13.29. The boiling-water reactor (BWR), as the name implies, generates steam directly in the core. Since this steam then goes directly to the turbine — generator, the system is inherently simpler than a PWR. However, a com­plication is the need to maintain the fraction of steam in the core less than about 14 percent by weight for stability purposes. Thus, the exiting water, which has not been vaporized, must be separated from the steam at the top of the core and recycled.

13.30. Since the core steam conditions are comparable to those on the secondary side of a PWR steam generator, the BWR core pressure is about one-half that of a PWR. Therefore, a larger core and lower power density are practical as well as larger-diameter fuel rods. In Table 13.3 are listed typical design specifications for a large BWR of the type marketed in the 1970s. The thermal and electrical powers are essentially the same as for

General	Thermal-Hydraulic
Power Thermal Electrical Specific power Power density
3830 MW 1330 MW 25.9 kW(th)/kg U 56 MW(th)/m3
Core
3.76 m (12.3 ft) —4.8 m (15.8 ft)
Length Diameter
Fuel
12.52 mm (0.493 in.) 0.864 mm (0.034 in.) 10.57 mm (0.416 in.) 16.3 mm (0.64 in.) 62 (8 x 8 array) 4.47 mm (14.7 ft.) 134 mm (5.28 in.) 0.01 m2 (15.5 in.2) 168 x 103 kg (3.69 x 105 lb)
Rod, OD Clad thickness Pellet diameter Rod lattice pitch Rods per assembly Assembly overall height Assembly width Coolant flow area/ assembly Fuel loading, 1Ю2
Control
Movable cruciform elements Overall length Poison section
193 4.42 m (14.5 ft) 3.66 m (12.0 ft)
Ave. feed enrichment —2.6% Ave. core enrichment —1.9% Burnup 2.38 TJ/kg (27,500 MW • d/t)

the PWR described in Table 13.1. Hence, the parameters quoted for the two types of light-water reactors may be compared.

13.31. A variety of BWR fuel designs are offered by various vendors. Worthy of note is a 9 x 9 lattice design containing 272 fueled rods and a central water channel (§10.57). An average linear heat rating of 15.9 kW/m is specified for rods having an outside diameter of 11 mm.

15.11. Passive systems based on light-water core designs benefit from the considerable design and operating experience of existing reactors and therefore are attractive to utilities. A PWR design, designated as AP600, has been developed by Westinghouse Electric Corp. in cooperation with the U. S. Department of Energy (DOE) and EPRI. Similarly, a 600-MW(el) BWR, known as the SBWR, has been developed by the General Electric Co., with DOE and EPRI cooperation. In this case, the smaller size permits natural recirculation of the coolant, a significant design simplification.

15.12. Gas-cooled reactors of various types have been used to generate electricity since the earliest days of the nuclear power industry, particularly in the United Kingdom. During more recent years, prototypes of the high — temperature, gas-cooled reactor (HTGR) concept have operated in the United States and in Germany. This type of helium-cooled, graphite­moderated system features fuel embedded in small spherical particles that retain fission products. The modular high-temperature gas-cooled reactor (MHTGR) is a passive, inherently safe concept which builds upon HTGR experience.

15.13. Sodium-cooled fast reactors have also received development at­tention for many years. Most notable among experimental reactors is EBR — II, which has been operating very satisfactorily since 1961 at a rated elec­trical output of 20 MW. Demonstrations of passive, inherent safety char­acteristics led to the development of the integral fast reactor (IFR) concept at Argonne National Laboratory. Subsequently, a team led by the General Electric Co. developed a modular reactor concept PRISM (Power Reactor, Innovative Small Module), which appears promising for future energy requirements.

15.14. The advanced passive PWR and BWR are likely to receive early favor by electrical utilities seeking a 600-MW(el)-size unit, particularly since they utilize technology familiar to them. The other candidates for commercialization are the MHTGR and PRISM concepts, which, should they receive utility support, are likely to come somewhat later. Finally, we will mention examples of other interesting concepts which have potential. One is the process inherent ultimate safety (PIUS), developed in Sweden, and the safe integral reactor (SIR), developed by a team led by Combustion Engineering based on earlier PWR designs intended for maritime use.

