ABWR Operation Experience at Kashiwazaki-Kariwa NPS

Tatsuro KOBAYASHI1*, Kimitoshi YAHAGI2, Masumi ISHIKAWA3, and Hideya KITAMURA4 1, 2, 3Kashiwazaki Kariwa Nuclear Power Station, TEPCO, 16-46 Aoyama-cho, Kashiwazaki-shi, 945-0393, 4Tokyo Electric Power Company, 1-3-1 Uchisaiwai-cho, Chiyoda-ku, 100-8560, Japan

The world’s first advanced boiling water reactors (ABWR), Kashiwazaki-Kariwa NPS unit 6 and 7 of twin 1,356 MWe, have continued safe and reliable operation contributing to not only supplying the Tokyo metropolitan area with electricity but also securing energy resources and reducing greenhouse gas emissions. The ABWR, which was jointly developed by GE, Toshiba, Hitachi and TEPCO based on the construction and operation experiences of conventional BWRs, could be integrated with various features, such as enhanced safety, reliability, improved operability, maneuverability, economy, reduced occupational radiation exposure and radioactive wastes. Those features have been demonstrated as expected through operation and outage experiences since 1996. Although we had experienced several unplanned shutdowns in unit 6 and 7, the root causes of these shutdowns were almost due to the conventional problems, such as the failure of electrical instrument, plant auxiliary system, or fuel failure. Thus, we will make every effort to continue excellent operation and we hope that the experience obtained at Kashiwazaki-Kariwa unit 6 and 7 is used for future development of the ABWR.

KEYWORDS: ABWR, RIP, RCCV, Digitalized Main Control Room

I. Introduction In Kashiwazaki-Kariwa NPS (Fig. 1), located 220 km northwest of Tokyo, 7 units of BWR have been in service since the start of commercial operation of unit 1 in 1985 (Fig. 2). The total electrical output is 8,212 MWe, producing approximate 60 TWh of electricity every year. There are two generations of BWR, such as the BWR-5 and the ABWR. The world’s first advanced boiling water reactors (ABWR), unit 6 and 7 of twin 1,356 MWe, have continued safe and reliable operation since their start of commercial operation in 1996 and 1997, respectively. Fig. 1 Kashiwazaki Kariwa NPS As of March 2003, operation and maintenance activities at the seven BWR units at Kashiwazaki-Kariwa NPS are The ABWR could be integrated with various features, carried out by about 950 TEPCO employees and 3 to 4 such as enhanced safety, reliability, improved operability, thousand subcontractor employees. Of these, about 60 maneuverability, economy, reduced occupational radiation operators and 20 dedicated O&M support team members exposure and radioactive wastes. Those features have are employed for the twin ABWRs. In addition, digital been demonstrated from construction through operation control engineering and maintenance group was organized stages. In this paper, we introduce ABWR operation in 1998 as a special team for the maintenance of the digital experience since the start of commercial operation of unit 6 equipment mainly adopted in the ABWRs. and 7.

* Corresponding author, Tel. +81-257-45-3131, Fax. +81-257-20-2529, E-mail: [email protected] 1/8 Unit 7: 1,356MW Feb-92 65months Jul-97

