Introduction

10.114. Whenever fissile material is handled, be it in the manufacture of fuel, loading of fuel assemblies, or storage of spent assemblies, nuclear criticality safety deserves attention. This topic normally refers to out-of­core operations. In-core reactivity safety considerations will be covered in Chapter 12.

10.115. All of the factors that influence the neutron balance affect crit­icality safety when fissile isotopes are being handled. The important con­siderations are the mass of the fissile isotopes, geometry, moderation, and reflection.

12.7. Defense in depth is the key design principle in nuclear reactor safety. The approach is to provide a series of philosophical and physical layers of protection against the release of radioactivity to the public. In other words, we have one barrier which, if breached, is “backed up” by a second barrier. The second barrier is backed up by a third barrier, and so on.

12.8. The first barrier, or goal, is philosophical, i. e., to prevent accidents from happening. Should an accident happen, the next layer of defense is to provide various countermeasures, in a sequential manner, to control the accident. In our defense, we then provide a series of physical barriers and further countermeasures to confine fission products that might be released. Finally, a containment structure which encloses the nuclear por­tion of the plant is provided “just in case” there is need for further backup. We will examine these principles in the following sections.

12.112. The noble gas fission products such as xenon are inert chemi­cally. During irradiation, they are partially released from the ceramic oxide

fuel matrix and occupy grain boundary regions and whatever other free space is available within the fuel rod. Upon failure of the cladding and, in turn, the coolant system boundary, they would travel into the containment and be available for transport to the environment without acting chemi­cally. Although the transition during this time period by radioactive de­cay to other elements that may be reactive is a possibility, the effect is small in light of other process uncertainties and may be neglected in approximation-level analyses.

12.146. Related to the system modeling codes there is the need to ex­amine the mechanical effects on components resulting from the strong depressurization waves during a LOCA. Since dimensionality is important for such analysis, the approach used is to start with a two-dimensional, homogeneous, equilibrium code to model the two-phase blowdown and then use a special-purpose code to model the motion of such components as the core barrel of a PWR [15].

9.9. The energy released in the core by fission appears in various forms, but mainly as the kinetic energy of the fission fragments, the fission neu­trons, and the beta particles resulting from radioactive decay of the fission products (see Table 1.2). The fission fragments are usually stopped within the fuel elements themselves; the small fraction that escapes into the clad­ding penetrates only about 0.01 mm. The beta particles of high energy may travel up to 2 mm in a cladding material such as zircaloy (§7.117), and so a large fraction of these particles may escape from the fuel element into the moderator or coolant, but they will not get out of the reactor core.

The fission neutrons lose most of their energy in the first few collisions with moderator atoms, and they travel distances ranging from some cen­timeters to a few feet. It is seem, therefore, that most of the heat from the three sources under consideration, comprising about 90 percent or more of the total energy generated, will be released within the reactor core.

9.10. The remaining 10 percent, or less, of the energy produced in fission appears as gamma rays which are distributed throughout the reactor core and surrounding components in a manner dependent on the specific materials and geometries involved. Thermal stresses as a result of gamma heating of the reactor vessel and surrounding radiation shielding must be considered by the designer (§7.30).

9.11. As we have seen (§2.213), heat generation by radioactive decay of the fission fragments continues after the fission reaction ceases. There­fore, provision must be made for cooling the fuel elements after shutdown. Particularly important are emergency core cooling features to be effective in the case of an accidental loss of normal cooling capability (§12.33). Also, the role of gamma heating affects the after-shutdown spatial distribution somewhat. For example, 1 hour after shutdown, the heat generation rate in the fuel elements will be about 1.5 percent of the operating value, whereas in the reflector and shield it will be approximately 10 percent of the rate during operation.

9.12. It was seen in Chapter 1 that the total energy released in fission, which ultimately appears as heat, is made up of contributions from a num­ber of sources. In general, the total energy, exclusive of the neutrino energy which is lost to the reactor system, may be expressed by

E « 191 + Ec (in MeV),

where Ec is the energy liberated as a result of various parasitic neutron capture processes, e. g., nonfission capture in uranium-235 and uranium — 238, and capture in moderator, coolant, structure, etc.; this includes the energy of the capture gamma radiations and the decay energies, i. e., the energies of the alpha and beta particles and gamma rays, of any radioactive species that are formed by parasitic neutron capture.[1] Since the value of Ec will obviously depend upon the nature of the materials present in the reactor core, it is evident that the total amount of heat produced by fission will vary, to some extent, from one type of reactor to another.

Example 9.1. Determine the total energy release in the core of a pressurized-water reactor (PWR) having volume fractions of uranium oxide (U02), water, and iron of 0.32, 0.58, and 0.10, respectively. The oxide fuel (density = 10.2 x 103kg/m3) has an average enrichment of 2.8 percent 235U, and the average cooling water density is 0.69 x 103 kg/m3. The energy released per neutron captured in uranium, water, and iron may be taken as 6.8, 2.2, and 6.0 MeV, respectively.

