14.46. As the economic desirability of extending the operating lifetime of existing plants became recognized, a longer view for nuclear plant main­tenance has been adopted. Rather than merely meeting regulatory re­quirements on an operating cycle-by-operating cycle basis, utilities have implemented aggressive long-term maintenance programs to avoid age — related degradation of safety-related systems, structures, and components. Such life-cycle management programs have as their objective the contin­uation of an “as new” plant condition so that a 20-year operating license extension can be applied for some years before the current 40-year license expires.

15.79. In this chapter we have looked forward to the rebirth of nuclear power and described some of the new technology that is being made avail­able to create a favorable climate for new plant construction. However, we cannot predict the future. Therefore, it is well to summarize here some of the nontechnical challenges that matter [10].

15.80. In the United States, a decision to build a new plant is made by an individual utility at the local level to meet the energy needs of its customers. National electricity usage growth forecasts do not necessarily apply at the local level. Nonnuclear options, including the upgrading of older units, are also available. Another consideration is the role of non­utility energy providers in developing generating facilities (§10.108).

15.81. Four topics are interrelated: investment risk, licensing stability, rate regulation, and public opinion. A utility can recover its investment only if initial cost projections are not changed by new licensing require­ments during construction and operation. Also, unreasonable limitations on rates charged customers must be prevented. In each of these areas, public opinion can play a role through political pressure on regulators. Thus, acceptance by local government and the local populance is very desirable as part of risk management strategy as well as good citizenship.

15.82. To conclude on a positive note, we should look ahead some years into the next century. Likely global population and economic growth will lead to a substantial increase in the demand for electricity. Solar power and other alternative energy sources may contribute to meet this demand, but only to a minor extent, leaving the burden to be shared by fossil and nuclear fuels. Assuming that unfounded public sensitivity to the risks as­sociated with high-level radioactive waste can finally be resolved and the environmental damage caused by fossil plants is generally realized, a gen­eral preference for nuclear power plants is likely to result [13].

REFERENCES

1. U. S. Senate Committee on Environmental and Public Works Hearing (101­861), “Role of Nuclear Energy in Meeting Future Energy Demands,” U. S. Gov’t Printing Office, 1990; “U. S. Electricity Needs and DOE’s Civilian Reactor Development Program,” U. S. General Accounting Office, 1990.

2. E. J. Bruschi and R. P. Vijuk, NucL Techno!., 91, 95 (1990); B. A. McIntyre and R. K. Beck, NucL Safety, 33, 36 (1992).

3. R. L. Huang et al., Trans. Am. NucL Soc., 63, 314 (1991).

4. G. Melese and R. Katz, “Thermal and Flow Design of Helium-Cooled

Reactors,” American Nuclear Society, 1984, Chap. 1.

5. F. A. Silady and A. C. Millunzi, NucL Safety, 31, 215 (1990).

6. P. R. Pluta et al., “PRISM,” Adv. NucL Sci. Technol., 19, 109 (1987).

7. С. E. Till and Y. I. Chang, “The Integral Fast Reactor,” Adv. NucL Sci.

Technol., 20, 127 (1988).

8. D. R. Pedersen and B. R. Seidel, NucL Safety, 31, 443 (1990).

9. K. Hanners et al., “The PIUS Principle and the SECURE Reactor Concepts,” Adv. NucL Sci. Technol., 19, 41 (1987).

10. L. R. Codey, “Proc. Topical Meeting on the Next Generation of Nuclear Power Plants,” American Nuclear Society, 1991.

11. H. A. Upton et al., Trans. Am. NucL Soc., 68, 355 (1993).

12. T. Matsuoka et al., NucL Safety, 33, 197 (1992).

13. C. Starr, Trans. Am. NucL Soc., 67, 46 (1992).

13.49. The advanced boiling-water reactor (ABWR) is a large, 1300- MW(el) system developed by an international team of BWR manufacturers benefiting from the extensive safety and performance experience of the many operating systems throughout the world. Two units are under con­struction in Japan, with commercial operation scheduled for the late 1990s. Some technical data are summarized in Table 13.4. The core power density has been reduced by about 10 percent compared with present BWR values given in Table 13.3. Various fuel assembly design options are available, but for orientation purposes, the fuel dimensions for the 8×8 lattice given in Table 13.3 are adequate.

15.25. Natural circulation, or more accurately, natural recirculation of the coolant, is the most interesting thermal-hydraulic feature of the SBWR.