Unit 6: 1,356MW Sep-91 62months Nov-96

Unit 5: 1,100MW Oct-83 78months Apr-90

Unit 4: 1,100MW Feb-88 78months Aug-94

Unit 3: 1,100MW Jul-87 73months Aug-93

Unit 2: 1,100MW Oct-83 83months Sep-90

Dec-78 82months Sep-85 Unit 1: 1,100MW

Fig. 2 Construction of Kashiwazaki-Kariwa NPS

II. Main Feature of the ABWR The ABWR, which was jointly developed by GE, Toshiba, Hitachi, TEPCO and other BWR operating utility companies in Japan, was evolved from the construction and Fig. 3 Design feature of the ABWR operation experiences of conventional BWRs. TEPCO decided to deploy the first-of-a-kind ABWR units at Kashiwazaki-Kariwa NPS in 1987. 1. Reactor Internal Pump (RIP) The advanced control & instrument system, fine-motion The introduction of RIPs eliminates the recirculation control rod drive (FMCRD), reactor internal pump (RIP), piping outside the vessel that results in the improvement of reinforced concrete containment vessel (RCCV), and large operability and safety. The reactor core will remain capacity turbine & moisture separator re-heater are main covered with water during any pipe break accidents due to features of the improvements adopted in the ABWR (Table elimination of recirculation piping (Fig. 4). This 1, Fig. 3). simplified piping of reactor recirculation was able to realize Safety has been further enhanced, compared with to be free from the SCC on PLR piping, which is recently conventional BWRs by the introduction of an ECCS detected and repaired in unit 1, 2, 3, 4 and 5. network consisting of three independent divisions with low One of 10 RIPs is maintained in each outage. and high pressure systems. Also, severe accident Maintenance and inspection of the RIPs are implemented mitigation systems have been originally added to its design. using handling equipment, which is also installed in the Reactor Maintenance Training Facility next to the NPS site. Table 1 Main specification of BWR-5 and ABWR In the training using the equipment, trainees are intended to Unit 5 Unit 6 acquire RIP’s assembly and disassembly techniques, Reactor type BWR-5 ABWR improving their maintenance and inspection skills. It is Start of commercial operation 1990 1996 possible to disassemble and assemble a RIP for less than 30 Rated thermal output (MWt) 3,293 3,926 hours, respectively. Primary containment vessel Mark II RCCV Rated electrical output (MWe) 1,100 1,356

Thermal efficiency (%) 33.4 34.5 Reactor Reactor P ressure V essel Fuel assemblies 764 872 P ressure V essel Control rods 185 205

Core ave. power density (kW/l) 50.0 50.6 Reactor Internal P u m p Coolant recirculation external internal Reactor C oolant ECCS system div. I LPCS+LPCI RCIC+LPFL R ecirc u lation Pum p div. II LPCI+LPCI HPCF+LPFL C urrent B W R ABWR div. III HPCS HPCF+LPFL Fig. 4 Reactor internal pumps

2/8 2. Core Design 4. Reinforced Concrete Containment Vessel (RCCV) The initial fuel design of 872 fuel assemblies was 8 by 8 In the previous BWR plant design, the reactor primary fuel with 39.5 GWD/t (average burn-up). And currently containment vessel (PCV) has been typically constructed they are replaced by 9 by 9 fuel with 45 GWD/t (average using thick steel. In the RCCV design adopted for the burn-up). In unit 6 and 7, more than half of the core has ABWR, a thin steel liner plate provides leakage protection been already replaced by 9 by 9 fuel (Table 2). As a while two-meter thick reinforced concrete contains the countermeasure against fuel failure by debris, fuel pressure (Fig. 6). assemblies with debris filter in the lower tie plate have been Unlike the freestanding structure of a conventional PCV, adopted since the 5th operation cycle in unit 6 in 2001. the RCCV was constructed simultaneously with reactor Spent fuel has been cooled and stored in the spent fuel building that led to reduce the construction period. In pool equipped with each unit until the transportation to the addition, this integrated configuration offers great flexibility reprocessing plant in Aomori prefecture. In each unit, in structural design because the reactor building and the about 400% capacity of a core can be reserved in the spent RCCV complement each other to withstand design loads. fuel pool, increasing storage capacity and/or narrowing Taking advantage of this flexibility, the plant general space between spent fuel assemblies. arrangement was optimized to reduce the reactor building volume. Table 2 Core design of the ABWRs

