14.20. During normahoperation, most nuclear plants run continuously at full rated power (base loaded). Should there be unexpected sudden transients or disturbances affecting the load, various procedures are pro­vided to accommodate them. Our purpose here is not to consider such

cases but to examine the inherent ability of a system to “follow” changes in load should the particular generating unit not be in base-loaded service.

14.21. An example of the natural load-following capability of a PWR is initiated by an opening of the steam throttle leading to the turbine in response to a sensor detecting a slight slowing of the generator as the load is increased. As the pressure is reduced on the secondary side of the steam generator, there is some flashing of water leading to additional heat re­moval from the primary system. As the cooler water returns to the core, the increase in reactivity resulting from the negative temperature coefficient causes an increase in reactor power. For a load reduction, the process is reversed. In practice, some “fine tuning” by the control system is usually

carried out involving control rod movement or boron concentration ad­justment. However, the picture is not quite so simple. The partial insertion of rods during part-load operation can lead to an undesirable upper core axial power shift on return to full power since the upper region xenon concentration had previously been reduced. The excess upper half power is known as axial offset. However, axial problems can be avoided and load following accomplished without boron concentration adjustment by bal­ancing the operation of two banks of control rods, with each bank con­taining some weak absorbers or “gray” rods. This is known as MSHIM (mechanical shim) strategy.

14.22. By contrast, in a BWR, an opening of the steam throttle reduces the core pressure, causing more boiling, which, in turn, reduces the reac­tivity as a result of the void coefficient. Therefore, BWR load following, generally in the range of 75 to 100 percent power, is accomplished by the control system, which can reverse this tendency by increasing the recir­culating flow rate and moving the rods an appropriate amount. Although the recirculating pumps operate at constant speed, valving is provided for flow rate adjustment.

15.57. Prior to about 1970, metallic fuel was thought to be unacceptable because of its poor irradiation behavior. This picture was reversed com­pletely by new alloy designs and very favorable irradiation experience at EBR-II during the 1970s. Metal fuel that can achieve a peak burnup of 12.9 TJ/kg (150,000 MW • d/t) now appears feasible.

15.58. The thermal expansion and high thermal conductivity of metallic fuel provides important safety advantages. For example, in a loss-of-flow — without-scram (LOFWS) event, the exit coolant temperature will increase during flow coast-down. The resulting thermal expansion of the fuel as­semblies provides a negative reactivity feedback which reduces the power. Although during power reduction, the stored Doppler reactivity effect appears as a positive contribution, it is not significant because of the high thermal conductivity of the fuel. This behavior was demonstrated dra­matically at EBR-II in 1986 when a deliberate LOFWS experiment resulted in automatic reactor shutdown within about 7 minutes, with a maximum coolant temperature rise of about 200°C, which occurred within the first minute [8].

15.59. The reference fuel for the ALMR is an alloy consisting of ura­nium with 27 percent Pu and 10 percent Zr. Blanket assemblies contain an alloy of depleted uranium with 10 percent zirconium. A heterogeneous core configuration includes 199 assemblies of various types as shown in Fig. 15.6 sodium-cooled fast reactors generally use hexagonal geometry for the assembly ducts instead of the square design employed in LWRs. Since the coolant is not a moderator, the tighter triangular rod pitch can be used to maximize the fuel volume fraction and improve heat transfer characteristics.

15.60. For the equilibrium fuel loading cycle, one-third of the core is replaced after each 18 months of operation. Since the metallic fuel yields a breeding ratio of about 1.2, more fissile material is produced than con­sumed. In a planned on-site facility, electrorefining using a liquid cadmium anode and a fused chloride salt will be used to extract the discharged fuel uranium-plutonium mixture from the dissolved mixture of fuel and fission products. New fuel rods would then be prepared by a casting process.

13.12. Operational reactivity control in a PWR is provided primarily by boric acid dissolved in the coolant water supplemented somewhat by control rods. In a procedure known as chemical shim (§5.187), the boric acid concentration is varied to control reactivity changes during the op­erating cycle, such as those resulting from fuel depletion and fission product buildup. However, selected fuel assemblies contain burnable absorber rods to limit power peaking (§10.31). Therefore, the soluble boron concentra­tion adjustment is made in accordance with the effects on reactivity con­tributed by these burnable absorber rods as shown in Fig. 13.3. Although

this figure is for a first core in which solid burnable poison is needed to substitute for the fission products accumulated in subsequent cycles, the general effects on soluble boron concentration are typical of boron burn­able absorber rod use [2]. Also shown in Fig. 13.3 is the residual poison reactivity penalty associated with this practice. Gadolinia rods behave somewhat differently. The control rods are used primarily for startup, safety shutdown, and to follow load changes. Some rods are partial length or partial strength to aid power distribution “shaping.”