The solution requires the calculation of the macroscopic neutron capture cross section for each component (except for the oxygen in U02 which is very small) as well as the 235U macroscopic fission cross section. The final result depends on ratios of cross sections rather than absolute values; hence, the tabulated values for 0.0253-eV neutron microscopic cross sections may be used in the calculations. The capture-to-fission ratios are then as given in the following summary:

The energy released as a result of nonfission captures is

Ec = (6.8 x 0.16) + (6.8 x 0.17) + (2.2 x 0.076) + (6.0 x 0.19) = 3.6 MeV

The total energy released per fission in this core is thus

E = 191 + 3.6 « 195 MeV = 3.1 x 1011 J.

9.121. Our discussion of flow boiling (§9.94 et seq.) introduced the complexity of two-phase flow. As we shall see, the ability to predict flow rates is of major importance in evaluating the safety of water-cooled re­actors. Therefore, developing a good understanding of two-phase flow phenomena has been the objective of much recent work [15]. However, space limits our presentation to a few introductory ideas.

9.122. In single component two-phase flow, we have a liquid flowing concurrently with its own vapor with evaporation or condensation from one phase to the other occurring in accordance with whatever heat transfer is taking place involving the system. The vapor phase, which is a gas, follows the laws of compressible flow as described in a significant branch of fluid mechanics. To develop a model for the system, a classical approach, as described in various texts [15], is to write equations for the conservation of mass, momentum, and energy over an element of the flow channel and to evaluate the pressure gradient. The necessary, rather complicated, set of relationships can best be managed by appropriate computer codes. For a reactor core, for which the picture is further complicated by intercon­nected flow regions, the calculational development is known as subchannel analysis (§9.135).

10.48. Advances in both computing methods and core modeling during recent years have led to automation of many of the iterative steps in the design process. Although the development of optimization procedures has received much attention at universities, the parameters that affect the “best” strategy tend to change over the long planning periods required. Therefore, electric utilities and fuel vendors tend to concentrate on the development of fast and efficient procedures that allow a wide range of loading scenarios to be explored to take advantage of changing economic and regulatory conditions.

The Haling Principle

10.49. In connection with both automation and optimization proce­dures, the Haling principle has proven to be a useful concept [13]. It was shown that if a power shape (power distribution) could be maintained constant throughout the operating cycle by proper management of soluble poison, solid absorbers, and movable control rods, the peak-to-average power value will be at a minimum. Although in practice it may be difficult to achieve such a goal, which would require matching control rod with­drawal with burnup effects, the concept is useful as a reference.

10.50. The Haling principle has been used in some fuel management schemes to separate the multicycle management problem from that of an individual cycle. However, the 1988 revision of regulatory emergency core cooling system modeling methods had the effect of increasing allowable core peaking factors (§12.132). Therefore, there has been a shift in design practice from attempting to achieve minimum core peaking to an emphasis on minimum fuel cycle costs. Such methods do not make use of the Haling principle.

Introduction

11.30. After about a З-year residence in the core, LWR spent-fuel as­semblies are discharged and stored in an on-site water pool. When most present reactors were built, about a 5-month cooling period was planned, after which the fuel was to be shipped to an off-site interim storage facility or to a reprocessing plant. However, since neither option has become available, spent assembly storage has presented a logistic problem that we will examine.

11.31. Consideration of the options available for managing spent fuel assemblies should begin with a study of such characteristics as their isotopic mix, contributions to radioactivity, heat generation, and physical features.

12.56. Transients caused by operational (and other) occurrences can cause an imbalance between the rate of heat generation in the fuel (or thermal power) and the rate of heat removal by the coolant. If the former exceeds the latter, there is danger of overheating and hence of damage to the fuel-rod cladding. In many cases, the imbalance can be rectified au­tomatically by the reactor control system without interruption of normal operation. The reactor protection system is designed to trip the reactor if the imbalance is too severe to be handled by the control system. The redundancy and diversity available in the control and protection systems make their total failure highly improbable.

12.57. The relationship between heat generation and heat removal rates may be considered in terms of the maximum heat flux (or heat generation rate) in a coolant channel and the critical heat flux (§9.98) at which some fuel cladding damage may be expected to occur. In order to prevent such damage, the critical flux must be well above the maximum (actual) flux, as depicted in Fig. 9.22, for example. An increase in the thermal power of a reactor will increase the (actual) flux, whereas a decrease in the coolant effectiveness will decrease the critical heat flux. Thus, transients capable of causing either (or both) of these effects will increase the possibility of cladding damage.

12.187. The destruction of Unit 4 of the Chernobyl Atomic Power Sta­tion in Ukraine on April 26, 1986 was the worst nightmare of nuclear engineers throughout the world. Attempts to convince a skeptical public that nuclear power is indeed a safe energy option would now be more difficult than ever, even though the reactor design is unique to the former Soviet Union and inherently unstable. A first step in discussing the accident and its impact is to become familiar with the system.

