design and development of gas turbine high temperature ... · htgr-gt. since 2001, the basic design...

GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1059

Design and Development of Gas Turbine High Temperature Reactor 300 (GTHTR300)

Kazuhiko Kunitomi1*, Shoji Katanishi1 , Shoji Takada1, Takakazu Takizuka1, Xing Yan1, Shinichi Kosugijyama1,

Hiroyuki Okada2 and Yoichi Sakuragi2 1Department of Advanced Nuclear Heat Technology, Oarai Research Establishment, Japan Atomic Energy Research Institute,

Oarai- machi, Ibaraki-ken, Japan 311-1394, Tel:+ 81-29-264-8892, Fax:+81-29-264-8710, E-mail: [email protected]

2Nuclear Power Engineering Department, Tokyo Electric Power Company,1-1-3 Uchisaiwai-cho, Chiyoda-ku ,Tokyo, Japan 100-0011

JAERI (Japan Atomic Energy Research Institute) started design and development of the high temperature

gas cooled reactor with a gas turbine electric generation system, GTHTR300, in April 2001. Design originalities of the GTHTR300 are a horizontally mounted highly efficient gas turbine system and an ultimately simplified safety system such as no containment building and no active emergency core cooling. These design originalities are proposed based on design and operational experiences in conventional gas turbine systems and Japan’s first high temperature gas cooled reactor (HTTR: High Temperature Engineering Test Reactor) so that many R&Ds are not required for the development. Except these original design features, devised core design, fuel design and plant design are adopted to meet design requirements and attain a target cost.

This paper describes the unique design features focusing on the safety design, reactor core design and gas turbine system design together with a preliminary result of the safety evaluation carried out for a typical severe event. This study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan.

KEYWORDS: HTGR, HTTR, Gas Turbine, Electric generation, safety, coated fuel particle, graphite

I. Introduction

Development of the High Temperature Gas-cooled Reactor with a gas turbine is underway in many countries(1-5). In Japan, the HTTR (High Temperature Engineering Test Reactor) with the outlet gas temperature of 950ºC and thermal power of 30MW was constructed at Oarai research establishment in JAERI. Its first criticality was attained in 1998(6) and its full power operation with outlet gas temperature of 850 ºC was successfully carried out on December 7, 2001(7).

In addition to the HTTR development, JAERI had undertaken the feasibility study on various types of HTGR-GT. Since 2001, the basic design of a Japan’s original gas turbine high temperature reactor (GTHTR300) shown in Fig.1 has been conducted. The GTHTR300 is expected to be deployed in 2010th as a safe and economically competitive electric generation system in Japan. Economical target of this system is the electricity cost of 4 Yen/kWh taking advantage of inherent safety of the HTGR and adopting unique design originalities. The basic design phase consisting of the system and component design, safety evaluation and economical assessment will be finished by the end of March 2004. Design modification will be conducted from 2004 to 2007 after a design review by utilities, and the basic design and development of the GTHTR300 including R&D for the gas turbine system will be finished by the end of March 2008.

In the following chapters, such original design features are described in detail, exemplified by safety, reactor

core design, fuel design and plant designs.

Turbine Generator

BlockCore

Compressor

Reactor

Heat ExchangerVessel

Power Conversion Vessel

Recuperator

Precooler

ControlValves

Figure 1 Plant layout of GTHTR300

Turbine Generator

BlockCore

Compressor

Reactor

Heat ExchangerVessel

Power Conversion Vessel

Recuperator

Precooler

ControlValves

Figure 1 Plant layout of GTHTR300 II. Basic Requirement for GTHTR300

User requirement for the GTHTR300 was determined with the help of Japanese utilities, industries and experts from national institutes and universities. The simple and unique GTHTR300 was designed to meet this requirement. Table 1 shows the major specification of the GTHTR300.


Table 1. Major specifications of GTHTR300

Reactor power 600 MWt Reactor inlet/outlet temp. 587/850℃ Gas turbine input pressure 7 MPa Gas turbine mass flowrate 439 kg/s Reactor core height ~8 m Inner/outer diameter of core ~3.6/5.5 m Fuel enrichment 14wt% Average power density 5.8 MW/m3 RPVinner diameter ~7.6m Inner dia. of gas turbine vessel ~5.7 m Inner dia. of heat exchanger vessel ~5.8 m Net efficiency 45.6% III. Safety philosophy

Severe accidents are defined as any conditions beyond design-base accidents, causing core damages with fission product releases to the environment, although all severe accident sequences are very low in probability. The new safety philosophy is to avoid most accidents, and to achieve a probability of severe accidents at least lower than 10-8/ry in internal events. Even in the worst event, fuel temperature exceeding its failure limit and excessive fuel oxidation by air ingress can be avoided because of the inherent safety features and the passive decay heat removal system. The severe accident caused by the external event is prevented by the same procedure as that of a new generation LWR.

