This chapter is intended to provide non-expert readers with basic information about nuclear power plant1 design, operation, and regulation in Japan and the United States. This information will be useful for understanding the technical discussions in subsequent report chapters. Expert readers may wish to skip ahead to Chapter 3.
This chapter is organized into five sections.
• Section 2.1 provides an overview of nuclear plant design and operation.
• Section 2.2 describes the design of boiling water reactors (BWRs) and their safety systems. (The reactors at the Fukushima Daiichi nuclear plant were BWRs.)
• Section 2.5 describes some key differences in BWR designs in Japan and the United States.
1 The terms “nuclear power plant” and “nuclear plant” are used interchangeably in this report.
Nuclear plants are used in the United States and many other countries primarily to meet baseload2 demands for electricity. These plants are especially well suited for this purpose because they can be operated for long periods without maintenance outages and can produce electricity at constant rates.
Nuclear plants generate electricity using the Rankine thermal cycle: the plant’s nuclear reactors produce heat that is used to convert water to steam. The steam drives a turbine that spins a generator to produce electricity. After passing through the turbine, the steam is cooled, condensed, and recirculated. This “steam engine” cycle is also used to produce electricity in other types of thermal power plants, particularly coal- and gas-fired plants.
The primary fuel for nuclear plants is slightly enriched uranium,3 usually in the form of 1-cm-long cylindrical uranium dioxide pellets. These pellets are encased in metal tubes, referred to as nuclear fuel cladding, each ~10 mm in diameter and about 4 m in length, made of various zirconium alloys (Zircaloy, ZIRLO, M5) containing 98 percent or more zirconium. The cladding provides structural support for the fuel pellets, serves as a barrier to the release of radioactive material from the fuel, and provides an efficient geometry for cooling. The ensemble of pellets and cladding is referred to as a nuclear fuel rod. Fuel rods are grouped into bundles, or fuel assemblies (Figure 2.1 [top]), each containing between about 64 and 300 rods.
The entire set of fuel assemblies, along with control rods and associated structural supports (Figure 2.1 [bottom]), constitutes the reactor core. The control rods contain materials that are highly neutron absorbing, such as hafnium, boron, or silver. The control rods can be used to shut down the reactor when fully inserted.
The reactor core is enclosed in a robust steel pressure vessel, the reactor pressure vessel (RPV) (Figure 2.2). The RPV contains numerous penetrations for steam and water lines, instrumentation, and controls. The robust RPV design allows the reactor to operate at high temperature and pressure to increase its thermal efficiency.4 It also provides a major barrier to the release of radioactive material from the reactor during an accident. Water
2 That is, the continuous demand for electricity from customers in regions served by the plant. Countries such as France have such a high percentage of nuclear power that the output of some of their plants is varied according to demand.
3 Enriched uranium contains uranium-235 in higher-than-natural abundances. Natural uranium contains about 0.7 percent uranium-235. Uranium used in most reactors contains 3-5 percent uranium-235.
4 BWRs operate at pressures and temperatures of about 7 MPa and 285ºC. Pressurized water reactors (PWRs) operate at pressures and temperatures of about 15 MPa and 315ºC. See Footnote 15 for a definition of MPa.
FIGURE 2.1 (Top) Side view of a BWR fuel assembly. Some fuel rods have been removed to reveal construction details. (Bottom) Cross-section showing four BWR fuel assemblies and a control rod. The control rod consists of four blades in the shape of a cross. It can be seen in cross section (red cross) in the figure. The control rods are moved into and out of the reactor core to control its power. SOURCE: ANS (2012, Fig. 4).
FIGURE 2.2 Schematic of a BWR/5 or BWR/6 reactor pressure vessel. SOURCE: ANS (2012, Fig. 5).
circulation through the RPV is used to control reactor pressure and temperature and generate steam for electricity production. The movement of water and/or steam out of the RPV is pressure-driven and is controlled by opening and closing valves.
The RPV is located within the containment of the building that houses the reactor (Figure 2.3). The containment is constructed of reinforced concrete a meter in thickness or a carbon steel shell a few centimeters thick and contains heavy metal bulkheads to allow access for maintenance work. Like the RPV, the containment also has numerous penetrations for steam and water lines, instrumentation, and controls. The containment serves as a barrier to the release of radioactive material to the environment during a severe accident (Sidebar 2.1).
Reactor power is regulated by manipulating the positions of the control rods in the reactor core.5 The reactor can be “started” by partially withdrawing the control rods from the core. This allows a sustained nuclear fission chain reaction to be initiated in the uranium fuel, which generates large quantities of heat. This heat is removed by the constant circulation of cooling water through the core. As noted previously, the reactor can be shut down by fully inserting the control rods into the core. A reactor is said to be scrammed when all of the reactor’s control rods are fully inserted and the fission process is halted after an off-normal condition is detected. Shutdown may occur automatically or can be initiated by reactor operators.
The operation of a reactor produces a wide range of radioactive isotopes:
• Fission of the uranium fuel results in the production of dozens of highly radioactive fission products, for example, cesium-137, iodine-131, and strontium-90. Some of these fission products, notably cesium and iodine, are volatile.
• Absorption of neutrons by materials in the reactor core produces transuranic elements such as plutonium-239 as well as neutron activation products such as cobalt-60.
These isotopes continue to decay and generate heat (referred to as decay heat) even after the reactor is shut down. Decay-heat generation immediately following reactor shutdown can be up to about 6 percent of the reactor’s licensed power. Heat generation decreases rapidly as short-lived isotopes (primarily fission products) decay (see Figure 2.4). Cooling is crucial in the first few days after the reactor is shut down and continues
5 Other means are used to regulate reactor power as well. The power in BWRs can be regulated by varying water flow through the core. The power in PWRs can be regulated by varying the concentration of boron, a neutron absorber, in reactor cooling water.
FIGURE 2.3 Simplified schematics for nuclear power reactors. (Top) Boiling water reactor. (Bottom) Pressurized water reactor. SOURCE: ANS (2012, Fig. 2 [Top] and Fig. 3 [Bottom]).
to be required for years (but at lower levels) to remove the heat generated from the decay of long-lived fission products. Reactor cooling systems are designed to remove this heat so as to prevent excessive temperature rise.
Reactor cooling is provided by several safety systems. Some safety systems operate during normal conditions to maintain RPV pressures and temperatures and water levels within a set range. Other safety systems are part of the emergency core cooling system (ECCS). These systems are used to cool the core during off-normal conditions. The effectiveness of these systems depends on their ability to remove decay heat through a combination of heating and boiling of water in the reactor while maintaining the water level in the RPV above the top of the reactor core.