Reactor th Building 5 operation cycle unit 6 unit 7 Steel

Containment

8 by 8: 483 497 Vessel

9 by 9: 212 179 Reinforced

70 m Concrete

Containment

9 by 9 with debris filter: 177 196 Vessel 58 m 48 m

Steel Liner 36 m

3. Fine Motion Control Rod Drive (FMCRD)

RIP Newly adopting the FMCRD contributed the reduction 29 m PLR Pump 29 m of plant start-up time with gang operation of control rods 51 × 49 m 59 × 56 m Conventional BWR ABWR (Fig. 5). Also, the use of the FMCRD in the ABWRs enables rod pattern adjustment under full power operation Fig. 6 Reinforced concrete containment vessel with almost no change in core flow while other BWRs needs considerable reactor power reduction on control rod pattern change, especially towards the end of operating 5. Digitalized Main Control Room cycles. TEPCO analyzed operator workloads during the ABWR Maintenance training for replacement of the FMCRD is development phase, in which heavy workloads, such as also possible to be practiced at the Rector Maintenance stabilizing the plant after an unplanned shutdown, were Training Facility. identified and automated. The automated control systems with integrated digital technology and improved man-machine interface of the ABWR-type main control panels, called third generation main control panels, dramatically improve plant operability (Fig. 7). Also, the use of digital control in safety grade systems has never caused safety related incidents so far. The characteristics of the third generation of main control room panels are as follows: (a) the compact main control console that allows operators to supervise and control in their seated position, (b) the large display panel to share information by operators, (c) the touch-operation switches for CRTs and flat Fig. 5 Fine motion control rod drive displays, 3/8 (d) the hierarchic arrangement of alarm displays, Kashiwazaki-Kariwa NPS is located or 500 ships of (e) the comprehensive digitalization of the control systems 200,000 tons oil tanker. Also, we can say that the release including safety systems, and of 430 million tons of carbon dioxide was prevented. As (f) the expansion of the scope of automation of the control of the end of March 2003, we have completed 50 outages at system based on operator workload analysis. Kashiwazaki-Kariwa NPS, 8 of which are implemented at Unit 6 and 7 share one common main control room unit 6 and 7. where main panels face each other. Two units are operated 700,000 99.7 100.0

88.8 by 3 shifts with ten crew members, in which one shift 84.1 85.7 85.7 86.8 86.6 82.4 90.0 84.3 84.0 84.6 600,000 82.6 80.7 80.6 80.4 supervisor is responsible for two units. 78.0 72.9 84.2 80.0 81.7 80.2 80.8 81.3 80.1 80.5 77.1 76.0 75.4 76.6 63.6 500,000 74.2 72.7 73.8 70.0 75.7 71.4 70.0

60.0 400,000 accum. generation ] K-7 h ] '97/7 W generation 50.0 ％ [ [G ave. availability in K.K. 300,000 K-6 ave. availability national '96/11 40.0

K-4 '94/8 200,000 30.0 K-3 K-2, 5 '93/8 20.0 K-1 '90/4,9 '85/9 100,000 10.0

0 0.0 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 [F.Y.]

Fig. 7 Main control room in unit 6 Fig. 8 Electricity output as of the end of March 2003

6. High Efficiency Turbine System Both low-pressure turbine using 52-inch last stage 1. Generation and Availability blades and moisture separator re-heaters have been adopted As of the end of March 2003, an accumulated to improve thermal efficiency of turbine system. Two generation of about 120 billion kWh and an availability of types of heater drain forward pumping systems have been 86% are achieved in unit 6 and 7 (Table 3, Table 4). also used to achieve higher thermal efficiency. Taking into consideration of the electric utility law, which In addition, unit 7 took off its electrical output limit of restricts operating cycle within 13 months at most, the 1,356MWe after its 4th outage in July 2002 in response to theoretical maximum availability is around 90%. Those the report to admit rated thermal power operation by the good performances are also the results of the improvements regulatory authority, Ministry of Economy, Trade and of maintenance works during the total 8 times of refueling Industry (METI), in 2001. As a result, the electrical outages, in which we have been successfully reducing its output could increase to around 1,400 MWe in winter duration. season, which was 3% increase of electrical output. In Table 3 Availability (as of the end of March 2003) Kashiwazaki-Kariwa NPS, unit 2, 5 and 7 have already begun rated thermal power operation. TEPCO plans to F.Y. BWR5 (K-1 to K-5) ABWR (K-6 & K-7) introduce rated thermal power operation to its own 17 1998 85.7% 89.0% BWRs one after another. 1999 88.9% 82.0% 2000 81.7% 83.9% III. Operation Records 2001 81.3% 89.8% In July 2002, the accumulated electricity by 7 units at 2002 57.4% 76.2% Kashiwazaki-Kariwa NPS reached 600 billion kWh, Total 81.6% 85.6% contributing to not only supplying the Tokyo metropolitan area with electricity but also securing energy resources and However, due to the inspection and repair works of reducing greenhouse gas emissions simultaneously (Fig. 8). reactor internals (shrouds) and PLR piping, following the This amount of electricity is equivalent to 200 years scandal of inspection data falsification, availability of all electricity consumption by where 4/8 seven units in the fiscal year 2002 fell to around 60%. 3. Troubles and Malfunctions Fortunately, in unit 6 and 7 no crack indication on their In Kashiwazaki-Kariwa NPS, we have experienced 35 shroud has been found because these shrouds have been troubles, which are defined in the electric utility law, since already taken countermeasures against residual stress. In the start of operation of unit 1 in 1985. addition, they don’t have external PLR piping. Although unit 6 and 7 keep good operation performance, a ratio of troubles happened in the ABWRs is slightly Table 4 Operation cycle and outage duration of the ABWRs higher than those of the national average (Fig. 10). We Unit 6 unit 7 had 6 times of unplanned shutdown in unit 6 and 7, all of (1st ope. cycle) (378) (329) which were classified as INES 0 or less (Table 5). Since 1st outage 61 55 the causes of the shutdowns were almost due to (2nd ope. cycle) (418) (425) conventional problems, such as the failure of electrical 2nd outage 44 45 instrument, plant auxiliary system, or fuel failure, they (3rd ope. cycle) (414) (419) suggest there are no ABWR specific problems. Of the 6 3rd outage 53 50 unplanned plant shutdowns, two cases were automatic plant (4th ope. cycle) (421) (422) shutdowns due to the actuation of 500 kV pilot wire 4th outage 66 111 protective relay and the generator exciter system failure. (5th ope. cycle) (421) (in operation) The frequencies of unplanned shutdown in the ABWRs are As of March 31, 2003 5th outage planned shutdown not as high as other first-of-a-kind plants in TEPCO.