THE AP600 [2]

Introduction

15.15. The AP600 (“advanced, passive”) reactor design features a two — loop Westinghouse PWR arrangement modified to have conservative safety margins and permit simplification of many supporting subsystems. The coolant loop arrangement and some of these features are shown in Fig.

15.1. In each loop, circulation is provided by two closely coupled canned motor pumps. As in the Combustion Engineering evolutionary PWR (§13.24), the steam generator has been enlarged to improve operating margins.

15.16. Some preliminary design specifications are summarized in Table

15.1. The core contains 145 fuel assemblies of the 17 x 17 lattice type with an active length of 3.66 m (12 ft), but similar to that described for the larger PWR in Table 13.1. However, the power density has been

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Close coupied pumps eliminate Hot-leg pipe small LOCA core

(31 in. ID) uncovery

reduced by about 30 percent to improve safety margins. Consistent with this reduction, a number of other specifications are less challenging than those listed in Table 13.1.

12.205. The role of these elements is clear. As a result of extensive studies and strategy development, some equipment modifications may ap­pear desirable. For example, one study showed that the addition of one or more high-pressure ECCS pumps reduces the core melt probability. Also, once management strategies have been developed, operating per­sonnel need to be trained in their use.

13.63. An evolutionary CANDU-type reactor system is being developed to be competitive in the international market with other medium-sized advanced concepts. This system, designated CANDU 3, is a 450-MW(el) plant that features a standardized design and numerous design features intended to achieve a short construction schedule. Prefabricated modules are extensively employed. The basic design is similar to that of the CANDU described previously, and thus takes advantage of proven technology. A new control room includes extensive human factors features similar to those incorporated into evolutionary LWRs in the United States. Design certi­fication by the US NRC is being pursued [8].

INTRODUCTION

14.1. As the nuclear power industry has matured, engineering emphasis has shifted from new construction-related design to operational consider­ations. Although many of the topics covered in previous chapters are rele­vant to plant operations, we are devoting a separate chapter to subjects that are particularly operations related.

14.2. Operational strategy to meet system load, refueling, and main­tenance requirements is an important area of utility engineering manage­ment. Control room operations and associated staff activities are another vital area. Although the subject of reactor control was introduced in Chap­ter 5, we can now provide additional depth, building upon the material of reactor safety and regulation presented in Chapter 12. Practically all aspects of operation is subject to strict licensing requirements, another topic worthy of at least introductory attention.

14.3. As plants approach the end of their original design lifetime, there are some special operational considerations primarily concerned with ma­terials behavior. Also, with no new reactors being built, there is a sub-

stantial economic incentive to extend the operating lifetime of existing reactors. Both of these related subjects will be discussed in this chapter. Finally, the engineering aspects of decommissioning presently operating plants will be treated.

15.43. The reactor core, pressure vessel, and steam generator arrange­ment is shown in Fig. 15.4. Major design specifications are summarized in Table 15.3. In the vessel, the helium coolant flows downward through the core, where it is heated to an average temperature of 687°C, then flows through an inner portion of a coaxial cross duct to the steam generator. The gas flows downward through a helical heat exchange bundle in which steam is produced in the tubes. This cooled gas then flows upward in an annulus compartment to the main circulator integral to the top of the steam generator vessel. Compressed cool gas returns to the reactor vessel through an outer annulus in the cross duct, then flows upward through another annulus between the core and vessel wall to the top of the core to complete the circuit.

15.44. Decay heat during shutdown may be removed either using the main heat transport loop or through a separate shutdown cooling system driven by a circulator at the bottom of the reactor vessel. In this system, heat is transferred to service water in a separate helium to water exchanger.

12.245. After an order has been placed for a certified plant to be located at an approved site, an application for a combined construction and op­erating license (COL) may be submitted to NRC. There is then an op­portunity to evaluate all remaining safety issues prior to the start of con­struction, including those that may arise in another public hearing conducted at this stage. After a combined license is issued, construction of the plant may proceed.