In the United States, nuclear fuel in a reactor must be replaced every 4 to 6 years depending on the reactor’s design and operation. U.S. reactors are typically shut down every 18-24 months6 for replacing a portion of the reactor fuel. The used (or spent) fuel is transferred from the reactor to a spent fuel pool. The pool has its own cooling system (the pool water is circulated through a heat exchanger) to remove decay heat from the fuel.
Nuclear plants can contain one or more reactors and their support systems, including water and electrical supplies, mechanical systems, and spent fuel pools. All nuclear plant sites in Japan and most in the United States have multiple reactors (see Sections 2.3 and 2.4 in this chapter).
More than 400 nuclear power reactors7 are currently operating throughout the world and 70 more are currently under construction. The large majority of nuclear power plants in the world and all plants in the United States and Japan are light-water reactors; these reactors are cooled and moderated by regular water.8 Two types of light-water reactors have been deployed worldwide for electricity production, including in Japan and the United States: BWRs and pressurized water reactors (PWRs). The design of these reactors is illustrated in Figure 2.3.
The primary difference between BWRs and PWRs is the mechanism for generating steam to produce electricity. BWRs produce steam directly in the core; that steam is separated, dried, and used to drive turbines and the electrical generators connected to them. PWRs produce high-temperature water that is circulated through a heat exchanger (referred to as the steam generator) to produce steam in a secondary water circulation loop. The steam in this secondary loop drives the turbines and their associated electrical generators. In both plant designs, the steam is condensed to water after
6 Japanese and some European reactors are shut down every 12 months for refueling.
7 Prior to the Fukushima nuclear accident, there were 442 operating power reactors in 30 countries (World Energy Council, 2012).
8 Moderation refers to the slowing down of fission neutrons to thermal energies to increase their nuclear fission cross-section. The CANDU (Canadian Deuterium Uranium) reactor uses heavy water as moderator and accounts for about 10 percent of reactors worldwide.
Nuclear reactors generate radioactive by-products, primarily fission products (e.g., iodine and cesium) and transuranic isotopes (e.g., plutonium), that build up inside the fuel and fuel rods during the course of operations. This radioactive material is contained completely within the fuel rods under normal operating conditions. In a severe reactor accident, however, the zirconium fuel rods can oxidize and rupture, the uranium fuel can overheat and melt, and radionuclides can be released into the reactor pressure vessel. If the reactor’s containment fails, radionuclides can be released into the reactor building and possibly into the environment. This is exactly the scenario that occurred in Units 1, 2, and 3 at the Fukushima Daiichi plant following the March 11, 2011, earthquake and tsunami.
Fission product releases, which normally constitute the largest fraction of radioactivity released during a severe accident, are usually in the form of gases (xenon and krypton), which are released when fuel rods rupture, and aerosols (iodinea and cesium), which are formed by condensation after vaporizing from hot fuel. Other fission products (strontium) and associated radioactive materials (uranium, plutonium) have very high vaporization temperatures and are largely retained in the reactor fuel, even when molten. Release of iodine-131 in any form—aerosol, molecular, or organic compound—is of particular concern because of its high activity (it has an 8-day half-life) and its ability to concentrate in the human thyroid gland if ingested. Children are particularly at risk of developing thyroid cancer as a result of exposure to radioactive iodine.
Once released to the atmosphere, cesium and iodine are transported by prevailing winds and can travel for considerable distances before wet or dry deposi-
passing through the turbines and the condensed water is recirculated. The water used to condense the steam is taken from a nearby ocean, river, or other water supply.
The reactors at the Fukushima Daiichi plant are BWRs. Consequently, the discussion in the remainder of this chapter focuses primarily on the design and operation of this reactor type.
BWRs were initially developed by General Electric Company during the 1950s and have evolved through “generations,” with each generation representing iterative evolutions in the design of steam systems, water recirculation systems, safety systems, and containments. In the United States, BWR containments are designated Mark I, Mark II, and Mark III (the oldest to most recent designs). These designs are illustrated schematically in
tion brings them to the ground or the surface of a water body such at a river, lake, or ocean. Prevailing winds at the time of the Fukushima Daiichi accident appear to have transported most of the radioactivity released from the damaged reactors out to the Pacific Ocean (Kawamura et al., 2011; Morino et al., 2011). However, sufficient quantities were also transported and deposited inland to contaminate large land areas (see Figure 1.4 in Chapter 1) to levels requiring long-term human use restrictions. The cesium isotopes Cs-134 (2-year half-life) and, more importantly, Cs-137 (30-year half-life) are the most important sources of long-term contamination. These isotopes have also contaminated the Fukushima Daiichi reactor buildings and will impede efforts to remove damaged fuel from the Unit 1, 2, and 3 reactors.
The most important function of the reactor containment during a severe accident is to prevent the release of iodine and cesium aerosols. If containment has to be deliberately vented to prevent excess pressure (see Sidebar 2.2), vented gases can be filtered through sand or water, if such filters are available, to reduce the quantity of aerosols that are released into the environment.
Venting BWR containments through the suppression chamber is preferred because the vent gases can be passed through the suppression pool to scrub out aerosols. The reactor building, which serves as a secondary containment, can also be used to reduce aerosol releases when the containment is bypassed or develops a leak as happened at the Fukushima Daiichi plant (see Chapter 4). However, hydrogen explosions in the Fukushima Daiichi reactor buildings reduced their effectiveness in filtering out radioactive aerosols.
a Iodine can also exist in other forms as well, for example, as elemental or organic gases (e.g., Gavrilin et al., 2004).
Figure 2.5. Additionally, there are six reactor generations denoted BWR/1,9 BWR/2, BWR/3, BWR/4, BWR/5, and BWR/6. The design of reactor cooling systems (see Section 2.2.3) has also evolved in these BWR reactor generations.
Newer BWR designs, the advanced boiling water reactor (ABWR) and economic simplified boiling water reactor (ESBWR), have been developed by General Electric. The ABWR design has been approved by the U.S. Nuclear Regulatory Commission (USNRC), and several ABWRs have been constructed in Japan (see Section 2.3). The ESBWR design has been submitted to the USNRC for approval and its review is nearing completion. The discussion in the remainder of this section focuses on first-generation BWRs (BWR/1-BWR/6).
Unit 1 at the Fukushima Daiichi plant is a BWR/3 with a Mark I containment: Units 2-5 are BWR/4 with Mark I containments; and Unit 6 is a
9 No first-generation (BWR/1) reactors are operating today.
FIGURE 2.4 Estimated thermal power output of reactor cores at the Fukushima Daiichi plant following shutdown. SOURCE: Based on methodology used in Gauntt et al. (2012b) and Phillips et al. (2012).