It is dependent on the difference in density between the cooler returning and feed fluid mix in the downcomer leg and the steam-water mixture in the core and chimney. As pointed out in §9.131, the flow rate depends on the relative influence of the density difference “driving force” and the frictional flow resistance. Hence, a short core of relatively low flow resist­ance is needed to assure an adequate recirculation rate. To assure reactor stability, the core flow must be maintained above certain levels which depend on the reactor power.

15.26. Reactor stability during normal operation and anticipated tran­sients is an important design consideration for all BWRs. For example, reactivity feedback instability of the reactor core could result in power oscillations. Hydrodynamic channel instability could inpede heat transfer to the moderator and also drive the reactor into power oscillations. Finally, the total system stability depends on the basic process dynamics. Stability criteria are stated in terms of a decay ratio, which is defined as the ratio of the magnitude of the second oscillatory overshoot to that of the first overshoot resulting from a step perturbation. The lower the ratio, the more stable the system. SBWR stability decay ratios have been modeled and found to be at least a factor of 2 lower than those for currently operating BWRs [3].

12.217. Fault tree analysis may be used to predict the probability of system failure based on data or estimates of failure at the subsystem level. For this purpose, fault tree models, like the one in Fig. 12.16, are con­structed and a probability is associated with each event; the results are then combined, along the following lines. Suppose that a top event A is connected through an AND gate to three events at the second level for which the probabilities are Въ B2, and B3, respectively. The probability of A is then Вг x В2 x B3. On the other hand, if the three second-level events are connected with A by an OR gate, the probability of A is Вг + В2 + B3. The second-level events are themselves connected with third-
level events and the probabilities of these events in turn determine Bu B2, etc. By proceeding in this way through all the levels, down to the lowest, of a fault tree, the probability of the top event is determined.

12.218. It is evident that when a tree consists of many levels, as is commonly the case, the calculation of the probability can become very complex. As a general rule, therefore, fault tree models are first simplified as far as possible by using Boolean algebra techniques. Such simplification is aided by eliminating events of low probability and combining related events into a single event. Although the approach is straightforward, com­puter representation is generally necessary, even after simplification. Prob­abilistic analysis also makes use of event trees which are discussed in the next section.

12.219. There are a number of uncertainties in fault tree analysis that place error limits on the results. For example, because of the lack of experience of an event, e. g., a double-ended pipe break in the primary coolant system of a reactor, the probability of failure cannot be established and must be estimated. Allowance should also be made for human error and the unavailability of systems or components because of test or main­tenance requirements. Other uncertainties may arise from possible omis­sion of important faults and from unforeseen common-mode failures. Fur­thermore, in the fault tree model a component or subsystem is regarded as either failed or not failed; there is thus no allowance for a partial failure in which a subsystem may be operative but not at its full efficiency.

14.10. Routine startups for LWRs may either be from a cold condition or from a so-called “hot standby” condition. When the reactor is started for the first time or following a shutdown that required the system to be cooled down and depressurized, we have a cold startup. Hot-standby startup refers to a return to power operation when the cooling system is still near operating temperature and pressure such as the condition following a re­actor trip resulting from a minor transient, instrument malfunction, or turbine trip.

14.11. Prior to initial “commercial” startup, extensive testing is pre­scribed by the NRC to meet licensing requirements. Such tests include not only functional tests of all components and instruments, but also physics tests at zero power and testing at various power levels.

14.12. Precriticality checks are carried out before any cold startup to assure that the plant is ready to function. In a PWR, the coolant system boron concentration is adjusted and various auxiliary systems prepared. Frictional heat from the circulating pumps is then used to raise the primary coolant temperature to about 200°C at a pressure of about 2.8 MPa over a period of about 6 hours. After the pressurizer is adjusted, systematic withdrawal of the control rods to approach criticality is initiated.

14.13. In connection with control rod withdrawal, the importance of neutron detection instrumentation must be emphasized. As pointed out in §5.216, appropriate different neutron detectors are used in the source range, intermediate range, and power range. These are monitored by con­trol room instrumentation, which has been significantly modernized during recent years (§14.25).

14.14. The control rods are withdrawn intermittently rather than con­tinuously, with the neutron level allowed to stabilize before the rods are withdrawn further. A neutron-level startup rate well below criticality of about 0.5 decade per minute, corresponding to about a 50-s period, is common. However, when criticality is approached, the neutron count rate no longer reaches equilibrium between control rod withdrawals as a result of the contribution of the core neutron source.

14.15. Reactor physics data are taken at criticality and the power is raised to about the 1 percent level, where various adjustments are made to the secondary system. Reactor power is increased by further manual control rod withdrawal and the pressure and temperature raised to oper­ating conditions while the power level is still relatively low. The reactor is now at essentially hot-standby conditions at zero electrical load. The turbine — generator may be brought up to speed and connected to the grid when the power level is about 15 percent. At this point, additional power increase is normally switched from manual to automatic control. The loading rate to full power is normally limited to 5 percent per minute, determined by turbine considerations. To compensate for the reactivity power defect, some dilution of the coolant boron concentration is needed before reaching 100 percent power. A cold startup requires a minimum of about 13 hours.