Since new data for fuel performance were or will be accumulated in HTTR operations, mechanistic source term is used to estimate radionuclide releases for plant siting evaluation instead of non-mechanistic source term based on AEC document TID-14844 and Japanese LWR or HTTR safety evaluation guideline(8). Initial failure of coated fuel particles in the HTTR in manufacturing is only 8×10-5

comparing with 10-2 in design, and no apparent failure was found in continuous irradiation tests up to 6.5% FIMA at Japan Material Testing Reactor (JMTR). A post irradiation test is also planed. In addition to these tests, data for long -term integrity of the fuel and plate out behavior will be accumulated in the HTTR operations. The full power operation of the HTTR was already finished, and no additional failure was found evidently.

When the mechanistic source term is used for the plant siting evaluation, the effective dose equivalent to whole body in the worst event can meet the dose guideline without the containment vessel.. However, in order to limit fuel failure by air ingress, a double-confinement is used in this system in lieu of the containment vessel. The double confinement consists of a building below and above the ground and can restrain the air ingress into the RPV.

The safety evaluation of the GTHTR300 has been made considering its inherent safety characteristics and outstanding engineering features of the system. The classification of the events to be evaluated is based on the HTTR and LWR safety evaluation guideline(8). In LWR and HTTR safety evaluations, the Anticipated Operation Occurrence (AOO) is defined as

the off-normal events with its frequency of roughly lower than approximately 10-2/ry, and the accident is defined as that with its frequency of lower than approximately 10-4/ry. The deterministic evaluation will be carried out to these AOOs and accidents. On the other hand, in the GTHTR300 safety evaluation, the accident is defined as the events with its frequency in the range from 10-2 /ry to 10-8/ry. The deterministic evaluation shall be made for these accidents. Full-scale PSA will be used as reference in a safety licensing of the GTHTR300.

IV. Safety evaluation 1. Acceptance criteria

Acceptance criteria for the GTHTR300 were determined based on those of the HTTR. Regarding the fuel and core damage criteria, the basically same criteria were used as those of the HTTR. Table 2 shows the acceptance criteria for the GTHTR300.

Table 2 Acceptance criteria for fuel in Anticipated

Operational Occurrence and Accident Anticipated Operational Occurrence

The maximum fuel temperature < 1600℃

Accident No severe damage in core Acceptance criteria for the primary boundary and

turbocompressor unit will be studied in the next phase of the design.

2. Selection of the events

Off-normal events to be postulated as AOOs and accidents were selected considering its frequency of occurrence and based on the investigation of main causes that affect each item of the acceptance criteria identified for the GTHTR300, that is (a) fuel temperature, (b) core damage, (c) pressure at the primary boundary, (d) missile damage from the turbocompressor unit, (f) pressure inside the confinement.

A major accident and hypothetical accident are selected and evaluated to endure the safety of the public in the case of serious accidents. The depressurization accident initiated by a break of the concentric cross duct between the RPV and PCV, or the reactivity insertion accident following the depressurization initiated by a stand pipe rupture are the severest accident in the GTHTR300. The major and hypothetical accident will be determined after the preliminary evaluation of both events.

3. Preliminary result of safety evaluation

In the final safety evaluation, the evaluation of all events selected in each evaluation category shall be carried out. However, the following two events were preliminary evaluated, the depressurization accident initiated by the break of the concentric cross duct between the RPV and PCV, and the reactivity insertion accident initiated by a stand pipe rupture. These two events represent the severest effect on the reactor system. The result of the depressurization accident is shown here.