The discussion of containment systems below focuses on the Mark I containment because of its relevance to the Fukushima Daiichi accident.
2.2.1 Containment System
The Mark I containment comprises the structure, referred to as the drywell, that houses the RPV. The drywell is connected to a water-filled chamber, referred to as the suppression chamber.10 The water pool in this
10 Also sometimes referred to as the wetwell and torus.
FIGURE 2.5 BWR containment designs. The location of the spent fuel pool is not shown in the Mark II containment. SOURCE: ANS (2012, Fig. 16).
chamber is referred to as the suppression pool and is designed to condense steam that is released from the RPV if it becomes overpressured. The pool is also used to remove (i.e., scrub) fission products in the vented gases when the reactor fuel is damaged. The RPV can be depressurized by opening safety relief valves (SRVs).
The suppression chamber can be cooled using various systems to maintain it within design pressures and temperatures. If cooling is lost, the suppression chamber can be vented to the atmosphere to reduce pressures and temperatures. The suppression pool water can be used to filter out radioactive material before venting (Sidebar 2.2).
The spent fuel pool resides outside of containment but inside the reactor building (Figure 2.5). It is located near the top of the drywell to allow fuel unloading to be performed under water. This requires that the spent fuel pool be elevated above ground level.
2.2.2 Pressure Control System
During normal BWR operations, steam produced in the reactor exits the RPV and flows to the main turbine. If the reactor is shut down, the main steam isolation valves (MSIVs) are closed, isolating the RPV from the power conversion system. Depending on the nature of the shutdown and associated operating procedures, the RPV can be depressurized by opening the SRVs; this allows steam to flow from the RPV into the suppression pool where it is condensed. (This cooling pathway is shown in the Sidebar 2.2 figure.) Depressurization is required before operators can activate the
TABLE 2.1 BWR Reactor Designs
|Reactor Function||Reactor Type and Fukushima Daiichi Unit|
|BWR/3, Unit 1||BWR/4, Units 2, 3, 4, 5||BWR/5, Unit 6|
|Reactor isolation pressure control||Isolation condenser and SRVs||All use SRVs; some have steam condensing mode of RHR||All use SRVs; some have steam condensing mode of RHR|
|Reactor isolation inventory control||Isolation condenser||RCIC||RCIC|
|ECCS high-pressure pumping||HPCI||HPCI||HPCS|
|ECCS high-pressure pump type||Turbine-driven HPCI||Turbine-driven||Motor-driven|
|ECCS low-pressure flooding delivery point||Recirculation pump discharge pipe||Recirculation pump discharge pipe or inside shroud (core region)||Inside core shroud, core region|
|Containment type||Mark I||Mark I||Mark II|
NOTES: The reactor designs included in this table are most pertinent to the Fukushima Daiichi and Daini plants. The table does not include advanced BWR designs such as the Advanced Boiling Water Reactor (ABWR) or Economic Simplified Boiling Water Reactor (ESBWR). The table shows general design features and may not be applicable to every BWR reactor. ECCS = emergency core cooling system, HPCI = high-pressure coolant injection, HPCS = high-pressure core spray, RFP = reactor feedwater pump, RCIC = reactor core isolation cooling, and SRV = safety relief valve.
SOURCE: Information taken from Boiling Water Reactor GE BWR/4 Technology Advanced Manual (available at http://pbadupws.nrc.gov/docs/ML0230/ML023010606.pdf).
low-pressure cooling systems to cool the reactor (see Section 126.96.36.199 in this chapter for a discussion of these cooling systems).
The SRVs will also automatically actuate through a purely mechanical function when pressures exceed preset values. This is a passive safety feature designed to protect the RPV from excessive pressures if operators are unable to actuate the SRVs.
The containment can be vented from the suppression chamber (see Sidebar 2.2) or drywell. This venting capability was enhanced (i.e., hardened) for BWR Mark I systems in the United States following the 1979 Three Mile Island nuclear accident11; hardened vents were also installed in
11 The reactors at the Three Mile Island plant are PWRs.
BWR Mark I reactors in Japan (see Section 2.5.2). This provided operators with a means to control containment pressures if they became elevated due to accident conditions. The enhancement was to typically install piping instead of sheet metal ducting as the pressure relief pathway. However, this enhancement was not made at all BWR plants.
TABLE 2.2 Mark I-BWRs in the United States That Are Similar to Units at the Fukushima Daiichi Plant
|Plant Name||Location (State)|
BWR/2, Mark I with IC
|Nine Mile Point 1||New York|
|Oyster Creek||New Jersey|
BWR/3, Mark I with IC (similar to Fukushima Daiichi Unit 1)
BWR/3, Mark I with RCIC
|Quad Cities 1||Illinois|
|Quad Cities 2||Illinois|
BWR/4, Mark I (similar to Fukushima Daiichi Units 2-4)
|Browns Ferry 1||Alabama|
|Browns Ferry 2||Alabama|
|Browns Ferry 3||Alabama|
|Brunswick 1||North Carolina|
|Brunswick 2||North Carolina|
|Hope Creek 1||New Jersey|
|Peach Bottom 2||Pennsylvania|
|Peach Bottom 3||Pennsylvania|
Containment venting requires manual operator action using emergency operating procedures. In U.S. nuclear plants the venting path is established through piping from above the suppression chamber that passes through the reactor building and exhausts into the atmosphere.
Venting involves the controlled release of gases from the containment of a nuclear plant to the environment in emergency situations, for example, after the failure of a reactor’s emergency core cooling system. In such situations, steam buildup in the containment can raise temperatures and pressures above design levels which, if unvented, could result in containment leakage or failure and uncontrolled releases of hydrogen and radioactive material into the reactor building and from there into the environment. Venting reduces containment pressures and temperatures; this reduces the potential for uncontrolled containment leakage and enables the reactor pressure vessel to be depressurized so that alternative means can be deployed to cool the core, for example, the injection of low-pressure cooling water from diesel-driven fire pumps or from fire trucks.
To vent a BWR reactor, operators must open motor-operated and air-operated valves (Figure S2.1). Motor-operated valves are typically opened (or “lined up”) using either AC or DC power; they can also be opened manually if operators can physically access them. Air-operated valves can be opened using compressed air and DC power. Once valve lineup is complete and containment pressure is high enough, a rupture disk in the vent line (if present) will activate and containment gases will be vented through the plant’s exhaust stack. These vented materials can contain radioactive materials (radioactive gases and fine particulate materials) and hydrogen from fuel cladding–steam reactions (see Sidebar 4.1 in Chapter 4).