13.13. The control rods are stainless steel tubes encapsulating an ab­sorber material such as hafnium, boron carbide, or a silver-indium-cadmium alloy. These rods are arranged in clusters which move within guide tubes (thimbles) which replace some fuel rods in the rod lattice of the fuel as­semblies. Perforations over a portion of the thimble length allow the escape of water as the rods are inserted. However, the lower end of the tubes is closed to decelerate the rods at the end of their drop. The clusters are operated in groups to accomplish their various functions. In assembly po­sitions not reserved for control rod insertion, the empty thimbles may be used for burnable absorber rods. However, other fuel designs provide for more flexibility (§10.32).

14.51. For all options, the reactor fuel is first removed. Loose contami­nation would be cleaned from accessible plant areas and radioactive wastes would be processed and shipped off-site. The subsequent strategy depends on the urgency to clear the site and comparative economics. A 20-year delay would reduce the residual radioactivity levels from major components by about a factor of 10. However, cobalt-60 activity would still require remote handling of equipment. A decay period of about 100 years would be needed to permit manual handling. Although a 20-year delay might simplify removal operations somewhat, the unavailability of the site could carry an economic penalty. Therefore, we have the option of prompt dis­mantling and removal or so-called “mothballing” for a period of years prior to dismantling. A third option is to enclose the vessel and other radioactive components with a concrete barrier and allow the plant to remain for an indefinite period.

14.52. All options would require a possessions-only license from the NRC which would call for security provisions, environmental surveys, and reports until all radioactive materials are removed from the site. Therefore, the expense of site maintenance and public relations considerations are likely to favor the early dismantling option.

13.54. Typical design specifications of an advanced CANDU type re­actor are given in Table 13.5, and the reactor core arrangement is shown in Fig. 13.12. The zircaloy pressure tubes containing the fuel rods are horizontal, thereby facilitating fuel handling (charge and discharge) while the reactor is operating at power. Each pressure tube contains 12 fuel bundles; in the older CANDU reactors there are 28 rods per bundle, but the number has been increased to 37 in the newer designs. The fuel rods, which are separated laterally by spacers, consist of natural uranium dioxide pellets clad with zircaloy. A number of vertical “booster” fuel rods con­taining slightly enriched uranium dioxide are not normally in the core; they are used only to override xenon poisoning if the reactor is restarted soon after an unexpected shutdown.

13.55. On-line refueling, i. e., without reactor shutdown, is conducted by means of two machines, one at each end of the core, operating in tandem. One machine inserts a fuel bundle in a pressure tube while the other machine removes the spent bundle displaced at the other end. Axial power flattening is achieved by limiting the number of bundles that are replaced at any time and by using bidirectional loading in adjacent chan­nels. Refueling strategy is determined on the basis of calculated channel fuel burnup and reactivity demands.

Introduction

15.32. Gas-cooled reactors, such as the HTGR, are quite different from LWRs. The moderator is graphite, a solid, while the coolant, normally pressurized helium, is chemically inert and does not change phase. The graphite moderator remains stable up to very high temperatures, and has a high heat capacity, characteristics that permit the reactor to accommodate accidental transients safely. However, since power densities tend to be much lower than those for LWRs, the cores are relatively large, a disad­vantage. By using the modular concept, in which several units of modest power rating and easily manageable core size collectively generate the desired electrical capacity, the major safety advantages of the HTGR con­cept can be practically realized.

15.33. Many commercial reactors cooled with carbon dioxide were built in the United Kingdom and in France starting in the 1950s. The early ones were fueled with metallic natural uranium clad with a magnesium- aluminum alloy called Magnox. Later plants, designated as advanced gas — cooled reactors (AGRs), used low-enriched uranium oxide clad with stain­

less steel as fuel and carbon dioxide as coolant. The AGRs are able to achieve high coolant temperatures, which provides a thermodynamic ef­ficiency of over 40 percent. Newer AGRs built in the United Kingdom featured the use of a prestressed concrete reactor vessel (PCRV) for con­taining the whole primary system, consisting of the core, steam generators, and circulators [4].