After the concentric cross duct rupture, the fuel temperature increases due to the loss of the forced cooling of the core though the reactor is shut down by the detection on the decrease of primary flowrate. The Reactor Core Cavity Cooling System (RCCS) installed around the RPV can remove the residual heat of the core from the RPV to the cooling panel of the RCCS by mainly radiation. Then, primary coolant helium gas including fission products blows out to the atmosphere. Shearing force produced by the helium gas lifts off plate-out fission products on the inner surface of the primary pipes, turbine blades and a heat exchanger core. Those fission products detached from the primary circuit are released from a stack to the environment. After the complete release of the helium gas, air ingress into the core starts. The air ingress into the core occurs by diffusion until the amount of air reaches the point that makes the driving force of natural circulation enough to induce the air into the core. However, the core temperature decreases below the oxidation reaction temperature before the natural circulation starts, and the oxidation of the core is limited.

Temperature transient in the depressurization accident was evaluated by the TAC-NC code(9). The code was verified before the safety evaluation of the HTTR. Figure 2 shows the maximum fuel temperature during the accident. The temperature does not exceed 1600℃ during the accident. The results proved that the RCCS can successfully cool the RPV as well that internal graphite structures absorb the decay heat.

Fig. 2 Maximum fuel temperature during depressurization . accident

V. Reactor core design

The reactor core consists of 90 annular fuel columns, 73 inner reflector columns (including control column), 48 outer reflector columns and 18 sectors of fixed reflectors, shown in Figure 3. An effective core size is about 5.5m in outer diameter and about 3.6m in inner diameter, and about 8m in height. Each fuel column consists of 8 layers of hexagonal fuel blocks, upper and lower reflector blocks. Each fuel block contains 57 fuel rods, and its size is about 0.4m across flat and about 1m in height. Coolant helium gas flows downward in annular space around fuel rods and removes heat from the fuel rods. Major characteristics of the

reactor core design are shown in Table 3. The feasibility of the reactor core design based on above proposal was confirmed using the same procedure for the HTTR core design(10-11). The power density is relatively flat during the operation and its peak is lower than 14 MW/m3. This low peak power can provide the maximum fuel temperatures almost below 1400ºC in normal operations shown in Fig.4. And, the maximum fuel temperature during the depressurization accident will become low because of a flat power density and low peak.

ReplaceableReflector Block

PermanentReflector Block

Fuel block

Control rodBlock

Figure 6 Cross sectional view of GTHTR300 core

ReplaceableReflector Block

PermanentReflector Block

Fuel block

Control rodBlock

Figure 6 Cross sectional view of GTHTR300 coreFig. 3 Cross sectional view of GTHTR300 core

200

400

600

800

1000

1200

1400

1600

1800

0 20 40 60 80 100 1Time (hour)

Tem

pera

ture

( ℃) Maximum Fuel Temperature

Temperature limit of 1600℃

RPV Temperature

20

Number of fuel block from the top

500

600

700

800

900

1000

1100

1200

1300

1400

1500

0 1 2 3 4 5 6 7 8

Number fuel block from the top

Fuel

tem

pera

ture

（℃）

　Coolant temperature

Fuel surface temperature

Maximum fuel temperature

1 2 3 4 5 6 7 8

Top Bottom

Figure ation


500

600

700

800

900

1000

1100

1200

1300

1400

1500

0 1 2 3 4 5 6 7 8


Fuel

tem

pera

ture

（℃）




1 2 3 4 5 6 7 8

Top Bottom


500

600

700

800

900

1000

1100

1200

1300

1400

1500

0 1 2 3 4 5 6 7 8


Fuel

tem

pera

ture

（℃）




1 2 3 4 5 6 7 8

Top Bottom

Figure ation 10 Fuel temperature profile during oper 10 Fuel temperature profile during operFig. 4 Fuel temperature profile during operation

Table 3 Major parameters of GTHTR300 reactor core EFPD 730days Fuel UO2 Number of fuel columns 90 Batch number 2 Fuel enrichment 14% Average burnup 114GWd/ton Amount of Initial Heavy Metal 7.69 ton Discharged fuel enrichment 5%


VI. Fuel design The fuel assembly is so called pin-in-block type which is composed of fuel compacts and a hexagonal fuel block. Fuel compacts are vertically piled up around a central rod and inserted into a fuel hole in the fuel block shown in Fig.5. The fuel assembly does not have a sleeve around the fuel compacts in order to keep the fuel temperature as low as possible. In the HTTR fuel design, radial temperature difference generated in the gap between the sleeve and compacts was thought to be the major cause of relatively high fuel temperature despite the average power density of 2.5 MW/m3. The fuel compact contains TRISO coated fuel particles consisting of a low-density porous pyrolytic carbon (PyC) buffer layer adjacent to the spherical fuel kernel, an isotropic PyC layer (inner PyC: IPyC), a silicon carbide (SiC) layer and an outermost PyC (outer PyC : OPyC) layer. Thickness of each layer is modified from that of the HTTR fuel particle so that the fuel particle retains fission product during the operation with the average burn up of 120GWd/ton and the maximum 150GWd/ton. Major specification of the fuel is shown in Table 4. The thickness is determined based on several R&D for a high burn up fuel, and its integrity was evaluated by the methodology developed for the HTTR fuel design(12).