12.246. An important concern of owner utilities is the role of any post­construction, pre-operation public hearing. The objective of both NRC and the utilities is to avoid any “unreasonable” delays in operational startup after the very substantial investment in plant construction has been made. Therefore, present plans are to provide for only an informal hearing, with consideration limited to whether the facility has been constructed, and will be operated, in conformity with the license. Furthermore, unless the NRC determines that there will not be adequate protection of the public health and safety, operation may proceed, with the rationale that minor issues of nonconformity can be resolved later.

12.247. During operation, the NRC continues to play a major regulatory role as described in various sections of this book (see, e. g., §14.47). Also, so-called Inspection, Tests, Analyses, and Acceptance Criteria (IT A AC) are specified in 10 CFR 52 “to provide assurance that the plant will operate in accordance with the design certification.”

14.36. The electric utility industry has recognized that there is a need for developing, testing, and applying expert systems. Quality assurance is a consideration since whenever subjective opinions are included in a knowl­edge base, the matter of confidence levels needs to be addressed. To meet a variety of needs, the Electric Power Research Institute established in 1989 the Knowledge-Based Technology Applications Center (KBTAC), located on the campus of Syracuse University.

Neural Network Development

14.37. Neural networks as an information-processing technology have attracted a great deal of recent developmental attention and appear to have significant applications to nuclear power plants, particularly in those areas in which expert systems are used, such as control diagnostics. Improve­ments in the monitoring of various systems as a result of neural network pattern recognition capability also appear likely.

15.70. Within the nuclear power industry, the term commercialization is used somewhat loosely. A power plant is said to be in commercial service when, after a suitable startup period, it furnishes electricity on a regular basis to a utility’s distribution grid. On the other hand, a new reactor type reaches commercialized status only after an unsubsidized plant of a size appropriate for a given utility grid operates reliably and economically for sufficient time to convince utilities that another plant should be built with reasonable financial risk. Some feel that the demonstration can be accom­plished by a subsidized plant.

15.71. The willingness of a utility to accept the financial risk of a new nuclear plant depends on several considerations in addition to the technical and economic merits of the concept. Clearly, a need must exist for the new capacity. A stable regulatory picture over the lifetime of the plant is es­sential. This applies to both the NRC and state level. We have discussed the major responsibility of the NRC in the licensing and operation of the plant. State regulators are responsible for a rate structure that will permit a utility to recover the expenses of energy production. Public acceptance and associated political pressures can play a significant role in this picture. However, we will not consider such considerations further here.

15.72. The competitive cost of a fossil-fueled or other type of power plant would also logically be studied by an electric utility planning new generation capacity. A comparison with a nuclear option is likely to be complicated with environmental, financial risk, and regulatory factors rel­evant (§15.81). Therefore, for our purposes, we will assume that advanced nuclear plants and other types of generating plants will be “competitive,” and concentrate on the relative potential of advanced nuclear plant designs.

15.73. Now that we have described the evolutionary and important advanced nuclear plant passive designs, it is helpful to examine them from the viewpoint of their commercialization potential. For example, the evo­lutionary designs are large while the advanced passive concepts are smaller.

The question of reactor size is one important issue that is relevant to their commercialization.

13.32. The internal structure of a typical BWR core and reactor vessel is depicted in Fig. 13.9. The core consists of almost 800 fuel assemblies in square 8×8 arrays; 62 fuel rods are contained in each assembly with two hollow central (“water”) rods designed so that water flows through them to provide additional moderation (§10.52). Each fuel assembly is contained in a four-sided channel separator or “can” to prevent crossflow of coolant between assemblies. A control rod (or element) with a cruciform cross section is located within a set of four assemblies (see Fig. 5.16).

13.33. The fuel assemblies rest in support pieces mounted on top of the control-rod guide tubes. The guide tubes, in turn, rest on the control-rod penetration nozzles in the bottom of the reactor vessel. The core plate (Fig. 13.9) merely provides lateral support. Orifices in the fuel-assembly support permit adjustment of the coolant flow distribution among the as­semblies in the core. These orifices can be changed, if necessary, but only by disassembly of the core structure.