12.188. The reactor, one of the RBMK-1000 type, was boiling light — water-cooled and graphite-moderated. Fuel element subassemblies, within pressure tubes through which the coolant flowed, contained 18 rods ar­ranged in concentric rings. At full power, the coolant channel exit steam quality was 14 percent. The core, 7 m high and 11.8 m in diameter, con­tained 1661 such channels. Refueling was accomplished at full power by a machine in a bay above the core, as shown in the Fig. 12.14 plant arrange­ment. A 2 percent enrichment was used, with 211 control rods of various types needed for reactivity compensation. The rated thermal power was 3200 MW.

12.189. As a result of the massive size of the reactor and fuel handling equipment, it was considered impractical to have a complete containment such as used for all U. S. reactors. Instead, an accident localization system, as indicated in Fig. 12.14, was provided. This consisted of various sealed compartments that enclose the circulating pumps, large piping, and other components, the failure of which could lead to a LOCA. This compartment system vents to a suppression pool. The cooling channel refueling con­nections were unprotected. Hence, the accident localization system was designed for a different type of accident, i. e., a large pipe LOCA, rather than the massive fuel channel failure that occurred.

12.190. Since graphite is the moderator in the RBMK-1000 reactor, voiding of the coolant, which is a neutron absorber, increases the reactivity. This positive reactivity coefficient was significantly greater at low power. Therefore, 700 MW(t) was specified as the minimum permissible contin­uous operating power level. An interesting positive reactivity effect is in­troduced when control rods are inserted as a result of the displacement of water at the bottom of the control rod channels by graphite control rod

Primary coolant boundary areas not protected by AtS Semi protected mm

Protected areas

followers. During post-accident analysis, it was also concluded that the emergency rod insertion rate was slow by western standards.

12.191. The accident occurred during a test of turbine-generator coast­ing-down power, which was used to drive an emergency feedwater pump for about 1 minute in the event of the loss of off-site power, with some of the desired electrical conditions simulated. Although the intended reactor power for the test was 700 MW(th), errors by the operators resulted in a power loss to 30 MW(th), where xenon growth, particularly at the bottom of the core, from the previous higher power operation made it difficult to increase the power without withdrawing almost all of the control rods. Even with such action, only a power level of 200 MW(th) could be achieved, a level in violation of operating procedures because of the inherent insta­bility of the reactor.

12.192. The test plan was initiated at this low power level by starting additional recirculating pumps as called for by an electrical simulation, which had the effect of reducing core voids and causing additional control rod withdrawal. As a result of operating difficulties, various protection devices were blocked out by the operators. When the recirculating pumps were allowed to coast down, as planned, coolant flow decreased and voids re-formed very rapidly in the pressure tubes, which increased reactivity because of the positive void coefficient, particularly at the bottom of the core. An emergency scram (trip) was initiated manually almost immedi­ately, but the almost fully withdrawn rods could not be inserted fast enough to prevent a prompt critical power excursion. In fact, the rod insertion introduced some additional reactivity at the core bottom as a result of water displacement by the graphite followers.

12.193. The rapid vaporization of the coolant in the pressure tubes generated a shock wave that ruptured most of the tubes. Apparently, there were two excursions, seconds apart, the second a result of almost complete coolant voiding. The fuel became molten and generated an immense quan­tity of steam, which blew the 1000-ton steel and concrete biological shield off the top of the reactor. Hydrogen, formed by the reaction of fragmented cladding and water, exploded, severely damaging the building. Pressure tube ruptures provided an inlet and an outlet for air to feed combustion of the graphite, which was probably ignited by the exothermic zirconium — niobium oxidation reaction. The fire continued for several days and cer­tainly complicated the management of the accident. The accident was cat­egorized as level 7 on the International Nuclear Event Scale (INES).[25]

12.194. In late 1986, after radiation levels had decreased somewhat, the damaged reactor building was enclosed by a concrete and steel shell, or “sarcophagus.” Several years later, explorations inside the sarcophagus revealed that about 96 percent of the original fuel is contained in solidified “lava flows” in chambers below the reactor vault and in the form of dust and particles distributed inside the building. Since it was necessary to construct the sarcophagus as an emergency measure upon damaged foun­dations, it is expected that replacement will eventually be necessary.

12.195. The total release of particulate fuel from the core is estimated at 3.5 percent of the original inventory. This corresponds to a radioactivity release into the environment on the order of 50 million curies. Fallout over parts of the former Soviet Union and other countries was widespread. Considering the most biologically sensitive fission products, 100 percent of the rare gases was released, about 20 percent of the iodine, and roughly 13 percent of the cesium. The estimated 2 million curie release of the 30- year half-life cesium-137 is the most significant long-term contamination contribution. The initial 32 fatalities all occurred on-site. The long-term consequences of the exposure received by about 200 plant workers treated for radiation sickness at the time of the accident and off-site exposure to fallout by some of the nearby general population remains uncertain.