1.2 2. Reduction of Start-up Time national ave. BWR-5 in KK The gang operation of FMCRDs and automated plant 1 ABWR operation reduce plant start-up time. In design, it takes 0.8 ] t i n u about 25 hours from the start of control rod withdrawal to / s 0.6 e l b

reach the rated output compared to start-up of a u 0.4 conventional BWR, which needs about 40 hours. From [tro the operating practice, it takes about 1 day and 20 hours 0.2 from the start of control rod withdrawal to generator 0 synchronization (Fig. 9). Particularly, from the start of CR 90 91 92 93 94 95 96 97 98 99 00 01 02 [F.Y.] withdrawal to criticality is further reduced to about 30 minutes, using the gang operation of the FMCRD. The time from generator synchronization through rated power Fig. 10 Transition of plant troubles output will diverse such as from 1 to 4 days, according to core design. Table 5 Number of unplanned shutdown in unit 6 and 7

K2 9th Manual Automatic Unit 6 4* 2

K7 4th Unit 7 3 0 (* inlc. period during the trial operation)

K6 1st Unit 6 Feb, ’96 Manual shutdown INES: 0- 0d0h 0d12h 1d0h 1d12h 2d0h 2d12h Due to the trip of the RIPs (during the trial operation) CR withdrawal to Criticality Aug, ‘96 Manual shutdown INES: 0- Criticality to Rated pressure Due to the increasing value of I-131 in the primary coolant Rated press. to Tb start (D/W inspection, etc.) Tb start to Gen synchro. system (during the trial operation) Aug, ‘98 Automatic scram INES: 0+ Fig. 9 Plant start-up time Due to the actuating 500 kV pilot wire protective relay