14.16. Startup from the hot-standby condition follows a similar proce­dure except rod withdrawal is initiated when the cooling system is close to operating conditions. However, the relative negative reactivity contribu­tions of the dissolved boron and control rods must first be balanced properly so that criticality will occur with the rods above a certain limit. In the event of a trip, there will then be adequate rod negative reactivity available. Hot-standby startup time can be as short as 1 hour.

15.51. The advanced liquid-metal-cooled reactor (ALMR) design is based on the PRISM (Power Reactor, Innovative Small Module) concept de­veloped by the General Electric Co., while the fuel system is based on the IFR (integral fast reactor) concept developed by Argonne National Lab­oratory [6,7]. Liquid-metal-cooled fast breeder reactors have attracted the attention of reactor designers since the earliest days of the nuclear power industry because of their ability to transform fertile uranium-238 into fissile plutonium-239 efficiently using fast neutrons (§10.65). Furthermore, so­dium cooling allows the reactor to be at atmospheric pressure. However, commercial development of such systems was deferred in the early 1980s as a result of proliferation concerns and economic conditions at the time.

15.52. During recent years, the outstanding performance of new me­tallic fuels in EBR-II, a 20-MW(el) fast reactor that has been operating very satisfactorily since 1961, resulted in the development of the IFR con­cept, which features an inherently safe core with passive features and proliferation-resistant on-site fuel recycling. The modular approach offers the flexibility, standardization, and construction economies described pre­viously for other advanced passive systems. Although funding for continued development appeared uncertain in 1994, the concept is described briefly here as an example of a fast-reactor option with future potential.

13.5. A description of the PWR system starts with a summary of the significant design specifications as listed in Table 13.1. As an aid in com­paring the systems described in this chapter, a consistent format has been followed which indicates general specifications, as well as values for the core, fuel, thermal-hydraulic, and control features. Specifications for dif­ferent nuclear steam supply system (NSSS) models offered by the same manufacturer vary somewhat. Those listed in Table 13.1 are for a four — coolant-loop Westinghouse model, as described in a reference safety analy­sis report submitted to the NRC in 1974 [2]. This was about the end of the period when new reactors were ordered. Many currently operating Westinghouse four-loop PWRs were ordered somewhat earlier and have a core 12 ft in length to provide a thermal output of about 3400 MW. However, the 3800-MW version, which operates at the maximum power that the NRC will authorize, takes advantage of economy-of-scale savings. Such savings arise from the principle that fixed and operating expenses scale up as only a weak function of size. Therefore, if the plant is base loaded, the generating cost, on a unit energy basis, would be less for the larger plant.

14.47. We have seen that licensing and other regulatory considerations affect practically every aspect of reactor operations. In addition to the activities described, several additional points are worthy of mention here. Adherence to regulations is monitored by on-site NRC personnel. Reactor

operators are licensed and must meet strict fitness for duty requirements. The industry has provided extensive training programs to qualify operating personnel which are subject to accreditation by INPO.

14.48. During maintenance operations, radiation exposures must be kept “as low as reasonably achievable” (ALARA) (§6.63). Radiation pro­tection and exposure times must be such that doses are as far below the numerical limits as is practical, as confirmed by appropriate monitoring and reports to the NRC.

14.49. In-service inspection of components and auxiliary systems is re­quired to assure the continued integrity of a reactor’s primary system and safety-related equipment. In connection with this and other license-related activities, reports to NRC are required.

The SI units are a coherent and consistent set of units that can be used in calculations without the need for conversion factors. For the present purpose two classes of SI units may be distinguished: base units and derived units.

There are seven dimensionally independent base units, but only the following five are used in this book:

periments, 3.157

[1]Although Ec should include the decay energies of all radioactive species formed as a result of nonfission neutron reactions, only those of relatively short half-life contribute to the energy release in the state of pseudo-equilibrium attained by the reactor soon after startup.

[2]Commercial power reactors have radial fuel regions with different enrichments in uranium — 235; as a result, the power distribution differs from that for uniform enrichment assumed here. Also, the axial power density is likely to be skewed for reload core loadings as a result of previous operational history (see Chapter 10).

[3]The quantity q/A, having the dimensions of heat/(time)(area), i. e., W/m2, is called the heat flux.