Bottom end spacer

Fuel compact

Coolant channel

Center rod

Upper end spacer

Fuel block(graphite

Co

Fuel kernelHigh desity PyC

SiC

Low density PyC

1mm

Coated fuel part

8.5mm

80mm

26mm

Figure 11 Sandwich shuffling of GTHTR300.

ated fuel particCoated Fuel Particle

Fuel rodFuel Rod

Fuel compFuel Compact

Fig. 5 Fuel Rod and Compact of GTHTR300

Table 4 Major specifications of GTHTR300 fuel

Fuel type Rod Fuel coating type TRISO Fuel kernel material UO2 Diameter of fuel kernel 550μm Thickness of buffer layer 140μm Thickness of inner PyC layer 25μm Thickness of SiC layer 40μm Thickness of outer PyC layer 25μm Diameter of coated fuel particle 1010μm Enrichment 14wt% Packing fraction of fuel compact 33vol%

Fuel cycle of the GTHTR300 is compliance with

Japanese policy, the recycling fuel cycle. Figure 6 shows the

scheme of the recycling. Fuels discharged from the core were once kept in fuel storages in a reactor site. After about a year of cooling, spent fuels are transferred to a recycling facility. In the recycling facility, fuel compacts are disassembled from fuel blocks and the fuel compacts are burnt to remove the outer layer of the CFP. Then, the SiC layers are mechanically milled by a jet grind method(13). The jet grind method developed by JAERI is superior in maintenability and durability because no blade for crushing the CFP is necessary. After the SiC layer was removed, the CFP is burnt again to remove the inner PyC layer and buffer layer. In the burning process, off-gas cycle using graphite-CO2 reaction was developed to reduce the CO emission(13). The removed UO2 fuels are sent to the PUREX facility for recycling. The PUREX process is basically the same as that used in LWR reactor fuels although radiation damage to solvent by high burnup fuels is a little bit higher. Basic technology for recycling is already established and no break through is necessary. However, demonstration test by using a pilot plant is inevitable.


VII. Plant design 1. Plant layout and non-intercooled cycle

The Power Conversion System (PCS) consists of the Heat Exchanger Unit (HEU) and Power Conversion Unit (PCU). The HEU vessel containing a pressurized water cooler (PWC) and a recuperator is placed at the lower level than the RPV in order to prevent water ingress into the core during an accident of heat transfer tube rupture in the PWC. The PCU vessel containing a turbo-compressor is allocated at the opposite side of the HEU. Since an intercooler and a water lubricant bearing are not used in the PCU, no water ingress accident occurs. Also, this plant layout greatly simplifies maintenance procedure comparing with the system whose PCU and HEX are all in one vessel.

The non-intercooled cycle loses plant total efficiency about 2% comparing with an intercooled cycle. However, it saves the capital cost almost 5% based on the preliminary evaluation(14). When the capital cost is normally assumed to be two thirds of the electricity cost, the non-intercooled cycle gains approximately 3.3% for the electricity cost. The cost estimation has relatively high margin of errors, and it may be dangerous to evaluate the electricity cost based on roughly determined design specifications. However, if the intercooler is adopted, all water related problems happened in Fort. St. Vrain in USA and AVR in Germany would be happened in the GTHTR300. In the HTTR operation, nitrogen gas used for pressurizing water was stagnated in an air cooler, and deteriorated thermal performance of the air cooler. Besides the operational complexity, a sophisticated inspecting machine is necessary since heat transfer tubes composing the primary boundary in the intercooler shall be inspected according to a Japanese regulation. That makes the O&M cost higher. Considering all effects, the non-intercooled design superior to the intercooled design.