5/8 May, ’99 Automatic scram INES: 0- relatively flat-shaped bottom shell of the RPV were Due to the generator exciter system failure suspected. However, the analysis was not able to May, ’00 Manual shutdown INES: 0- confidently support these premises. Reload fuel Due to the increasing value of I-131 in the primary coolant assemblies with debris filter are introduced and expected to system eliminate this concern. June, ‘01 Manual shutdown INES: 0- Due to the leakage from the reactor auxiliary cooling 4. Outages system in the primary containment vessel TEPCO also sought better maintainability in the design of the ABWR. For example, the FMCRD main body is Unit 7 designed to be maintenance-free and the gland packing, the May, ’97 Manual shutdown Out of INES only part requiring periodic replacement, is located in a Due to the abnormal sounds from the low-pressure turbine separated small housing attached to the lowest end of the (B) during the trial operation main body. Mar, ‘99 Manual shutdown INES: 0- The 1st outage of unit 7 was conducted in 55 days, Due to the increasing values in the off-gas radiation monitor shorter than previously attained at 1st outage of BWR-5s in July, ‘99 Manual shutdown INES: 0- Kashiwazaki-Kariwa NPS. The 55 days was determined Due to the trip of one of the RIPs by the regulation that requires full-section turbine inspections in the 1st outage. We are trying to reduce the (1) RIP related incidents outage duration, in which maintenance of the reactor In July 28, 1999, during the normal operation of unit 7, facility will determine the critical time schedule. an unplanned shutdown occurred due to the failure of a RIP 100 power supply cable terminal by vibration. Although 90 BWR5 in KK ave. continued operation was possible since the ABWR can 80 ABWR ave. operate with one of their ten pumps out of service (and can 70 60 ] s y operate at reduced power with up to three pumps out of a 50 [d service,) the reactor was shutdown to investigate the root 40 30 cause. The design of the cable terminal was modified 20 immediately. 10 0 In February 23, 1996, the failure of the RIP power 1st 2nd 3rd 4th supply controller happened during pre-operation of unit 6. Again, continued operation was possible by re-starting the Fig. 11 Outage duration of 1st through 4th outages tripped RIP and utilizing the redundant controller, the plant had been manual shutdown for 19 days to investigate the In order to operate nuclear power plants as long as root cause. possible in all kinds of power plants, we make outage plan taking into account for the total days, for example about (2) Fuel failure 200 days in four consecutive outages for the ABWR. The We have experienced 7 fuel failures, 3 of which were duration of outages is reduced in the ABWR compared to subject to post-irradiation examination by which the cause early stage in the BWR-5 (Fig. 11). of failure was found to debris-induced fretting corrosion. At present, we set a standard work schedule of outages Because iodine levels in the reactor coolant had been under for the BWR-5 as well as the ABWR as 45 days (Table 6). the limits of the operation technical specifications, The shortest outage in TEPCO was achieved by 32 days in operation had been continued using power suppression the 9th outage of Kashiwazaki-Kariwa unit 2 in 2002. On techniques in two cases. average, it takes 72 days for unit 1 through unit 5, and 61 Since the population of fuel failures is higher than that days for unit 6 and 7 so far. of the other TEPCO’s reactors, an analysis has been In ordinary, plant equipment is planned to make executed to see if there is any design-specific cause maintenance in time-based maintenance (TBM). The including debris failure. Drain-forward system design in maintenance cycle is extended longer in the ABWR than in the balance of plant, no annular region in the RPV and the BWR-5, such as control rod drives (Table 7). Also, in 6/8 the ABWR, the maintenance interval of the RIPs was level ever achieved in TEPCO. extended from 5 to 10 outages since the 3rd outage of unit 7. Improvements in reactor and piping materials and appropriate work space to maintain sufficient distance from Table 6 Standard work schedule of outages major radiation sources contribute to the reduction in BWR-5 K2-9th ABWR K6-2nd radiation exposure. This is attributable to no primary loop Rx head off 5.0 4.0 4.5 5.0 recirculation piping and consequentially less ISI and a Fuel unloading 3.0 2.0 2.5 1.0 better working environment inside the containment. The RIP disassembly 2.0 2.0 reduction of maintenance requirements for reactor LPRM replacement, 2.0 2.0 component, such as the FMCRD, is also the reason to CRD inspection, 4.0 6.0 reduce radiation exposure. and CR replacement 11.0 10.0 The inspection of submerged valves will be taken place Fuel loading, 7.0 6.0 7.0 in every 4 outages. In addition, we investigated the cause th and RIP assembly 7.0 2.0 of a small scratch on a RIP impeller in the 4 outage of unit th Core verification 2.0 1.0 2.0 1.0 7. Thus, the radiation exposure at the 4 outage of the Rx vessel restoration 5.0 4.0 5.0 6.0 ABWRs is relatively high (Fig. 13). PRV L/T 1.0 1.0 1.0 1.0 9,000 9 PCV restoration 3.0 1.0 3.0 3.0 8,000 8 7,000 7 n] PCV L/T 2.0 2.0 2.0 2.0 6,000 6 ma v/

Pre-ope. test 3.0 2.0 3.0 3.0 ] 5,000 5 n mS

a r o [m

4,000 4 ,

System lineup 1.0 1.0 1.0 1.0 v S n

3,000 3 a

Startup 2.0 2.0 2.0 2.0 m 2,000 2 [ Total (days) 45.0 32.0 45.0 44.0 1,000 1 0 0 92 93 94 95 96 97 98 99 00 01 02 [F.Y]

Table 7 Comparison of maintenance cycle num of employees in KK total dose in KK [man Sv] ave. dose in KK [mSv/man] ave. dose national [mSv/man]