15.34. While the AGR was being developed, attention was also given in various countries, including the United States, to the concept for an all­ceramic, helium-cooled, graphite-moderated, reactor using the thorium fuel cycle. This became the basis for the HTGR, the U. S. development of which was led by General Atomics. Other development efforts have been active in other countries, particularly in Germany. A 40-MW(el) demon­stration plant operated successfully at Peach Bottom, Pennsylvania from 1967 to 1974 using a prismatic core concept (§15.41). This was superseded by a 330-MW(el) commercial-size demonstration plant at Fort St. Vrain, Colorado, which operated from 1979 to 1989. During this period, numerous mechanical problems were encountered, all of which were resolved or well understood prior to shutdown. Meanwhile, in Germany, a 300-MW(el) plant, the THTR-300, demonstrated the “pebble-bed” core concept from 1987 to 1989. During the early 1970s, a number of orders were placed by U. S. utilities for large HTGRs. However, as a result of economic pressures existing at the time, these orders were canceled.

12.231. To assess risks, we need to relate failure probabilities with con­sequences. For this purpose, the event tree (§12.220) links an initiating event, with a probability determined by fault tree analysis, through a chain of events with evaluated probabilities, leading to radioactive release to the environment and transport to populated areas. Finally, the dosage received and known health effects provides a measure of the consequences of the initiating event. An example of such a result is shown in Fig. 12.18, which is a risk curve from the 1975 Reactor Safety Study. Here, the ordinate is the frequency (or probability) that fatalities of magnitude X or greater will be produced.

12.232. Modeling of the various steps leading to risk assessment results in a considerable challenge. We need to consider the source term and severe accident progression steps, atmospheric transport, and finally, health effects. Therefore, results such as those shown in Fig. 12.18 have substantial uncertainty bands, the magnitude of which have been debated. Typically, such bands cover about two orders of magnitude. However, since the failure probabilities are so small, the reactor risks are still orders of magnitude less than natural or other industrial risks.

12.233. As mentioned, the pioneering reactor risk assessment study was the 1975 Reactor Safety Study. Subsequent critiques identified various weaknesses in the methodology and questioned the confidence levels es­timated and some of the results. However, follow-up studies tended to confirm the general order of magnitude levels of the RSS results. Consid­ering that the RSS prediction of a meltdown probability of once in 20,000 reactor-years of operation, it is logical to conclude that the systems are acceptably safe, even if there is an order-of-magnitude uncertainty in the probability prediction.

12.234. A comprehensive risk assessment study with emphasis on severe accidents was completed in 1990 [31]. This effort, commonly referred to as the NUREG-1150 study was the most ambitious since the 1975 Reactor Safety Study (RSS). It utilized many methodology improvements devel­oped since the earlier study, particularly in the areas of core damage eval­uation, source terms, containment event tree development, and conse­quence modeling. New codes such as CRAC2 for consequence modeling,

including atmospheric dispersion, were utilized [32]. An important new feature was the use of an expert system approach as part of the uncertainty analysis.

12.235. The results, developed for five representative LWRs, predict risk levels somewhat lower than those estimated in the RSS. However, the overall uncertainty range envelopes the results of both studies. A great deal of detail was provided regarding the behavior of each of the plants studied and possible safety-related backfits to reduce some of the uncer­tainties identified.

12.236. Although general confirmation of earlier study meltdown prob­ability predictions was important, the NUREG-1150 study’s primary sig­nificance was to provide a much improved picture of plant component performance under severe accident conditions that provides guidance for design and regulation. Possible equipment backfits for each plant were identified and their cost-effectiveness determined. Information on the harsh environment during an accident provided a new basis for equipment qual­ification as part of the regulatory process. In fact, the study provides a basis for the reexamination of many safety-related equipment licensing requirements.

14.23. PWR shutdown to either a hot-standby or cold condition is ini­tiated by unloading the turbine-generator and corresponding plant systems automatically. For a hot-standby subcritical condition, the control rods are inserted, cooling loop flow is maintained with two of the four pumps, and the residual heat removal system is placed in service. To place the plant in cold shutdown condition, numerous adjustments to plant systems are needed, including blocking the action of the safety injection systems and increasing the dissolved boron concentration.

