hydrogen production rate 1.9–2.4 t/h

The process heat required for the S–I process is provided in form of hot helium gas from the high temperature nuclear reactor and used in various steps of the process stream concentration and decomposition. The electricity is generated in-house by the same nuclear reactor and used to power the process electrolysers for stream concentration, gas circulators including the ones used in the helium gas loop to transport the heat from the nuclear reactor to the hydrogen process plant, the process fluid pumps and other utilities.

According to the energy and material balance of the S–I process, the gross thermal input is 175 MW(th), of which 5 MW(th) is input from the helium circulator gas compression heating of the heat transport loop that connects the reactor to the hydrogen plant. The net thermal input to the process is 168.9 MW(th). The net electricity consumption is 25.4 MW(e) accounting for all major usages of electricity including process electric utility (pumps and electrolyzer), and the helium gas circulation power consumption of the helium heat transport loop. Assuming a conversion efficiency of 48.8%, the hydrogen production rate is 30 655 Nm³/h (or 66.1 t/d). By-product is oxygen produced at a rate of 15 328 Nm³/h.

5.2. BOUNDARY CONDITIONS