BWR-5 ABWR Fig. 12 Transition of radiation exposure PLR pump, RIP all (2) pumps all (10) pumps by 10 o/a by 10 o/a CRD, FMCRD all (185) units 25% 1.4 0.0200 BWR5 in KK ave.[manSv] 0.0180 by 7 o/a by 10 o/a 1.2 ABWR ave.[manSv] BWR5 in KK ave.[mSv/(man day)] 0.0160 Spool piece N.A. all (205) pieces 1 ABWR ave.[mSv/(man day)]

0.0140 ] y) a

] 0.0120

by 10 o/a v 0.8 n d S a n 0.0100 a m ( (o/a: outages) / [m 0.6 v

0.0080 S m 0.4 0.0060 [ 0.0040 IV. Radiation and Radioactive Wastes 0.2 The record of radiation exposure has been continued 0.0020 0 0.0000 low in Kashiwazaki-Kariwa NPS compared to national 1st 2nd 3rd 4th average (Fig. 12). One of the reasons comes from the introduction of the ABWRs, which have improved work environments, such as spacious RCCV, automated Fig. 13 Radiation exposure during outages equipment for the replacement of the FMCRDs and the maintenance of the RIPs. 2. Reduction of Radioactive Wastes 1. Dose during the Outages The volumes of low-level radioactive wastes during the The first outage of unit 7 in 1998 recorded a total plant operation as well as outages have been decreasing as radiation exposure of 0.153 manSv, which is the lowest expected in the ABWRs (Fig. 14, Fig. 15).

7/8 In order to reduce the low-level radioactive wastes, requirements, hollow fiber filters for condensate filtering system, (c) small radiation exposure and less radioactive wastes, non-regeneration use of condensate demineralizer resins and and incineration processing of combustible solid materials (d) high plant thermal efficiency. and spent resin have been adopted. The radioactivity on the surface of drums is less than 0.05 mSv for most drums VI. Other Achievements in storage. Other achievements are obtained since the introduction of the ABWRs: 160 12,000 BWR5 in KK ave. ABWR ave. (a) We have a bilateral agreement with Chooz-B NPP of 140 drums in storage 10,000 EDF in France to exchange technical information, 120 ]

W 8,000 especially in the field of digital I&C. So far, 3 times e] M 100 g a 0 r 0 o 0 t

, of technical exchange meetings took place in France s

1 n

r 80 6,000 i e s p m and Japan. s ru m 60 [d ru 4,000

[d (b) More than 100 thousand visitors are coming to 40 2,000 Kashiwazaki-Kariwa NPS every year. Many of them 20 make a field tour to unit 6 and 7. 0 0 1997 1998 1999 2000 2001 2002 [F. Y.] VII. Conclusion

We have accumulated more than total 12 years of ABWR operation experience including 8 times of outages Fig. 14 Number of drums per 1,000MW since the start of commercial operation of unit 6. From these experiences, we have obtained following conclusions in this paper: 7.0 (a) Integrity of newly developed equipment was 6.0 confirmed. 5.0

] (b) Enhanced safety and reliability were confirmed. y

a 4.0 d / s (c) Reduced occupational radiation exposure and m u r 3.0 [d radioactive wastes were achieved. 2.0 (d) Excellent operability and maneuverability were 1.0 achieved. 0.0 (e) Improved plant economy was confirmed. 1u 2u 3u 4u 5u 6u 7u Thus, we will make every effort to continue excellent operation and we hope that the operation experience Fig. 15 Drums per day from an outage through the obtained at Kashiwazaki-Kariwa unit 6 and 7 is used for following operation cycle in the recent 4 cycles future development of the ABWR.

VIII. References V. Improved Economy 1) A. Omoto, “Ten reactor years of ABWR experience,” In the ABWR, economy had been already improved in Nuclear Engineering International, May 2002, construction stage by the reduction in building volumes and 20(2002). construction period compared to the BWR-5 by about 20%. 2) TEPCO Kashiwazaki-Kariwa NPS, “Annual Report Also, in operation and outage stages, O&M costs will be 2001,” (2002). further improved by the following reasons: 3) Tokyo Electric Power Co., “Advanced Boiling Water (a) high availability by decreasing unplanned shutdowns Reactor,” (2000). and reducing plant startup time, 4) Tokyo Electric Power Co., “Development of Advanced (b) short outage duration by less maintenance Boiling Water Reactor,” (1999).

8/8