13.34. An important BWR consideration is the power stability based on the coupled hydrodynamic-neutronic feedback response resulting from the formation of steam in the core (§5.109). Power oscillations, which would cause fluctuations of the steam voids in the moderator, and hence in the reactivity, should be damped by a proper combination of feedback parameters (§5.132 et seq.); a detailed discussion is beyond the scope of this treatment. Stability tends to be improved by a decrease in power density, increase in coolant flow rate, and increase in rod diameter. Stability considerations govern the thermal power-coolant flow combination during startup and operation as shown in a “power map” (Fig. 14.1) which defines the acceptable operating region. Generally, a BWR of present design can be operated on natural circulation up to about 25 percent of rated power.

13.35. As with the PWR (§13.10) the H/U ratio in a BWR affects the fuel utilization. However, as a result of the low hydrogen density in the vapor fraction of the coolant (—0.4 by volume), the H/U ratio tends to be slightly lower than that for a PWR although the H20/U02 volume ratio (—2.4) is higher.* The void (vapor) fraction, and the H/U ratio which

Water in the channels between fuel assemblies is included in the calculation of the ratio.

depends on it, varies both axially and radially in the core, as well as with burnup.

13.36. A stainless steel cylindrical shroud within the reactor vessel sur­rounds the core and separates the upward flow of coolant through the core from the downward flow in the annulus between the shroud and the vessel. The jet pumps (see §13.39) within the annulus are supported by the shroud.

13.37. Liquid water entrained in the steam-water mixture leaving the fuel assemblies is separated in the upper part of the reactor vessel. The mixture first enters vertical standpipes in the shroud head and then passes through axial flow centrifugal separators. The swirling motion drives the water droplets to the outer wall from which they flow back to the core via the downcomer annulus outside the shroud (Fig. 13.9). The steam, still containing some moisture, passes on through a dryer assembly of vanes and troughs; here most of the remaining moisture is separated and returned to the downcomer.

13.38. The two-stage system represents a design challenge since the exit steam must contain no more than 0.1 percent water by weight. Steam leaving the first stage contains less than 10 percent water by weight. Since the system must process the entire two-phase mixture from the core, the drying capacity of the system has been one of the constraints on core power.

15.17. The passive features center on the operation of the safety injec­tion systems and provisions for removing core decay heat. Passive systems are defined as those which are self-contained or self-supported. To perform their safety functions, they rely on gravity, as in natural circulation cooling, or stored energy, such as that in compressed gases. Emergency electrical needs must be met by battery power rather than by standby diesel gen­erators as is current practice. Valves to initiate safety system operation should be check valves or those activated by stored energy. Passive systems are expected to perform their safety functions for up to 72 hours after the initiating event independent of operator action and off-site power. Thus, should there be an emergency shutdown, decay heat must be passively removed from the core. Also, coolant must be replaced if there is a loss — of-coolant accident.

15.18. In the AP600, the safety injection systems are integrated with the reactor coolant makeup and residual heat removal systems. Decay heat is removed by a natural circulation system utilizing a heat exchanger sub­merged in a large refueling water tank within the containment as shown in Fig. 15.2. The tank volume is sufficient to absorb decay heat for about 2 hours, after which the water would start to boil. Then, the generated steam would condense on the inside of the containment and drain back into the refueling tank. Gravity-driven injection or pressurized accumu­lators replace water lost in the coolant system. As shown in Fig. 15.2, two core makeup tanks containing borated water at full cooling system oper­ating pressure are available to inject water by gravity in the event of a small-break LOCA. For larger breaks, two accumulators pressurized to 4.93 MPa(a) (715 psia) by nitrogen are available for injection. Finally, additional water can be injected by gravity from the in-containment re­fueling water storage tank provided that the cooling system pressure has first been reduced to about 1.72 MPa(a) (25 psia). This passive safety injection system eliminates the need for various pumps and support com­ponents such as diesel generators, thus significantly reducing the number of plant components required.

15.19. Passive cooling is also provided for the containment. Figure 15.2 shows an inner steel containment vessel which serves to transfer heat to outside air moving by natural circulation between the steel shell and the outer concrete shield. Additional cooling is provided by water stored in a tank at the top of the concrete shield. This water is allowed to evaporate

on the outside of the steel vessel. This water tank is sized for 3 days of operation, after which refilling is expected. However, if the water supply is not replenished for some reason, the containment pressure will rise to a maximum in about 2 weeks, but will then reach only 90 percent of the design pressure.