In BWRs, gases can be vented through the suppression pool prior to release to “scrub out” some of their radioactive constituents. (In PWRs, gases can be vented into containment and scrubbed using water sprays.) Scrubbing is not 100 percent effective in removing radioactive constituents from the vented gases, however. Consequently, the venting of a reactor with damaged fuel would likely result in the release of some radioactive materials into the environment (e.g., noble gases). Decisions on venting and appropriate protective actions for the public need to balance the benefits of maintaining the integrity of containment to prevent large-scale radioactive releases with the consequences of immediate but smaller releases.
At present, no U.S. reactors have filtered vents, but the USNRC has initiated a rulemaking process to determine whether such vents should be required for BWR Mark I and Mark II reactors. Eighty non-BWR nuclear plants in Western Europe have filtered vents. Eighteen non-BWR reactors in Canada have filtered vents or have committed to installing them. Only 13 BWRs in the world have filtered vents (Borchardt, 2012a, Enclosure 3). Only a few nuclear plants in Japan have filtered vents but all have now committed to install them.
FIGURE S2.1 Simplified illustration of the containment venting system for Units 1-3 at the Fukushima Daiichi plant. The SRVs are opened to depressurize the RPV into the suppression pool in the suppression chamber (green line). The suppression chamber is vented to the exhaust stack by opening two valves and activating the rupture disk, if present (red line). NOTE: D/W = drywell, PCV = primary containment vessel, RPV = reactor pressure vessel, S/C = suppression chamber, SRV = safety relief valve. SOURCE: Courtesy of TEPCO (Available at http://www.tepco.co.jp/en/nu/fukushima-np/review/review1_2-e.html. Accessed on June 3, 2014.)
2.2.3 Core Cooling System
BWRs have various engineered safety features to cool their cores depending on their generation. Three systems played key roles in the Fukushima nuclear accident (see Chapter 4):
• Isolation condenser (IC) system: Used in BWR/2s and BWR/3s, including Unit 1 at the Fukushima Daiichi plant.
• Reactor core isolation cooling (RCIC) system: Used in BWR/4s, including Units 2-4 at the Fukushima Daiichi plant, BWR/5s, BWR/6s, and the Advanced Boiling Water Reactor.
• High-pressure coolant injection (HPCI) system: Used in BWR/3s and BWR/4s, including Units 2-4 at the Fukushima Daiichi plant.
These systems are designed to remove decay heat from the reactor in the absence of AC power. They require DC power for control purposes but in some situations can operate for extended periods without any power. These systems are described in subsequent sections of this chapter. More complete descriptions can be found in technical information documents such as the Reactor Concepts Training Manual (USNRC, 2012b).
AC power is required to operate other safety systems. These include the core spray, residual heat removal (RHR), and containment spray systems. Containment and suppression pool spray systems also can be powered by the diesel-driven fire protection system or emergency water sources. These systems played little or no role in the Fukushima nuclear accident and so they are described only briefly in the next section.
188.8.131.52 Low-Pressure Core Cooling Systems
Low-pressure core cooling systems comprise two separate and independent systems: the core spray system and the low-pressure coolant injection (LPCI) system of the RHR system (Figure 2.6). These systems require AC power to operate pumps, controls, and valves.
The core spray system pumps water from the suppression pool into the RPV (to remove decay heat) using two separate and independent pumping loops. The core spray system sprays water from above the core onto the tops of the fuel assemblies. Water is supplied by AC-powered, high-volumetric flow pumps. The core spray system and the LPCI mode of the RHR system operate only when the RPV is at low pressure.
The RHR system is a multipurpose system that uses AC-powered, high-volumetric flow pumps in different configurations to supply plant needs. The RHR system is normally aligned in the LPCI configuration to supply water makeup to the RPV for core cooling under loss-of-coolant condi-
tions.12 During LPCI operation, RHR pumps take water from the suppression pool and discharge it into the RPV directly or after flowing through a heat exchanger that transfers heat to the ultimate heat sink.
184.108.40.206 Isolation Condenser System (Fukushima Daiichi Unit 1 Reactor)
The IC system (Figure 2.7) is used to remove decay heat and conserve reactor water inventory when the RPV becomes isolated from the power conversion system (i.e., the turbine and condenser; see Figure 2.3) at or near operating pressures. It has two trains of equipment (labeled “Train A” and “Train B” in Figure 2.7), each consisting of a large heat exchanger and associated piping. The secondary (shell) side of the heat exchanger, basically a large tank, contains enough water to remove decay heat from the RPV for several hours. The shell-side water can be replenished using the makeup-water or fire protection systems or fire trucks.
The system can operate without electrical power or operator intervention as long as the system valves are open and there is water in the shell side of the heat exchanger. The system operates by gravity flow: Steam enters the heat exchanger via a steam line from the RPV, and condensate is returned to the RPV through a recirculation pump line.
As shown in Figure 2.7, there are four valves for each IC train. The two valves outside of containment are operated by DC power from batteries; the two valves inside containment are operated by AC power. If DC power is lost, a separate DC-powered interlocking logic circuit causes all four valves in each train to close, effectively shutting down the IC system. Once closed, the valves inside containment cannot be reopened unless AC power is available.13 This system logic affected the operation of the valves for the IC in Unit 1 of the Fukushima Daiichi plant during the accident (see Chapter 4).
220.127.116.11 Reactor Core Isolation Cooling System (Fukushima Daiichi Unit 2 and 3 Reactors)
The RCIC system (Figure 2.8) is designed to make up water inventory losses from the RPV caused by water boil-off when the RPV is isolated from the turbine and condenser. It is designed to operate independently of auxiliary AC power, service air, or external cooling-water systems and can provide adequate makeup water to the RPV in the following circumstances:
12 For example, when water levels in the RPV drop below acceptable levels because of a pipe break or other off-normal condition.
13 There are several possible sources of AC power: offsite AC power, onsite emergency diesel generators, and onsite DC sources via inverters.
FIGURE 2.6 Schematic of the (A) core spray system for Unit 1 and (B) residual heat removal (RHR) system for Units 2 and 3 at the Fukushima Daiichi plant. Motor-operated (MO) valves are indicated by connected triangles. SOURCE: Government of Japan (2011b, Figs. IV-2-1 and IV-2-9).