2. Conventional steel vessel Spent fuel

Fuel compact

SiC particle

Fuel kernel

Oxide powder

Purex process

disassemble

burn

jet grind

re-burn

dissolution

off-gas recycle

Spent fuel

Fuel compact

SiC particle

Fuel kernel

Oxide powder

Purex process

disassemble

burn

jet grind

re-burn

dissolution

off-gas recycle

Figure 300 fuel

A reactor pressure vessel (RPV) can be fabricated of code certified, low-cost steel SA 533. This is made possible by a plant layout (see Fig. 1) in which coolant is circulated through reactor via a pair of leveled coaxial cross duct. Figure 7 shows the flow path inside the RPV. The helium gas of 140ºC from the compressor flows up cooling the inner surface of the RPV to keep the RPV temperature below its creep range during the normal operations. After cooling the RPV, it enters the core through the gap between the control rod and core barrel. When 0.5 % of total primary coolant flowrate is used for the RPV cooling, the temperature of the RPV during the normal operation is kept lower than the limit temperature of 375ºC stipulated in ASME section III(15) or Japanese structural standard(16). A ring with small holes placed at the space between the RPV and core barrel is used for the adjustment of flowrate for the RPV cooling.

14 Recycling process of GTHTRFig. 6 Recycling of GTHTR300 fuel

Coolant flowfor RPV coolingCore inlet

coolant flow

Core outletcoolant flow

Coolant flowfrom compressor

Coolant flowto recuperator

Coolant flowfor RPV coolingCore inlet

coolant flow

Core outletcoolant flow

Coolant flowfrom compressor

Coolant flowto recuperator

Fig. 7. Flow path in RPV

Use of SA533 reduces the capital cost by a large

margin comparing with use of high-temperature, more costly steels such 21/4Cr-Mo and 9Cr-1Mo-V. The large saving in material and ease fabrication transportation contributes to the cost advantage of the PRV design.

The preliminary design for the RPV was finished, and structural integrity during life time was confirmed. As for fast neutron damage, the total maximum neutron fluence of 3 X 1023 n/m2 is almost equivalent to that of the current LWR (3.46 X 1023 n/m2). Although the minimum temperature limit of the RPV during the high pressure operation increases, integrity of the RPV is maintained. As for the structural integrity, a pressurized accident was selected as the worst event for the evaluation. In the pressurized accident initiated by an inner pipe rupture in a concentric hot gas duct, the RPV temperature is the highest and the initial internal


pressure of the RPV remains after the accident occurs. The pressurized accident is expected to occur twice in the lifetime and classified as the ASME Service C condition. However, in this evaluation it is conservatively classified as ASME Service B condition. Temperature transient during the pressurized accident was evaluated by the thermal transient analysis code TAC-NC(9). As a result, the maximum RPV temperature reaches the peak temperature of 510ºC, keeps almost the same temperature for 50 hours and decreases below the creep range temperature of 370ºC at 440 hours after the RPV temperature becomes its peak. The thickness of the RPV is determined 166mm so that the initial stress of the RPV is kept 140 N/mm2, and the creep damage remains below 0.9 even the case that the internal pressure continues to be the initial pressure of 8.0 MPa. Practically, the internal pressure decreases rapidly after the equilibrium of primary circuit pressure and decreases further as the average helium gas temperature decreases. Considering these effects, the creep damage is far below the limit of 1.0.

3. PCS design

Figure 8 shows the schematic drawing of the Power Conversion Unit (PCU). A horizontal single-shaft helium turbocompressor is mounted in a pressure vessel directly connected to the RPV. The turbocompressor consisting of a six-stage turbine and a twenty-stage of compressor drives a synchronous generator with a shaft speed of 3600rpm corresponding to 60 Hz electrical service. The stage reaction of the compressor is selected to be 75% to achieve the highest aerodynamic performance with the small loss of secondary flow. On the other hand, the stage reaction of the turbine is designed to be 43-45% because effect of the secondary flow is negligible. A three-dimensional computational fluid dynamic analysis was carried out to

optimize the blade shape of the turbine and compressor. Due to the adoption of 3D blades, polytropic efficiency of the

turbine and compressor is 93% and 90.5%, respectively. However, construction and operation experience for the closed cycle helium gas turbine has never been accumulated in Japan. And, German experience is not enough for this system. In order to demonstrate aerodynamics performance, an one-third scale compressor test will be constructed and its performance will be confirmed.