[4]For values of bla < 2, the quantity In (b/a), which appears here and in heat-transfer equations for clad fuel rods (§9.47 et seq.), is well approximated by 2(b — a)/(b + a).

[5] Certain fluids, e. g., suspensions, which do not obey the Newton equation, are referred to as non-Newtonian fluids.

[6]The saturation temperature is the temperature of the saturated vapor, i. e., saturated steam, at the existing pressure.

[7]The designation “W-3” arises from the fact that it is the third such correlation developed by the Westinghouse organization.

[8]If English units are used in equation (9.39) or in subsequent equations, gc, the factor relating force and mass, must be used as appropriate.

Tn the equivalent Darcy-Weisbach equation a friction factor equal to 4/is used.

[9] Since the velocity head is equal to the kinetic energy of a unit mass of flowing fluid, the term arises from the similarity with the potential energy of a height (or “head”) of a unit fluid mass.

[10]

[11]For further information, standard reference books and manufacturers’ catalogs should be consulted. The loss coefficients Kx are often expressed in terms of an “equivalent” length of straight pipe.

[12]In SI units, R = 8.314 J/K • mol, and the molecular weight (mass of one mole) is in kilograms (§1.12). Thus, the velocity of sound in helium at 1000 К is

[13]For a fuel rod of given dimensions and fuel composition, the heat flux, i. e., the heat flow per unit area q/A, is proportional to the local heat-generation rate (or heat source) per unit volume Q the latter is equivalent to the local power density which is determined by 2/ф [cf. equation (2.54)].

+Other parameters, e. g., fuel cladding temperatures, become limiting under emergency conditions (Chapter 12).

[14]This postulate implies that the following analysis gives a good approximation provided dtldx « dtldn, where n is a coordinate normal to x.

[15] Since the average core outlet temperature in a PWR is below the saturation temperature, little or no vapor leaves the reactor vessel. However, there is substantial subcooled (local) boiling in the flow core to the heated surface, and bulk boiling may occur in the “hot channels.”

[16]It can be readily shown that the heat flux at the outer surface of a clad fuel rod is Qa2l 2b, where Q is the average rate of heat generation (by fission) per unit volume of fuel, a is the radius of the fuel pellet, and b is the outer radius of the clad rod. These three quantities are affected by the parameters in Table 9.2.

Statistical subfactors Based on ± Зет

Rod pitch and bowing 1.045

Clad rod diameter 1.042

Statistical combination 1.062

Nonstatistical subfactors

Inlet flow maldistribution 1.03

Internal mixing, and boiling flow redistribution and diversion 1.10

Flow mixing 0.90

[17]The plant capacity factor is the ratio of the actual electrical energy (in kW • h) generated in a given period (usually one year) to the amount that could be produced if the plant operated continuously at its rated power during the whole period.

[18]A burnup of 2.85 TJ/kg U is equivalent to 33,000 megawatt-days/metric ton U (see Example 1.2).

[19]Another measure of relative biological hazard is the water (or air) dilution volume, sometimes called the toxicity, defined as the volume of water (or air) required to dilute a given amount of radioisotope to a concentration for drinking (or continuous exposure). However, this measure does not take into account solubility or ingestion processes.

[20]The NRC recommends that, unless a guaranteed and reliable source of water is available, e. g., an ocean, large river, or large lake, the ultimate heat sink should consist of two separate sources, such as two independent cooling towers or spray ponds, or a tower and a spray pond, or one of these together with some other water source.

[21]This tank of borated water is normally used for temporary storage of spent fuel elements during refueling operations.

[22]Some severe accident scenarios lead to containment failure (§12.99). However, contain­ment designs for future reactors are likely to include severe accident mitigation features.

[23]In a cold-leg break, the desired flow of injected water down the downcomer and up through the core would be resisted by the blowdown flow in the opposite direction. If the resistance is large, ECC water bypass could occur, in which the injected fluid flows around the downcomer annulus and out through the pipe break.

[24] According to the data in Table 6.1, the mass energy-absorption coefficient of gamma rays in soft body-tissue is roughly 1.1 times that in air over a wide range of energies.

[25]The scale developed by the International Atomic Energy Agency provides for seven levels, with each step representing a factor of 10 in severity. Thus a level 3 incident would be roughly 10,000 times less serious than the level 7 Chernobyl accident.

[26]The term fault as used here refers to a mechanical failure or malfunction but also includes any type of abnormal situation, such as deviation from operational procedures.

[27]HVAC = heating, ventilation, and air conditioning.

[28]For further information, see “The International System of Units (SI),” National Bureau of Standards Special Publication 330; “Standard for Metric Practice,” American Society for Testing and Materials.