FIGURE 2.7 Schematic of the isolation condenser (IC) system for Unit 1 at the Fukushima Daiichi plant. Motor-operated (MO) valves are indicated by connected triangles. Black indicates valve closed during normal operations; white indicates valve open during normal operations. Power sources to operate the valves (AC or DC power) are indicated. SOURCE: ANS (2012, Appendix F, Fig. 1).
• RPV is isolated from the power conversion system (turbine and condenser) and is being maintained at operational pressures and temperatures.
• Reactor is shut down and at high pressure14 with loss of normal feedwater.
• AC power is lost.
The RCIC system consists of a steam-driven turbine pump and associated piping, valves, and instrumentation necessary to implement several
14 That is, before the reactor is depressurized to a level where the low-pressure cooling systems can be operated. However, it is also possible to lower the reactor pressure to below the shutoff head of the RHR/LPCI or LPCS so that one of these sources of water can be injected to the RPV while still having enough RPV pressure to provide steam for RCIC operation (150- to 200-psi range).
flow paths. The system is driven by steam produced by decay heat in the RPV. Steam exits through isolation valves and is routed through the turbine pump to provide the motive force for pumping makeup water into the reactor. The steam is exhausted to the suppression pool after exiting the turbine. Makeup water can be supplied from either the condensate storage tank (CST) or the suppression pool, with the preferred source being the CST. Makeup water enters the RPV through the feedwater injection line (see Figure 2.8).
FIGURE 2.8 Schematic of the reactor core isolation cooling (RCIC) system for Units 2 and 3 at the Fukushima Daiichi plant. Valves are indicated by connected triangles. Black indicates valve closed; white indicates valve open. Power sources (AC or DC power) for motor-operated (MO) valves are indicated. Hydraulically operated (HO) valves are controlled either automatically or manually via a DC-powered control system. The electronic governor regulator (EGR) controls the HO valve and throttles steam flow to the reactor core isolation cooling (RCIC) turbine. Water is supplied from the suppression pool (S/P) or condensate storage tank (CST). SOURCE: ANS (2012, Appendix F, Fig. 2).
As shown in Figure 2.8, the valve outside of containment is operated by DC power supplied by batteries; but the valve inside containment is operated by AC power. If DC power is lost, a separate DC-powered interlocking logic circuit causes both the DC and AC valves to close, effectively shutting down the RCIC system. This logic circuitry was specifically intended as an isolation function to prevent leakage from the containment if a break occurs in the RCIC piping. Once closed, the valve inside containment cannot be reopened unless AC power is available (see Footnote 13).
The RCIC system is designed to operate over a wide range of RPV pressures—from full operating pressures (~7 MPa15) to ~1 MPa. The suppression pool acts as the heat sink for steam generated by reactor decay heat. Decay heat can be removed from the suppression pool using the heat exchangers in the RHR system when AC power is available.
The continued operation of the RCIC system following a loss of DC and AC power depends on the timing of the power losses in the AC and DC circuits that control the valves and the “failsafe” control logic—similar to the IC system operation that was described in Section 18.104.22.168. In the case of an extended loss of AC power, such as occurred at the Fukushima Daiichi plant following the tsunami, the RCIC system may stop operation for the following reasons:
• DC power for the failsafe logic control has failed, causing the system’s valves to close (if motive power for the valves is still available).
• Suppression pool temperature is too high, possibly leading to failure of the turbine and pump bearings.
• Containment pressure is too high, causing the RCIC system turbine to shut down.16
22.214.171.124 High-Pressure Coolant Injection System (All Fukushima Daiichi Reactors)
The HPCI system (Figure 2.9) is similar to the RCIC system in function except that it has about seven times the flow capacity (680-1,270 m3/h). It is designed to operate when the RPV remains at high pressure; such conditions might occur when a small pipe break causes water levels in the RPV to drop, but the diameter of the broken pipe is not large enough to depressurize the RPV. The HPCI can also act as a backup to the RCIC system. The same types of actuation signals initiate and terminate both the HPCI
15 Megapascals (106 pascals). Pascal is the SI-derived unit for pressure and is equal to 1 N/m2; 1 MPa ≈ 145 pounds per square inch (psi).
16 BWR emergency operating procedures are being rewritten to override the high containment backpressure trip for the RCIC turbine.
FIGURE 2.9 Schematic of the high-pressure core injection (HPCI) system for Units 1-3 at the Fukushima Daiichi plant. Valves are indicated by connected triangles. Black indicates valve closed; white indicates valve open. Power sources (AC or DC power) for motor-operated (MO) valves are indicated. Hydraulically operated (HO) valves are controlled either automatically or manually via a DC-powered control system. The electronic governor regulator (EGR) controls the HO valve and throttles steam flow to the high-pressure coolant injection (HPCI) turbine. Water is supplied from the suppression pool (S/P) or condensate storage tank (CST). SOURCE: ANS (2012, Appendix F, Fig. 3).
and RCIC, and DC power is needed to operate the HPCI pump and some HPCI system valves.
2.2.4 Emergency Power Systems
Nuclear plants are designed with multiple power sources to run pumps, valves, and controls to remove the decay heat from the reactor core. AC
power is normally provided from offsite sources and is brought into the plant through multiple independent power lines. If offsite power is lost, AC power can be generated by onsite emergency diesel generators. These generators are designed to start up automatically in the first minute following a loss of offsite power. Each reactor at a nuclear plant has at least two diesel generators for redundancy. There is enough fuel onsite to last for several days if power and operable pumps are available to move it from large onsite storage tanks to smaller tanks that supply the diesel generators.
Large batteries (or banks of batteries) are situated onsite to provide emergency DC power for a select set of valves, instruments, lighting, and communications; these batteries are designed to supply power for about 8 hours under typical load conditions. As noted previously, DC power is used to operate critical valves and monitoring instrumentation for the IC and RCIC systems. Consequently, it is essential to protect the batteries and circuits used to carry DC power through the plant so that these will continue to function even when AC power is lost.
Prior to the Fukushima Daiichi accident, Japan had 54 operating nuclear power reactors at 16 sites (see Figure 2.10 and Table 2.3). These reactors provided about 30 percent of Japan’s electricity supply. In early 2011, Japan was the world’s third-largest producer of electricity from nuclear power, after the United States and France. Tokyo Electric Power Company, the owner/operator of the Fukushima Daiichi plant, owns 17 nuclear reactors at three sites: Fukushima Daiichi (6 reactors), Fukushima Daini (4 reactors), and Kashiwazaki Kariwa (7 reactors) (see Figure 2.10). Collectively, these reactors supplied about a third of Japan’s nuclear power–generated electricity before the accident.