Rotordynamics of the turbocompressor also important The rotor of the generator is connected with that of the turbocompressor with a flexible diaphragm coupling which does not transfer vibration effect from the turbocompressor to generator or the generator to turbocompressor. Each rotor was supported with two sets of journal magnetic bearings at each rotor end. One thrust magnetic bearing is installed at the turbocompressor rotor end to absorb the trust unbalance between the gas turbine and compressor. The journal magnetic bearings bolt the rotor weight of 45 ton and 60 ton in the turbocompressor side and generator side, respectively. The finite element analysis for the rotor vibration mode proved that the rotor has a 20 % margin from the critical speeds (see Table 5). Although the rotor must pass the first bending mode, the 3rd critical speed, in acceleration to the rated speed, the calculated damping Q factor is 2.8. The Q factor is small enough and no difficulty is expected to pass the critical speed. The second bending mode, the forth critical speed (124% of the rated speed) is well over the maximum rotor speed even during the loss of off site power accident. The rotor speed is controlled to be less than 110% of the rated rotor speed using a by-pass control system.

Table 5 Critical speed and damping Q factor of turbo

compressor and generator Mode Turbo compressor Generator Critical

speed (rpm)

Damping Q factor

Critical speed (rpm)

Damping Q factor

1 641 1.8 824 4.7 2 1700 1.5 1007 5.6 3 2587 2.8 1812 1.8 4 4446 7.7 2882 5.1

VIII. Economic assessment

The GTHTR300 cost estimation was carried out by referring economics of current LWRs presented by Agency of Natural Resources and Energy(17). Capital cost, operating and maintenance cost (O&M), fuel cycle cost and decommission cost were estimated individually. Table 6 provides the capital cost of each component in the GTHTR300 plant consisting of four units with common facilities. Table 7 shows electric cost.

The capital cost includes design, material, manufacturing, construction and contingency. Labor cost necessary for cost estimation was estimated based on in-house standards of MHI (Mitsubishi Heavy Industries).

Fig. 8 Schematic drawing of Power conversion unit


Table 6 Preliminary cost evaluation for GTHTR300 Reactor components 10,159 Power conversion system 15,909 Auxiliary system 7,713 Radioactive waste treatment system 450 Fuel handling system and storage 1,895 Control and instrumentation system 4,000 Radiation protection system 1,260 Ventilation system and building 10,260 Electric system 2,200 Total 53,846 (Unit million yen)

The O&M cost and decommission cost were

estimated referring to those costs for current LWRs in Japan. The fuel cycle cost includes the capital cost for a fuel manufacturing facility, material cost including Uranium and graphite, manufacturing, inspection. Inspections for the GTHTR300 fuel were streamlined taking advantage of the fuel manufacturing experience in the HTTR. The total capital cost was 53.8 billion Japanese Yen and 197,000 Japanese Yen/kWe. It was almost two-third of the capital cost for a previous HTGR-GT system. The major reason for this cost reduction is the conventional material usage for the RPV, elimination of inter-cooler, and the horizontal turbocompressor designed based on existing technology. The total electric cost was 4.4 Yen/kWh which was close to the target cost of 4 Yen/kWh. Due to further cost reduction studies considering the optimization of the GTHTR300 design and rationalization of the construction process, the total cost is expected to reduce about 10%.

Table 7 Electric cost of GTHTR300

Calculation condition Efficiency (%) 45.7 Construction

cost(Million Yen)

53.7

Life time (year) 60

Decommissioning Cost(Yen/kWh)

0.2

Capital cost (Yen/kWh) 1.61 O & M cost (Yen/kWh) 1.48 Fuel cost (Yen/kWh)* 1.07

Total cost (Yen/kWh) 4.4 *Including intermediate storage, not including recycling

IX. Development items and GTHTR300 development schedule

The first basic design phase including the safety evaluation and economical assessment will finish at end of

FY-2003. During this design phase, Check & Review (C&R) by a special board consisting of members from utilities, universities, industries and the other national research laboratories has been performed every six months so that the design can meet the requirement from private sectors. The final C&R on the first basic design phase will be conducted in 2004 and technical suggestions from this committee will be reflected in the second basic design phase to be performed from FY-2004 to FY-2007.

In addition to the basic design, the development of the compressor and magnetic bearing for the gas turbine system started in FY-2001. These key components still have technical uncertainties such as aerodynamics in the compressor and control performance of the magnetic bearing. In the compressor development, 1/3 scale compressor with four stages was fabricated and will be tested. In the magnetic bearing development, 1/3 scale rotor with the simulated weight of turbocompressor and generator will be manufactured, and the magnetic bearing performance will be confirmed. The tests will finish at the end of FY-2007. A control and operational performance test using 1/3 scale turbocompressor system is also planned. A small helium gas with its electric heater capacity of 5 MW will be constructed and the control and operational performance of the system will be confirmed.