14.24. A BWR operational shutdown is one of short duration in which the reactor remains critical at about 0.01 percent of rated power and the turbine-generator is removed from the grid. Pressure is maintained and the steam produced is bypassed to the condenser, as needed. In an isolated shutdown, the reactor is not critical but the system is maintained at pressure with the residual heat removal system activated. For cold shutdown, the pressure is slowly reduced, with the cooling rate limited to about 50°C per hour and the residual heat removal system utilized.

15.61. Reactor behavior, as inferred from EBR-II tests and extensive analyses, indicates an ability to safety accommodate a wide variety of accident scenarios. However, traditional sodium-cooled reactor concerns regarding the positive sodium void reactivity effect and the possibility of fuel failure propagation in the event of fuel assembly blockage may have

/ Gas Expansion Module 3 Ш Shield 48

Reflector 42

Radial Blanket 33

(3 Driver Fuel 42

Internal Blanket 24

ф Control 6

^>3 Ultimate Shutdown 1

Total: 199

Fig. 15.6. ALMR reference metal core (General Electric Co.).

to be addressed. Also, an additional containment structure enclosing the primary system may be required.

15.62. The advantages of modular design yield cost estimates that are competitive with those for other advanced reactors and fossil-fueled sys­tems after about four full-size power stations are built and operated. Since the need to produce fissile material by breeding appears to be at least several decades away, there is time for a prototype plant test program to test further passive safety features and refine the system design. This could be done by first building a single module or as an alternative, a single commercial power block, if necessary funds are made available.

13.14. Most Westinghouse PWR systems have either three or four cool­ant loops, each with its own circulation pump and steam generator (Fig. 13.4). These pumps are normally of the single-stage, vertical centrifugal type, each driven by a 6-MW (8000-hp) motor. They are rated at about 6.2 m3/s (98,000 gpm), with a pressure developed of 7.4 x 105 Pa (108 psi). Not shown in the simplified coolant loop figure are connections to numerous auxiliary systems, such as those for safety injection, residual heat removal, and chemical and volume control.

13.15. A typical 1000-MW(el) plant steam generator design, with an overall height of about 20.7 m (68 ft) and an upper shell diameter of 4.4 m (14.5 ft), is shown in Fig. 13.5. Heated primary system coolant from the reactor vessel passes through the inverted U-shaped tubes, and satu­rated steam at 7.6 x 106 Pa(a), i. e., 1100 psia, is formed in the outer “shell” side. The upper part of the steam generator consists of a steam­drum section where centrifugal moisture separators and a steam dryer remove entrained water from the steam. The steam temperature is 291°C (556°F).

13.16. An important component of the steam-generation system is the system pressurizer which is connected to one of the primary coolant loops (see Fig. 13.5). A typical pressurizer is a cylindrical tank, about 16.2 m (53 ft) high and 2.44 m (8 ft) in diameter. During normal steady-state

operation it contains roughly 60 percent of liquid water and 40 percent steam (by volume). Electric immersion heaters in the lower part of the vessel can provide heating, whereas cold water sprayed into the steam space can provide cooling, as required. Both heaters and cooling sprays are operated by pressure signals. If, for any reason, the system pressure should drop, a low-pressure signal would actuate the heaters. The steam generated would then increase the system pressure. On the other hand, if the pressure should rise, a high-pressure signal would operate the cooling water spray. Some of the steam in the pressurizer would be condensed and the system pressure would decrease, if the pressure should increase beyond the control capability of the sprays, one or more motor-operated relief valves would open automatically and steam would be vented to a quench tank partially filled with water. Should the pressure in the quench tank exceed the design value, a rupture disc would allow the tank to vent to the containment sump. An important contribution to the Three Mile Island accident (§12.179) was the failure of the pressurizer relief valve to close

when the pressure was restored to normal. Therefore, assurances that valves would operate reliably have received a great deal of attention since that time.

13.17. Steam from the generators passes to the turbine system which consists of high-pressure and low-pressure stages on the same shaft. Par­tially expanded steam leaving the high-pressure turbine goes through moisture-separator and reheater units before further expansion in the low — pressure turbines. Steam drawn from the high-pressure section and from the steam generator serve as the heat sources for reheating. The exhaust steam from the low-pressure stage passes to the condensers, normally lo­cated under the turbines. The condensate provides the feedwater for the steam generators. Before entering the generators, however, the condensate temperature is increased in a series of feedwater heaters which use steam drawn from the turbines to heat the water.