The nuclear plant fleet in Japan consists of 24 PWRs and 26 BWRs. All but four of these plants are Generation II designs.17 Four ABWRs, at Hamaoka, Kashiwazaki-Kariwa, and Shika (Figure 2.10), are Generation III designs.
Figure 2.11 shows the operating electrical generating capacity of nuclear plants in Japan in 2011 and 2012. There was a decrease in capacity following the Fukushima Daiichi accident in March 2011 as reactors were
17 Reactor generation terminology was developed by the U.S. Department of Energy. Generation II reactors were constructed beginning in the 1960s. They have mechanically and electrically operated safety systems that can be started automatically or by operator control. Most of the world’s current reactor fleet consists of Generation II reactors. Generation III reactors were constructed beginning in the 1990s. They incorporate more passive safety systems and have other design improvements. See Goldberg and Rosner (2011) for additional information.
FIGURE 2.10 Locations of nuclear power plants in Japan. SOURCE: Nuclear Energy Agency (available at http://www.oecd-nea.org/press/2011/NEWS-02.html. Accessed on June 3, 2014).
taken offline for scheduled maintenance and were not allowed to restart. All Japanese reactors were shut down by April 2012. Two of the reactors at the Kansai Electric Power Co.’s Ohi plant in western Japan (Figure 2.10) were allowed to restart in July 2012 because of concerns about power shortages in the Kyoto region. These reactors were subsequently shut down for scheduled maintenance in September 2013 and were not allowed to restart.
All nuclear reactors in Japan must undergo a safety review by the new nuclear plant regulator (Nuclear Regulation Authority; see next section) before they can be restarted. These reviews are currently under way, and no completion date has been announced.
TABLE 2.3 Operating Nuclear Plants in Japan Prior to the Fukushima Daiichi Accident
|Plant Name||Unit||Reactor and Containment Type||Thermal Capacity (MWt)||Commercial Operation||Licensee|
|Fukushima Daiichi||1||BWR-Mark I||1,380||1971||TEPCO|
|Fukushima Daiini||1||BWR-Mark II||3,293||1982||TEPCO|
|2||BWR-Mark II (Improved)||3,293||1984|
|3||BWR-Mark II (Improved)||3,293||1985|
|4||BWR-Mark II (Improved)||3,293||1987|
|Hamaoka||3||BWR-Mark I (Improved)||3,293||1987||Chubu|
|4||BWR-Mark I (Improved)||3,293||1993|
|Higashidori||1||BWR-Mark I (Improved)||3,293||2005||Tohoku|
|2||BWR-Mark II (Improved)||3,293||1990|
|3||BWR-Mark II (Improved)||3,293||1993|
|4||BWR-Mark II (Improved)||3,293||1994|
|5||BWR-Mark II (Improved)||3,293||1990|
|2||BWR-Mark I (Improved)||2,436||1995|
|3||BWR-Mark I (Improved)||2,436||2002|
|Shika||1||BWR-Mark I (Improved)||1,593||1993||Hokuriku|
|2||BWR-Mark I (Improved)||2,436||1989|
|Plant Name||Unit||Reactor and Containment Type||Thermal Capacity (MWt)||Commercial Operation||Licensee|
NOTES: ABWR = advanced boiling water reactor; BWR = boiling water reactor; PWR = pressurized water reactor.
SOURCES: Nuclear Regulation Authority, written communication; IAEA (2014b).
FIGURE 2.11 Electrical generating capacity from operating Japanese nuclear power plants prior to and following the Fukushima Daiichi accident. SOURCE: Electrical generating capacity from Nuclear Regulation Authority, written communication; reactor shutdown dates from NRA (2013a).
2.3.1 Regulation of Nuclear Plants in Japan
Prior to the Fukushima Daiichi accident, the Nuclear and Industrial Safety Agency (NISA) within the Ministry of Economy, Trade and Industry (METI) was responsible for nuclear plant regulation in Japan. NISA was overseen by the Nuclear Safety Commission (NSC), a senior government body responsible for formulating safety policy, and the Atomic Energy Commission (AEC), which was responsible for nuclear power and research policy. Both the NSC and AEC were part of the Cabinet Office18; however, they were advisory and neither had direct authority over nuclear plant regulation.
Following the Fukushima Daiichi accident, NISA’s association with METI was seen to compromise its independence and pose a conflict of interest because METI also promotes nuclear energy. The Japanese government decided to eliminate NISA and establish a new organization in its place. This new organization, the Nuclear Regulation Authority (NRA), was
18 The Cabinet Office is Japan’s executive branch of government. It consists of the Japanese prime minister and other state ministers.
FIGURE 2.12 (Left) Japanese government nuclear organizations prior to the Fukushima Daiichi nuclear accident. (Right) Organization of new Japanese nuclear regulator (NRA) effective September 2012. JAEA = Japan Atomic Energy Agency. JNES = Japan Nuclear Energy Safety Organization. NIRS = National Institute of Radiological Sciences. SOURCE: Task Force for the Reform of Nuclear Safety Regulations and Organisations (n.d).
established as an extra-ministerial organization of the Ministry of Environment in September 2012. NRA combines the roles of NISA and NSC and also assumed the nuclear-related functions of the Ministry of Education and Science (see Figure 2.12).
The NRA is headed by a five-member commission composed of a chairman and four commissioners who are appointed by the Japanese prime minister and confirmed by the National Diet for 5-year terms. A secretary general directs the activities of the Secretariat of the NRA and carries out the policies and decisions of the commission. Most of the staff of NRA were transferred from METI and the Ministry of Education, Culture, Sports, and Science and Technology (MEXT); they will not be allowed to return to METI or MEXT in the future because they were hired by the NRA under a “no-return” rule.
FIGURE 2.13 Locations and names of currently operating nuclear plants in the United States. SOURCE: USNRC (2013b, Fig. 16).