X. Summary

The GTHTR300 is expected to be a new electric generation system in 2010s due to salient safety and economical features. Especially, the unique design such as the horizontal turbocompressor unit, long refueling interval, and double confinement is effective for improvement of the performance and the cost reduction. Remaining technical uncertainty on the turbocompressor will be solved by several R&Ds being conducted in JAERI. Also, operational experiences related to the reactor technology will be obtained from the HTTR operation. JAERI will finish the GTHTR300 program by the end of FY2007.

Acknowledgement

Authors would like to thank Mr. Minatsuki and Mr. Tanihira in Mitsubishi Heavy Industries, and Mr. Ohashi in Fuji electric for their kind assistance.

References

1) Nicholls D. R., “Utility Requirements for HTGRs”,

Proc. Technical Committee Mtg. High Temperature Gas Cooled Reactor Technology Development , Commercializing the HTGR, Johannesburg, Republic of South Africa, November 13-15, 1996, International Atomic Energy Agency.

2) Van Heek A. I., “Status of the HTR-Programme in the Netherlands”, Proc. Workshop of the Role of Modular HTRs in Netherlands, Petten, The Netherlands, November 28-30, 1994, p.83(1994).

3) Simon W. A. and Shenoy A. S., ”International


cooperation in developing the GT-MHR evolution and program status”, Proc. of an IAEA Technical Committee Meeting, Petten, the Netherlands, 10-12 November 1997, P67-80.

4) A. Kadak, et al., “Pebble Bed Modular Reactor”, ANS Winter meeting Reno, Nevada, November 11-15, 2001.

5) Xu. Y. and Zuo K., “Studies on High temperature Research Reactor in China”,. Proc. of the sixth Asian symposium on research reactors”, Oarai Research Establishment, Japan, March 29-31, 1999, p9-16.

6) Fujimoto N., Nakano M., Nojiri N., Takeuchi S., Fujisaki K., and Yamashita k., “First criticality prediction of the HTTR by 1/M interposition method”, Proc. of the sixth Asian symposium on research reactors”, Oarai Research Establishment, Japan, March 29-31, 1999, p328-333.

7) Shiozawa S., Iyoku T. and Tachibana Y., “Present status and perspective of HTGR in Japan”, HTR-TN2002, 1st International Topical meeting on HTR, Petten the Netherlands, April22-24, 2002.

8) Nuclear Safety Commission, “Guidelines for Safety Evaluation of LWR Power Plants (Safety Evaluation Guideline)”, 1998 (in Japanese).

9) Kunitomi, K., Nakagawa S., Suzuki K. and et al., “ Two-dimensional Thermal Analysis Code “TAC-NC” for High Temperature Engineering Test Reactor and its Verification”, JAERI-M 89-001, 1989(in Japanese).

10) Yamashita K., Shindo R., Murata I, and et al., “Nuclear Design of the High Temperature Engineering Test Reactor (HTTR)”, Nucl. Sci. and Eng. 122 p212-228, 1996.

11) Harada H. and Yamashita K., ”The reactor core analysis code CITATION 1000P for high temperature engineering test reactor”, JAERI-M 89-135, Japan Atomic Research Institute, 1989 (in Japanese).

12) Saito S. and et al., “ Design of High Temperature Engineering Test Reactor (HTTR)”, JAERI 1332, Japan Atomic Energy Research Institute, 1994.

13) Sawa K., Sumita J. and Takashi W., “Fuel Failure and Fission Gas Release Analysis Code in HTGR”, JAERI-Data/Code 99-034, 1999(in Japanese).

14) Takada S. Takizuka T, K. Kunitomi and et al., “ J. At. Energy Soc. Japan, Vol.1. No.4, 2002( to be published) (in Japanese).

15) ASME Boiler and Pressure Vessel Code Sec. III. Div.1,”Rules for Construction of Nuclear Facility Components-Class 1 Components”, (2001 edition), 2001.

16) MITI, “ Structural standard for Nuclear power generation plant (MITI standard 501)”, 1994(in Japanese).

17) Agency of Natural Resource and Energy, ”Economics of nuclear power generation “, distributed document in the 70th meeting of Nuclear Group. (1999).

design and development of gas turbine high temperature ... · htgr-gt. since 2001, the basic design...

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