TABLE 2.4 Operating Nuclear Plants in the United States (June 2014)
|Plant Name||Unit||Reactor and Containment Type||Thermal Capacity (MWt)||Commercial Operation||Licensee|
|Browns Ferry||1||BWR-Mark I||3,458||1974||Tennessee Valley Authority|
|Brunswick||1||BWR-Mark I||2,923||1977||Carolina Power & Light|
|Calvert Cliffs||1||PWR-DRYAMB||2,737||1975||Calvert Cliffs|
|Catawba||1||PWR-ICECND||3,411||1985||Duke Energy Carolinas|
|Columbia||2||BWR-Mark II||3,486||1984||Energy Northwest|
|Cooper||BWR-Mark I||2,419||1974||Nebraska Public Power|
|Diablo Canyon||1||PWR-DRYAMB||3,411||1985||Pacific Gas & Electric|
|D. C. Cook||1||PWR-ICECND||3,304||1975||Indiana Michigan Power|
|Duane Arnold||BWR-Mark I||1,912||1975||FPL Energy Duane Arnold|
|E. I. Hatch||1||BWR-Mark I||2,804||1975||Southern|
|Fermi||2||BWR-Mark I||3,430||1988||DTE Electric|
|Fort Calhoun||1||PWR-DRYAMB||1,500||1973||Omaha Public Power|
|Grand Gulf||1||BWR-Mark IIIa||4,408||1985||Entergy|
|H. B. Robinson||2||PWR-DRYAMB||2,339||1971||Carolina Power & Light|
|Hope Creek||1||BWR-Mark Ib||3,840||1986||PSEG Nuclear|
|J. A. FitzPatrick||BWR-Mark I||2,536||1975||Entergy FitzPatrick|
|J. M. Farley||1||PWR-DRYAMB||2,775||1977||Southern|
|Plant Name||Unit||Reactor and Containment Type||Thermal Capacity (MWt)||Commercial Operation||Licensee|
|McGuire||1||PWR-ICECND||3,411||1981||Duke Energy Carolinas|
|Monticello||1||BWR-Mark I||1,775||1971||NSP Minnesota|
|Nine Mile Point||1||BWR-Mark I||1,850||1969||Nine Mile Point Nuclear|
|North Anna||1||PWR-DRYSUB||2,940||1978||Virginia Electric & Power|
|Oconee||1||PWR-DRYAMB||2,568||1973||Duke Energy Carolinas|
|Oyster Creek||1||BWR-Mark I||1,930||1969||Exelon|
|Palo Verde||1||PWR-DRYAMB||3,990||1986||Arizona Public Service|
|Peach Bottom||2||BWR-Mark I||3,514||1974||Exelon|
|Point Beach||1||PWR-DRYAMB||1,800||1970||NextEra Point Beach|
|Prairie Island||1||PWR-DRYAMB||1,677||1973||NSP Minnesota|
|Quad Cities||1||BWR-Mark I||2,957||1973||Exelon|
|River Bend||1||BWR-Mark IIIa||3,091||1986||Entergy|
|R. E. Ginna||PWR-DRYAMB||1,775||1970||R. E. Ginna Nuclear|
|Saint Lucie||1||PWR-DRYAMB||3,020||1976||Florida Power & Light|
|Sequoyah||1||PWR-ICECND||3,455||1981||Tennessee Valley Authority|
|Shearon Harris||1||PWR-DRYAMB||2,900||1987||Carolina Power & Light|
|South Texas||1||PWR-DRYAMB||3,853||1988||STP Nuclear|
|Surry||1||PWR-DRYSUB||2,857||1972||Virginia Electric & Power|
|Plant Name||Unit||Reactor and Containment Type||Thermal Capacity (MWt)||Commercial Operation||Licensee|
|Susquehanna||1||BWR-Mark II||3,952||1983||PPL Susquehanna|
|Three Mile Island||1||PWR-DRYAMB||2,568||1974||Exelon|
|Turkey Point||3||PWR-DRYAMB||2,644||1972||Florida Power & Light|
|Vermont Yankee||1||BWR-Mark I||1,912||1972||Entergy|
|V. C. Summer||1||PWR-DRYAMB||2,900||1984||South Carolina Electric & Gas|
|Watts Bar||1||PWR-ICECND||3,459||1996||Tennessee Valley Authority|
|Wolf Creek||1||PWR-DRYAMB||3,565||1985||Wolf Creek Nuclear|
a Mark III containments have a concrete secondary containment (also known as a Shield Building).
bHas concrete secondary containment unlike other BWRs of this type.
NOTES: Reactor types: BWR = boiling water reactor; PWR = pressurized water reactor. Containment types: DRYAMB = dry, ambient pressure; DRYSUB = dry, subatmospheric pressure; ICECND = wet, ice condenser; Mark I = wet, Mark I; Mark II = wet, Mark II; Mark III = wet, Mark III.
SOURCES: USNRC (2013b), IAEA (2014b).
2.4.1 Regulation of Nuclear Power in the United States
The U.S. Nuclear Regulatory Commission (USNRC) is responsible for nuclear reactor and materials safety in the United States and U.S. territories. The USNRC was established by the Energy Reorganization Act of 1974 to be an independent19 agency in the executive branch of the U.S. government. Before the USNRC was established, nuclear safety regulation and nuclear power promotion were the responsibility of the Atomic Energy Commission (AEC). The Energy Reorganization Act dissociated AEC’s responsibilities: USNRC assumed the AEC’s regulatory responsibilities and the Energy Research and Development Administration assumed AEC’s responsibilities for nuclear promotion. ERDA was later reorganized into the U.S. Department of Energy (USDOE).
The USNRC is overseen by five Commissioners, one of whom is designated as chairman, who are appointed by the president of the United States and confirmed by the U.S. Senate to serve 5-year terms. The Commission formulates policies and regulations for nuclear reactor safety, issues orders to licensees, and adjudicates legal matters brought before it. The USNRC is headquartered in Rockville, Maryland, and has four regional offices (in Pennsylvania, Georgia, Illinois, and Texas) to provide direct links to individual nuclear plants through resident inspectors.
The Atomic Energy Act of 1954 specifies that U.S. nuclear energy facilities can be licensed for an initial period of 40 years and that such licenses are renewable. USNRC regulations permit licenses to be renewed for periods not to exceed 20 years (10 CFR § 54.3120). Most of the currently operating nuclear plants in the United States have received or are seeking 20-year license renewals, which would extend their operating lives to 60 years. The USNRC and nuclear industry are examining the feasibility of an additional 20-year renewal to extend plant operating lives to 80 years.
19 Independent agencies in the U.S. government are run by commissions or boards with oversight from the U.S. Congress. The members of the commissions and boards are appointed by the president; some appointments require U.S. Senate confirmation.
20 This section of Title 10 of the Code of Federal Regulations (CFR) is available at http://www.nrc.gov/reading-rm/doc-collections/cfr/part054/part054-0031.html. Accessed on June 3, 2014.
Twenty-one BWRs in the United States have the same reactor and containment designs as the Fukushima Daiichi units (see Table 2.2).21 Six are of the same design as the Fukushima Unit 1 (BWR/3 Mark I) and 15 are the same design as Fukushima Units 2, 3, and 4 (BWR/4 Mark I).22 Several safety enhancements have been made to Japanese and U.S. Mark I BWRs since they began operating; some of these enhancements are described in the following sections.
2.5.1 Fire Protection
After a 1975 fire in Unit 1 at the Browns Ferry Nuclear Plant, fire protection requirements in the United States were enhanced. Reactor safety shutdown systems were physically separated to provide redundancy and independence during any single fire event. However, as discussed in Chapter 7 (see Section 7.3.3), not all U.S. reactors have adopted these measures. These measures also have not been adopted in Japan.
2.5.2 Hardened Containment Vents
Installation of hardened containment vents in Mark 1 BWRs was recommended by the USNRC23 following the 1979 Three Mile Island accident; the U.S. nuclear industry committed to voluntarily comply with this recommendation. All but one24 Mark I BWR in the United States are currently equipped with hardened vents, but vent designs are plant specific. Hardened vents were also installed at all eight of the Mark I BWR plants in Japan. Following the Fukushima Daiichi accident, the USNRC issued a new order25 to Mark I and Mark II BWR licensees to design and install “Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions” (see Chapter 5 and especially Appendix F).
21 Although the reactors and containments have a standard design, the design, of the remainder of the plant, including reactor buildings, control rooms, and locations of safety systems, are not standardized.
22 In addition, there are two BWR/2 Mark I plants in the United States: Nine Mile Point Unit 1 (New York) and Oyster Creek (New Jersey).
23 Installation of a Hardened Wetwell Vent (Generic Letter 89-16). Available at http://i2.cdn.turner.com/cnn/2012/images/02/16/nrc.gl.89.16.pdf. Accessed on June 3, 2014.
24 The exception is the James A. FitzPatrick plant, which is located in New York.
25 Compliance with Order EA-13-109, Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions (JLD-ISG-2013-02). Available at http://pbadupws.nrc.gov/docs/ML1330/ML13304B836.pdf. Accessed on June 3, 2014.
2.5.3 Containment Inerting
The USNRC also required the inerting26 of containments in BWRs with Mark I and Mark II designs following the 1979 Three Mile Island accident. This practice was adopted worldwide. Following the Fukushima Daiichi accident, the USNRC examined the need for additional hydrogen control measures but decided not to take immediate action (see Chapter 5).
2.5.4 Other Containment Modifications
In the early 1980s, Mark I BWR containment systems were modified to improve their safety margins in loss-of-coolant accidents. Modifications included reinforcements to the suppression chamber and associated structures.27 Japanese regulations and plant modifications closely followed U.S. practice, and so these changes may have been implemented in Japanese plants as well.
2.5.5 Control Room Improvements
The 1979 Three Mile Island accident prompted the nuclear industry to enhance control room process and design. Access to control rooms was limited, safety alarms were improved, and changes in control and display systems were made. Some of these changes were likely implemented in Japan.28
2.5.6 Station Blackout
In 1988, the USNRC issued a station blackout rule29 that required nuclear plants to maintain highly reliable onsite AC power; ensure that plants can cope with station blackout (defined as the loss of both offsite AC power and onsite emergency AC power) for a predetermined period of time using battery backup power; develop procedures and training for restoring offsite AC power and onsite emergency AC power; and make other modifications to plants as needed.
26 Nitrogen is used to displace air within the primary containment vessel when the reactor is operating. By reducing the concentration of oxygen to less than 4 percent, it is possible to prevent explosions or fires within the containment even if hydrogen is generated and released from the RPV into containment. BWR Mark III containments are not inerted. They have hydrogen control systems that are designed to burn hydrogen at low concentrations.
27 The modifications are described in Borchardt (2012a, Enclosure 2).
28 It is difficult to make a direct comparison between Japan and the United States; control rooms in Japan are different in terms of reliance on computer controls and number of staff.
29 10 CFR § 50.63, Loss of all alternating current power. Available at http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0063.html. Accessed on June 3, 2014.
The Japanese Nuclear Safety Commission’s Regulatory Guide for Reviewing Safety Design of Light Water Nuclear Power Reactor Facilities (August 1990) provides the following guidance on station blackout:
Guideline 27. Design Considerations against Loss of Power
The nuclear reactor facilities shall be so designed that safe shutdown and proper cooling of the reactor after shutting down can be ensured in case of a short-term total AC power loss.30
According to an unofficial translation of the Nuclear Safety Commission’s Special Committee on Nuclear Safety Standards and Guides, “short term” has been routinely interpreted as meaning DC battery capacity to maintain residual heat removal for 30 minutes under station blackout conditions (NSCJ, 2012).
Nuclear plants throughout Japan have AC emergency power and coping capabilities similar to U.S. nuclear plants with multiple redundant AC power sources (e.g., 13 emergency diesel generators at the Fukushima Daiichi plant) and backup batteries (trains of battery-powered 125VDC and 250VDC power sources). The battery coping time (i.e., the length of time that station batteries can provide power under a specified load) at Japanese nuclear plants is comparable to the 4- to 8-hour coping time typical for U.S. nuclear plants.
2.5.7 Improved Mitigation Capabilities
The September 11, 2001, terrorist attacks led to an extensive review of accident scenarios beyond then-current plant design standards. The USNRC issued an Interim Compensatory Measures Order31 in 2002 that directed nuclear plant licensees to develop mitigation strategies to cope with large fires and explosions from any cause, including aircraft impacts. These strategies are intended to use readily available resources to maintain or restore core cooling, containment, and spent fuel pool cooling. A final rule was issued in March 2009 (USNRC, 2009a).
Japanese utilities did not implement these measures because they were unaware of the details of the U.S. program. Japanese regulatory agencies that were aware of this program did not discuss it with utilities or impose
30 The quoted material is taken from an unofficial translation of the guide.
31 Interim Compensatory Measures Order EA-02-026 (this document is designated as Safeguards Information and is not available to the public).
similar requirements.32 However, following the 2007 Niigata Chuetsu-Oki Earthquake and fire at the Kashiwazaki-Kariwa plant, additional water supplies, fire pumper trucks, and external connections to the reactor building fire protection system were required at all nuclear plants in Japan.
32 Section B.5.b of the USNRC Order for Interim Safeguards and Security Compensatory Measures was designated by the USNRC as Safeguards Information, and so it was exempt from public release. Consequently, TEPCO would not have had direct access to this information. However, as discussed in Chapter 7, the USNRC shared some B.5.b information with Japanese government authorities. Moreover, the USNRC requirements were made publicly available in 2009 (USNRC, 2009a) and were incorporated into reactor designs that were being developed by Japanese vendors for sale in the United States.